te_1596

IAEA-TECDOC-1596-CD

Improvement of Technical Measures to Detect and Respond to Illicit Trafficking

of Nuclear and Radioactive Materials

Results of a Coordinated Research Project 2003–2006

July 2008

The originating Section of this publication in the IAEA was:

Surveillance, Seals and Remote Monitoring Section International Atomic Energy Agency

Wagramer Strasse 5 P.O. Box 100

A-1400 Vienna, Austria

IMPROVEMENT OF TECHNICAL MEASURES TO DETECT AND RESPOND TO ILLICIT TRAFFICKING OF NUCLEAR AND RADIOACTIVE MATERIALS

IAEA, VIENNA, 2008 IAEA-TECDOC-1596-CD

ISBN 978–92–0–156308–8 ISSN 1011–4289

© IAEA, 2008

Printed by the IAEA in Austria July 2008

FOREWORD

Equipment to detect illicit trafficking of nuclear and other radiological materials at borders and in a country has its own specific requirements and is very different from equipment used in other radiation monitoring cases. Automated and manual measurements need to be done in the field, often outdoors, at land or sea border crossing points or at airports. The free flow of goods and passengers must not be greatly impacted, thus requiring that the measurement time be short. The design needs to take into account that the users of the equipment are not experts in radiation detection, thus the results of the measurements should be easy to understand. This coordinated research project (CRP) on Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials was undertaken in 2002 to address technical difficulties in these areas, and to form a consensus regarding the most important technical requirements for border monitoring equipment. A problem which proved to be very troubling to users of border monitoring equipment, was becoming obvious at the start of this CRP: the radiation detection systems routinely generated a considerable number of genuine radiation alarms (from, e.g. naturally occurring radioactive material (NORM) in the shipped materials), which were of no significance to illicit trafficking but nevertheless required a response. Without effective tools in the hands of the customs officers or border guards to quickly categorize the isotope, the entire process of border monitoring would not work. Many contract and agreement holders worked on ways aspects of resolving this problem. As a result, improved handheld instruments were designed (several of which have thus far been commercialized), a set of guidelines on simplifying user interfaces was developed, and techniques for categorizing alarms in various circumstances were studied. In addition to the above, other investigations relating to specific issues in border monitoring were undertaken, such as ways to detect shielded nuclear materials, and ways to verify the isotopic and activity level of contents of legal shipments. Specific proposals to achieve the desired goals were made, and demonstrated. The following is a compilation of the final results presented at the third and final research coordination meeting, held in Vienna, 24–28 April 2006. In summary, one can say that this CRP has contributed significantly to the improvements of equipment needed to combat illicit trafficking of nuclear and other radioactive materials. This CRP was supported by the IAEA Office of Nuclear Security, Department of Nuclear Safety and Security with funds from the Nuclear Security Fund, and by the Division of Technical Support, Department of Safeguards. The IAEA officers responsible for this publication were R. Arlt and K. Baird, both of the Division of Technical Support, Department of Safeguards.

EDITORIAL NOTE

This CD-ROM has been prepared from the original material as submitted by contributors. Neither the IAEA nor its Member States assume any responsibility for the information contained on this CD-ROM. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.

CONTENTS

Foreword

Summary of the Coordinated Research Project

DETAILED REPORTS

Feasibility of Using Hand-Held Isotope Identifiers for Attribute Testing of Radioactive Sources in Shipment Containers N. Kravchenko, G. Yakovlev, D. Babich, V. Klyuchkin, A. Kravtsov, D. Danko, A. Lekomtsev, A. Churakov, M. Karetnikov

Verification of the Design Information of Shipment Containers Using Gamma-Spectrometry S. Korneyev, A. Khlimanovich, B. Martsynkevich, E. Bystrov

The vehicle monitor for the detection of radioactive materials Q. Zhang, J. Ma, M. Han, Y. Huang

Development and Demonstration of the New Methods for the Detection of Hidden HEU L. Meskhi, L. Kurdadze, I. Takadze, G. Cirekidze, N. Gogitidze

Comparative Study of New Scintillation Materials in Application to the Border Detection Equipment M. Moszynski, A. Syntfeld, M. Gierfik, A. Nassakski, T. Szczesniak

Analysis of procedures of testing the technical means for detection, localization and identification of nuclear materials and radioactive substances G. Yakovlev, I.Bannykh, V. Bojko, V. Bezverbniy, I. Chirkina, A. Gromov, N. Kravchenko, D. Danko, a. Lekomtsev, I. Bryagin, M. Karetnikov

Investigation of the Alternative Gamma-Spectropic Detectors for Quantitative Determination of Activity Isotopes "TYK" (Standard Shipping Containers) S. Ulin, V. Dmitrenko, K. Vlasik, Z. Uteshev, N. Kravchenko, I. Bannyh, I. Chirkina, Y. Popkov, A. Korotkov, A. Dorin, N. Ivanova, A. Ischenko, R. Ibragimova

Using Associated Particle Technique for Detection ff Shielded Nuclear Material A. Kuznetsov, D. Vakhtin, M. Zubkov, D. Vakhtin, a. Evsenin, O. Osetrov, I. Gorshkov

Elaboration of Methodical Recommendations on Application of Technical Facilities at Customs Control Over the Fissionable and Radioactive Materials Trafficking Across the Customs Border A. Borisenko, V. Kustov, L. Eliseenko, V. Temchenko, O. Aliokhina, B. Semerkov, N. Kravchenko, I. Bannykh, D. Danko

Research and Development of a Hand-Held Neutron Monitor, and Feasibility Studies of a 6LiI(Eu) Crystal Based RID and SPRD M. Majorov, A. Lebedev, O. Kraev, A. Kazimov, S. Kolchev

Radiation Monitoring with NORM Detection of Vehicles at Borders at Stand-Still V. Petrenko, Y. Karimov, N. Shipilov, A. Podkovirin, M. Fazilov, B. Yuldashev

Ruggedized Detection Probe for Field Use with Large Volume Coplaner CZT Detector L. Grigorjeva, V. Gostilo, A. Loupilov, I. Lisjutin, R. Rjabchikov

Investigation of Different Scenarios that can be Used to Mask Nuclear Material with Other Gamma Emitters M. Reinhard, D Prokopovich, D. alexiev, N. Dytlewski, D. Hill

Reduction of the Frequency of Innocent Alarms in Border Monitors H. Boeck, M. Swoboda, M. Schrenk, S. Hengster, V. Schwarz

Illicit trafficking: Monte Carlo modelling of shielded uranium source gamma spectra from the Nal detector S. Abousahl, P. Ragan

Gamma Spectra Measurements of Various Radiation Sources Using a LaBr3 P. Mortreau, A. Fernandes, R. Berndt

SHORT SUMMARIES

Feasibility Study of a Wrist Watch Based Radiation Survey Meter (Phase I) and Isotope Identying Gamma Spectrometer (Phase II) A. Wolf, F. Gabriel

Procedures for, and Testing of, Border Monitoring Equipment R. Kouzes

Testing and Analysing Instrumentation and Procedures for Radiation Detection and Identification P. Beck, G. Sdouz

Capacity Building for Detection and Response to Illicit Trafficking of Radioactive Materials M. Melich, M. Dobiaš

Detection of Nuclear Materials, Including Concealed Highly Enriched Uranium, Using Enhanced Detection Methods J. L. Jones, D. R. Norman

Inspection of Shipping Containers for Undisclosed Radioactive Materials V. Valković, S. Blagus, D. Sudac

Characterization of Various Survey Meters through Car-Borne Survey in Java Island as a Basis Data for Searching Orphan Sources U. Pande Made, I. Yuwono, Y. R. Akhmad, S. Sjamsoe, S. Subiharto

Development and Utilization of Nuclear Analytical Techniques for Analysis of Nuclear Forensic Materials H. Demirel, P. Arikan

Development of Recommendation, Guidelines and Methods for Coast Guard Officers on Measurement Using Portable and Hand-Held Isotope Measurement Devices M. Divizinyuk, O. Bleshenko, T. Dudar, Yu. Sabulonov, V. Nasarenko, Yu. Kostenko, G. Lisitshenko

ANNEX I: List of the Participants in the Coordinated Research Project ANNEX II: Usability Guide for Manufacturers of Radiation Monitoring Devices

Summary of the Coordinated Research Project

Radiation monitoring for safety and security purposes has a long history. The first detection systems were developed for, and used by, military and later civil nuclear facilities, as well as for geological uranium exploration. During the 1980s, the Chernobyl accident produced a major release, which added to the already widespread fallout from atmospheric nuclear weapons tests in the atmosphere. This prompted many countries to use radiation monitoring systems on their territories. Automated radiation monitoring systems were also developed for, and installed in, nuclear facilities to monitor the material flow as a nuclear safeguards measure. In addition, many measurement stations were installed around the world to monitor airborne radiation, which would be an indicator of nuclear weapons tests.

Starting in the late 1980s and throughout the 1990s, the concerns over (i) abandoned radioactive sources in scrap, and (ii) reported cases of illicit trafficking in nuclear and other radioactive materials, prompted the desire for additional development and deployment of specialized radiation monitoring equipment, with the main focus on border crossing points. These concerns were strongly enhanced after the terrorist attacks of September 11 2001 in the USA, as these events strengthened the fears that terrorists might combine explosives with radioactive sources (dirty bombs) or even improvised nuclear weapons to disrupt normal life in a society or cause severe damage.

Equipment to detect illicit trafficking at borders and in a country, the topic of this coordinated research project (CRP), has its own specific requirements and is, although similarities exist, very different from equipment used in other radiation monitoring cases. Automated and manual measurements need to be done in the field, often outdoors, at land or sea border crossing points or at airports. Changing background conditions need to be taken into account when alarms are triggered. The free flow of goods and passengers should be impacted to the minimum extent possible, thus requiring that the measurement time be short. The equipment design needs to take account of the fact that the users of the equipment are not experts in radiation detection, thus the results of the measurements should be easy to understand: ideally, pass/fail messages would be given immediately and unambiguously — a challenging task for developers and vendors.

In 2003, shortly after the CRP had commenced, it had already been demonstrated that the sensitivity of portal monitoring systems had become adequate, in particular after a series of national and international tests. However, a new problem had become obvious, which was troublesome to the users. The sensitive detection systems routinely picked up a considerable number of radiation alarms which were of no significance to illicit trafficking but nevertheless required a response. They were caused either by medical isotopes in persons or naturally occurring radioactive materials (NORM) in transported goods. Without effective tools in the hands of the responsible parties to quickly categorize the isotope which had caused the alarm the concept of border monitoring would not work.

Gamma spectrometry and isotope identification using hand-held gamma spectrometers or even automated spectrometric systems appeared to be the way to solve this dilemma. While gamma spectrometry under laboratory conditions was already a mature technology in 2003, this was not the case for small, manually operated hand-held gamma spectrometers. Early instruments were plagued by instabilities and nonlinearities of the energy scale, leading to failures in the identification of the isotopes. The identification software was struggling with the low statistics of the gamma spectra taken with short measurement time, with low resolution of the scintillation detectors used and small gamma peaks on a high background of scattered gammas. Therefore agreements and contracts to support the improvement of this important equipment class became a focus of the CRP.

Furthermore, the slow throughput, two step detection/categorization response bothered users who wanted a combination of alarm and immediate categorization. This was particularly needed in cases where frequent innocent or nuisance alarms caused unnecessary delay and excitement of the public. Common examples include the highly visible response to the detection of a medical isotope in a person at an airport, or the high frequency of alarms at border crossing points caused by trucks transporting NORM. In addition the ease of use of the instruments was often quite poor, resulting in the frustration of the users at remote border crossings, etc., who were experts in other fields but not experts in radiation measurements.

Under the CRP significant scientific/technical contributions were made by contract and agreement holders and invited experts to address the above described problems. Below is a list of topics investigated, with reference to the reports or summaries contained herein:

• Detection materials and detector response

o Illicit trafficking: Monte Carlo modelling of shielded uranium source gamma spectra from the Nal detector (Abousahl et al.).

o Gamma spectra measurements of various radiation sources using a LaBr3 (P. Mortreau et al.).

o Comparative study of new scintillation materials in application to the border detection equipment (Moszynski et al.).

o Characterization of various survey meters through car-borne survey in Java Island as a basis data for searching orphan sources (Pande Made et al.).

o Investigation of different scenarios that can be used to mask nuclear material with other gamma emitters (Reinhard et al.).

• New instrumentation development

o Reduction of the frequency of innocent alarms in border monitors (Boeck et al.).

o Ruggedized detection probe for field use with large volume coplaner CZT detector (Grigoreva et al.).

o Research and development of a hand-held neutron monitor, and feasibility studies of a 6LiI(Eu) crystal based RID and SPRD (Majorov et al.).

o Radiation monitoring with NORM detection of vehicles at borders at standstill (Petrenko et al.).

o Investigation of the alternative gamma-spectropic detectors for quantitative determination of activity isotopes ‘TYK’ (standard shipping containers) (Ulin et al.).

o Feasibility study of a wristwatch based radiation survey meter and as a isotope identifying gamma spectrometer (Wolf et al.).

o The vehicle monitor for the detection of radioactive materials (Zhang).

• Testing and implementation procedures for border monitoring equipment and support facilities

o Testing and analysing instrumentation and procedures for radiation detection and identification (Beck et al.).

o Elaboration of methodical recommendations on application of technical facilities at customs control over the fissionable and radioactive materials trafficking across the customs border (Borisenko et al.).

o Development and utilization of nuclear analytical techniques for analysis of nuclear forensic materials (Demirel et al.).

o Development of recommendations, guidelines and methods for coast guard officers on measurement using portable and hand-held isotope measurement devices (Divizinyuk).

o Procedures for and testing of border monitoring equipment (Kouzes).

o Capacity building for detection and response to illicit trafficking of radioactive materials (Melich).

o Analysis of procedures of testing the technical means for detection, localization and identification of nuclear materials and radioactive substances (Yakovlev et al.).

• Verification of contents of sealed shipment containers (involving gamma radiation probes or detection)

o Detection of nuclear materials, including concealed highly enriched uranium, using enhanced detection methods (Jones et al.)

o Verification of the design information of shipment containers using gamma-spectrometry (Korneyev, et al.).

o Feasibility of using hand-held isotope identifiers for attribute testing of radioactive sources in shipment containers (Kravchenko, et al.).

• Verification of contents of sealed shipment containers (involving neutron radiation probes or detection)

o Using the associated particle technique for detection of shielded nuclear material (Kuznetsov et al.).

o Development and demonstration of the new methods for the detection of hidden HEU (Meskhi et al.).

o Inspection of shipping containers for undisclosed radioactive materials (Valcovic et al.).

The yearly research coordination meetings (RCMs) were also used to develop a set of technical specifications for border monitoring instruments, discussing and agreeing them with a group of users, developers, including experts of the IEC and ANSI standards committees, and other standards drafting groups. This activity was supported by a workshop at the JRC Ispra, where multiple instruments were used to evaluate test specifications and associated test procedures. As a result of this work, a technical specifications publication, IAEA Nuclear Security Series No. 1, Technical and Functional Specifications for Border Monitoring Equipment (2006) was produced. In these same RCMs, it was repeatedly noted by uses that the instruments developed by manufacturers were not ideally suited for use by non-experts. Based upon these complaints, a usability guide for user interface design and instrument ergonomics was developed, presented and discussed. This document is included herein (Annex II).

Identified needs for further research

During discussions at the third and final RCM for this CRP (April 24–28 2006), extensive discussions took place regarding the current state of the art for radiation border monitoring and the challenges that the community currently faces. As in 2003 the main issue involved the generation of alarms by non-threat isotopes. The possibility of discriminating against medical isotopes had been examined during this CRP, and initial successes were documented It was the consensus of the RCM meeting that the biggest problem at this time still is the discrimination of alarms caused by NORM from those due to illicit movement of materials. These NORM alarms, typically found at sites where cargo routinely passes, impede the natural flow of commerce and limit the effectiveness of the radiation monitors. Thus, improved equipment and techniques are needed to (i) reduce the number of NORM alarms and (ii) to allow inspection agents to more quickly and effectively determine the nature of radiation alarms. It is for these topics that the greatest need for further research was identified.

Reducing NORM alarms through the use of spectroscopic portals, which investigate the isotopic signature of the alarm, was one identified method for achieving these goals. The spectroscopic portal has the advantages of larger detectors and more sophisticated algorithms over hand-held instruments. However, before these instruments can be deployed effectively, their limitations due to statistics and ‘masking’ must be studied carefully. Masking is when the larger radiation signature of either NORM, industrial or medical material is used to cloak the smaller signature of illicit nuclear material. In-field installation of these instruments without understanding their limitations could result in the compromising the effectiveness of the entire illicit trafficking monitoring system.

Due to the large number of plastic based radiation portal monitors that have already been deployed worldwide, combined with their much lower unit cost, research to improve the information obtainable from plastic based radiation monitors should also be performed. The capability needs to be developed and implemented so that the inspection agent has more information at his or her disposal to quickly decide if further inspection of an alarming vehicle is warranted (such as accurately identifying which vehicle generated the alarm, cross-checking the cargo manifest of that vehicle, and the ‘radiation profiles’ (time history of the measured radiation levels) of the individual detectors). The development of the capability of automatically supplying this information to the inspection agent would dramatically increase the agent’s ability to resolve alarms.

Spectroscopic portals are not applicable to all monitoring situations and hand-held instruments will still play an important role in resolving radiation alarms. Thus, research to improve hand-held radioisotope identifier devices (RIDs) should also be continued. This research should include (i) the pursuit of better identification algorithms (both in identification capability and reliability) and (ii) the investigation of higher resolution crystals.

The development of a hand-held neutron search instrument that further approaches the detection capability of the radiation portal monitor would also be of great assistance. Although there are no NORM neutron sources, the operation of neutron detector systems results in statistical false alarms that require verification. The radiation portal monitors have much more detection capability than existing hand-held instruments. Therefore, the verification of neutron alarms can only be quickly performed by passing the vehicle through the monitor a second time, which is not always practical. A hand-held neutron search instrument would be a significant asset in resolving these alarms.

A final research area that was discussed was the need to increase the capability of the inspection agents to resolve alarms by developing improved ‘reach-back’ capabilities. The agents desperately need an enhanced capability to acquire help from experts with a higher level of training in the use of radiation detection equipment and the interpretation of data from these devices. Operational experience has shown that no amount of training can prepare the field agent for all of the radiation alarm scenarios. The ‘reach-back’ capability is essential to effective operation of the monitoring systems. The capability could be improved by developing better standardized data formats, data transmission proficiency, and data analysis tools. The establishment of an international database for radiation monitoring information was also discussed.

A final topic of discussion, which is not related directly to research and development, was the need to establish guidelines and techniques of the measurement and documentation of the international shipment of commercial radioactive material. International commercial shipments of radioactive material can be used to hide the shipment of illicit isotopes or quantities of radioactive material. International agreements and techniques need to be developed to prevent these legal shipments from being used by nuclear terrorists to acquire illicit material.

The conclusion of the discussions that took place during the final RCM was that, although significant progress has been made in the development of tools to combat nuclear smuggling, further research in several different areas is still required. The

detection capability of existing systems is not the area where more development is needed. The development of the capability to discriminate between real illicit trafficking and the movement of NORM material is the area that requires the most improvement. This research should not only concentrate on better isotopic identification systems but should also include the development of tools for better information distribution and communication.

The IAEA has agreed to pursue those issues raised by the RCM and, to this end, has initiated a follow-up Coordinated Research Project in this area.

Feasibility of Using Hand-Held Isotope Identifiers for Attribute Testing of Radioactive Sources

in Shipment Containers

N. Kravchenko Russian Research Center Kurchatov Institute,

Moscow, Russian Federation

Abstract

The project “Feasibility of using hand-held isotope identifiers for attribute testing of radioactive sources in shipment containers” was being carried out in the framework of IAEA Research Contract # 12444\Nuclear Security Multi-donors Fund by Russian Research Center "Kurchatov Institute" from March 15, 2003, till May 31, 2006. The scope of work included:

1. Analysis of technical characteristics of hand-held isotope identifiers with respect to the determination of transport index and the verification of shielded sources.

2. Establishing the list of most frequently shipped radioactive sources and nuclear materials.

3. Analysis the design information of standard, approved shipping containers used for the transportation of nuclear materials and radioactive substances

4. Performing test measurements on various containers using hand-held devices. Assessment of performance of the devices with respect to the scope of the project and draw conclusions with respect to the technical feasibility of: ⎯ Confirmation of transport index of nuclear and radioactive materials

shipped in various shipping packaging containers; ⎯ Identification of nuclear and radioactive materials shipped in various

shipping packaging containers; ⎯ Definition of isotopic composition of gamma nuclides in their various

combinations

The results of the project execution are as follows:

1. The information on the most frequently shipped radioactive sources and nuclear materials and design of standard shipping containers for the transportation of nuclear materials and radioactive substances was systemized; 2. The spectrometric analysis of most frequently shipped isotopes was carried out;

3. The spectrometer capabilities were verified in terms of the definition of isotopic composition of gamma nuclides in their various combinations;

4. The spectrometer capabilities were verified in terms of the definition of isotopic composition of gamma nuclides when placed in various protective containers.

5. The capabilities of spectrometers (identifiers) on transport index confirmation were verified.

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SUMMARY

The project “Feasibility of using hand-held isotope identifiers for attribute testing of radioactive

sources in shipment containers” was being carried out in the framework of IAEA Research

Contract # 12444\Nuclear Security Multi-donors Fund by Russian Research Center "Kurchatov

Institute" from March 15, 2003, till May 31, 2006. The chief scientific investigator is Nikolay

E. Kravchenko. The scope of work included:

1. Analysis of technical characteristics of hand-held isotope identifiers with respect to the

determination of transport index and the verification of shielded sources.

2. Establishing the list of most frequently shipped radioactive sources and nuclear materials.

3. Analysis the design information of standard, approved shipping containers used for the

transportation of nuclear materials and radioactive substances

4. Performing test measurements on various containers using hand-held devices.

Assessment of performance of the devices with respect to the scope of the project and

draw conclusions with respect to the technical feasibility of:

⎯ Confirmation of transport index of nuclear and radioactive materials shipped in

various shipping packaging containers;

⎯ Identification of nuclear and radioactive materials shipped in various shipping

packaging containers;

⎯ Definition of isotopic composition of gamma nuclides in their various combinations

The results of the project execution are as follows:

1. The information on the most frequently shipped radioactive sources and nuclear materials

and design of standard shipping containers for the transportation of nuclear materials and

radioactive substances was systemized;

2. The spectrometric analysis of most frequently shipped isotopes was carried out;

3. The spectrometer capabilities were verified in terms of the definition of isotopic

composition of gamma nuclides in their various combinations;

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4. The spectrometer capabilities were verified in terms of the definition of isotopic

composition of gamma nuclides when placed in various protective containers.

5. The capabilities of spectrometers (identifiers) on transport index confirmation were

verified.

The results were reported on the «Research Co-ordination Meeting of the International Atomic

Energy Agency’s Coordinated Research Project on Improvement of Technical Measures to

Detect and Respond to Illicit Trafficking of Nuclear and radioactive Materials» held in Vienna in

2003 and 2006, and in Dagomys (Russia) in 2004, on the “Countering Nuclear Terrorism”

meeting in Yerevan (Armenia) in 2005, and on the “Spectrometric Analysis and Data

Processing” seminars in Obninsk, Russian, in 2004 and 2005.

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TABLE OF CONTENT

1. TECHNICAL CHARACTERISTICS OF HAND-HELD DEVICES FOR DETERMINATION OF TRANSPORT INDEX AND VERIFICATION OF SHIELDED SOURCES........................................................................................................6 1.1. Survey dosimeter DRS-PM1401 .........................................................................................6 1.2. Survey device ISP-PM1401K-1...........................................................................................7 1.3. Personal dosimeter DKG-PM1621 ......................................................................................9 1.4. Personal dosimeter DKG-PM1203 ......................................................................................9 1.5. X-ray and gamma dosimeter EL-1119...............................................................................10 1.6. Radiometer-spectrometer MKS-A02-01............................................................................11 1.7. Radiometer-spectrometer MKS-A03.................................................................................12 1.8. Radiometer-spectrometer RSU-01 “Signal”......................................................................13 1.9. Spectrometer GAMMA-1C/NB.........................................................................................14 1.10. Radiometer-spectrometer PM1401K ...............................................................................15 1.11. Spectrometer MKS – AT 6101 ........................................................................................16 1.12. Survey identifier (Smartphone) PM1802.........................................................................17 1.13. Semiconductor spectrometer SKS-50..............................................................................19 1.14. InSpector 1000.................................................................................................................20 1.15. Spectrometer IdentiFINDER – Ultra (Target) .................................................................22 1.16. Identifier – Exploranium GR-135 Radioactive Isotope Identification Device ................23

2. ESTABLISHING THE LIST OF MOST FREQUENTLY SHIPPED RADIOACTIVE SOURCES AND NUCLEAR MATERIALS...........................................................................25

3. SYSTEMATIZATION THE DESIGN INFORMATION OF STANDARD SHIPPING CONTAINERS FOR NUCLEAR AND RADIOACTIVE MATERIALS...............................28

4. TEST MEASUREMENT PROGRAM .....................................................................................32 4.1. General provisions .............................................................................................................32 4.2. Test conditions used for gamma-sources...........................................................................33 4.3. Test conditions used for neutron sources...........................................................................33 4.4. Requirements to the means of control ...............................................................................34 4.5. Requirements to radioactive and nuclear materials ...........................................................34

5. CONFIRMATION OF TRANSPORT INDEX OF NUCLEAR AND RADIOACTIVE MATERIALS SHIPPED IN VARIOUS SHIPPING PACKING CONTAINERS ..................35

6. IDENTIFICATION OF NUCLEAR AND RADIOACTIVE MATERIALS SHIPPED IN VARIOUS SHIPPED PACKING CONTAINERS ..................................................................41 6.1. Experimentations in 2003 at RRC “Kurchatov Institute” on the definition of isotopic

composition of gamma nuclides........................................................................................41 6.2. Experimentations in 2005 at RRC “Kurchatov Institute” on the definition of isotopic

composition of gamma nuclides........................................................................................43 6.2.1. Identification of nuclear and radioactive materials in various packaging containers..45 6.2.2. Identification of nuclear and radioactive materials in their various combinations......54 6.2.3. Detection of nuclear materials by neutron emission ...................................................55

6.3. Additional processing of spectra using software ...............................................................57 6.3. Summary on capabilities of the devices for identification of nuclear and radioactive

materials ............................................................................................................................81 6.3.1. Spectrometer InSpector 1000 ......................................................................................81 6.3.2. Spectrometer MKS – AT 6101....................................................................................81 6.3.3. Spectrometer MKS-A03..............................................................................................81 6.3.4. Spectrometer IdentiFINDER – Ultra (Target).............................................................82 6.3.5. Spectrometer Exploranium GR-135 ............................................................................83 6.3.6. Spectrometer MKS-PM1401K ....................................................................................83 6.3.7. Spectrometer Smartphone PM1802.............................................................................84

CONCLUSIONS ...........................................................................................................................84

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ABBREVIATIONS

ADC – analogue-to-digital converter;

FRM – fissionable and radioactive materials;

GS – gamma-sources;

IR – ionizing radiation;

SIR – source of ionizing radiation;

IAEA – International Atomic Energy Agency;

EDR – equivalent dose rate;

SW – software;

TAP – total absorption peak;

SCD – semiconductor detector;

PRC – primary radiation control;

FWHP – full width at half peak;

SPC – shipping packaging container;

FCS – Federal Customs Service;

NM – nuclear materials.

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1. TECHNICAL CHARACTERISTICS OF HAND-HELD DEVICES FOR DETERMINATION OF TRANSPORT INDEX AND VERIFICATION OF

SHIELDED SOURCES

1.1. Survey dosimeter DRS-PM1401 Survey microprocessor-based dosimeter DRS-PM1401 is intended for survey (detection and identification) of radioactive materials by recording gamma-rays emitted by them. The device is recommended for control of shipped nuclear and radioactive materials under conditions of sea and river ports, railway and motor transport, pedestrian and transport portals.

Fig. 1.1. Survey dosimeter DRS-PM1401.

Dosimeter measures the equivalent gamma-radiation dose rate. It is graduated in μSv/h units. The principal feature of the device is the detector (scintillator) type. The sensitivity of the detector is much higher for low energies (the sensitivity at the energy of 50–60 keV is higher by the order of magnitude then that for 1.5–3.0 MeV). It makes possible to detect nuclear materials (NM) most efficiently. The scintillation unit of PM 1401 consists of CsI(Tl) scintillator, photodiode, and transducer amplifier. The vibrating indicator is withdrawn off the casing. It is connected to the dosimeter through the cable, and it induces mechanic shocks inside the indicator at the exceeding the certain operation threshold in the search mode. It makes possible to detect gamma-sources secretly or operate in very noisy conditions. The sound indicator is intended to alarm at the exceeding the preset alarm level in the testing and search modes. The sound indicator is disabled when the vibration indicator is switched on. The information on the test results, mode of operation, and measuring results in the test and search modes is displayed on the LCD panel. Three basic modes of dosimeter operation are 1) the testing, 2) calibration against background, and 3) search modes. Two additional modes are used for setting the standard deviation (n coefficient) and for the power supply control. The transition between various modes is handled sequentially and automatically. The testing mode starts just after power-up and used for testing the basic units of the device, especially, LCD, sound indicator, and processor. After successful testing (it takes about 7 seconds) the dosimeter passes to the “calibration against background” mode. But before this mode, LCD displays the value of standard deviation (n coefficient) set at the previous operation.

Device specifications

Relative energy resolution by the 662 keV (Cs-137) gamma-line. ± 20% Range of recorded gamma-ray energy 0.06–3 MeV Measuring time 0.25 sec Operating temperature range From –30 to +50 °C Overall dimensions 22 x 82 x 120 mm Weight 400 g

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1.2. Survey device ISP-PM1401K-1 Survey dosimeter-radiometer MKS-PM1401K is designed for searching, detecting and localizing radioactive and nuclear materials by means of registering gamma- and x-rays (hereinafter referred to as the photon), neutron, alpha- and beta-rays, measuring the equivalent dose rate (EDR) of photon radiation, density of alpha- and beta-particles flow (control over the surface contamination), accumulating and storing gamma-spectra.

MKS-PM1401K is a combination of a detector unit and processing unit (moderator of neutrons) designed to be hand-held. The processing unit is constructively and electrically connected to the detector unit. The device is shock-resistant and water-resistant, as it is designed to be operated in severe conditions.

The dosimeter belongs to portable means of measurement of ionizing radiation and can be operated both in the laboratory and in the field environment.

The dosimeter is designed as a portable device, with the front panel equipped with an LCD panel, four-button keyboard, IrDA port, and a removable clip. On the dosimeter top end side there is a scintillating detector of the scintillating detector unit. On the dosimeter back side there is Geiger-Muller counter of the gamma-ray detection unit with removable filters used for measuring alpha-, beta- or gamma-radiation, as well as spacer rings used for measuring alpha-, beta- or gamma-radiation. Neutron radiation detector of the neutron detection unit is located inside the dosimeter. On the dosimeter bottom end side there are: a socket to connect vibrating alarm indicator, audio alarm indicator, and a replaceable cover of the battery compartment, where a battery is located.

Fig. 1.2. Survey radiometer-spectrometer DRS-PM1401K.


Energy dependency relative to the energy of 662 keV (Cs-137) in the photon radiation measurement mode within: - in the energy range between 0.015 and 0.045 MeV - in the energy range between 0.045 and 20.0 MeV

±40% ±30%

Range of recorded gamma-ray energy 15–2,000 keV Count time in the following modes: - search mode - calibration mode

2 sec 36 sec

Operating temperature range From –30 to +500 °C Weight 0.65 kg

Various types of ionizing radiation are measured using the built-in detector units.

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EDR of photon radiation and the density of alpha- and beta-particles flow are measured using the built-in all-purpose detector unit based on the Geiger-Muller counter.

Neutron radiation (count speed) is recorded using the built-in neutron radiation detector unit based on the slow neutron counter.

Photon radiation (count speed) is recorded in the search mode using the detector unit based on the CsI scintillator. The same detector unit is used to record scintillation spectra of photon radiation.

Each detector unit is operated by separate microprocessor controllers, whose information is transferred to the main microprocessor controller.

The operation modes are selected and the dosimeter is programmed using a four-button keyboard through the screen menu. Measurement results and dosimeter operation modes are displayed on the LCD panel.

If connected to a PC the operation modes are selected, the dosimeter is programmed, and the measurement results are transferred to the PC using an IrDA-compatible interface or a radio channel through Bluetooth.

Dosimeter-radiometer MKS-PM1401K has the following basic operation modes:

• Testing mode (testing);

• Calibration mode (calibration);

• Menu display mode (menu);

• Search for gamma-ray and neutron sources (γ, n search);

• Neutron registration mode (n registration);

• Mode of search for gamma- alpha- and beta-ray sources (γ, α, β search);

• Mode of measurement of EDR of photon radiation (γ measurement);

• Mode of measurement of density of alpha- and beta-particles flow (α, β measurement);

• Mode of registration of scintillation spectra of photon radiation (spectrum);

• Settings mode (settings);

• PC connection mode (IR exchange, Bluetooth).

The dosimeter monitors power voltage in any of the operation modes.

LCD panel backlight is available in any of the dosimeter operation modes.

Survey dosimeter-radiometer MKS-PM1401K has a pre-installed library with the following isotopes: for special nuclear materials (U-233, U-235, Np-237, Pu-239, Pu-241), for medical isotopes (Cr-51, Ga-67, Tc-99m, Pd-103, In-111, I-125, I-123, I-131, Tl-201, Xe-133), for industrial isotopes (Co-57, Co-60, Ba-133, Cs-137, Ir-192, Nl-204, Se-75, Ra-226, Am-241), for naturally occurring isotopes (K-40, Ra-226, Th-232 and daughters, U-238 and daughters. Complementary subgroup isotopes: (Bl-207, Cd-139, Eu-152, Ir-192, Mn-54, Na-22, Sn-113, Na-22, Th-228, Tl-44, Y-88, Zn-65).

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MKS-PM1401K has the Cs-137 calibration source located inside the detector casing, which makes manual calibration unnecessary each time before starting spectrometric measurements. Minimization of an operator’s actions and the device operation simplicity in the “express analysis” mode are the main advantages of the device.

1.3. Personal dosimeter DKG-PM1621 Personal dosimeter DKG-PM1621 is intended for

• continuous measurement of equivalent dose rate of external gamma- and X-radiation;

• continuous measurement of time for dose acquisition;

• continuous measurement of equivalent dose of external gamma- and X-radiation;

• transfer of obtained data stored in permanent memory to PC through the IR adapter.

The transition between various modes is handled sequentially and automatically. The testing mode starts just after power-up and used for testing the basic units of the device, especially, LCD, sound indicator, and processor

Fig. 1.3. Personal dosimeter DKG-PM1621.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. ±30% Range of recorded gamma-ray energy 0.01–20 MeV Actuation time 5 sec Operating temperature range From –20 to +60 °C Overall dimensions 87 x 72 x 60 mm Weight 150 g

1.4. Personal dosimeter DKG-PM1203 Personal dosimeter DKG-PM1203 is intended for

• measurement of equivalent dose rate (EDR) of external gamma-radiation;

• measurement of equivalent dose of external gamma-radiation;

• setting of alarm threshold of equivalent dose rate and equivalent dose. At crossing this threshold, the sound alarm is switched on.

• measurement of time for dose acquisition.

• tranfer of obtained data stored in permanent memory to PC through the IR adapter.

The built-in time indicator advances serviceability. The device is graduated in International units of equivalent dose (mSv) and equivalent dose rate (mSv/h). The accompanying sound and current time indication are provided. These devices are used for primary radiation monitoring and personal radiation control by custom officers.

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Relative energy resolution by the 662 keV (Cs-137) gamma-line. ± 25% Range of recorded gamma-ray energy 0.06–1.5 MeV Measuring time (automatically set) 1-36 sec Operating temperature range From –20 to +60 °C

The current equivalent dose rate is continuously displayed on LCD panel in the equivalent dose rate measurement mode. In the equivalent dose measurement mode, the LCD panel displays the value of equivalent dose obtained from time of power up, time of threshold setting or time of threshold crossing. At crossing the equivalent dose rate threshold, the device induces sound alarm and passes to the equivalent dose rate measurement mode. The device alarms until the equivalent dose rate will decrease below the threshold or the higher threshold will be set.

Fig. 1.4. Personal dosimeter DKG-PM1203.

1.5. X-ray and gamma dosimeter EL-1119 X-ray and gamma dosimeter EL-1119 is the multifunctional sensitive broad-band hand-held device with digital indication and microprocessor control. It is used for control of X-ray devices, inspection of luggage, goods, and for primary and additional radiation control of goods and carriers. The dosimeter is intended for measurements of:

• exposure dose rate;

• rate of dose absorbed in air;

• equivalent dose rate of gamma- and X-ray;

• exposure dose;

• dose absorbed in air;

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• equivalent dose of gamma- and X-ray.

Fig. 1.5. X-ray and gamma dosimeter EL-1119.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. ±35% Range of recorded gamma-ray energy 20–3000 keV Actuation time No more then 30 sec Operating temperature range From –30 to +40 °C Overall dimensions 220 x 80 x 60 mm Weight 800 g

1.6. Radiometer-spectrometer MKS-A02-01 Radiometer-spectrometer MKS-A02-01 consists of main MKS-A02 unit with built-in gamma- and neutron channels and remote alpha- and beta-detectors BDK-AB1. The device can be powered from 220 V or, optionally, from built-in accumulators to use it as hand- held device. The total weight of the radiometer is 3.6 kg. The device is operated with the keyboard on the front panel providing rather simple and informative menu. In the search mode, MKS-A02 uses both neutron and gamma-channels. MKS-A02 permits to measure exposure dose rate from neutron sources (Pu-Be only) as well as from gamma-ray sources.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. No more then 8% Range of recorded gamma-ray energy 503000 keV Actuation time: At the identification mode At all other modes

No more then 30 min No more then 2 min

Operating temperature range From –20 to +50 °C Overall dimensions 290 x 160 x 135 mm Weight 3.6 kg

The reference Th-232 gamma-source (included in the device kit) is used for the device calibration. The nuclide library in the memory of MKS-A02 can be refreshed and expended with data on gamma-rays emission of various nuclides. It makes possible to identify unknown nuclides in a short time if the data on these nuclides are in the library. The typical library for the custom control includes the following nuclides: Th232, Pu239, Ga67, I131, Cs137, Co60,

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Ra226, U233, U235, U238, Bi207, In111, Tl201, Pd103, Ba133, K40, Co57, Th99m, Am241, Y88. Optionally, the spectrum can be transferred to the external PC for analysis with the special software. In this case, MKS-A02 is operating as a full-scale spectrometer of Gamma 1C/NB type. The surface alpha- and beta-radioactivity are measured with the remote BDK-AB1 detector inserted with the cable to the special connector on the front panel of the device. The alpha- and beta-particles are recorded simultaneously, and the device does not need any special calibration. Radiometer-spectrometer MKS-A02 is well adjusted for measuring the isotopic composition of the inspected object. For small size sources (when the source activity is low enough that it can be unpacked), MKS-A02 can define the isotopic composition and assess the activity of each nuclide with use of special software (supplied with the device).

1.7. Radiometer-spectrometer MKS-A03 The radiometer consists of main MKS-A03 unit with built-in gamma- and neutron channels. The device can be powered from 220 V or, optionally, from built-in accumulators (providing for 16 hours of continuous operation) to use it as a quite convenient hand-held device. The radiometer is designed as a compact portable device with the front panel equipped with an LCD display, functional buttons, LED indicators, power adapter socket, and interface cable socket. The functional buttons designation depends upon the pre-installed software.


Relative energy resolution by the 662 keV (Cs-137) gamma-line No more than 8% Range of recorded gamma-ray energy 50–3,000 keV Actuation time: In the identification mode In all other modes

Not specified No more than 30 min No more than 2 min

Operating temperature range From –20 to +500 °C Weight 3 kg In the search mode, MKS-A03 uses both neutron and gamma-channels simultaneously. MKS-A03 permits to measure equivalent dose rate from neutron sources (Pu-Be sources only) as well as from gamma-ray sources.

The reference Th-232 gamma-source (included in the device kit) is used for the device calibration. One or several nuclide libraries can be added into the memory of MKS-A02 and expanded, as required, with data on gamma-rays emission of various nuclides. It makes possible to accurately identify nuclides in an unknown sample within a short time if the data on energy peaks of these nuclides are in the library.

The library to be used for the “express analysis” includes the following nuclides:

Ga-67, Tc-99m, Pd-103, In-111, I-123, I-125, I-131, Tl-201, Ag-110m, Bi-207, Cd-109, Cr-51, Na-22, Co-57, Co-57, Co-58, Co-60, Ba-133, Cs-134, Cs-137, Eu-152, Eu-155, Fe-59, K-40, Mn-54, Se-75, Ir-192, Ra-226, Th-232, Am-241, U-238, Zn-65, U-233, U-235, Pu-239, Np-237, Pu-240. Radiometer-spectrometer MKS-A03 is sufficiently well adjusted for measuring the isotopic composition of the inspected object. For small size sources (when the source activity is low enough that it can be unpacked), after measuring gamma-spectrum MKS-A03 can define the isotopic composition of the sample and assess the activity of each nuclide with the use of special software (supplied with the device).

The radiometer has the following operation modes to search for gamma-ray and neutron sources and measure equivalent dose rate: Search/EDR, Analysis and Monitor.

The radiometer enters the “Search/EDR” mode after being switched on. In this mode, it continuously measures and displays an equivalent dose rate of gamma-ray and neutron sources.

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Equivalent dose rates are presented both graphically and digitally. The measurement unit is µSv/h. The search is enabled by histograms in the bottom part of the display showing the nature of changes in the count speed of gamma-ray and neutron detectors. Count speed-dependent audio and visual alarm is available to simplify search procedures.

Fig. 1.6. Radiometer-spectrometer MKS-A03.

In the “Analysis” mode the radiometer measures gamma-spectrum, searches for peaks in the spectrum and compares the positions of the identified peaks with the library data. In case of an exact match, the radionuclide name and type (industrial, natural or medical) is displayed.

In the “Monitor” mode the radiometer continuously measures count speeds (in p/sec) for gamma-ray and neutron detectors, compares them with threshold values and switches on visual and audio alarm at crossing them. Threshold values are calculated during the calibration against background.

In the search mode, the device constantly makes 1-second-long measurements and calculates count speeds for gamma-ray and neutron channels. The first 20 measurements are averaged to assess the count speed for the background and calculate the detection threshold (3 sigmas above the background), whereas the next measurements are compared against the threshold. Audio or visual alarm is switched on at crossing the count speed threshold.

If the frequency of the indicator response increases to a level making any further frequency increase undetectable, one should pass to the next measurement range by pressing “B” and “C” buttons for gamma- and neutron radiation, respectively.

The radiometer continuously measures an equivalent dose rate. The measured values are displayed in the “Search/EDR” mode.

1.8. Radiometer-spectrometer RSU-01 “Signal” The device is intended for detection and identification of fissile and radioactive materials at the custom inspection or for use in the «Customs mobile posts of radiating control (CMPRC)”», and also for the personal radiating safety control. The program incorporated in RSU-01 provides the following operations:

• Search of sources of ionizing radiation by external gamma- or neutron radiation.

• Definition of the source activity directly in-situ with the gamma-detection block.

• Definition of specific activity of inspected object.

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• Definition of equivalent doze rate of gamma- and neutron radiation.

• Definition of density of alpha, beta particles and neutron fluxes.

In the "express analysis" mode the device allows to identify and determine activity of the following nuclides: Сs-137, Тh-232, Ra-226, K-40. The measured spectrum can be written down as a file and then analyzed with the special software at the external computer. In this case the radiometer is operating as a full-scale spectrometer of Gamma 1C/NB type. With the software "LsrmCustoms" the following problems can be also solved:

• measurement of open source activity,

• measurement of activity of the source inside the protective transport container,

• the expanded search of nuclides and their identification.

Fig.1.7. Radiometer-spectrometer RSU-01 “Signal”.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. No more then 35% Range of recorded gamma-ray energy 200–3000 keV Actuation time: No more then 15 min Operating temperature range From –20 to +40 °C Overall dimensions 180 x 140 x 75 mm Weight 1.5 kg

1.9. Spectrometer GAMMA-1C/NB Spectrometer GAMMA-1C/NB is the portable hand-held gamma-spectrometer in dust-and-water proof embodiment. It is designed to work with a laptop computers. The spectrometer is based on the principle of transformation of gamma-ray energy to the electric pulses of proportional amplitude. The electric pulses are recorded by the multichannel peak analyzer and processed on a laptop computer. Spectrometer GAMMA-1C/NB is intended for:

- search of sources of gamma-radiation;

- measurements of equivalent doze rate of gamma-radiation;

- measurements of effective specific activity of natural nuclides;

- measurements of gamma-spectrum;

- the analysis of the measured spectrum;

- storages of the measured data in non-volatile memory;

- data exchange with a computer.

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The device can be used for inspection of legally transported fissile materials with known parameters and isotope structure, and illegally trafficking materials with unknown isotope structure. Opening of transport and security packages is not required. The spectrometer GAMMA-1C/NB provides the mode of searching the radioactive sources and measurements of exposure dose rate of gamma-rays. It allows using the spectrometer as the survey device for the additional radiation control and the profound radiation inspection. Besides, it can be applied for the express analysis of radiating cargo. For that task, the nuclide library in the memory of MKS-A02 can be refreshed and expended with data on gamma-rays emission of various nuclides. It makes possible to identify unknown nuclides in a short time if the data on these nuclides are in the library. The library for the custom control includes the following nuclides: Th232, Pu239, Ga67, I131, Cs137, Co60, Ra226, U233, U235, U238, Bi207, In111, Tl201, Pd103, Ba133, K40, Co57, Th99m, Am241, Y88. Optionally, the spectrum can be transferred to the external PC for analysis with the use of special software.

Fig. 1.8. Spectrometer GAMMA-1C/NB.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. 8% Range of recorded gamma-ray energy 50–3000 keV Actuation time: No more then 30 min Operating temperature range From –20 to +50 °C Weight 20 kg

The spectrometer GAMMA-1C/NB is supplied with the laptop computer with the specialized software for determination of isotopic composition and nuclide activity inside the special shipping packing container. Moreover, the software allows controlling the enrichment of uranium and uranium compounds without opening the transport containers and packages.

1.10. Radiometer-spectrometer PM1401K Radiometer-spectrometer consists of main PM1401K unit with built-in gamma- , alpha- , beta- and neutron channels and minicomputer PAL. The device can be powered from 220 V or from built-in accumulators to use it as hand-held device. The total weight of the radiometer is 0.65 kg. The device is operated with the keyboard on the front panel providing rather simple and informative menu. In the search mode, PM1401K can employ neutron and gamma-channels simultaneously. PM1401K allows to measure exposure dose rate of gamma-rays, flux density of alpha- and beta-particles. The built-in Cs-137 source of gamma radiation incorporated in the device casing is used for calibration.

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PM1401K can be used for nuclide identification. For express analysis, the memory of device contains more than 40 spectrums of gamma-emission of various nuclides. In the “expert mode” the measured spectrum can be written down as a file and then analyzed with the special software installed on the external computer. In this case the radiometer is operating as a full-scale spectrometer.


Relative energy resolution by the 662 keV (Cs-137) gamma-line No more then 7% Range of recorded gamma-ray energy 15–3000 keV Actuation time: At the identification mode At all other modes

No more then 30 min No more then 2 min

Operating temperature range From –20 to +50 °C Overall dimensions 290 x 160 x 135 mm Weight 0.65 kg

The measurements of surface alpha- and beta-radioactivity need not any additional calibration. Radiometer-spectrometer PM1401K is well adjusted for measuring the isotopic composition of the inspected object. It can assess the activity of identified nuclide at the known thickness of the transport package shielding.

1.11. Spectrometer MKS — AT 6101 Spectrometer MKS — AT 6101 is a portable gamma-radiation spectrometer. The spectrometer is based on the principle of transformation of gamma-ray energy falling within the sensitivity of the scintillating detector into the electric pulses of proportional amplitude, which are then recorded by the peak analyzer and processed by a computer.

Spectrometer MKS — AT 6101 is intended for:

• Search of gamma-sources;

• Radionuclide identification;

• Measurements of gamma-ray energy distribution;

• Equivalent gamma-radiation dose rate.

The MKS-AT6101 detector unit consists of NaI(Tl) scintillator. The detector is withdrawn off the casing and is connected to the spectrometer through the cable. Four basic modes of the spectrometer operation are: calibration against background, search, identification of radioactive isotopes, and spectrum. The transition between various modes is handled manually. The K-40 gamma-source (included in the device kit) is used for the device calibration. The spectrometric information is displayed on the LCD panel with a backlight function and resolution of 128х64. There is an audio and visual alarm available, which is switched on at crossing the dose rate threshold during the search for gamma radionuclides and their identification. Spectrometric measurement of the dose rate is available using the “Spectrum-dose” transformation operator. Continuous automatic stabilization of the spectrometer energy scale is available using light-emitting diodes. Digital compensation of the spectrometric channel is available using the incorporated temperature sensor.

• The memory allows recording and storing up to 300 spectra.

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Fig. 1.9. Radiometer-spectrometer AT 6101.


Energy resolution by the 662 keV (Cs-137) gamma-line No more than 9% Range of recorded gamma-ray energy 20–1,500 keV Actuation time Not specified Operating temperature range From –20 to +500 °C Weight Less than 1.5 kg

MKS-AT6101 has a pre-installed library with the following isotopes: for special nuclear materials (U-233, U-235, Np-237, Pu-239, Pu-241), for medical isotopes (Cr-51, Ga-67, Tc-99m, Pd-103, In-111, I-125, I-123, I-131, Tl-201, Xe-133), for industrial isotopes (Co-57, Co-60, Ba-133, Cs-137, Ir-192, Nl-204, Se-75, Ra-226, Am-241), for naturally occurring isotopes (K-40, Ra-226, Th-232 and daughters, U-238 and daughters. Complementary subgroup isotopes: (Bl-207, Cd-139, Eu-152, Ir-192, Mn-54, Na-22, Sn-113, Na-22, Th-228, Tl-44, Y-88, Zn-65).

1.12. Survey identifier (Smartphone) PM1802 Two-channel survey device for the detection and identification of fissionable and radioactive materials PM1802, designed as a mobile phone (Smartphone) with identification and remote communication functions.

The device features include:

• Serviceable by one specialist.

• Simultaneous gamma- and neutron radiation detection.

• “Timer/counter” mode for neutron and gamma-channels aimed at improving detection efficiency.

• Durable, designed to be used outdoors in a broad range of temperature and humidity conditions.

• Display backlight providing for operation in the darkness; also readable in bright sunlight.

• Battery charge is sufficient for at least 8 hours of continuous operation.

• Automatic (convenient) and manual (for specialists) operation modes, their switcher having special protection.

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Fig. 1.10. Survey identifier ISP-PM1802


Energy resolution by the 662 keV (Cs-137) gamma-line 7.5% Range of recorded gamma-ray energy (based on a special order) 25–3б000 keV

Actuation time – Operating temperature range From –20 to +500 °C Weight 0.24 kg Spectrum analyzer features of the monitor:

• At least 1,000 analog-to-digital converter (ADC) channels.

• Stabilization of dependency from the energy (e.g. that of temperature sensor, LED or radioactive source, and recalibration mode in the presence of naturally occurring radionuclides).

• Linearization of dependency from the energy for scintillating detectors.

• Energy calibration adjustment using a radioactive source, preferably a naturally occurring isotope.

• Memory for at least 50 spectra with 1,000 channels in each.

• Standard industrial PC connection, standard spectral data format.

• User-friendly PC-support software designed for specialists.

• Internal memory and PC connection.

• PC connection using standard interface with communications devices (e.g. RS232, USB, IR or WIFI, the port is protected from water drops) or adaptation with a PC-based device. Data transferability to another device, in particular using the mobile phone radio channel.

• It is desirable to store at least 100 measured values of alarm signal (e.g. date and time, signal value, etc.).

• Standardized information stored in a file, including all the relevant settings and diagnostic data, as well as measurement results.

Smartphone PM1802 has a pre-installed library with the following isotopes: for special nuclear materials (U-233, U-235, Np-237, Pu-239, Pu-241), for medical isotopes (Cr-51, Ga-67, Tc-99m, Pd-103, In-111, I-125, I-123, I-131, Tl-201, Xe-133), for industrial isotopes (Co-57, Co-60, Ba-133, Cs-137, Ir-192, Nl-204, Se-75, Ra-226, Am-241) and for naturally occurring isotopes (K-40, Ra-226, Th-232 and daughters, U-238 and daughters. Complementary subgroup isotopes: Bl-207, Cd-139, Eu-152, Ir-192, Mn-54, Na-22, Sn-113, Na-22, Th-228, Tl-44, Y-88, Zn-65).

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1.13. Semiconductor spectrometer SKS-50 Spectrometer SKS-50 is the most modern and powerful spectrometer available to custom control services. It is used for:

- identification and control of activity of radioisotope sources crossing the customs border of Russian Federations in protective containers of shipping packing container (SPC) type and without them;

- control of enrichment degree of uranium compounds in protective containers of SPC type and without them;

- definition of nuclide composition of plutonium samples in protective containers and without them.

The software of the spectrometer provides automatic and manual modes of processing and analysis of the information in-situ.

Fig. 1.11. Spectrometer SKS-50.


Relative energy resolution by the 662 keV (Cs-137) gamma-line. 8% Range of recorded gamma-ray energy 50–3000 keV Actuation time: No more then 30 min Operating temperature range From –20 to +50 °C Weight No more than 20 kg

The СКС-50 system allows to define required parameters of radioactive and fissile materials inside SPC with the following accuracy:

• the error of definition of activity of radioactive materials in the range from 3 to 30% (depends on the nuclide and container types (the error of definition of activity depends on the decay scheme, completeness of the description of a controllable source, contaminations in the basic controllable material, completeness of data on time of manufacturing of controllable material, position of the source in the container, completeness of the description of parameters of the container, degree of conformity of the real container to its description, degree of external and internal pollution of the container with unwanted radioactive substances, etc.),

• the error of definition of enrichment of uranium from the uranium-235 isotope lays in the range from 0,5 up to 5% for well determined samples and items, for other — 10–15% (depends on completeness of data on physical parameters of measured material, the sizes of measured object, completeness of the description of parameters of the container, degree of conformity of the real container to its description, degree of external and internal pollution of the container with unwanted radioactive substances, etc.),

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• the error of definition of isotopic composition of plutonium from the plutonium-239 isotope lays in the range from 2 up to 5% (depends on completeness of data on physical parameters of measured material, the sizes of measured object, completeness of the description of parameters of the container, degree of conformity of the real container to its description, etc.)

The СКС-50 system is a gamma-ray spectrometer with high metrological and operational parameters. The spectrometer is equipped with the software package for qualitative and quantitative processing of complex gamma-spectrum. The usage of germanium detector provides high selectivity of recording individual gamma — lines from which the nuclides are identified. The nuclide activity, enrichment of uranium and isotope composition of plutonium are determined through the quantitative processing of gamma-spectrum.

1.14. InSpector 1000 InSpector 1000 is a digital multi-channel analyzer, which can be used in field operations for:

• Search of radioactive sources;

• Measurements of equivalent dose rate (EDR) of gamma-radiation;

• Nuclide identification and measurements of effective specific activity level;

• Measurements of gamma-spectra;

• Analysis of the measured spectra;

• Storage of the measured data in non-volatile memory;

• Data exchange with a computer.

Fig. 1.12. Radiometer-spectrometer (identifier) — CanberraInSpector 1000.

The analyzer measures equivalent dose rate of gamma-radiation and is graduated in the following user-selectable units: µSv/h, mSv/h, µR/h, mR/h, R/h, µrem, and mrem.

The detector unit consists of NaI scintillator, high voltage source, and transducer amplifier.

Four basic modes of the device operation are: dose, search, nuclide identification, and spectrum. There is also an additional mode: adjustment.

The “Dose” mode displays both the instant dose rate and the accumulated dose in one of the several display modes in such units as sieverts, roentgens or rems.

If any other operation mode is on, all the functions of the “Dose” mode are still performed in the background mode, such as: continuous calculation of the dose and dose rate, monitoring of the alarm and warning thresholds, and activation of the alarm and warning signals at crossing this threshold.

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The dose rate display mode is operator-selectable and includes a simple, expanded, energy, logarithmic, and linear form.

The “Search” mode displays a time-dependent histogram of the radiation rate registered by the device allowing for the localization of radioactive sources.

The operator can also select a linear diagram of count speed or dose rate in this mode. At crossing the alarm threshold for the dose rate or the accumulated dose, the color of the linear scale changes and an audio alarm is signaled.

The “Nuclides” mode provides for real-time identification of radioactive isotopes and calculation of their activity level in Bq or µCi. The results of the isotope identification are summarized in a table, while the selection of an advanced report form displays a linear indicator of the dose rate as well. The data contained in the advanced report can be sorted by atomic numbers in an ascending order, by data in a descending order, or by the magnitude of error in an ascending order.

The “Spectrum” mode allows for acquiring and processing radionuclide spectra.

If the first column is selected records are sorted by atomic numbers.

The analyzer supports two types of spectra: directly acquired data (“live spectrum”) and data loaded from a file (previously acquired and saved spectrum).

If two or more nuclides generate peaks with approximately the same energy, InSpector is unable to determine, which of the nuclides has caused the peak. In this case the “?” sign is displayed in front of the nuclide name.

The “Adjustment” mode allows for establishing the general system parameters and customized parameters for each of the four measurement modes.

The device software envisages automated and manual processing and analysis of the information directly acquired at the place of measurement.

Device specifications Energy resolution by the 662 keV (Cs-137) gamma-line Range of recorded gamma-ray energy 50–3,000 keV Actuation time Not specified Operating temperature range From –10 to +550 °C Weight Less than 2.2 kg

InSpector 1000 has the following pre-installed libraries: Nal-NORM.nlb for natural radionuclides (K-40, Ra-226, Th-232, U-235, U-238), Nal-SNM.nlb for special nuclear materials (U-233, U-235, U-238, Pu-239), Naldemo.nlb for a typical mixed gamma-source (Co-57, Co-60, Sr-85, Y-88, Cd-109, Sn-113, Cs-137, Ce-139, Hg-203), Nal-MED.nlb for medical nuclides (Ga-67, Tc-99m, Pd-103, In-111, I-123, I-125, I-131, Xe-133, Ir-192, Tl-201), Nal-PeakLocate.nlb limiting the search of peaks for specified nuclides (Co-60, Cs-137) Nal-INDU.nlb for industrial nuclides (Na-22, Co-57, Co-60, Ba-133, Cs-137, Eu-152, Ir-192, Ra-226, Th-232, Am-241), Nal-ANSI.nlb complying with the ANSI 42.34 standard (Ra, Th, K-40, Co-57, Co-60, Ga-67, Tc-99m, In-111, I-123, I-125, I-131, Ba-133, Xe-133, Cs-137, Tl-201, U-233, U-235, Np-237, U-238, Pu-239, Am-241, Pu-241).

The advantages of the device include a multi-color sensor display and an ergonomic six-button keyboard, which allows easily selecting any of the operation modes and switching between the modes by pressing one button.

22

Its disadvantages include the need to constantly re-calibrate the device during its operation to obtain correct measurements.

1.15. Spectrometer IdentiFINDER — Ultra (Target) The device consists of the main unit with built-in gamma- and neutron channels and is a digital multi-channel analyzer, which can be used in field operations for:








In the search mode, the device uses both neutron and gamma-channels simultaneously.

The device can be powered from 220 V or, optionally, from built-in accumulators (providing for 10 hours of continuous operation) to use it as a quite convenient hand-held device. The radiometer is designed as a compact portable device with the front panel equipped with an LCD display, 4 sensor buttons, LED indicators, as well as a power adapter socket and interface cable socket, which are located at its side.

Measurement ranges of equivalent dose rate are from 10 nSv/h to 1 Sv/h and from 1 µrem/h to 100 rem/h. Dose measurement ranges are from 10 nSv/h to 1 Sv/h and from 1 µrem/h to 100 rem/h.

Fig. 1.13. Spectrometer (identifier) — IdentiFINDER with NaI detector.

The detector unit consists of NaI (Tl) scintillator, photoelectronic multiplier, spectrometric amplifier, and power unit. The spectrometer has the following basic operation modes to search for gamma-ray and neutron sources and measure equivalent dose rate: test, search, identification of radioactive isotopes, and spectrum analysis. The transition between various modes is handled manually. The spectrometric information is displayed on the LCD panel with a backlight function and resolution of 61х43. There is an audio and visual alarm available, which is switched on at crossing the dose rate threshold during the search for gamma radionuclides and their identification.

23

Temperature stabilization function is incorporated into the device. The memory allows recording and storing up to 100 spectra.

Device specifications:

Energy resolution by the 662 keV (Cs-137) gamma-line Range of recorded gamma-ray energy 20–3,000 keV Actuation time Operating temperature range From –15 to +550 °C Weight Less than 1.25 kg

IdentiFINDER (Target) has the following pre-installed libraries: for local inspectors (Ba-140, Cd-115, Ce-141, Ce-144, I-132, La-140, Mo-190, Nb-95, Nd-147, Pr-144, Nuc Pu, Ru-103, Sb-125, Te-132, Xe-131m, Med Xe-131, Xe-133m, Xe135, Zr-95, Med I-131, Tc-99m, Ga-67), for special nuclear materials (U-233, U-235, Pu-239), customs library (Med Xe-133, Med I-131, Med Tc99m, Med Ga-67, Med In-111, Med Pd-103, Med Tl-201, Am-241, Ba-133, Bi-207, Co-57, Co-60, Cs-134, Cs-137, Eu-152, K-40, Mn-54, Na-22, Ra-226, Th-232, U-238, Pu-239, U-233, U-235, Np-237), for medical nuclides (Ga-67, Tc-99m, Pd-103, In-111, I-123, I-125, I-131, Tl-201), for industrial nuclides (Ag-110m, Bi-207, Cd-109, Cr-51, Na-22, Co-57, Co-57, Co-58, Co-60, Ba-133, Cs-134, Cs-137, Cs-137, Eu-152, Eu-155, Fe-59, K-40, Mn-54, Se-75, Ir-192, Ra-226, Th-232, Am-241, U-238, Zn-65), for security (Nuc Pu, Med Xe-133, Med I-131, Med Tc-99m, Ga-67, Med I-123, Med I-125, Med In-111, Med Tl-201, Am-241, Ba-133, Bi-207, Co-57, Co-60, Cs-134, Cs-137, Eu-152, Ir-192, K-40, Mn-54, Na-22, Ra-226, Se-75, Th-232, Nuc U-238, Nuc U-233, Nuc U-235, Nuc Pu, DU-238, Nuc Np-237, Nuc Pu-240).

1.16. Identifier — Exploranium GR-135 Radioactive Isotope Identification Device This is a radioactive isotope identification device which can be used in field operations for:








24

Fig. 1.14. Identifier — Exploranium GR-135 Radioactive Isotope Identification Device.

The device can be powered from 220 V or, optionally, from built-in accumulators (providing for 8 hours of continuous operation) to use it as a quite convenient hand-held device. The radiometer is designed as a compact portable device with the front panel equipped with an LCD display, one-button joystick switch, with a power adapter socket, and interface cable socket located at its side, and LED indicators located on the docking station. Measurement ranges of equivalent dose rate are from 1 µR/h to 1 R/h or from 10 nSv/h to 10 mSv/h. Dose measurement ranges are from 1 µR/h to 1 R/h or from 10 nSv/h to 10 mSv/h. The detector unit consists of NaI scintillator and Geiger-Muller counter. A neutron detector (solid state lithium glass) and cadmium zinc telluride (CZT) detector can be attached. The spectrometer has the following basic operation modes: search, identification of radioactive isotopes, stabilization, and spectrum analysis. The transition between various modes is handled manually. The spectrometric information is displayed on the LCD panel with a backlight function and resolution of 65х65. There is an audio and visual alarm available, which is switched on at crossing the dose rate threshold during the search for gamma radionuclides and their identification. Temperature stabilization function is incorporated into the device.


Energy resolution by the 662 keV (Cs-137) gamma-line No more than 9% Range of recorded gamma-ray energy 50–3,000 keV Actuation time – Operating temperature range From –10 to +500 °C Weight 2.4 kg The reference Cs-137 gamma-source (included in the device kit) is used for the calibration.

Exploranium GR-135 has a pre-installed 2-level library of nuclides: for special nuclear materials (U-233, U-235, Pu-239) and others (Bi-207, Cf-252, Cd-109, Ce-139, Cr-51, DEP-U, Eu-152, Eu-154, F-18, Fe-59, Ir-192, Kr-85, Mn-54, Mo-99, Na-22, Na-24, Pd-103, Se-75, Sm-153, Ra-226, Th-232, K-40, Co-57, Co-58,Co-60, Ga-67, Tc-99m, In-111, I-123, I-125, I-129, I-131, Ba-133, Xe-133, Sr-85, Y-88, Y-90m, Cd-109, Sn-113, Ce-139, Cs-137, Tl-201, U-233, U-235, Np-237, U-238, Pu-239, Am-241, Pu-241).

25

2. ESTABLISHING THE LIST OF MOST FREQUENTLY SHIPPED RADIOACTIVE SOURCES AND NUCLEAR MATERIALS

Practice of custom registration and custom control of fissile and radioactive materials for last years shows that more than 50 types of fissile and radioactive materials cross the customs border of Russian Federation annually. The list of these isotopes is shown in theTable 2.1. Table 2.1. List of fissile and radioactive materials crossing the customs border of Russian

Federation.

n/n Nuclide designation Nuclide Symbol Half-life Exempt activity in SPC of

“A” type (Ci) 1 Americium − 241 241 Am 433 years 0.008 2 Americium − 243 243 Am 7.38×103 years 0.008 3 Actinium − 225 225Ac 10 days 0.003 4 Barium − 133 133 Ba 10.7 years 10 5 Bismuth − 207 207 Bi 38.0 years 10 6 Carbon − 14 14 C 5730 years 100 7 Calcium − 45 45 Ca 163 days 40 8 Cadmium − 109 109 Cd 453 days 70 9 Cerium − 139 139 Ce 140 days 100 10 Cerium − 144 144 Ce 284 days 7 11 Californium − 252 252 Cf 2.64 years 0.002 12 Curium − 244 244 Cm 18 years 0.01 13 Cobalt − 56 56 Co 77 days 5 14 Cobalt − 57 57 Co 270 days 90 15 Cobalt − 58 58 Co 71 days 20 16 Cobalt − 60 60 Co 5.25 years 7 17 Chromium − 51 51 Cr 27.8 days 600 18 Cesium − 137 137 Cs 30 years 9 19 Iron − 55 55 Fe 2.72 years 1000 20 Iron − 59 59 Fe 45 days 10 21 Gallium − 67 67 Ga 3.26 days 7 22 Gallium − 68 68 Ga 288 days 7 23 Gadolinium − 153 153 Gd 242 days 100 24 Germanium − 68 68 Ge 288 days 1000 25 Tritium 3 H 12.34 years 1000 26 Iodine − 124 124 I 4.18 days 100 27 Iodine − 125 125 I 60 days 70 28 Iodine − 129 129 I 1.57×107 years 2 29 Indium − 113 113 In 130 days 60 30 Iridium − 192 192 Ir 74 days 20 31 Potassium − 40 40 K 1.28×109 years 10

32 Krypton − 85 85 Kr

27

The List of Shipped Isotopes

100

0,42 0,84 5,

190,

42 0,84

8,84 9,40

25,1

10,

28 0,84

10,3

81,

96 2,66

1,12

10,9

48,

705,

892,

951,

68 6,31

5,05

2,95

15,2

90,

560,

000,

426,

170,

42 0,84

0

10

20

30

40

50

60

70

80

90

100

Уран

Плутоний

H-3

C-1

4

Na-

22

Mn-

54

Co-

60

Sr-8

9

Co-

57

Tc-9

9m

Ru-

106

Cs-

137

Ba-

133

Cm

-244

Th-2

32

Am

-241

P-33

I-12

5

Cd-

109

Fe-5

5

Gd-

153

Ge-

68

I-12

4

Ir-1

92

Kr-

85

Ni-6

3

Np-

237

Se-7

5

Sr-8

5

Y-8

8

Ratio, %

Fig. 2.1. Relative quantity of isotopes crossing the border.

Table 2.2. List of most frequently shipped isotopes.

Natural uranium Enriched uranium Germanium-68 Carbon-14 Cobalt-57 Iodine-125 Cesium-137 Krypton -85 Europium-152 Radium-226 Tritium-3

Cobalt-60 Palladium-103 Iridium-192 Phosphorus-32 Selenium-75 Gadolinium-153 Yttrium-90 Barium-133 Curium-244 Plutonium-239 Prometium-147

Listed in Table 2.2 nuclides (shipped in transport packages) can be ranked by activity (in Ci) on the following groups:

- nuclides with the activity of millions of Ci (Cobalt-60, Cesium-137);

- nuclides with the activity of hundreds of kCi (Iridium-192, Tritium-3);

- nuclides with the activity of dozens of kCi (Selenium-75, Prometium-147);

- nuclides with the activity of thousands of Ci (Krypton-85, Phosphorus-32, Yttrium-90);

- nuclides with the activity of hundreds of Ci (Carbon-14, Iodine-125, Palladium-103).

The activity of other nuclides (except those given in Table 2.2) are less then 100 Ci. This group has the following features:

28

- Cadmium-109, Cobalt-57, Cobalt-58, Iron-55, Galium-68, Germanium-68, Phosphorus-33, Ruthenium-106, Strontium-89, barium-133 are shipped regularly;

- Bismuth-207, Calcium-45, Cerium-139, Cerium-144, Californium-252, Curium-244, Galium-67, Nickel-63, Phosphorus-32, Radium-226, Strontium-85, Tangsten-188, Ittrium-88, Ittrium-90 are shipped periodically;

- Indium-113, Molybdenium-99, Sodium-22, Neptunium-237, Samarium-151, Technetium-99m are shipped occasionally.

Sources of radiation are widely used in industry, agriculture, medicine; they production is commercialized by manufactures of Russian Federation. Their customers include well-known firms from Great Britain, France, USA, Germany, etc. Thus the problems of custom inspection are rather urgent. For example, the alpha-sources are used in automatic systems for fire-fighting and smoke detectors, neutralizers of static electricity, gas chromatography, and gas analyzers. The beta-sources are applied in anti-icing systems for planes and helicopters, thickness meter, densimeters, radiation installations. Gamma-sources are used in medical radiation installations, industrial defectoscopy, control instrumentation, aviation and space systems. Neutron sources are employed for express-analysis of rocks and ores, humidity control, oil well logging.

3. SYSTEMATIZATION THE DESIGN INFORMATION OF STANDARD SHIPPING CONTAINERS FOR NUCLEAR AND RADIOACTIVE MATERIALS

The standard shipping containers approved for transportation of nuclear materials and radioactive substances were analyzed. The results are summarized in Tables 3.1–3.3.

Table 3.1. Stainless steal standard containers.

n/n Type of shipping packaging container

Container design specifications Transported material

1. Model 48Y

Stainless steel, 21 mm wall thickness Uranium hexafluoride with enrichment by U-235 up to 1,0%

2. Model 30B

Stainless steel, 16 mm wall thickness Uranium hexafluoride with enrichment by U-235 up to 5%

3. Mound 1 KW

Protective container is made of stainless steel, 38 mm wall thickness

Plutonium 238, 239

4. TK-S 15/1 Stainless steel, 10 mm wall thickness Fuel assemblies

26

33 Manganese − 54 54 Mn 312 days 20

34 Molybdenium − 99 99 Mo 67 days 100

35 Natrium − 22 22 Na 2.4 years 8

36 Nickelь − 63 63 Ni 100.1 years 100

37 Neptunium − 237 237 Np 2.14×106 years 0.005

38 Phosphorus − 32 32 P 14.,3 days 30 39 Phosphorus − 33 33 P 25.4 days 100

40 Palladium − 103 103 Pd 17 days 700 41 Prometium − 147 147 Pm 2.6 years 80 42 Radium − 226 226 Ra 1600 years 0.05

43 Ruthenium − 106 106 Ru 1 years 7

44 Selenium − 75 75 Se 118.45 days 40 45 Samarium − 151 151 Sm 90 years 90

46 Tin − 113 113 Sn 115 days 60

47 Tin − 119 m 119m Sn 293 days 100

48 Strontium − 82 82 Sr 25 days 80

49 Strontium − 85 85 Sr 64.8 days 30

50 Strontium − 90 90 Sr 29 years 0.4 51 Zinc − 65 65 Zn 244 days 30

52 Technetium − 99m 99m Tc 2.13×105 years 100

53 Tungsten − 188 188 W 69.4 years 40

54 Yttrium − 88 88 Y 107 days 30

55 Yttrium − 90 90 Y 2.67 days 10

56 Uranium (hexafluoride or other compounds)

235 U

7.04×108 years

0.2

57 Plutonium − 239 239 Pu 2.41×104 years 0.002

Uranium hexafluoride, uranium oxides and other uranium compounds are the most frequently shipped nuclear materials. Americium-241 is also shipped in large amounts; for example, more then 20590 Ci of this isotope crossed the border in 2002.

Some nuclear materials are shipped in a very low amount such as Californium-252 (only 1.52612 Ci/year), Neptunium-237 (0.007 Ci/year), and Plutonium-239.

The materials most frequently crossing the border of Russian Federation are printed in bold in Table 1 and summarized in Table 2.2 that displays the list of isotopes that are shipped most frequently (90%). The data are also displayed in Fig. 14.

29

Table 3.2. Standard containers with lead shielding.

n/n Type of shipping packaging container

Container design specifications

Transported material Most frequently shipped nuclide

1. UKT1B-250M

Lead protective container, 250 mm wall thickness

Cobalt-60 Phosphorus-32

Со-60

2. UKT1A-5 Lead protective container, 5 mm wall thickness

Al-26; P-32, Cr-51, Co-57, Fe-55, Ti-44 Cd-109, Mo-93, Pd-103, In-111, W-181, Au-195, Bi-207, As-73

Со-57, Cd-109, Pd-103, P-32

3. UKT1A-10 Lead protective container,10 mm wall thickness

Al-26; P-32, Cr-51, Co-57, Fe-55, Ti-44 Cd-109, Mo-93, Pd-103, In-111, W-181, Au-195, Bi-207, Be-7, Na-22, Ga-67,Ge-68, As-73,74, Y-88, Sr-85, Tc-95m, In-111, Sn-119, I-124,123, Ba-133, Re-183,184,Tl-201, Se-85

Se-85, Со-57, Cd-109, Pd-103


Al-26; P-32, Cr-51, Co-57, Fe-55, Ti-44 Cd-109, Mo-93, Pd-103, In-111, W-181, Au-195, Bi-207, Be-7, Na-22, Ga-67,Ge-68, As-73,74, Y-88, Sr-85, Tc-95m, In-111, Sn-119, I-124,123, Ba-133, Re-183,184,Tl-201

Se-85, Со-57, Cd-109, Pd-103


Al-26; P-32, Cr-51, Co-57, Fe-55, Ti-44 Cd-109, Mo-93, Pd-103, In-111, W-181, Au-195, Bi-207, Be-7, Na-22, Ga-67,Ge-68, As-73,74, Y-88, Sr-85, Tc-95m, In-111, Sn-119, I-124,123, Ba-133, Re-183,184,Tl-201

Se-85, Со-57, Cd-109, Pd-103

6. UKT1B-120 Lead protective container, 120 mm wall thickness

С-14 P-32 S-35 Ca-45 Cr-51 Mn-54 Sc-46 Co-58,60 Ni-63 Zn-65 Fe-55,59 Se-75 Sr-85,89,90 Y-90 Cd-109 In-114

Co-60, Y-90, Cd-109


С-14 P-32 S-35 Ca-45 Cr-51 Mn-54 Sc-46 Co-58,60 Ni-63 Zn-65 Fe-55,59 Se-75 Sr-85,89,90 Y-90 Cd-109 In-114 Pu-239

Co-60, Y-90, Cd-109



Co-60, Y-90, Cd-109



Co-60, Y-90, Cd-109, In-114

30



Co-60, Y-90, Cd-109, In-114

11. UKT1A-5М1 Lead protective container, 5 mm wall thickness; polystyrene with 24 mm wall thickness


Pu-239

12. UKT1B-ZG


Co-60 Сs-137

Co-60 Сs-137

13. 1ВU 180 Lead protective container, 180 mm wall thickness

Ir-192 Ir-192

14. TR-1/t


Sr-89,90 Sr-89,90

15. GT-2m Lead protective insert inside the generator, 35 mm wall thickness

Tc-99m Tc-99m

16. GT-3 Lead protective insert inside the generator, 30 mm wall thickness

Tc-99m Tc-99m

17. GI-1 Lead protective insert inside the generator, 30 mm wall thickness

In-113m In-113m

18. Shipping packaging container of Ga-67 generator

Lead protective insert inside the generator, 25 mm wall thickness

Ga-67, Ge-68 Ga-67, Ge-68.

Table 3.3. Standard containers with shielding from depleted urnium. n/n Type of

shipping packaging container

Container design specifications

Transported material Most frequently shipped nuclide

1. UKT 1V-05/0050

Protective container of depleted uranium, 53.3 mm wall thickness

Ir-192 Ir-192

2. UKT 1V-10000/0185

Protective container of depleted uranium, 57 mm wall thickness

Сs-137 Ir-192 Ir-192

3. 3300A Protective container of depleted uranium, 163 mm wall thickness

Co-60 Сs-137

Co-60

4. UKT PV (U) (313-1) UKT PV (U) (495-1) UKT PV (U) (725-3)

Protective container of depleted uranium, wall thickness of 313 mm, 495 mm, and 725 mm, respectively.

Am-241 Сm-244 Cf-252 Cf-252

31

5. UKT 1V (U)-96-7


Co-58,60 Se-75 Ir-192 I-131 Mn-54 Fe-55,59 Mb-99

Co-60 Ir-192

6. UKT 1V -60-1 Protective container of depleted uranium, 75 mm wall thickness

Tl-204 Ir-192 W-188 Cs-137 Ba-133 I-131 Ru-106 Zn-65 Cd-109, 115 Sr 85,89,90 Sn-113,117,119,121 Ni-63 Co-58,60 Mn-54 Cr-51 Ca-45 P-32 C-14

Co-60 Ir-192 Cs-137

7. 2843 Protective container of lead with 226 mm wall thickness and depleted uranium with 25 mm wall thickness

Co-60 Сs-137

Co-60 Сs-137

8. ATEA 333/335 Protective container of depleted uranium, 55 mm wall thickness

Ir-192 Ir-192

9. D 80457 (S 289)


Co-60 Ir-192 Se-75 Co-60 Ir-192 Se-75

10. Amersham-Model-702


Сs-137 Ir-192 Se-75 Yb-169 Co-60 Ir-192 Se-75

11. UK-50-S Protective container of depleted uranium, 98 mm wall thickness

Co-60 Ir-192 Se-75 Sr-90 Cs-134,137 Ra-226 Аm-241,243, mixtures

Co-60 Ir-192 Se-75 Cs137

12. UK-12-S Protective container of depleted uranium, 62.5 mm wall thickness

Co-60 Se-75 Sr-90 Ir-192 Ra-226 Cs-134,137 Ам-241,243, mixtures

Co-60 Ir-192 Se-75 Cs137

13. Teatron Protective container of lead with 216 mm wall thickness and depleted uranium with 10 mm wall thickness

Со-60 Со-60

14. HU-GP 20 HU-GP 40

Protective container of depleted uranium, wall thickness of 14 mm and 25 mm, respectively.

С-14 P-32,33 S-35 Ca-45 Cr-51 Mn-54 Fe 55,59 Сo-58, Co-60 Ni-63 Zm-65 Se-75 Sr-85,89 Pd-103 Rh-106 Cd-109 In-113 Sn-113,117,119,121 Ba-133 Eu-152,154, U-235, Ам-241,243 Pu-238,239 Cf-252 Gd-153

Co-60 Se-75

15. UKT-PV-1 Additional shielding of paraffin with 200 mm wall thickness

Pu-238 Am-241 Cm-244 Cf-252 Po-210

Am-241 Cm-244 Cf-252 Po-210

32

4. TEST MEASUREMENT PROGRAM

4.1. General provisions Test measurements of portable radiation control devices used by customs officers are made in order to assess technical characteristics of these devices in checking the compliance of radioactive cargo transported in various shipping packaging containers (SPCs) and bringing down the risk of illegal trafficking of nuclear and radioactive materials:

• - Identification of nuclear and radioactive materials shipped in various shipping packaging containers;

• - Verification of spectrometer (identifier) capabilities in terms of the definition of isotopic composition of gamma nuclides in their various combinations;

• - Verification of spectrometer (identifier) top capabilities in terms of the definition of isotopic composition of gamma nuclides when placed in various protective containers.

The tests are carried out by representatives of the Federal Customs Service (FCS) of the Russian Federation together with specialists of the RRC “Kurchatov Institute,” the commission being chaired by an FCS representative. The arrangement and test procedures are to comply with the requirements established based on those provided in GOST 15.001-88 national standard. The results of the test measurements are intended to be used in the development of recommendations concerning the application of various radiation control devices in solving the tasks of customs control of the fissionable and radioactive materials.

Test measurements of radiation control devices used by customs officers are performed in the natural atmospheric environment of experimental room (at the site) of RRC “Kurchatov Institute.” The natural gamma background level should not exceed 0.2 µSv/h.

Each test measurement is to be made within 3 minutes after the completion of the previous measurement. During the intervals between the measurements, the samples of radioactive (nuclear) materials are to be withdrawn from the room (site) for a distance of at least 5 meters.

It is not allowed during the test measurements to make any adjustments to the radiation control devices, which are not envisaged in operational manuals.

Any abrupt changes occurring in the natural background at the site during the measurements should not exceed ± 5% per minute.

Radiation control devices are to be prepared for work in accordance with the requirements of the operational manuals (passports) for the selected modification.

When working with radioactive sources and samples of nuclear materials, radiation safety requirements should be met, which are contained in the Rules and Regulations:

• - OSPORB-99 – “Basic Sanitary Rules for Ensuring Radiation Safety. SP 2.6.12.799-99”;

• - NRB-99 – “Radiation Safety Regulations, SP 2.61.758-99”.

The following should also be complied with during test measurements:

• - Requirements of the “Rules of Technical Operation of Consumers’ Electric Facilities” (sections E1, E2, 13) approved by the Head of the State Inspection Agency for Energy (Glavgosenergonadzor) in 1992;

• - Requirements of the “Inter-industry Rules on Labor Safety (Safety Rules) when Operating Electric Facilities. POT RM-016-2001, RD 153-34.0-03.150-00”.

33

All the results of test measurements are recorded in protocols, which are to be signed by those engaged in test measurements in the presence of the commission members.

4.2. Test conditions used for gamma-sources Before taking spectrometric measurements, spectrometers may require energy calibration in accordance with Operation Manuals for the specific types of devices. Sample calibration gamma-sources included in device kits are used for these purposes. The procedures are accompanied by the assessment of the calibration convenience and its technological efficiency. When assessing spectrometric measurement procedures applicable to the customs control technologies the following principles should be followed:

• Maximum simplicity in operating the device;

• Maximum automation of adjustment procedures when making measurements;

• Maximum automation of calibration procedures;

• Minimum risk of an operator’s faulty actions.

The spectrum is measured at a distance of 10-15 cm from the source of radiation provided that the measurement dead time should not exceed 50%, whereas the measurement time, which depends upon the source activity level, protective layers thickness and material, as well as the detector type, should not exceed 5 minutes.

Before performing each measurement, all sources of ionizing radiation are removed from the room except for the inspected one. The spectra of the inspected samples are measured successively by each of the tested spectrometers. The measurement results are recorded in a Protocol. One sample may be measured by several devices simultaneously provided that the measurement conditions are identical. Nuclides are automatically identified in the analyzed SPC without using any additional software processing.

4.3. Test conditions used for neutron sources 1) Neutron source samples are placed successively on a flat surface so that the minimum distance from the center of the source to the path of the detector geometrical center would be (1±0.01) m (to determine the detection threshold of the monitor neutron channel);

2) The device is moved away to such a distance from the source that the source radiation would not affect the neutron background at the device location, i.e. to a recommended distance of 5 to 10 meters. In order to prevent the response of the gamma channel, the source is placed into a lead container with the wall thickness of 5 cm.

3) The device is switched on and operated in accordance with the device Operation Manual;

4) In order to determine the detection threshold of the device neutron channel, the device is carried along the preset path after it has entered the search mode at a speed of 0.5 ± 0.05 m/s against the neutron source. The speed should remain constant all along the path during the whole test series. Each time the device has completed the preset path it is moved away to a distance of 5 to 10 meters from the source for the device alarm indicators to go out and its count speed to set to values approximating the count speed level of the neutron background.

34

5) The devices are placed at the operator’s belt with their face to the source of radiation. The operator’s weight and size characteristics should approximate those of a polymethylmethacrylate phantom of a size of 30 x 30 x 15 cm.

6) The test results are considered satisfactory if the device gave at least nine alarm signals (audio or vibrating ones) when moved ten times along the preset path.

7) The said operations are repeated for all the above listed neutron sources to determine the minimum detection threshold of the device with regard to neutron radiation.

8) Before making detection tests using the neutron channels the devices should be calibrated in the natural neutron background (i.e. in the absence of neutron sources).

9) If negative result is obtained at a distance of 1 meter, the test conditions are repeated in an effort to determine the minimum distance making it possible for the device to detect the neutron source samples offered for the tests.

Note. Such test conditions for the device neutron channel are taken from GOST R51635-2000 Russian national standard “Radiation Monitors of Nuclear Materials” keeping in mind that similar test conditions related to the movement of the device relative to the source most closely match the technology of applying such types of devices to detect illegal trafficking of nuclear and radioactive materials and are actively used by the Russian customs.

4.4. Requirements to the means of control The following equipment and means of control are used in making measurements:

• Set of samples of radioactive materials, plutonium and uranium;

• Standard dosimetric equipment to control gamma-radiation dose rates (gamma- and x-ray dose rate meter AT1121 (Belarus) is used);

• Stopwatch in accordance with TU 25.1819.0021;

• Ruler in accordance with GOST 7502 national standard.

Measurement devices and means of measurements used when making special measurements should have technical passports and be pre-checked, which should be confirmed by a record in the relevant documents.

4.5. Requirements to radioactive and nuclear materials The work is carried out using various radioactive materials, which are sealed radionuclide sources of ionizing radiation, and NM consisting of plutonium and highly enriched uranium with an indication of their mass and isotope composition added by an indication of activity level for radioactive materials.

Plutonium-based NM should contain at least 98% of plutonium mass fraction with a total mass of other impurities within 2% (no more than 0.1% of the impurities for elements with a number less than ten) and at least 20% of Pu-239.

Uranium-based NM should contain at least 99.75% of uranium mass fraction with a total mass of other impurities within 0.25% and at least 0.1% of U-235.

NM should be kept in a leak-proof manner. Plutonium-based NM should be placed in a one- or two-layer protective capsule produced out of corrosion-resistant steel or nickel with a total wall thickness within 3 mm.

Uranium-based NM should be placed in a leak-proof protective aluminium casing with a thickness within 1.5 mm. Leak-proof protective casings produced out of other materials may be applied if the capsule wall thickness is equivalent to a 1.5-mm-thick Al wall in terms of weakening the basic U-235 gamma-line (186 keV).

35

The activity level of a wipe sample taken from the NM surface should not exceed 2⋅10-2Bq/cm2.

5. CONFIRMATION OF TRANSPORT INDEX OF NUCLEAR AND RADIOACTIVE MATERIALS SHIPPED IN VARIOUS SHIPPING PACKING

CONTAINERS

The objects of the investigations on control of transport index confirmation were the shipping packing containers (SPC) of various configuration, made of materials of various density with the following radioactive sources inside. The specifications of materials and containers are summarized in Table 5.1. Designation and specification of the devices for radiation control used for experimentation are given in Table 5.2. The results on measurements of exposure dose rate (EDR) are summarized in Table 5.3 and displayed in Fig. 5.1–5.3. The configuration of experimental equipment is given in Fig. 5.4.

The problem of transport index verification is most correctly solved with devices measuring EDR by the “reconstructed” spectrum of gamma-radiation:

• Radiometer-spectrometer MKS-A02-01;

• Spectrometer GAMMA-1C/NB;

• X-ray and gamma dosimeter EL-1119,

And devices measuring gamma-radiation with the Geiger counters:

• Radiometer-spectrometer PM1401K;

• Personal dosimeter DKG-PM1203;

• Personal dosimeter DKG-PM1621;

• Radiometer-spectrometer RSU-01 “Signal”.

• Survey dosimeter DRS-PM1401;

• Survey device ISP-PM1401K-1;

Theses devices measure the EDR correctly in the whole range of gamma-ray energies. Note: For the spectrometer GAMMA-1C/NB and radiometer-spectrometer MKS-A02-01, the measurements are correct only after required calibration of these devices by the calibrating gamma-sources (included in device kit). For radiometer-spectrometer RSU-01 “Signal”, the function of EDR measurement is called after device calibration. This function is available through the "alternative" menu, that essentially complicates actions of custom officers at the operative estimation of EDR with the help of this device. At the same time measurements are correct enough.

36

Tab

le 5

.1. S

peci

ficat

ion

of so

urce

s use

d fo

r exp

erim

enta

tion.

n/n

Sour

ce

desi

gnat

ion

SPC

type

M

ater

ial a

nd w

all t

hick

ness

of S

PC

Isot

ope

desi

gnat

ion

Isot

ope

activ

ity

(Bq)

Ura

nium

w

eigh

t

1 S1

K

T 1-

10

Lead

, 10

mm

Ва1

33

3.54×1

05

2 S2

K

T 1-

10

Lead

, 15

mm

Ва1

33

3.54×1

05

3 S3

K

T 1-

10

Lead

, 10

mm

С

s137

2.

24×1

04

4 S4

K

T 1-

15

Lead

, 15

mm

С

s137

2.

24×1

04

5 S5

K

T 1-

5 Le

ad, 5

mm

Со6

0+Со5

6 2.

0×10

4 +1.2×1

04

6 S6

U

KT

11v-

24

Cl-2

.8

Lead

, 30

mm

St

eel,

10 m

m

Eu15

2 9.

9×10

7

7 S7

K

3-1

(Alu

min

.)

Alu

min

ium

, 1 m

m

Cm

244

9.9×

108

8 S8

SU

\013

\B(U

) U

rani

um d

eple

ted

by U

235

isot

ope

dow

n to

0.

1%, 6

0 m

m, S

teel

, 10

mm

U

-238

(U-2

35 –

0.1

%)

65

kg

9 S9

U

\B(U

) U

rani

um d

eple

ted

by U

235

isot

ope

dow

n to

0.

2%, 3

0 m

m, S

teel

, 10

mm

U

-238

(U-2

35 –

0.2

%)

65

kg

10

S10

St

eel,

6 m

m

UO

2 (U

-235

– 0

.71%

)

198

g 11

S1

1

Stee

l, 6

mm

U

3O8 (U

-235

–5.

4%)

17

9 g

12

S12

St

eel,

6 m

m

U3O

8 (U

-235

– 2

1.6%

)

180

g

S13-

1

Lead

, 3 m

m

Stee

l, 3

mm

U

O2 (

U-2

35 –

36%

)

120

g 13

S1

3-2

St

eel,

1 m

m

UO

2 (U

-235

– 3

6%)

12

0 g

14

S14

St

eel,

6 m

m

U3O

8 (U

-235

– 4

7.1%

)

180

g 15

S1

5

Bra

ss, 3

mm

M

etal

(U-2

35 –

75%

)

538

g 16

S1

6

Stee

l, 6

mm

U

O2 (

U-2

35 –

89.

1%)

12

0 g

S17-

1

Stee

l, 1

mm

U

rani

um m

onon

itrid

e (U

N) (

U-2

35 –

93%

)

16 g

17

. S1

7-2

U

rani

um d

eple

ted

by U

235

isot

ope

dow

n to

0.

1%, 6

0 m

m, S

teel

, 10

mm

U

rani

um m

onon

itrid

e (U

N) (

U-2

35 –

93%

)

1106

g

18.

S18

St

eel,

3 m

m

Plut

oniu

m d

ioxi

de (P

u-23

9 –

75.6

%)

5

g 19

. S1

9

Stee

l, 3

mm

Pl

uton

ium

dio

xide

(Pu-

239

– 88

%)

5

g 20

. S2

0

Stee

l, 1

mm

(P

u-23

9 –

93%

)

7 g

37

Tab

le 5

.2. D

esig

natio

n an

d sp

ecifi

catio

n of

the

devi

ces f

or ra

diat

ion

cont

rol u

sed

for e

xper

imen

tatio

n.

Spec

ifica

tions

n/

n Sh

ort

desi

gnat

ion

of d

evic

e

Full

desi

gnat

ion

of d

evic

e T

ype

of

reco

rded

ra

diat

ion

Det

ecto

r ty

pe

Ran

ge o

f rec

orde

d ga

mm

a-ra

ys

Act

uatio

n tim

eM

easu

ring

tim

e

Poss

ibili

ty fo

r ex

pres

s id

entif

icat

ion

1 D

1 Pe

rson

al m

icro

proc

esso

r-ba

sed

dosi

met

er

DK

G-P

M12

03

γ G

eige

r-M

ulle

r (G

M)

coun

ter

0.06

-1.5

MeV

45

sec

5 se

c –

2 D

2 Pe

rson

al d

osim

eter

of X

-ray

s and

gam

ma-

rays

D

KG

-PM

1621

γ,X

-ray

G

M c

ount

er

0.01

–20

MeV

N

o m

ore

than

41

sec

1 m

in

–

3 D

3 Su

rvey

dos

imet

er D

RS-

PM14

01

γ C

sI (T

l) sc

intil

lato

r 0.

06–3

MeV

N

o m

ore

than

41

sec

0.25

sec

– 4

D4

Surv

ey m

icro

proc

esso

r-ba

sed

sign

al d

osim

eter

IS

P-PM

1401

K-1

γ,

neut

ron

CsI

(Tl)

scin

tilla

tor,

neut

ron

coun

ter

0.06

–3 M

eV

Cal

ibra

tion

time:

36

sec

No

mor

e th

an

41 se

c –

5 D

5 X

-ray

s and

gam

ma-

rays

dos

imet

er E

L-11

19

γ,X-r

ay

Plas

tic sc

intil

lato

r 0.

02–1

0 M

eV

Cal

ibra

tion

time:

40

sec

5 se

c –

6 D

6 M

ulti-

purp

ose

radi

omet

er-s

pect

rom

eter

M

KS-

A02

-01

γ, ne

utro

n,

alph

a, b

eta

N

aI (T

l) sc

intil

lato

r 0.

06–1

0 M

eV

Cal

ibra

tion

time:

40

sec

No

mor

e th

an5

min

20

7 D

7 H

and-

held

radi

omet

er-s

pect

rom

eter

RSU

-01

“Sig

nal”

γ,

neut

ron,

al

pha

CsI

(Tl)

ZnS

(Ag)

Li

F sc

intil

lato

rs

0.1–

10 M

eV

Cal

ibra

tion

time:

120

sec

No

mor

e th

an5

min

4

8 D

8 R

adio

met

er-s

pect

rom

eter

MK

S PM

1401

K

γ, X

-ray

, al

pha,

bet

a

GM

cou

nter

, the

rmal

ne

utro

n co

unte

r, C

sI

scin

tilla

tor

0.01

5–20

MeV

0.

15–3

.5 M

eV (f

or

mea

surin

g th

e flu

x de

nsity

of b

eta-

parti

cles

: 120

sec

36 se

c 40

9 D

9 Sp

ectro

met

er G

AM

MA

-1C

/NB

γ

NaI

(Tl)

scin

tilla

tor

0.05

–3 M

eV

300

sec

No

mor

e th

an5

min

20

10

R

efer

ence

Se

mic

ondu

ctor

spec

trom

eter

SK

S-50

γ

Pure

ger

man

ium

de

tect

or

0.05

–3 M

eV

300

sec

No

mor

e th

an

30 m

in

–

38

Tab

le 5

.3. R

esul

ts o

n m

easu

rem

ents

of e

xpos

ure

dose

rate

from

nuc

lear

and

radi

oact

ive

mat

eria

ls b

y th

e ra

diat

ion

cont

rol d

evic

es a

dapt

ed fo

r cus

tom

in

spec

tion

tech

nolo

gies

.

Mea

sure

d va

lue

of e

xpos

ure

dose

rate

, µSv

/h

Sour

ce

desi

gnat

ion

D1

D2

D3

D4

D5

D6

D7

D8

D9

S1

0.76

0.

7 0.

58

1.98

0.

8 0.

8 0.

8 0.

7 0.

8 S2

0.

22

0.16

6 0.

21

0.4

0.28

0.

3 0.

2 0.

2 0.

3 S3

0.

63

0.61

0.

39

0.58

0.

6 0.

6 0.

5 0.

5 0.

6 S4

0.

38

0.29

0.

156

0.44

0.

29

0.3

0.3

0.3

0.3

S5

11.8

10

.3

4.4

0.96

12

.7

13

11

12

13

S6

111.

6 66

.3

31

37.3

11

0 11

0 90

90

11

2 S7

11

.6

15.3

6

3.9

18.6

18

10

11

17

S8

8.

4 9.

76

13.3

15

.16

6.36

6.

5 8

9 6

S9

12.2

6 12

.53

10.8

11

.1

5.56

6

11

11

7 IS

0 1.

7 1.

66

1.56

1.

5 1.

65

1.7

1.5

2 1.

6 S1

1 1.

26

1.76

1.

75

1.7

1.82

1.

8 1

1.5

1.7

S12

2.34

2.

87

4.1

4.1

2.38

2.

4 2.

2 2.

4 2.

3 S1

3 0.

23

0.23

4.

26

4.46

4.

13

4.1

0.2

3 4

S14

8.1

4.48

6.

73

4.7

6.46

6

7 7

6 S1

5 7.

3 3.

6 16

.82

20.9

4.

52

5.5

4.2

7 5.

6 S1

6 5.

58

5 7.

64

9.1

5.6

5.6

5.1

5.4

6 S1

7 3.

54

2.18

4.

46

20.6

3 4.

12

4.4

3 3

4 S1

8 9.

1 6.

4 4.

23

9.86

8.

8 8

9 8

8.3

S19

7.07

4.

53

1.23

11

.4

6.3

6 7

7 6

S20

2.1

1.36

1.

46

2 4.

2 4

2.2

2 4.

3

39

D1 D2 D3 D4 D5 D6 D7 D8 D9S8

S11

S14

S17S20

0

5

10

15

20

25

EDR

, mkS

v/h

S8S9S10S11S12S13S14S15S16S17S18S19S20

Fig.5.1. Integrated diagram of EDR measurements by above devices.

D1 D2 D3 D4 D5 D6 D7 D8 D910mm Pb

0

0,5

1

1,5

2

EDR,

mkS

v/h

10mm Pb 15mm PB

Fig. 5.2. Results of measuring Ba-133 with activity of 0.354 MBq.

40

D1 D2 D3 D4 D5 D6 D7 D8 D9

Ba133 10mm PbBa133 15mm Pb

Cs137 10mm PbCs137 15mm Pb0

0,2

0,4

0,6

0,8

1

1,2

1,4

1,6

1,8

2

EDR,

mkS

v/h

Ba133 10mm Pb Ba133 15mm Pb Cs137 10mm Pb Cs137 15mm Pb

Fig. 5.3. Relative changes in EDR measurements. The devices least adapted for EDR measurement in the energy range from 50 keV up to 3000 keV are:

• Survey dosimeter DRS-PM1401;

• Survey device ISP-PM1401K-1.

This is a result of usage of CsI (Tl) scintillating detector which is not compensated by energy. The last is adapted first of all for the detection of nuclear materials (NM). This fact that is well visible on considerably increased EDR values in the energy range typical for NM. At the same time the measured values of EDR are much lower in the 50–60 keV energy range.

Fig.5.4. Configuration of experimental equipment.

41

6. IDENTIFICATION OF NUCLEAR AND RADIOACTIVE MATERIALS SHIPPED IN VARIOUS SHIPPED PACKING CONTAINERS

6.1. Experimentations in 2003 at RRC “Kurchatov Institute” on the definition of isotopic composition of gamma nuclides

For study of identification of the nuclear and radioactive materials transported in various shipping packing containers (SPC), the following custom devices of the radiating control with the gamma-spectrometer channel were tested:

• Radiometer-spectrometer MKS-A02-01;

• Radiometer-spectrometer RSU-01 “Signal”;

• Spectrometer GAMMA-1C/NB;

• Radiometer-spectrometer PM1401K.

The work was performed with of various SPC configurations, made of materials of various density with the following radioactive sources inside (see Table 5.1):

• KT 1-10 (Ва-133 source, activity: 3.54·105 Bq);

• KT 1-15 (source: Ва-133, activity: 3.54·105 Bq);

• KT 1-10 (source: Сs-137, activity: 2.24·104 Bq);

• KT 1-15 (source: Сs-137, activity: 2.24·104 Bq);

• KT 1-5 (source: Со-60, activity: 2.0 ·104 Bq; source: Со-56, activity: 1.2*104 Bq)

• KT –11v-24Cl-2.8 (source: Eu-152, activity: 9.9·107 Bq);

• K3-1 (aluminium) (source: Cm-244, activity: 9.9·108 Bq)

• Container SU\013\B(U) of U-238 (enrichment by U 235: 0.1%), mass of U: 65 kg.

• Container U\B(U) of U-238 (enrichment by U 235: 0.2%), mass of U: 65 kg,

and also samples of nuclear materials in various packages:

• U sample, enrichment by U 235: 0.71%; mass of UO2: 198 g;


• U sample, enrichment by U 235: 21.6%; mass of U3O8: 180 g;


• U sample, enrichment by U 235: 47.1%; mass of U3O8: 180 g;

• U sample, enrichment by U 235: 75. 0%; metall-538 g;



• U sample, enrichment by U 235: 93.0%; mass of uranium nitride (UN): 16 g;

• U sample, enrichment by U 235: 93.0%; mass of uranium nitride (UN): 1106 g;

• Pu-239 sample (75.6%); mass of plutonium dioxide (PuO2): 5 g;



42

For identity of measurements, the devices were settled down on a circle around the object of control at the identical distance. The distance was chosen from the manufacturer recommendations on counting load for devices. As a rule, the distance from the package to the source was within 15-20 cm. A) All the devices were tested on the capability of the express - analysis in the "simplified" mode of measurement with the use of built-in nuclide libraries. The device capabilities on the further processing the measured spectrum through the computer were separately checked, using the additional software supplied with these spectrometers. The devices MKS-A02-01 and GAMMA-1C/NB require the special calibration by Th-232 prior to the express-analysis in the "simplified" mode of measurement with the use of built-in nuclide libraries. The device RSU-01 “Signal” needs the calibration by Na-22. At the same time for further processing the measured spectrum with the computer, the additional calibration by Eu-152 is required. It is sufficiently complicated the calibration and increases the spectrum processing time. The total statistics of identification at the express-analysis is given in Fig.6.1.

0

5

10

15

20

1234

Fig.6.1. Statistics of identification in the express-analysis mode.

B) After calibration, the devices MKS-A02-01 and GAMMA-1C/NB could identify only “simple” spectrums such as Cs-137 and Co-60. After additional calibration by Eu-152 and by processing the measured spectrum through the computer, all nuclides were identified excluding Cm-244 and Am-241. C) In the “express-analysis” mode, the device RSU-01 “Signal” could identify and assess activity of all four nuclides available in its Data Bank: Th-232, Ra-226, K-40, Cs-137. Other nuclides were not identified.

0

5

10

15

20

1234

Fig.6.2. Statistics of identification at the additional software processing of measured spectrum.

D) The total statistics of identification at the additional software processing of measured spectrum is given in Fig.6.2. The survey radiometer-spectrometer PM1401K has the calibration Cs-137 source inside the casing. It allows avoiding manual calibration. The main advantages of

43

the device are minimization of manual operations, simplicity of device operation in the “express-analysis” and “expert” modes. In the “express-analysis” mode, the device could identify most control nuclides (17 from 22; all uranium samples were preliminary identified. At the same time, the samples from low- and highly enriched uranium were well distinguished which is unique for the device of this class). E) The measurements demonstrated that isotopes, which spectrum of radiation lays in soft energy area (highly enriched U-235, line 186 keV, Am-241, line 56 keV, etc.) are rather difficult for identification. Nevertheless, at the combined shielding with 3 mm of lead and 3 mm of steel the identification of uranium sample with enrichment of 36% (by U-235) is possible. For the sample (1106 g) of highly (93%) enriched U-235 (Fig. 6.3), at the shielding of reduced-enriched uranium (thickness of 60 mm), the U-235 has not been identified.

Fig. 6.3. Measurements of highly shielded (thickness of 60 mm) enriched U-235.

6.2. Experimentations in 2005 at RRC “Kurchatov Institute” on the definition of isotopic composition of gamma nuclides

The following radiation control devices used by customs officers and equipped with a spectrometric channel to register gamma-radiation are tested for the purpose of identifying nuclear and radioactive materials shipped in various shipping packaging containers:

• - Radiometer-spectrometer MKS-A03 (Russia);

• - Radiometer-spectrometer AT 6101 (Belarus);

• - Radiometer-spectrometer (identifier) CanberraInSpector 1000 (USA);

• - Radiometer-spectrometer (identifier) IdentiFINDER (Target) with NaI;

• - Radiometer-spectrometer (identifier) Exploranium GR-135;

• - Survey radiometer-spectrometer PM1401K;

• - Survey identifier (Smartphone) PM1802.

44

Spectrometric measurements are performed using radioactive sources and nuclear materials placed beyond various protection layers. The samples of uranium and plutonium sources used in the experiments ate given in Fig. 6.4 and 6.5.

Fig. 6.4. Uranium sources.

1) Ba-133 source, activity: 3⋅105 Bq,

2) Cs-137 source, activity: 3⋅104 Bq,

3) Eu-152 source, activity: 3⋅105 Bq,

4) Co-60 source, activity: 3⋅105 Bq,

5) Ra-226 source, activity: 2⋅105 Bq,

6) Th-232 source, activity: 3⋅103 Bq,

7) K-40 source, activity: 5⋅102 Bq,

8) -U sample (reactor-type U with Cs-137) No. 5c, enrichment by U-235: 0.71%; mass of

UO2: 187 g

9) -U sample (reactor-type U with Cs-137) No. 6c, enrichment by U-235: 4.4%; mass of

U3O2: 79 g

10) -U sample No. 4, enrichment by U-235: 21.6%; mass of U3O8: 180 g

11) -U sample No. 3, enrichment by U-235: 47.1%; mass of U3O8: 180 g

12) -U sample No. 2, enrichment by U-235: 88.9%; metal: 10.4 g

13) -U sample No. 1, enrichment by U-235: 90.1%; metal: 3.1 g

14) -Pu-239 sample No. 7 (96.7%); Pu metal: 0.3 g

45

15) -Pu-239 sample No. 8 (90.0%); Pu dioxide: 5.9 g

16) -Pu-239 sample No. 9, irradiated reactor-type Pu; Pu dioxide: 60 g

Fig. 6.5. Plutonium sources.

The samples of protective containers are shown in Fig. 6.6.

6.2.1. Identification of nuclear and radioactive materials in various packaging containers

The results on the definition of isotopic composition of radioactive substances and radioactive materials placed in various protective containers are summarized in Table 6.1 and Table 6.2, correspondingly, and displayed in Fig. 6.6–6.8.

Fig. 6.6. The snap-shots of experiments on identification of sources.

46

Table 6.1. Definition of isotopic composition of radioactive substances placed in various protective containers

Identification results (names of identified isotopes) Samples to be

identified MKS-A03

АТ6101 CanberraInSpector 1000

IdentiFINDER (Target)

Exploranium GR-135

PM1401K (Smartphone) PM1802

Identification conditions

Ba-133 with

activity level of

3⋅105 Bq, no

protection

Ва-133 (90 sec)

Ва-133 (100 sec)

Ва-133 (60 sec)

Ва-133 (100 sec)

No (300 sec)

Ва-133 (90 sec)

Ва-133 (100 sec)

Devices

equivalent dose rate:

1.22 µSv/h

Ba-133 with

activity level of 3⋅105 Bq in KT1-5

(5 mm Pb)

Ва-133 (40 sec)

Ва-133 (23 sec)

Ra-226 K-40

(300 sec)

Ва-133 (60 sec)

No (300 sec)

Ва-133 (60 sec)

Ва-133 (180 sec)

Devices


0.85 µSv/h

Ba-133 with

activity level of 3⋅105 Bq

in KT1-10 (10 mm

Pb)

Ва-133 (180 sec)

Ва-133 (45 sec)

Ra-226 K-40

(300 sec)

Ва-133 Eu-152

(100 sec)

No (300 sec)

Ва-133 (90 sec)

Ва-133 (240 sec)

Devices


0.25 µSv/h

Ba-133 with


in KT1-15 (15 mm

Pb)

Ва-133 (240 sec)

Ва-133 (100 sec)

Ra-226 K-40

(300 sec)

No (300 sec)

U-235 (300 sec)

Ва-133 (100 sec)

Ва-133 (240 sec)

Devices


0.15 µSv/h

Ba-133 with


in KT1-20 (20 mm

Pb)

К-40 (300 sec)

К-40 (300 sec)

Ra-226 K-40

(300 sec)

No (300 sec)

U-235 К-40

(300 sec)

Ва-133 Cs-137

(300 sec)

No (300 sec)

Devices


0.11 µSv/h

Cs-137 with

activity level of

3⋅104 Bq, no

protection

Cs-137 (30 sec)

Cs-137 (23 sec)

Cs-137 (60 sec)

Cs-137 (30 sec)

Xe-133 (300 sec)

Cs-137 (30 sec)

Cs-137 (50 sec)

Devices

equivalent dose rate: 4 µSv/h

Cs-137 with


(5 mm Pb)

Cs-137 (50 sec)

Cs-137 (23 sec)

Cs-137 (60 sec)

Cs-137 (30 sec)

Weapon-grade Pu

K-40

Cs-137 (60 sec)

Cs-137 (60 sec)

Devices


47

Cs-137 with


in KT1-10 (10 mm

Pb)

Cs-137 (30 sec)

Cs-137 (23 sec)

Cs-137 (60 sec)

Cs-137 (60 sec)

No (300 sec)

Cs-137 (30 sec)

Cs-137 (30 sec)

Devices


1.45 µSv/h

Cs-137 with


in KT1-15 (15 mm

Pb)

Cs-137 (35 sec)

Cs-137 (30 sec)

No (300 sec)

Cs-137 (60 sec)

No (300 sec)

Cs-137 (35sec)

Cs-137 (20 sec)

Devices


0.95 µSv/h

Cs-137 with


in KT1-20 (20 mm

Pb)

Cs-137 (30 sec)

Cs-137 (32 sec)

No (300 sec)

Cs-137 (60 sec)

No (300 sec)

Cs-137 (30 sec)

Cs-137 (30 sec)

Devices


0.45 µSv/h

Eu-152 with

activity level of

3⋅105 Bq, no

protection

Eu-152 (90 sec)

Eu-152 (60 sec)

Eu-152 (300 sec)

Eu-152 (90 sec)

No (300 sec)

Eu-152 (40 sec)

Eu-152 (90 sec)

Devices

equivalent dose rate: 0.8

µSv/h

Eu-152 with


(5 mm Pb)

Eu-152 (90 sec)

Eu-152 (90 sec)

Eu-152 (300 sec)

No (300 sec)

No (300 sec)

Zn-65 (40 sec)

No (300 sec)

Devices

equivalent dose rate: 0.4 µSv/h

Eu-152 with


in KT1-10 (10 mm

Pb)

No (300 sec)

No (300 sec)

No (300 sec)

No (300 sec)

No (300 sec)

No (300 sec)

No (300 sec)

Devices


Co-60 with

activity level of

3⋅105 Bq, no

protection

Co-60 (40 sec)

Co-60 (10 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (50 sec)

Co-60 (30 sec)

Devices


Co-60 with


(5 mm Pb)

Co-60 (25 sec)

Co-60 (10 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (25 sec)

Co-60 (20 sec)

Devices


48

Co-60 with


in KT1-10 (10 mm

Pb)

Co-60 (30 sec)

Co-60 (10 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (30 sec)

Co-60 (30 sec)

Devices


µSv/h

Co-60 with


in KT1-15 (15 mm

Pb)

Co-60 (40 sec)

Co-60 (15sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (30 sec)

Co-60 (50 sec)

Devices


0.98 µSv/h

Co-60 with


in KT1-20 (20 mm

Pb)

Co-60 (30 sec)

Co-60 (30 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (60 sec)

Co-60 (30 sec)

Co-60 (40 sec)

Devices


µSv/h

Ra-226 with

activity level of

2⋅105 Bq, no

protection

Ra-226 (30 sec)

Ra-226 (40 sec)

Ba-133 Ga-67 Tl-204

(120 sec)

Ra-226 (60 sec)

Ra-226 Pu U

(120 sec)

Ra-226 (25 sec)

Ra-226 (22 sec)

Devices


µSv/h

Ra-226 with

activity level of

2*105 Bq in KT1-5

(5 mm Pb)

Ra-226 (240 sec)

Ra-226 (150 sec)

No (300 sec)

Ra-226 (120 sec)

Ra-226 Pu

K-40 (60 sec)

Ra-226 (180 sec)

Ra-226 (120 sec)

Devices


0.24 µSv/h

Th-232 with

activity level of

3⋅103 Bq, no

protection

Th-232 (60 sec)

Th-232 (56 sec)

U-235 U-238

(60 sec)

Th-232 (60 sec)

Th-232 (60 sec)

Th-228 (60 sec)

Th-232 Th-228

(120 sec)

Devices


0.32 µSv/h

Th-232 with


(5 mm Pb)

Th-232 (120 sec)

Th-232 (120 sec)

U-235 U-238

(150 sec)

Th-232 (120 sec)

No (300 sec)

No (300 sec)

Ra-226 (120 sec)

Devices


0.17 µSv/h

K-40 with activity level of

5⋅102 Bq, no

protection

K-40 (120 sec)

K-40 (120 sec)

No (300 sec)

K-40 (120 sec)

U-235 K-40

(120 sec)

K-40 Cs-137

(300 sec)

K-40 (300 sec)

Devices


0.19 µSv/h

49

0

0.5

1

1.5

2

Lead container thickness, cm

Ba-133

Cs-137

Eu-152

Ra-226

Co-60

Th-232

АТ61

01

GR

-135

Targ

et

Iden

t-100

0

МКС

-А03

РМ18

02

РМ14

01К

Isotopes

Device type

АТ6101

GR-135

Target

Ident-1000

МКС-А03

РМ1802

РМ1401К

Fig.6.7. Identification of radioactive materials in lead containers.

0

1

2

3

4

5

6U-235-90%

U-235-88%,

U-235-48%

U-235-21%

U-235-0,71%, Cs-137

U-235-4.4%, industrial

АТ61

01

GR-1

35

Targ

et

Iden

t-100

0

МКС

-А03

РМ18

02

РМ14

01К

АТ6101

GR-135

Target

Ident-1000

МКС-А03

РМ1802

РМ1401К

Fig.6.8. Identification of uranium in lead containers.

50

Pu-239-90% и U-235-90%

Pu-239, industrial+Cs137

АТ61

01

GR

-135

Targ

et

Iden

t-100

0

МКС

-А03

РМ18

02

РМ14

01К

0

2

4

6

8

10

12

14

16

АТ6101

GR-135

Target

Ident-1000

МКС-А03

РМ1802

РМ1401К

Fig.6.9. Identification of plutonium in lead containers.

51

Tabl

e 6.

2. D

efin

ition

of i

soto

pic

com

posi

tion

of n

ucle

ar m

ater

ials

in v

ario

us p

rote

ctiv

e co

ntai

ners

Id

entif

icat

ion

resu

lts (n

ames

of i

dent

ified

isot

opes

) Sa

mpl

es to

be

iden

tifie

d M

KS-

A03

A

T61

01

Can

berr

aInS

pect

or

1000

Id

entiF

IND

ER

(T

arge

t)

Exp

lora

nium

G

R-1

35

PM14

01K

(S

mar

tpho

ne)

PM18

02

Iden

tific

atio

n co

nditi

ons

U-2

35 –

90%

Sa

mpl

e N

o. 1

, 3 g

N

o pr

otec

tion

U-2

35, s

peci

al,

(180

sec)

U

-235

, wea

pon-

grad

e,

(23

sec)

U-2

35,

(300

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35,

Pu,

(120

sec)

U-2

35, w

eapo

n-gr

ade,

(2

5 se

c)

U-2

35, w

eapo

n-gr

ade,

(1

20 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.3

3 µS

v/h

U-2

35 –

90%

Sa

mpl

e N

o. 1

, 3 g

Pr

otec

tion:

1 m

m,

Cu

U-2

35, s

peci

al,

(40

sec)

U

-235

, wea

pon-

grad

e,

(25

sec)

U-2

35,

(300

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35,

Pu,

K-4

0 (1

50 se

c)

U-2

35, w

eapo

n-gr

ade,

(3

0 se

c)

U-2

35, w

eapo

n-gr

ade,

(6

0 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.2

9 µS

v/h

U-2

35 –

90%

Sa

mpl

e N

o. 1

, 3 g

Pr

otec

tion:

2 m

m,

Cu

U-2

35, s

peci

al,

(40

sec)

U

-235

, wea

pon-

grad

e,

(23

sec)

U-2

35,

(300

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35,

Pu,

K-4

0 (1

50 se

c)

U-2

35, w

eapo

n-gr

ade,

(2

0 se

c)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.26

µSv/

h

U-2

35 –

90%

Sa

mpl

e N

o. 1

, 3 g

Pr

otec

tion:

1.5

mm

, Fe

U-2

35, s

peci

al,

(40

sec)

U

-235

, wea

pon-

grad

e,

(27

sec)

U-2

35,

(300

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35,

Pu,

K-4

0 (1

80 se

c)

U-2

35, w

eapo

n-gr

ade,

(3

0 se

c)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.23

µSv/

h

U-2

35 –

88.

9%

Sam

ple

No.

2, 1

0 g

Prot

ectio

n: 1

.5 m

m,

Fe

U-2

35, s

peci

al,

(40

sec)

U

-235

, wea

pon-

grad

e,

(15

sec)

U-2

35,

(300

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35,

(300

sec)

U

-235

, wea

pon-

grad

e,

(30

sec)

U-2

35, w

eapo

n-gr

ade,

(2

5 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.4

µSv

/h

U-2

35 –

88.

9%

Sam

ple

No.

2, 1

0 g

Prot

ectio

n: 5

mm

, Pb

U-2

35, s

peci

al,

Co-

57

(300

sec)

Th-2

28,

(160

sec)

N

o (3

00 se

c)

No

(300

sec)

U

with

Pb

prot

ectio

n,

K-4

0 (1

80 se

c)

No

(300

sec)

N

o (3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.1

5 µS

v/h

U-2

35 –

88.

9%

Sam

ple

No.

2, 1

0 g

Prot

ectio

n: 5

mm

, Pb

, 1

mm

, Cu

U-2

35, s

peci

al,

K-4

0 (3

00 se

c)

Th-2

28,

(133

sec)

N

o (3

00 se

c)

Bi-2

07

(100

sec)

U

with

Pb

prot

ectio

n,

K-4

0 (2

40 se

c)

No

(300

sec)

N

o (3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.1

4 µS

v/h

U-2

35 –

47%

Sa

mpl

e N

o. 3

, 158

g;

Prot

ectio

n: 3

mm

, Fe

U-2

35, s

peci

al

(60

sec)

U

-238

, nat

ural

Th

-228

, in

dust

rial

(30

sec)

U-2

35

(120

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

U-2

38

Pu

(60

sec)

Th-2

32,

indu

stria

l (6

0 se

c)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 1

.2 µ

Sv/h

52

U-2

35 –

47%

Sa

mpl

e N

o. 3

, 158

g;

Pro

tect

ion:

3 m

m,

Fe,

1 m

m, C

u

U-2

35, s

peci

al

(60

sec)

U

-238

, nat

ural

Th

-228

, in

dust

rial

(47

sec)

U 2

35

(60

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

U-2

38

Pu

(60

sec)

Th-2

32,

indu

stria

l (6

0 se

c)

Th-2

32,

Th-2

28,

indu

stria

l (6

0 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 1.1

µSv

/h

U-2

35 –

21%

Sa

mpl

e N

o. 4

, 158

gPr

otec

tion:

3 m

m,

Fe

U-2

38,

U-2

35

(60

sec)

U-2

38,

U-2

35

(30

sec)

U 2

35

(60

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

U-2

38

Pu

(60

sec)

Pu-2

40,

(240

sec)

U

-238

, (2

40 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.6

3 µS

v/h

U-2

35 –

21%

Sa

mpl

e N

o. 4

, 158

g;

Pro

tect

ion:

3 m

m,

Fe,

1 m

m, C

u

U-2

38,

U-2

35

(120

sec)

U-2

38,

U-2

35

(30

sec)

U 2

35

(60

sec)

U

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

U-2

38

Pu

(60

sec)

U-2

35,

U-2

38

(70

sec)

U-2

38

(100

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.62

µSv/

h

U-2

35 –

0.7

1%

Sam

ple

No.

5, 1

73 g

R

eact

or-ty

pe sa

mpl

e (w

ith C

s-13

7);

Prot

ectio

n: 3

mm

, Fe

Cs-

137,

U

-238

, (5

0 se

c)

Cs-

137,

U

-238

, (4

0 se

c)

No

(300

sec)

C

s-13

7,

U w

ith n

o in

dica

tion

of

isot

ope

type

, (6

0 se

c)

U-2

35

U-2

38

(60

sec)

Cs-

137,

U

-238

, (2

00 se

c)

Cs-

137,

U

-238

, (1

50 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.4

4 µS

v/h

U-2

35 –

4.4

%

reac

tor-

type

, Sa

mpl

e N

o. 6

, 79

g Pr

otec

tion:

5 m

m,

Fe

Cs-

137,

U

-238

, U

-235

(1

20 se

c)

Cs-

137,

U

-235

U

-238

, (3

0 se

c)

Cs-

137,

U

-235

(1

20 se

c)

Cs-

137,

(3

00 se

c)

Cs-

137,

U

-235

(1

20 se

c)

Cs-

137,

(2

40 se

c)

Cs-

137,

(2

40 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.9

µSv

/h

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

; N

o pr

otec

tion

Pu-2

39 W

G, s

peci

al

Am

-241

(6

0 se

c)

Am

-241

, Pu

-239

(8

0 se

c)

U-2

35

(120

sec)

Pu

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

Pu w

ith n

o in

dica

tion

of

isot

ope

type

(6

0 se

c)

Pu-2

39

wea

pon-

grad

e (1

80 se

c)

Pu-2

39 w

eapo

n-gr

ade

(280

sec)

Dev

ices

equ

ival

ent

dose

rate

: 0.9

µSv

/h

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

; Pr

otec

tion:

1.5

mm

, C

u

Pu-2

39 W

G, s

peci

al

Am

-241

(6

0 se

c)

Pu-2

39

(45

sec)

U

-235

(1

20 se

c)

Pu w

ith n

o in

dica

tion

of

isot

ope

type

(6

0 se

c)

U-2

35

Pu w

ith n

o in

dica

tion

of

isot

ope

type

(6

0 se

c)

Pu-2

39

wea

pon-

grad

e (1

80 se

c)

Sn-1

13

(60

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.5 µ

Sv/h

53

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

; Pr

otec

tion:

1.5

mm

, C

u,

1 m

m, F

e

Pu-2

39 W

G, s

peci

al

(60

sec)

Pu

-239

(4

5 se

c)

U-2

35

(120

sec)

Pu

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

U-2

35

Pu w

ith n

o in

dica

tion

of

isot

ope

type

(6

0 se

c)

Cs-

137,

(1

50 se

c)

Sn-1

13

Co-

57

(60

sec)

Dev

ices

equ

ival

ent

dose

rate

: 0.4

1 µS

v/h

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

; Pr

otec

tion:

5 m

m, P

b

Pu-2

39 W

G, s

peci

al

(120

sec)

Pu

-239

K

-40

(260

sec)

No

(300

sec)

N

o (3

00 se

c)

U-2

35

(60

sec)

N

o (3

00 se

c)

Sn-1

13

(150

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.14

µSv/

h Pu

-239

– 9

6.7%

Sa

mpl

e N

o. 7

, 0.3

g;

Prot

ectio

n: 1

0 m

m,

Pb

K-4

0,

Pu-2

39 W

G, s

peci

al

(300

sec)

No

(300

sec)

N

o (3

00 se

c)

No

(300

sec)

U

-235

Pu

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

No

(300

sec)

N

o (3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.1

1 µS

v/h

Pu-2

39 –

90%

Sa

mpl

e N

o. 8

, 5.9

g;

No

prot

ectio

n

Pu-2

39 W

G, s

peci

al

Am

-241

(6

0 se

c)

Am

-241

, Pu

-239

nuc

lear

, (4

0 se

c)

Am

-241

, (1

00 se

c)

Am

-241

, Pu

with

no

indi

catio

n of

is

otop

e ty

pe

(60

sec)

Pu w

ith n

o in

dica

tion

of

isot

ope

type

(6

0 se

c)

No

(300

sec)

Sn

-113

(1

50 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 1.2

µSv

/h

Pu-2

39 –

90%

Sa

mpl

e N

o. 8

, 5.9

g;

in K

T1-5

(5 m

m P

b)

Pu-2

39 W

G, s

peci

al

(60

sec)

N

o (3

00 se

c)

No

(300

sec)

N

o (3

00 se

c)

U-2

35

(60

sec)

Pu

-239

C

s-13

7,

(300

sec)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.16

µSv/

h Pu

-239

– 9

0%

Sam

ple

No.

8, 5

.9 g

; in

KT1

-10

(10

mm

Pb

)

Pu-2

39 W

G, s

peci

al

(120

sec)

N

o (3

00 se

c)

Am

-241

, (1

20 se

c)

No

(300

sec)

Pu

with

no

indi

catio

n of

is

otop

e ty

pe

U-2

35

(60

sec)

Cs-

137,

(3

00 se

c)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.145

µSv

/h

Pu-2

39 –

90%

Sa

mpl

e N

o. 8

, 5.9

g;

in K

T1-1

5 (1

5 m

m

Pb)

Pu-2

39 W

G, s

peci

al

K-4

0 (2

40 se

c)

No

(300

sec)

N

o (3

00 se

c)

No

(300

sec)

N

o (3

00 se

c)

Cs-

137,

(3

00 se

c)

No

(300

sec)

D

evic

es e

quiv

alen

t do

se ra

te: 0

.14

µSv/

h

Pu-2

39 ir

radi

ated

, re

acto

r-ty

pe, 5

mm

Sa

mpl

e N

o. 9

, 60

g

Pu-2

39 W

G, s

peci

al

Pu-2

39 re

acto

r-ty

pe,

Am

-241

(3

00 se

c)

Cs-

137,

C

r-51

(1

20 se

c)

No

(300

sec)

N

o (3

00 se

c)

Am

-241

U

-235

(1

20 se

c)

Cs-

137,

(3

00 se

c)

Cs-

137,

(3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 5 µ

Sv/h

54

6.2.

2. Id

entif

icat

ion

of n

ucle

ar a

nd ra

dioa

ctiv

e m

ater

ials

in th

eir v

ario

us c

ombi

natio

ns

The

resu

lts o

n th

e de

finiti

on o

f iso

topi

c co

mpo

sitio

n of

radi

oact

ive

subs

tanc

es a

nd ra

dioa

ctiv

e m

ater

ials

in th

eir v

ario

us c

ombi

natio

ns a

re s

umm

ariz

ed

in T

able

6.3

. Ta

ble

6.3.

Id

entif

icat

ion

resu

lts (n

ames

of i

dent

ified

isot

opes

) Sa

mpl

es to

be

iden

tifie

d M

KS-

A03

A

T61

01

Can

berr

aInS

pect

or

1000

Id

entiF

IND

ER

(T

arge

t)

Exp

lora

nium

G

R-1

35

PM14

01K

(S

mar

tpho

ne)

PM18

02

Iden

tific

atio

n co

nditi

ons

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

U

-235

– 9

0%,

Sam

ple

No.

1, 3

g

No

prot

ectio

n

Pu-2

39 w

eapo

n-gr

ade,

U

-235

(6

0 se

c)

Pu-2

39 w

eapo

n-gr

ade,

U

-235

(3

0 se

c)

U-2

35

(60

sec)

U

, with

no

indi

catio

n of

is

otop

e (6

0 se

c)

U-2

35

(120

sec)

U

-238

, U

-235

(1

20 se

c)

Sn-1

13,

U-2

35

(120

sec)

Dev

ices

equ

ival

ent

dose

rate

of P

u-23

9:

0.6

µSv/

h D

evic

es e

quiv

alen

t do

se ra

te o

f U-2

35: 0

.6

µSv/

h Pu

-239

– 9

6.7%

Sa

mpl

e N

o. 7

, 0.3

g

Ba-

133

with

act

ivity

of

3*1

05 B

q N

o pr

otec

tion

Pu-2

39 w

eapo

n-gr

ade,

A

m-2

41,

Ba-

133

(60

sec)

Ba-

133

(30

sec)

I-

131

Am

-241

(1

20 se

c)

Ba-

133,

Pu

with

no

indi

catio

n of

is

otop

e (6

0 se

c)

Pu-2

39

(120

sec)

B

a-13

3 (1

20 se

c)

Ba-

133

(120

sec)

D

evic

es e

quiv

alen

t do

se ra

te o

f Pu-

239:

1.

5 µS

v/h

Dev

ices

equ

ival

ent

dose

rate

of B

a-13

3:

1.5

µSv/

h Pu

-239

– 9

6.7%

Sa

mpl

e N

o. 7

, 0.3

g

Ba-

133

with

act

ivity

of

3*1

05 B

q Pr

otec

tion:

1 m

m, F

e

Ba-

133,

Pu

-239

wea

pon-

grad

e,

Am

-241

, (1

80 se

c)

Ba-

133

(30

sec)

I-

131

(120

sec)

B

a-13

3 (6

0 se

c)

Pu-2

39

Np-

237

(120

sec)

No

(300

sec)

N

o (3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

of P

u-23

9:

0.9

µSv/

h D

evic

es e

quiv

alen

t do

se ra

te o

f Ba-

133:

0.

8 µS

v/h

Pu-2

39 –

96.

7%

Sam

ple

No.

7, 0

.3 g

B

a-13

3 w

ith a

ctiv

ity

of 3

*105

Bq

Prot

ectio

n: 2

mm

, Fe

Ba-

133,

Pu

-239

wea

pon-

grad

e,

Am

-241

, (1

20 se

c)

Ba-

133

(60

sec)

I-

131

(120

sec)

B

a-13

3 (6

0 se

c)

Pu-2

39

Ba-

133

(120

sec)

Ba-

133

(120

sec)

B

a-13

3 (1

20 se

c)

Dev

ices

equ

ival

ent

dose

rate

of P

u-23

9:

0.5

µSv/

h D

evic

es e

quiv

alen

t do

se ra

te o

f Ba-

133:

0.

5 µS

v/h

U-2

35 –

4.4

%,

rege

nera

ted

with

Cs-

137

Sam

ple

No.

6, 7

9 g

Prot

ectio

n: 5

mm

, Fe

Cs-

137,

U

-238

, U

-235

(1

20 se

c)

Cs-

137,

U

-238

, U

-235

(3

0 se

c)

Cs-

137,

U

-238

, (1

20 se

c)

Cs-

137,

(3

00 se

c)

Cs-

137,

U

(1

80 se

c)

Cs-

137,

(3

00 se

c)

Cs-

137,

(3

00 se

c)

Dev

ices

equ

ival

ent

dose

rate

: 0.9

µSv

/h

55

6.2.3. Detection of nuclear materials by neutron emission

Neutron channels undergo additional tests in the following devices:

• - Radiometer-spectrometer MKS-A03 (Russia);

• - Radiometer-spectrometer (identifier) IdentiFINDER (Target) with NaI;

• - Survey radiometer-spectrometer PM1401K;

• - Survey identifier (Smartphone) PM1802.

The neutron channel sensitivity to the following nuclear materials is determined:

• - Am-Li sample (neutron source obtained from the irradiated Pu) with neutron radiation of 5⋅104 n/sec during the tests;

• - Cm-244 sample (neutron source) with neutron radiation of 1.6⋅104 n/sec during the tests;

• - Cf-252 sample (neutron source) with neutron radiation of 9⋅104 n/sec during the tests.

The configuration of experiment on neutron channel testing is displayed in Fig. 6.10.

Fig. 6.10. Configuration of neutron measurements.

56

Table 6.4. Testing of neutron channels of the devices.

Measurement results (identification, cm) Measurement conditions Samples to be identified MKS-A03 IdentiFINDER

(Target) PM1401K (Smartphone)

PM1802

Am-Li (neutron source obtained from the irradiated Pu) with intensity of 5⋅104 n/sec

160 15 120 40 45 µSv/h from the surface of the source container

Cm-244 with intensity of 1.6⋅104 n/sec


Cf-252 with intensity of 9⋅104 n/sec


Note:

17) The devices IdentiFINDER (Target), PM1401K and (Smartphone) PM1802 were placed

at the operator’s belt with their face to the source of radiation.

18) Since MKS-A03 is not designed to be attached to the belt it was placed in the operator’s

hand and held at a distance of 20 to 30 cm from his body.

19) The devices PM1401K, (Smartphone) PM1802, and MKS-A03 were moved along the

sources of neutron radiation at a speed of 0.5 m/sec.

20) In order to get the response of the gamma channel from IdentiFINDER (Target) at the

said distances, the device was moved at a reduced speed of 0.3 to 0.4 m/sec.

Am-Li , intensity 5*E4 1/sec

Cm-244 intensity 1.6*E4 1/sec

Cf-252intensity 9*E4 1/sec

МКС-А03РМ1401К

TargetРМ1802

0

20

40

60

80

100

120

140

160

180

200

МКС-А03РМ1401КTargetРМ1802

Fig.6.11. Results of identification of neutron source at various distances.

57

6.3. Additional processing of spectra using software In understanding that the measurements taken at the customs control posts with the help of radiometer-spectrometers yield evaluative and preliminary results needed to take a decision as to whether the identified SIR should be sent for further expert evaluation, the following was assumed to be sufficient:

• The spectrum identification result should be shown as a letter-and-digit abbreviation;

• Complex-isotope spectra should contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively;

• The library of nuclides used for additional processing of spectra should have approximately the same volume, which, however, should exceed that of the libraries embedded in the radiometer-spectrometers for express identification.

• The library of nuclides should be editable;

• The measurement result protocol should be printable;

• The spectrum should be printable.

• The software should identify the spectrum automatically (if the parameter is chosen in the parameter window) when the spectrum is loaded (a desirable option for facilitating the customs FRM control). No additional adjustments should be required. The measurement result should be shown on the screen.

• There should also be an option of manual identification of the nuclide through a menu item (e.g. “identification”).

• The following isotopes were included into the Libraries of Spectra of the tested radiometer-spectrometers as of the time of processing (32 items): Am-241, Ba-133, Co-57, Co-60, Cs-137, Cs-134, Cd-109, Eu-152, Ir-192, Mn-54, Na-22, Np-237, Se-75, Th-228, Th-232,Cr-51, Ga-67, I-123, I-125, I-131, In-111, Tc-99m, Tl-201, Xe-133, K-40, Ra-226, U-238, U-233, U-235, Pu-239, U-RG Pu-239, U-WG Pu-239.

The results obtained in the course of processing the spectra with additional software are summarized in the following tables:

Table 6.5. Additional identification of the spectra of the inspected radioactive material samples in protective containers (SPCs);

Table 6.6. Additional identification of the spectra of the inspected nuclear material samples in protective containers (SPCs);

Table 6.7. Additional identification of the spectra of the inspected radioactive material samples in various combinations.

The left column of the Tables contains summarized details of the measured isotope and measurement parameters. The next column presents the express analysis results obtained in the laboratory of the RRC “Kurchatov Institute” in 2005. Column three specifies the identification results obtained with the help of additional software (this column is in bold type). Positive results obtained with the help of additional processing are highlighted.

The summary Table 6.8 and charts contain summarized results of the measurements.

58

Tab

le 6

.5. A

dditi

onal

iden

tific

atio

n of

the

spec

tra o

f the

insp

ecte

d ra

dioa

ctiv

e m

ater

ial s

ampl

es in

pro

tect

ive

cont

aine

rs

Prel

imin

ary

and

addi

tiona

l Ide

ntifi

catio

n re

sults

(ide

ntifi

ed is

otop

es)

АТ

610

1 МКС

-А03

Id

entiF

IND

ER

(Tar

get)

РМ

1401К

(S

mar

tfon

e) РМ

1802

Sam

ples

for

iden

tific

atio

n

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al (f

ile,

resu

lt)

Prel

imin

ary

Add

ition

al (f

ile,

resu

lt)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Ва-

133

activ

ity

3*10

5 Bq

with

out

shie

ldin

g

Ва-

133

(90

s)

Ва-

133

K-4

0

# 38

Ва-

133

(90

s)

Ba-

133

004

Ва-

133

(100

s)

Not

iden

tifie

d

#002

Ва-

133

(90

s)

# 11

Ba-

133

Eu-

152

Pd-1

03

Hg-

203

Bi-2

10

In-1

11

Tl-2

01

Pu-2

40

Pu-2

42

Th-

231

Ва-

133

(100

s)

# 11

Ba-

133

Pd-1

03

Hg-

203

In-1

11

Tl-2

01

Ва-

133

activ

ity

3*10

5 Bq

in КТ1

-5

(5m

m P

b)

Ва-

133

(40

s)

Ва-

133

K-4

0

# 39

Ва-

133

(40

s)

Ba-

133

0005

Ва-

133

(60

s)

In-1

11

U-2

35

Tl-2

01

# 00

3

Ва-

133

(60

s)

# 12

Pd-1

03

Zr-

95

Ce-

143

Ва-

133

(180

s)

# 12

Pd-1

03

Ba-

133

Ва-

133

activ

ity

3*10

5 Bq

in

КТ1

-10

(10

mm

Pb)

Ва-

133

(180

s)

Ва-

133

K-4

0

# 40

Ва-

133

(180

s)

Ba-

133

0006

Ва-

133

Eu-1

52

(100

s)

In-1

11

U-2

35

Tl-2

01

# 00

4

Ва-

133

(90

s)

# 13

Pd-1

03

Zr-

95

Ва-

133

(240

s)

# 13

Pd-1

03

Ba-

133

Ва-

133

activ

ity

3*10

5 Bq

in

Ва-

133

(240

s)

Ва-

133

K-4

0

Ва-

133

(240

s)

Ba-

133

K-4

0

Not

iden

tifie

d

(300

s)

In-1

11

U-2

35

Ва-

133

(100

s)

# 14

Cs-

137

Ва-

133

(240

s)

# 14

Pd-1

03

59

КТ1

-15

(15

mm

Pb)

# 41

0007

T

l-201

# 00

5

I-13

2

Ag-

110

Pd-1

03

Ce-

143

Ba-

133

Ba-

133

Ва-

133

activ

ity

3*10

5 Bq

in

КТ1

-20

(20

mm

Pb)

К-4

0

(300

s)

Ва-

133

K-4

0

# 42

К-4

0

(300

s)

Ba-

133

K-4

0

0008

Not

iden

tifie

d

(300

s)

In-1

11

U-2

35

Tl-2

01

# 00

6

Ва-

133

Cs-

137

(300

s)

# 15

Pd-1

03

Ba-

133

I-13

2

I-13

1

Cs-

137

Sb-1

27

not i

dent

ified

(300

s)

# 15

Pd-1

03

Ba-

133

I-13

1

Cs-

137

activ

ity

3*10

4 Bq,

with

out

shie

ldin

g

Cs-

137

(30s

)

Cs-

137

# 43

Cs-

137

(30s

)

Cs-

137

0009

Cs-

137

(30s

)

Not

iden

tifie

d

# 00

7

Cs-

137

(30s

)

# 16

Сs-

137

Ac-

228

Ra-

226

Br-

82

Co-

60

I-13

2

Cs-

137

(50s

)

# 16

Сs-

137

U-2

35

Y-9

0

Cs-

137

activ

ity

3*10

4 Bq

in

КТ1

-5

(5 m

m P

b)

Cs-

137

(50s

)

Cs-

137

# 44

Cs-

137

(50s

)

Cs-

137

0010

Cs-

137

(30s

)

Not

iden

tifie

d

# 00

8

Cs-

137

(60s

)

# 17

Cs-

137

Y-9

0

Eu-

152

U-2

35

Te-

129

Pu-2

41

Cs-

137

(60s

)

# 17

Сs-

137

U-2

35

Tl-2

01

60

I-13

2

Ga-

67

Cs-

137

activ

ity

3*10

4 Bq

in

КТ1

-10

(10

mm

Pb)

Cs-

137

(30s

)

Cs-

137

# 45

Cs-

137

(30s

)

Cs-

137

0011

Cs-

137

(60s

)

Not

iden

tifie

d

# 00

9

Cs-

137

(30s

)

# 18

Cs-

137

Y-9

0

Pu-2

41

U-2

35

Te-

129

Sb-1

25

Cs-

137

(30s

)

# 18

Cs-

137

Cs-

137

activ

ity

3*10

4 Bq

in

КТ1

-15

(15

mm

Pb)

Cs-

137

(35s

)

Cs-

137

К-4

0

# 46

Cs-

137

(35s

)

Cs-

137

0012

0013

Cs-

137

(60s

)

Not

iden

tifie

d

# 01

0

Cs-

137

(35s

)

# 19

Cs-

137

Cs-

137

(20s

)

# 19

Сs-

137

U-2

35

Y-9

0

Cs-

137

activ

ity

3*10

4 Bq

in

КТ1

-20

(20

mm

Pb)

Cs-

137

(30s

)

Cs-

137

# 47

Cs-

137

(30s

)

Cs-

137

0014

Cs-

137

(60s

)

Not

iden

tifie

d

# 01

1

Cs-

137

(30s

)

# 20

Cs-

137

Cs-

137

(30s

)

# 20

Сs-

137

I-13

2

Pr-1

44

Eu-1

52 a

ctiv

ity

3*10

5 Bq

with

out

shie

ldin

g

Eu-1

52

(90s

)

Eu-

152

# 48

Eu-1

52

(90s

)

Eu-

152

0017

Eu-1

52

(90s

)

In-1

11

Pu-2

39

Co-

57

Am

-241

# 01

3

Eu-1

52

(40s

)

# 22

Cs-

136

Ra-

226

Ac-

228

Ba-

133

Pb-1

03

Eu-

152

Eu-1

52

(90s

)

# 21

Eu-

152

Cs-

134

Ba-

133

Pd-1

03

U-2

34

Co-

57

Pu-2

39

61

Ce-

144

In-1

11

Xe-

133

Th-

232

Ra-

226

I-13

2

Ac-

228

Pa-2

34

Bi-2

07

Nb-

95

Ag-

110

U-2

38

Eu-1

52 a

ctiv

ity

3*10

5 in КТ1

-5

(5 m

m P

b)

Eu-1

52

(90s

)

Eu-

152

# 49

Eu-1

52

(90s

)

Eu-

152

0018

Not

iden

tifie

d

(300

s)

Not

iden

tifie

d

# 01

6

Zn-6

5

(40

s)

# 23

Ag-

110

Cs-

137

I-13

2

Ac-

228

Sb-1

27

Not

iden

tifie

d

(300

s)

# 22

Cs-

136

Ba-

133

Pd-1

03

I-13

1

Ac-

228

Pa-2

34

Ra-

226

Eu-1

52 a

ctiv

ity

3*10

5 in КТ1

-10

(10

mm

Pb)

Not

iden

tifie

d

(300

s)

Eu-

152

# 50

Not

iden

tifie

d

(300

s)

Eu-

152

0019

Not

iden

tifie

d

(300

s)

Not

iden

tifie

d

# 01

7

Not

iden

tifie

d

(300

s)

# 24

Ra-

226

Ba-

133

Pd-1

03

I-13

1

Ga-

67

Not

iden

tifie

d

(300

s)

# 23

Br-

82

Nb-

95

Cs-

136

K-4

0

Cs-

134

62

U-2

35

Eu-

152

Eu-

152

I-13

2

Na-

24

Co-

60 a

ctiv

ity

3*10

5 Bq

with

out

shie

ldin

g

Co-

60

(40s

)

Co-

60

# 53

Co-

60

(40s

)

Co-

60

0022

Co-

60

(60s

)

# 01

8

Cs-

137

Th-

232

Bi-2

07

Co-

60

(50s

)

# 26

Co-

60

Co-

60

(30s

)

# 26

Co-

60

Co-

60 a

ctiv

ity

3*10

5 Bq

in КТ1

-5

(5 m

m P

b)

Co-

60

(25s

)

Co-

60

# 54

Co-

60

(25s

)

Co-

60

0023

Co-

60

(60s

)

# 01

9

Cs-

137

Th-

232

Bi-2

07

Co-

60

(25s

)

# 27

Co-

60

Co-

60

(20s

)

# 27

Co-

60

Co-

60 a

ctiv

ity

3*10

5 Bq

in

КТ1

-10

(10

mm

Pb)

Co-

60

(30s

)

Co-

60

# 55

Co-

60

(30s

)

Co-

60

0024

Co-

60

(60s

)

# 02

0

Cs-

137

Th-

232

Bi-2

07

Co-

60

(30s

)

# 28

Co-

60

Co-

60

(30s

)

# 28

Co-

60

Co-

60 a

ctiv

ity

3*10

5 Bq

in

KТ1

-15

(15

mm

Pb)

Co-

60

(40s

)

Co-

60

# 56

Co-

60

(40s

)

Co-

60

0025

Co-

60

(60s

)

# 02

1

Cs-

137

Th-

232

Bi-2

07

Co-

60

(30s

)

# 29

Со-

60

Xe-

133

Sr-8

5

Pb-2

12

Th-

232

In-1

11

Sb-1

27

Co-

60

(50s

)

# 29

Co-

60

Co-

60 a

ctiv

ity

3*10

5 Bq

in

Co-

60

(30s

)

Co-

60

# 57

Co-

60

(30s

)

Co-

60

Co-

60

(60s

)

# 02

2

Cs-

137

Co-

60

(30s

)

# 30

Co-

60

Co-

60

(40s

)

# 30

Co-

60

63

KТ1

-20

(20

mm

Pb)

0026

T

h-23

2

Bi-2

07

Ra-

226

activ

ity

2*10

5 Bq

with

out

shie

ldin

g

Ra-

226

(30

s)

Ra-

226

# 51

Ra-

226

(30

s)

Ra-

226

Ba-

133

0020

Ra-

226

(60

s)

#014

Co-

57

Ce-

139

Am

-241

Bi-2

07

Ra-

226

(25

s)

# 31

not i

dent

ified

Ra-

226

(22

s)

# 24

Ra-

226

I-13

1

Pd-1

03

Ba-

133

Pm-1

48

Eu-

152

Ra-

226

activ

ity

2*10

5 Bq

in

КТ1

-5

(5 m

m P

b)

Ra-

226

(240

s)

Ra-

226

# 52

Ra-

226

(240

s)

Ra-

226

0021

Ra-

226

(120

s)

# 01

5

Co-

57

Ce-

139

Am

-241

Bi-2

07

Ra-

226

(180

s)

# 32

Tl-2

08

V-5

0

La-

140

Ra-

226

(120

s)

# 25

Pm-1

48

Cs-

134

Y-8

8

Ag-

110

Br-

82

Ra-

226

Th-2

32 a

ctiv

ity

3*10

3 Bq

with

out

shie

ldin

g

Th-2

32

(60

s)

Th-

232

# 58

Th-2

32

(60

s)

Th-

232

0027

Th-2

32

(60

s)

# 02

3

Co-

57

Ce-

139

Co-

60

Be-

7

Th-2

28

(60

s)

# 33

U-2

35

Th-

232

Tl-2

08

La-

140

Mn-

54

I-13

2

Cs-

137

Bi-2

07

Sb-1

22

Th-2

32

Th-2

28

(120

s)

# 31

Ac-

228

La-

140

Th-

232

Tl-2

08

Xe-

133

Pb-2

12

Te-

132

Ba-

133

Pd-1

03

64

Ag-

110

Ra-

226

Eu-

152

Th-2

32 a

ctiv

ity

3*10

3 Bq

in

КТ1

-5

(5 m

m P

b)

Th-2

32

(120

s)

Th-

232

# 59

Th-2

32

(120

s)

Th-

232

0028

Th-2

32

(120

s)

# 02

4

Co-

57

Ce-

139

Am

-241

Bi-2

07

Not

iden

tifie

d

(300

s)

# 34

Tl-2

08

Th-

232

U-2

35

Mn-

54

Sb-1

22

Ru-

103

Rh-

106

Kr-

85

La-

140

Br-

82

Ra-

226

(120

s)

# 32

Tl-2

08

V-5

0

Th-

232

Cs-

134

La-

140

Ra-

226

Br-

82

K-4

0 ac

tivity

5*1

02

Bq

with

out

shie

ldin

g

K-4

0

(120

s)

K-4

0

# 88

K-4

0

(120

s)

K-4

0

0058

K-4

0

(120

s)

# 05

1

not i

dent

ified

.

K-4

0

Cs-

137

(300

s)

# 82

Th-

234

Am

-241

K-4

0

Pu-2

40

Pu-2

42

U-2

34

U-2

38

K-4

0

(300

s)

# 59

K-4

0

Co-

60

65

Tab

le 6

.6. A

dditi

onal

iden

tific

atio

n of

the

spec

tra o

f the

insp

ecte

d nu

clea

r mat

eria

l sam

ples

in p

rote

ctiv

e co

ntai

ners

Pr

elim

inar

y an

d ad

ditio

nal I

dent

ifica

tion

resu

lts (i

dent

ified

isot

opes

) Sa

mpl

es fo

r

iden

tific

atio

n АТ

610

1 МКС

-А03

Id

entiF

IND

ER

(Tar

get)

РМ

1401К

(S

mar

tfon

) РМ

1802

Pr

elim

inar

y A

dditi

onal

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

U-2

35 –

90%

Sam

ple

# 1,

3 g

With

out

shie

ldin

g

U-2

35,

wea

pon-

grad

e,

(23

s)

Th-

228

U-2

35 (9

0%)

# 60

U-2

35,

spec

ial,

(180

s)

U-2

35 (9

0%)

U-2

38

0030

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 02

5

Cd-

109

U-2

35,

wea

pon-

grad

e,

(25

s)

# 58

U-2

35

U-2

35,

wea

pon-

grad

e,

(120

s)

# 33

Tl-2

01

Th-

231

Th-

234

Cd-

109

U-2

35

Pu-2

42

Th-

232

Tl-2

08

Eu-

152

K-4

0

Pu-2

40

Bi-2

07

Co-

60

Ag-

110

Am

-241

Ra-

226

Xe-

133

66

U-2

35 –

90%

Sam

ple

# 1,

3 g

Shie

ldin

g 1m

m

Cu

U-2

35,

wea

pon-

grad

e,

(25

s)

Th-

228

U-2

35 (9

0%)

# 61

U-2

35,

spec

ial,

(40

s)

U-2

35 (6

0.9%

)U

-238

Th-

232

0031

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 02

6

Cd-

109

U-2

35,

wea

pon-

grad

e,

(30

s)

# 59

U-2

35

Am

-241

Th-

234

Pu-2

40

Pu-2

42

U-2

34

U-2

35,

wea

pon-

grad

e,

(60

s)

# 34

Th-

231

Th-

234

Tl-2

01

U-2

35

Pu-2

42

Th-

232

Cd-

109

Tl-2

08

Eu-

152

K-4

0

Pu-2

40

Am

-241

Ra-

226

Bi-2

07

U-2

35 –

90%

Sam

ple

# 1,

3 g

Shie

ldin

g 2

mm

Cu

U-2

35,

wea

pon-

grad

e,

(23

s)

Th-

228

U-2

35 (9

0%)

# 62

U-2

35,

spec

ial,

(40

s)

U-2

35 (6

0.4%

)

Th-

238

U-2

38

0032

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 02

7

Cd-

109

U-2

35,

wea

pon-

grad

e,

(20

s)

# 60

U-2

35

Not

iden

tifie

d

(300

s)

# 35

Th-

231

Am

-241

Th-

234

U-2

35

Pu-2

42

Pu-2

40

Cd-

109

Tl-2

01

67

U-2

35 –

90%

Sam

ple

# 1,

3 g

Shie

ldin

g

1.5

mm

Fe

U-2

35,

wea

pon-

grad

e,

(27

s)

Th-

228

U-2

35 (9

0%)

# 63

U-2

35,

spec

ial,

(40

s)

U-2

35 (9

7%)

U-2

38

0033

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 02

8

Cd-

109

U-2

35,

wea

pon-

grad

e,

(30

s)

# 61

U-2

35

Not

iden

tifie

d

(300

s)

# 36

Th-

231

Th-

234

Pu-2

42

Am

-241

Th-

234

U-2

35

Pu-2

40

Cd-

109

Tl-2

01

Cs-

134

Ra-

226

U-2

35 –

88.

9%

Sam

ple

# 2,

10 g

Shie

ldin

g

1.5

mm

Fe

U-2

35,

wea

pon-

grad

e,

(15

s)

Th-

228

U-2

35 (9

0%)

# 64

U-2

35,

spec

ial,

(40

s)

U-2

35 (9

5.9%

) T

h-23

2

0034

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 02

9

Cd-

109

U-2

35,

wea

pon-

grad

e,

(30

s)

# 62

U-2

35

U-2

35,

wea

pon-

grad

e,

(25

s)

# 37

Tl-2

08

Th-

231

La-

140

Th-

234

Pu-2

42

Am

-241

Th-

234

U-2

35

Pu-2

40

Cd-

109

Tl-2

01

Cs-

134

Ra-

226

68

Xe-

133

U-2

35 –

88.

9%

Sam

ple

# 2,

10 g

Shie

ldin

g 5

mm

Pb

Th-2

28,

(160

s)

Th-

228

# 65

U-2

35,

spec

ial,

Co-

57

(300

s)

U-2

35

Th-

232

0036

Not

iden

tifie

d

(300

s)

# 03

0

Co-

60

K-4

0

Not

iden

tifie

d

(300

s)

# 63

Cs-

134

Th-

232

Tl-2

08

Sb-1

27

I-13

2

Cs-

137

Ag-

110

Ra-

226

Br-

82

Not

iden

tifie

d

(300

s)

# 38

Th-

232

Tl-2

08

Bi-2

07

Kr-

85

Sr-8

5

Ru-

103

Ba-

140

Sb-1

22

La-

140

Cs-

134

Ra-

226

U-2

35 –

88.

9%

Sam

ple

# 2,

10 g

. Shi

eldi

ng

5 m

m P

b, 1

mm

Cu

Th-2

28,

(133

s)

Th-

228

# 66

U-2

35,

spec

ial,

K-4

0

(300

s)

U-2

35

Th-

232

0037

Bi-2

07

(100

s)

# 03

1

Co-

60

K-4

0

I-13

1

Not

iden

tifie

d

(300

s)

# 64

Th-

234

Pu-2

40

Pu-2

42

Cs-

134

Th-

232

Tl-2

08

Sb-1

27

I-13

2

Not

iden

tifie

d

(300

s)

# 39

Cs-

134

Pa-2

34

U-2

38

Tl-2

01

Pu-2

42

Pu-2

40

Th-

232

Tl-2

08

69

Cs-

137

Ag-

110

Ra-

226

U-2

38

Bi-2

07

Kr-

85

Sr-8

5

Ru-

103

Ba-

140

Sb-1

22

La-

140

Cs-

134

Ra-

226

Ac-

228

U-2

35 –

47%

Sam

ple

# 3,

158

g

Shie

ldin

g 3

mm

Fe

U-2

38, n

at.

Th-2

28,

indu

stria

l

(30

s)

Th-

228

U-2

35 (3

7%)

# 67

U-2

35,

spec

ial

(60

s)

K-4

0

U-2

35 (7

.1%

)

Ra-

226

0038

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

2

Cd-

109

Th-2

32,

indu

stria

l

(60

s)

# 65

Pa-2

34

Cs-

137

Y-9

0

U-2

35

Pu-2

41

Ga-

67

Not

iden

tifie

d

(300

s)

# 40

Co-

60

Pu-2

40

Pu-2

42

Tl-2

01

Na-

24

Th-

234

Th-

234

Ra-

226

Am

-241

Eu-

152

U-2

35 –

47%

Sam

ple

# 3,

158

g

Shie

ldin

g 3

mm

Fe, 1

mm

Cu

U-2

38, n

at.

Th-2

28,

indu

stria

l

(47

s)

Th-

228

U-2

35 (3

7%)

# 68

U-2

35,

spec

ial

(60

s)

U-2

35 (6

5.7%

)

0039

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

3

Cd-

109

Th-2

32,

indu

stria

l

(60

s)

# 66

Th-

232

Tl-2

08

Ac-

228

Bi-2

07

Th-2

32,

Th-2

28,

indu

stria

l

(60

s)

# 41

Tl-2

01

Pu-2

42

Pu-2

40

Th-

234

70

Th-

231

Pa-2

34

U-2

38

Th-

232

Tl-2

08

Ra-

226

Cd-

109

Eu-

152

Ac-

228

Rh-

106

Co-

60

U-2

35 –

21%

Sam

ple

# 4,

158

g

Shie

ldin

g 3

mm

Fe

U-2

38,

U-2

35

(30

s)

Th-

228

U-2

35 (1

8%)

# 69

U-2

38,

U-2

35

(60

s)

U-2

35

0040

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

4

Cd-

109

Pu-2

40,

(240

s)

# 67

Th-

232

Tl-2

08

Bi-2

07

Sb-1

22

Ag-

110

Pa-2

34

U-2

38

U-2

38,

(240

s)

# 42

Pu-2

42

U-2

38

Pa-2

34

Pu-2

40

Tl-2

01

Th-

234

Th-

231

Cd-

109

Am

-241

U-2

35 –

21%

Sam

ple

# 4,

158

g

Shie

ldin

g 3

mm

Fe, 1

mm

Cu

U-2

38,

U-2

35

(30

s)

Th-

228

U-2

35 (1

7%)

# 70

U-2

38,

U-2

35

(120

s)

U-2

38

U-2

35 (5

5%)

0041

U w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

5

Cd-

109

U-2

35,

U-2

38

(70

s)

# 68

U-2

35

Ac-

228

U-2

38

(100

s)

# 43

Pu-2

42

Pu-2

40

U-2

38

Pa-2

34

71

La-

140

Tl-2

01

Th-

234

Na-

24

Th-

231

Am

-241

Eu-

152

Co-

60

U-2

35 –

0.7

1%

Sam

ple

#5,

173

g R

eact

or-

type

(с C

s-13

7)

Shie

ldin

g 3

mm

Fe

Cs-

137,

U-2

38,

(40

s)

Сs-

137

U-2

35

(less

then

1%)

# 71

# 89

Cs-

137,

U-2

38,

(50

s)

U-2

35

0042

Cs-

137,

U w

ithou

t

isot

ope

desi

gnat

ion,

(60

s)

# 03

6

Cr-

51

Be-

7

Th-

232

Cs-

137,

U-2

38,

(200

s)

# 70

U-2

35

Pa-2

34

U-2

38

Cs-

137,

U-2

38,

(150

s)

# 44

Pa-2

34

Ga-

67

Pu-2

41

Th-

231

Tl-2

01

Pu-2

42

Pu-2

40

U-2

35

Am

-241

Pu-2

39

U-2

33

U-2

38

Co-

57

I-13

2

Eu-

152

72

U-2

35 –

4.4

%

reac

tor-

type

,

Sam

ple

# 6,

79 g

Shi

eldi

ng

5 m

m F

e

Am

-241

,

Pu-2

39

(80

s)

Pu-2

39

U-R

G

#74

Cs-

137,

U-2

38,

U-2

35

(120

s)

U-2

35

U-2

38

0060

Cs-

137,

(300

s)

# 05

3

Cr-

51

Cs-

137,

(240

s)

# 69

U-2

35

Zr-

95

Nb-

95

Pa-2

34

U-2

38

Mo-

99

I-13

2

Br-

82

Cs-

137,

(240

s)

# 61

Cs-

137

Pa-2

34

Pu-2

40

Pu-2

42

Pu-2

41

Am

-241

U-2

34

U-2

33

U-2

35

Co-

57

U-2

38

Th-

234

Th-

231

Pu-2

39

Cd-

109

Tl-2

01

Eu-

152

Ac-

228

Pu-2

39 –

96.

7%

Sam

ple

# 7,

0.3

g

With

out

shie

ldin

g

Pu-2

39

(45

s)

Pu-2

39

U-R

G

#72

Pu-2

39

WG

,spec

ial

Am

-241

(60

s)

Pu-2

39

0043

Pu w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

7

Pu-2

39

U-2

35

Pu-2

39

wea

pon-

grad

e

(180

s)

# 71

I-13

1

I-13

2

Pd-1

03

Ba-

133

Cs-

137

Y-9

0

Pu-2

39

wea

pon-

grad

e

(280

s)

# 45

Th-

234

Pu-2

42

Cd-

109

Pd-1

03

Ba-

133

U-2

33

73

Pu-2

41

U-2

35

Mo-

99

Te-

129

Ir-1

92

Th-

231

Am

-241

I-13

1

Pu-2

40

Pu-2

39 –

96.

7%

Sam

ple

#7,

0.3

g

Shie

ldin

g

1.5

mm

Cu

Pu-2

39

(45

s)

Pu-2

39

U-R

G

#73

Pu-2

39

WG

,spec

ial

Am

-241

(60

s)

Pu-2

39

0044

Pu w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

8

Pu-2

39

U-2

35

Cr-

51

Cd-

109

Pu-2

39

wea

pon-

grad

e

(180

s)

# 72

Zr-

95

Nb-

95

I-13

2

Pd-1

03

U-2

33

Ir-1

52

Ba-

133

Y-9

0

Pu-2

41

Am

-241

Mo-

99

Eu-

152

Sn-1

13

(60

s)

# 46

Pu-2

42

Th-

234

Cd-

109

Na-

24

Eu-

152

Ba-

133

Pd-1

03

I-13

1

Th-

231

Co-

60

U-2

33

Ir-1

52

Am

-241

Pu-2

40

Pu-2

39 –

96.

7%

Sam

ple

#7,

0.3

g Sh

ield

ing

1.5

mm

Cu,

1 m

m F

e

Pu-2

39

К-4

0

(260

s)

Pu-2

39

U-R

G

#75

Pu-2

39

WG

,spec

ial

(60

s)

Pu-2

39

Pu-2

41

0045

Pu w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 03

9

Zn-

65

Cs-

137,

(150

s)

# 73

I-13

1

I-13

2

Pu-2

40

Pu-2

42

Ba-

133

Sn-1

13

Со-

57

(60

s)

# 47

Pu-2

40

Pu-2

42

Am

-241

Th-

234

U-2

34

74

Cs-

137

U-2

33

Ag-

110

Pd-1

03

Ir-1

92

Sb-1

27

Th-

231

Pu-2

39 –

96.

7%

Sam

ple

#7,

0.3

g

Shie

ldin

g 5

mm

Pb

Not

iden

tifie

d

(300

s)

К-4

0

Pu-2

39

U-R

G

#76

Pu-2

39

WG

,spec

ial

(120

s)

Pu-2

39

0046

Not

iden

tifie

d

(300

s)

# 04

0

Pu-2

39

U-2

35

Ir-1

92

I-13

1

Not

iden

tifie

d

(300

s)

# 74

Th-

232

Tl-2

08

Ce-

144

Mo-

99

Tc-

99

Pu-2

39

I-13

2

Cs-

137

U-2

35

Ag-

110

Pd-1

03

Eu-

152

Sb-1

27

Sn-1

13

(150

s)

# 48

Eu-

152

Ra-

226

Cs-

134

Sb-1

25

Pm-1

48

In-1

11

Tl-2

01

U-2

35

Th-

232

Tl-2

08

Br-

82

Ga-

67

Pu-2

39 –

96.

7%

Sam

ple

#7,

0.3

g

Shie

ldin

g 5

mm

Pb

Not

iden

tifie

d

(300

s)

# 77

not i

dent

ified

Pu-2

39

WG

,spec

ial

(120

s)

Pu-2

39

0047

Not

iden

tifie

d

(300

s)

# 04

1

not i

dent

ified

Not

iden

tifie

d

(300

s)

# 57

Sb-1

27

Te-

129

I-13

2

Y-9

0

Pu-2

41

Sn-1

13

(150

s)

# 49

Eu-

152

Br-

82

Pa-2

34

Cs-

134

Sb-1

25

75

Am

-241

Pm

-148

In-1

11

Tl-2

01

U-2

34

Th-

232

Tl-2

08

Pu-2

39

La-

140

Pu-2

39 –

96.

7%

Sam

ple

#7,

0.3

g Sh

ield

ing

10 m

m P

b

Am

-241

,

Pu-2

39

nucl

ear,

(40

s)

Pu-2

39

U-R

G

# 81

K-4

0,

Pu-2

39

WG

,spec

ial

(300

s)

Pu-2

39

Pu-2

38

Pu-2

40

Pu-2

41

0049

Not

iden

tifie

d

(300

s)

# 04

2

not i

dent

ified

Not

iden

tifie

d

(300

s)

# 50

Sb-1

27

I-13

2

Pr-1

44

Te-

129

Not

iden

tifie

d

(300

s)

# 50

not i

dent

ified

Pu-2

39 –

90%

Sam

ple

# 8,

5.9

g

With

out

shie

ldin

g

Not

iden

tifie

d

(300

s)

Pu-2

39

U-R

G

# 82

Pu-2

39

WG

,spec

ial

Am

-241

(60

s)

Pu-2

39

Am

-241

0051

Am

-241

,

Pu w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 04

5

U-2

35

Pu-2

39

I-12

5

Not

iden

tifie

d

(300

s)

# 75

Am

-241

Th-

231

Th-

234

U-2

34

Co-

57

Pu-2

40

Pu-2

42

Pu-2

39

I-13

2

Cs-

137

I-13

1

Sb-1

27

Sn-1

13

(150

s)

# 53

Th-

234

Pu-2

42

Pu-2

40

Cd-

109

Na-

24

Pu-2

41

Y-9

0

I-13

1

Ba-

133

Th-

231

Am

-241

Eu-

152

76

Ag-

110

Co-

60

Pu-2

39 –

90%

Sam

ple

# 8,

5.9

g

in КТ1

-5

(5 m

m P

b)

Not

iden

tifie

d

(300

s)

Not

iden

tifie

d

# 83

Pu-2

39

WG

,spec

ial

(60

s)

Pu-2

39

0052

Not

iden

tifie

d

(300

s)

# 04

6

U-2

35

Pu-2

39

Cr-

51

Pu-2

39

Cs-

137,

(300

s)

# 76

I-13

1

I-13

2

Cs-

137

Not

iden

tifie

d

(300

s)

# 54

In-1

11

Tl-2

01

U-2

35

Ga-

67

I-13

1

Pu-2

39 –

90%

Sam

ple

#8,

5.9

g

in КТ1

-10

(10

mm

Pb)

Cs-

137,

Cr-

51

(120

s)

Not

iden

tifie

d

# 84

Pu-2

39

WG

,spec

ial

(120

s)

Pu-2

39

0053

Not

iden

tifie

d

(300

s)

# 04

7

U-2

35

Pu-2

39

Zn-

65

Cs-

137,

(300

s)

# 77

I-13

1

Ba-

133

I-13

2

Cs-

137

Pd-1

03

Ag-

110

Not

iden

tifie

d

(300

s)

# 56

Th-

231

Am

-241

Ce-

143

Cs-

137

I-13

2

Ag-

110

Pd-1

03

Sb-1

27

Ba-

133

Th-

234

Pu-2

40

Pu-2

42

U-2

33

U-2

34

Co-

57

Pu-2

39

77

I-13

1

Sb-1

25

Pu-2

39 –

90%

Sam

ple

# 8,

5.9

g in

КТ1

-15

(15

mm

Pb)

Pu-2

39

WG

,spec

ial

Pu-2

39

reac

tor-

type

,

Am

-241

(300

s)

U-W

G

Pu-2

39

# 93

Pu-2

39

WG

,spec

ial

K-4

0

(240

s)

N

ot id

entif

ied

(300

s)

# 04

8

U-2

35

Ce-

139

Tc-

99m

Cs-

137,

(300

s)

# 78

K-4

0

Sb-1

27

I-13

2

Cs-

137

Not

iden

tifie

d

(300

s)

# 51

not i

dent

ified

.

Pu-2

39

irrad

iate

d ,

reac

tor-

type

Sam

ple

# 9,

60 g

Shie

ldin

g 5

mm

Fe

U-2

35,

wea

pon-

grad

e,

(23

s)

Th-

228

U-2

35 (9

0%)

# 60

Pu-2

39

WG

,spec

ial

Pu-2

39

reac

tor-

type

,

Am

-241

(300

s)

N

ot id

entif

ied

(300

s)

# 05

5

U-2

33

Th-

232

Cs-

137,

(300

s)

# 54

Th-

234

Pu-2

40

Pu-2

42

Pd-1

03

Am

-241

Cs-

137,

(300

s)

# 63

Cs-

137

Ir-1

92

Be-

7

Y-9

0

U-2

35

In-1

11

Pu-2

41

Sb-1

25

Ru-

103

78

Tab

le 6

.7. A

dditi

onal

iden

tific

atio

n of

the

spec

tra o

f the

insp

ecte

d nu

clea

r mat

eria

l sam

ples

in v

ario

us c

ombi

natio

ns.

Prel

imin

ary

and

addi

tiona

l Ide

ntifi

catio

n re

sults

(ide

ntifi

ed is

otop

es)

АТ

610

1 МКС

-А03

Id

entiF

IND

ER

(Tar

get)

РМ

1401К

(S

mar

tfon

) РМ

1802

Sam

ples

for

iden

tific

atio

n

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Prel

imin

ary

Add

ition

al

(file

, res

ult)

Pu-2

39 –

96.

7%

Sam

ple

# 7,

0.3

g

U-2

35-9

0%,

Sam

ple

# 1,

3 g

With

out

shie

ldin

g

Pu-2

39 W

G,

U-2

35

(30

s)

U-R

G P

u-

239

U-2

35 (9

0%)

Th-

228

# 78

Pu-2

39 W

G,

U-2

35

(60

s)

Pu-2

39

U-2

35

0050

U, w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 04

4

Pu-2

39

Cd-

109

Ga-

67

U-2

38,

U-2

35

(120

s)

# 53

U-2

35

I-13

1

Pd-1

03

Ba-

133

Sn-1

13,

U-2

35

(120

s)

# 52

Th-

231

Th-

234

Pu-2

42

Tl-2

01

Pu-2

40

Cd-

109

U-2

35

Th-

232

Tl-2

08

Cs-

134

Ra-

226

Bi-2

07

Am

-241

Br-

82

Pu-2

39 –

96.

7%

Sam

ple

# 7,

0.3

g

Ва-

133

activ

ity

3*10

5 Bq

Ba-

133

(30

s)

U-P

u-23

9

Ва-

133

# 85

Pu-2

39 W

G,

Am

-241

,

Ba-

133

(60

s)

Pu-2

39

Ba-

133

0054

Ba-

133,

Pu- w

ithou

t

isot

ope

desi

gnat

ion

(60

s)

# 04

8

U-2

35

Ce-

139

Tc-

99m

Ba-

133

(120

s)

# 80

Pd-1

03

Sb-1

27

I-13

2

Ce-

143

Ba-

133

(120

s)

# 57

Th-

231

Th-

234

Cd-

109

Pd-1

03

79

With

out

shie

ldin

g

Cs-

137

Pu-2

42

Pu-2

40

I-13

1

U-2

33

Am

-241

Ba-

133

Pu-2

39 –

96.

7%

Sam

ple

# 7,

0.3

g

Ва-

133

activ

ity

3*10

5 Bq

Shie

ldin

g 1

mm

Fe

Ba-

133

(30

s)

Ва-

133

# 86

Ba-

133,

Pu-2

39 W

G,

Am

-241

,

(180

s)

Pu-2

39

Pu-2

40

Ba-

133

0055

Ba-

133

(60

s)

# 04

9

U-2

35

Ce-

139

Not

iden

tifie

d

(300

s)

# 81

Pd-1

03

Сd-

109

Pu-2

42

Th-

231

Sb-1

27

I-13

2

Ce-

143

Cs-

137

Not

iden

tifie

d

(300

s)

# 58

Pd-1

03

I-13

1

Ba-

133

Pu-2

39 –

96.

7%

Sam

ple

#7, 0

.3 g

Ва-

133

activ

ity

3*10

5 Bq

Shie

ldin

g 2

mm

Fe

Ba-

133

(60

s)

Ва-

133

# 87

Ba-

133,

Pu-2

39 W

G,

Am

-241

,

(120

s)

Pu-2

40

Pu-2

39

Ba-

133

0057

Ba-

133

(60

s)

# 05

0

U-2

35

Ce-

139

Ba-

133

(120

s)

# 83

U-2

38

Pa-2

34

U-2

35

Zr-

95

I-13

2

Pu-2

41

Ba-

133

(120

s)

# 61

not i

dent

ified

U-2

35 –

4.4

%,

rege

nera

ted

with

Cs-

137

Sam

ple

#6, 7

9 g

Cs-

137,

U-2

38,

U-2

35

(30

s)

U-2

35 (6

%)

Cs-

137

# 90

Cs-

137,

U-2

38,

U-2

35

(120

s)

U-2

38

Cs-

137

U-2

35 (7

%)

Cs-

137,

(300

s)

# 05

3

Cr-

51

Cs-

137,

(300

s)

# 84

Cs-

137

Eu-

152

U-2

38

Cs-

137,

(300

s)

# 62

Сs-

137

Ir-1

92

Be-

7

80

Shie

ldin

g 5

mm

Fe

0060

Pa

-234

Te-

129

Sb-1

25

I-13

2

Ra-

226

Tl-2

01

Ru-

103

Tab

le 6

.8. S

umm

ary

of th

e re

sults

on

expr

ess a

nd a

dditi

onal

iden

tific

atio

n.

R

esul

ts o

f exp

ress

iden

tific

atio

n / r

esul

ts o

n ad

ditio

nal i

dent

ifica

tion

(usi

ng so

ftw

are)

Sour

ces

MK

S-A

03

PM14

01K

(S

mar

tfon

) РМ

1802

Iden

tiFIN

DE

R

(Tar

get)

Id –

100

0 G

R-1

35

АТ

610

1

Rad

ioac

tive

mat

eria

ls(2

3 sa

mpl

es)

21/2

3 19

/10

19/9

19

/0

9/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

8/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

21/2

3

Nuc

lear

mat

eria

ls

(23

sam

ples

) 24

/21

10/9

5/

2 14

(with

out i

soto

pic

com

posi

tion)

/6

(onl

y Pu

-239

)

9/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

10/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

14/1

8

Com

bina

tions

of

radi

oact

ive

and

nucl

ear m

ater

ials

(2

3 sa

mpl

es)

5/5

0/0

0/2

1/0

1/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

2/

addi

tiona

l ide

ntifi

catio

n

was

not

car

ried

out

2/3

81

6.3. Summary on capabilities of the devices for identification of nuclear and radioactive materials

6.3.1. Spectrometer InSpector 1000

The InSpector 1000 sample provided for testing was subject to periodical additional calibration. Considering the customs control procedures in field conditions, long adjustment procedures considerably affect the use of the device. User interface is quite complicated.

In the “express analysis” mode the device was able to identify:

• 4 out of 6 radioactive sources offered for testing;

• 5 out of 6 uranium product samples offered;

• None of the plutonium product samples offered.

The device is not equipped with a neutron channel.

The software is sufficiently comprehensive with a high level of detail as to the SPC modeling elements. The program weakness includes difficulties in working with it, which may impede its use during the customs FRM control.

6.3.2. Spectrometer MKS – AT 6101

The MKS – AT6101 sample provided for testing is capable of energy calibration using the K-40 source, which is not covered by the Radiation Safety Regulations due to its activity level. This is a considerable advantage of the device in terms of operation convenience and simplicity when identifying radioactive cargo in field conditions. The user interface is simple and easy to use.


• All the radioactive sources offered for testing (it also showed one of the best results when these isotopes were screened by protective materials);


• 2 out of 3 plutonium product samples offered.

The device is not equipped with a neutron channel.

- The spectrum identification result is shown as a letter-and-digit abbreviation; - Complex-isotope spectra contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively; - The library of nuclides is editable; - The measurement result protocol is printable; The spectrum is printable. The convergence rate for the results obtained in the course of the express analysis and the analysis using additional software is very high and is one of the best ones. Considering the fact that this radiometer-spectrometer showed one of the best express analysis results, especially for the complex-isotope spectra, and the best result in additional processing of complex-isotope spectra with the help of software indicating the U-235 and Pu-239 enrichment, this radiometer-spectrometer meets the requirements of the customs FRM control technology in the full amount. Its additional benefits include maximum simplicity in printing out the measurement results and user-friendly interface.

6.3.3. Spectrometer MKS-A03

The MKS-A03 sample provided for testing is equipped with the Th-232 calibration gamma-source located on a special platform, therefore the device automatically performs additional

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calibration of its spectrometric channel when in a “stand-by” mode. The device was not subject to additional calibration during the tests. This is a considerable advantage of the device in terms of operation convenience and simplicity when identifying radioactive cargo in field conditions. The user interface is simple and easy to use.


• All the radioactive sources offered for testing (it also showed one of the best results when these isotopes were screened by protective materials);

• 6 out of 6 uranium product samples offered (in the “express analysis” mode the device quite correctly distinguished between the low enriched and highly enriched uranium samples, which is unique for this class of equipment);

• 3 out of 3 plutonium product samples offered, which is unique for this class of equipment.

• Being equipped with a neutron channel MKS-A03 showed the best result in identifying nuclear materials based on their neutron radiation.

- The spectrum identification result is shown as a letter-and-digit abbreviation; - Complex-isotope spectra contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively; - The library of nuclides is editable; - The measurement result protocol is printable; The spectrum is printable. The convergence rate for the results obtained in the course of the express analysis and the analysis using additional software is very high and is one of the best ones. Considering the fact that this radiometer-spectrometer showed the best express analysis result especially for the complex-isotope spectra and mixtures, this radiometer-spectrometer meets the requirements of the customs FRM control technology in the full amount. There is a minor weakness related to some difficulties in printing out the measurement results.

6.3.4. Spectrometer IdentiFINDER – Ultra (Target)

The IdentiFINDER – Ultra (Target) sample provided for testing is capable of energy calibration using a LED-generated photo peak, therefore the device was not subject to additional calibration during the tests. This is a considerable advantage of the device in terms of operation convenience and simplicity when identifying radioactive cargo in field conditions. The user interface is simple and easy to use.


• The basic number of radioactive sources offered for testing (6 out of 6 were identified; only non-protected Eu-152 sample could be pre-identified);


• 2 out of 3 plutonium product samples offered;

• IdentiFINDER – Ultra (Target) showed the worst result in identifying nuclear materials based on their neutron radiation.

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- The spectrum identification result is shown as a letter-and-digit abbreviation; - Complex-isotope spectra do not contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively; - The library of nuclides is editable; - The measurement result protocol is not printable; The spectrum is not printable. In spite of high express analysis results shown by the identifier, especially for radioactive materials the use of additional software did not provide for additional information about the spectra measured.

6.3.5. Spectrometer Exploranium GR-135

The GR-135 sample provided for testing is equipped with the Cs-137 calibration gamma-source located on a special platform. The device was subject to additional calibration during the tests. User interface is rather complicated.


• Only 2 out of 6 radioactive sources offered for testing;


• 2 out of 3 plutonium product samples offered;

• The device is not equipped with a neutron channel.

It was not possible to open the saved files with a .txt extension containing information about the spectra measured.

6.3.6. Spectrometer MKS-PM1401K

The PM1401K sample provided for testing is equipped with the Cs-137 calibration gamma-source located on a special platform, therefore the device was not subject to additional calibration during the tests. This is a considerable advantage of the device in terms of operation convenience and simplicity when identifying radioactive cargo in field conditions. The user interface is simple and easy to use.



• 6 out of 6 uranium product samples offered (in the “express analysis” mode the device quite correctly distinguished between the low enriched and highly enriched uranium samples, which is unique for this class of equipment);

• Only 1 out of 3 plutonium product samples offered.

As it is equipped with a neutron channel PM1401K showed one the best results in identifying nuclear materials by their neutron radiation.

- The spectrum identification result is shown as a letter-and-digit abbreviation; - Complex-isotope spectra contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively; - The library of nuclides is editable; - The measurement result protocol is printable; The spectrum is printable. The convergence rate for the results obtained in the course of the express analysis and the analysis using additional software is low. The results obtained in the course of additional processing of complex-isotope spectra with the help of software contain information as to the U-235 and Pu-239 enrichment; however, it is not always correct. A list of identified isotopes is

84

shown in the measurement protocol. The list often contained 10-20 items, which cannot be described as a correct identification. Most of the isotopes are within triplets and multiplets irresolvable by the software used. This radiometer-spectrometer does not meet the requirements of the customs FRM control technology in the full amount. Its additional benefits include maximum simplicity in printing out the measurement results and a user-friendly interface.

6.3.7. Spectrometer Smartphone PM1802

The Smartphone PM1802 sample provided for testing was not equipped with the Cs-137 calibration gamma-source. The device was not subject to additional calibration during the tests since the producer’s representative recommended not carrying out such work but using the calibration performed at the factory instead.




• None of the 3 plutonium product samples offered.

PM1802 showed one the worst results in identifying nuclear materials based on their neutron radiation.

- The spectrum identification result is shown as a letter-and-digit abbreviation; - Complex-isotope spectra contain information as to the U-235 enrichment for U spectra and the Pu-239 enrichment for Pu spectra, respectively; - The library of nuclides is editable; - The measurement result protocol is printable; The spectrum is printable. The convergence rate for the results obtained in the course of the express analysis and the analysis using additional software is extremely low. The device showed poor results in the express analysis of the spectra obtained. The results obtained in the course of additional processing of complex-isotope spectra with the help of software contain information as to the U-235 and Pu-239 enrichment; however, it is not always correct. A list of identified isotopes is shown in the measurement protocol. The list often contained 10-20 items, which cannot be described as a correct identification. Most of the isotopes are within triplets and multiplets irresolvable by the software used. The software used in the device is similar to that used in the Survey radiometer-spectrometer PM1401K. This device does not meet the requirements of the customs FRM control technology in the full amount. Its additional benefits include maximum simplicity in printing out the measurement results and a user-friendly interface.

CONCLUSIONS

The information on the most frequently shipped radioactive sources and nuclear materials and design of standard shipping containers for the transportation of nuclear materials and radioactive substances was systemized.

The capabilities of spectrometers (identifiers) on transport index confirmation were verified. All tested devices are capable to solve a problem of transport index confirmation

The spectrometric analysis of most frequently shipped isotopes was carried out. 3. The spectrometer capabilities were verified in terms of the definition of isotopic composition of gamma nuclides in their various combinations, and isotopic composition of radioactive substances and nuclear materials placed in various protective containers.

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For the decision of the problem of radioisotope identification, it is expedient to develop devices with extremely simplified procedure of measurements and identification and friendly user interface. Information about each saved spectrum is stored in a special file. Each software developer currently uses its own format for presenting the information, which extremely impedes adaptation and unification of the measured spectra obtained with the help of different radiometer-spectrometers. Each spectrum file contains the necessary calibrations as to the energy and resolution of the spectrometer and each ADC channel stores a fixed number of pulses; yet the information is presented in quite different forms and the files have different extensions. In order to unify the customs FRM control procedure related to the spectra identification, the exchange of spectrometric information (including that with customs services in other countries) in a single format seems crucial.

The software should become as understandable and user-friendly as possible to be used in the customs FRM control technologies, considering that measurements are taken almost in field conditions and within a minimum period of time. The software product should move from a survey instrument category into a practical one.

In preparing a library of nuclides for resolving the tasks of the customs FRM control, the spectrum is to be modeled for each nuclide both without any protection and with its possible distortion by the protection. Each nuclide spectrum should be described with account taken of the lines adjusted to their weakening by the protection, the total absorption peaks, the backscatter peaks, the leakage peaks, the Compton edges, etc.

The results obtained in the course of this work as well as positive results in the field of customs FRM control identified in such radiometer-spectrometers as АТ-6101 and MKS-А03 (using additional software to obtain additional supportive information as to the measured spectra) are recommended to be used by IAEA to standardize the spectrometric measurement procedures.

It is advisable to start developing the relevant IAEA Recommendations and to continue the work related to further improvement of this category of spectrometric equipment and transfer of the existing software into a more practical field of use.

Verification of the Design Information of Shipment Containers Using Gamma-Spectrometry

S. Korneyev Joint Institute for Power and Nuclear Research, NAS, Belarus, Minsk

A. Khilmanovich, B. Martsynkevich

B.I. Stepanov Institute of Physics, NAS, Belarus, Minsk

E. Bystrov ATOMTEX Enterprise, Belarus, Minsk

Abstract

The possibility of using the Compton parts in a spectrum to determine the thickness of

possible shielding materials was analyzed. Measurements were made, using 137Cs and

60Co sources, that showed that it is possible to do this using data from either NaI(Tl)

or HPGe detectors. The accuracy of the method was analyzed, and found to be

promising. An algorithm based upon this work was developed, and inserted into

commercially-available software. At present time the correctness of software has been

tested for lead, steel and a tungsten shields with thicknesses from 3 to 50 mm for

152Eu, 137Cs, 60Co, 133Ba, 88Y, and 22Na. This developed software can be used in

practice by customs services and can be easily implemented for routine procedures.

1

INTRODUCTION As usual for storage and transportation of radioactive sources and nuclear materials

are used containers made of materials with high Z for effective shielding of the gamma-radiation. Declared value of activity and isotope content of radioactive source transported in the container can be controlled by means of measurements of an external gamma-ray field on a surface of the transport container, but for definition of real activity of transported radioactive material it is necessary to know precisely thickness of a wall of the container and a material of which the container is made. In this work we study of possibilities to extract information about container design by analyzing of gamma radiation of transported source inside. Because of transformation of the original spectra of a source of radiation by shielding material there is opportunity to measure container thickness using shape of pulse-height spectra measured outside a container.

In the 1st year of the project, we used Monte Carlo simulation simultaneously with

experiments to assess possible approaches for the solving of formulated task. For this 1st year, data from a HPGe detector and from a NaI(Tl) detector were compared to calculations and Monte Carlo Simulation. In the 2nd year of the project, we, in cooperation with staff from the company Atomtex, implemented these results using NaI hardware and commercially-available software.

Year 1 activities 1. Experimental and calculation method 1.1. Experiment setup

The geometry used for the initial calculations and experiments is shown in a Fig. 1

Fig. 1. Experiment layout.

Measurements were carried out in geometry when the detector was on distance of

50 and 100 mm from a radioactive source which was placed behind lead shield with thickness which varied up to 50 mm. Such variations in thickness of shield have been chosen from a reason of that these thicknesses are characteristic for typical designs of containers intended for transportation low activity radioactive sources. In experiments standard gamma ray point type sources manufactured by VNIIM Russia were used.

2

Intensity of a source 137Cs was 6.63·104 gammas per second and 60Cs 4.57·104 gammas per second.

Measurements were carried out with HPGe and NaI(Tl) detectors with the following characteristics: energy resolution of detector NaI(Tl) manufactured by Bicron with the size of the crystal 2"x2" is 7 % on 662 keV; energy resolution of semi-conductor detector GC2018 manufactured by Canberra is 1.8 keV on 1332 keV and efficiency of registration of 20 % relative to NaI(Tl) 3"x3" on 1332 keV on 25 cm distance. For the analysis of spectra portable analyzer InSpector fully controlled by PC was used. The measured spectra were analyzed by Genie-2000 software [1–2]. If necessary, additional processing of pulse-height spectra (smoothing, subtraction of a background) was made. Parameters of energy calibration and parameters of peak Gaussian broadening were used at processing and as input data at Monte Carlo simulation. Acquired data was stored in 4096 channels for HPGe detector and 512 channels for NaI(Tl) detector [3–4].

1.2. Monte-Carlo simulation

General-purpose Monte Carlo computer code MCNP [5] was used for simulation of

pulse-height spectra. So-called pulse height tally is built in MCNP code for modeling of detector response and this possibility was used in simulation process. The pulse height tally is analogous to a physical detector and represents distribution of the absorbed energy in spectrometer channels due to physical processes of radiation interaction with detector material. Measurements of pulse-height spectra of radioactive sources 241Am, 57Co, 133Ba, 137Cs, 60Co for different distances was performed for creation and test of adequate mathematical models of detectors. Then were calculations with the purpose of adjustment of parameters of mathematical models to bring to conformity of experimental and modeled parameters. Schematic drawings of used models are presented in a Figs 2–3. Well known, that the data of manufacturers of semi-conductor detectors on their sizes are not absolutely exact and the detailed model demands the account of its geometrical form, the data on depth of an internal aperture, thickness of a dead layer besides influence of heterogeneity of an electric field inside the detector is essential that results in non-uniformity in sensitivity volume of the detector. The account of these data is very important at carrying out of mathematical calibrations of the detector for absolute measurements [6–7]. The model were adjusted to provide difference between experimental efficiencies and calculated within the limits of 5%. Much more attention paid on as possible to satisfy to the requirement of similarity of the shape of the calculated and measured spectra.

Fig. 2. Sectional view of HPGe detector model used in Monte Carlo calculations.

3

Fig. 3. Sectional view of NaI(Tl) detector model used in Monte Carlo calculations.

In Tables 1 and 2 are presented measured and Monte Carlo simulated efficiencies. It is necessary to pay attention, that the consent of calculation and experiment for the semi-conductor detector in the chosen range is better, this fact can be explained more developed mathematical model and more simple procedure of the analysis of the pulse-height spectra.

Table 1. Experimental and calculated full-energy peak efficiency for HPGe detector, 20% efficiency relative to NaI(Tl)

10 cm 20 cm E, keV exp ,%σ calc ,%σ exp ,%σ calc ,%σ

59 3.81 310−⋅ 2.1 4.15 310−⋅ 1< 1.06 310−⋅ 2.2 1.14 310−⋅ 1<

122 8.22 310−⋅ 9.8 8.51 310−⋅ 1< 2.31 310−⋅ 9.9 2.38 310−⋅ 1<

276 5.24 310−⋅ 2.2 4.94 310−⋅ 1< 1.56 310−⋅ 2.4 1.51 310−⋅ 1<

303 4.68 310−⋅ 2.0 4.48 310−⋅ 1< 1.40 310−⋅ 2.1 1.37 310−⋅ 1<

356 4.01 310−⋅ 2.0 3.96 310−⋅ 1< 1.21 310−⋅ 2.0 1.17 310−⋅ 1<

384 3.71 310−⋅ 2.2 3.48 310−⋅ 1< 1.15 310−⋅ 2.4 1.08 310−⋅ 1<

662 2.23 310−⋅ 2.1 2.15 310−⋅ 1< 6.40 410−⋅ 2.2 6.47 410−⋅ 1<

1174 1.28 310−⋅ 2.1 1.31 310−⋅ 1< 3.97 410−⋅ 2.2 3.99 410−⋅ 1<

1332 1.14 310−⋅ 2.1 1.19 310−⋅ 1< 3.51 410−⋅ 2.2 3.62 410−⋅ 1<

4

Table 2. Experimental and calculated full-energy peak efficiency for NaI(Tl) 2"x2" detector

10 cm 20 cm E, keV exp ,%σ calc ,%σ exp ,%σ calc ,%σ

59 1.14 210−⋅ 2.2 1.13 210−⋅ 1< 3.02 310−⋅ 2.7 3.15 310−⋅ 1<

122 — — 1.19 210−⋅ 1< — — 3.37 310−⋅ 1<

662 3.41 310−⋅ 2.1 3.42 310−⋅ 1< 1.13 310−⋅ 2.2 1.06 310−⋅ 1<

1174 1.53 310−⋅ 2.2 1.88 310−⋅ 1< 4.69 410−⋅ 2.3 5.81 410−⋅ 1<

1332 1.46 310−⋅ 2.2 1.67 310−⋅ 1< 4.45 410−⋅ 2.3 5.24 310−⋅ 1< A number of calculations with shielded and unshielded sources of gamma rays with

wide energy range have been performed to show adequacy of the description of the shape of calculated pulse height spectra to measured ones. Monte Carlo simulated pulse-height spectra in comparison with experiment are presented on Figs 4–11. Comparison of the simulated and measured spectra shows, that model calculation does not allow to describe the shape of tails of spectra from zero up to energy close to Compton edge in case of detector NaI(Tl) and in soft area of energies in case HPGe. The principal cause of distinctions will be that multiple scattered gamma rays appear in real experiment. The account of this can be carried out by detailed modeling of structure of a design of the detector and detailed reproduction of conditions of experiment. For the problems of the research there is no such necessity. Moreover, this result specifies undesirability of use of a low energy tail of a spectrum because its shape is very sensitive and can vary depending on measurement conditions.

Fig. 4. A comparison of the measured NaI(Tl) spectrum for gamma-rays from 137Cs with Monte

Carlo simulated spectrum. Lead Shield Thickness: 10 mm.

Fig. 5. A comparison of the measured NaI(Tl) spectrum for gamma-rays from 60Co with Monte


5

Fig. 6. A comparison of the measured NaI(Tl) spectrum for gamma-rays from 137Cs with Monte


Fig. 7. A comparison of the measured NaI(Tl) spectrum for gamma-rays from 60Co with Monte


Fig. 8. A comparison of the measured HPGe spectrum for gamma-rays from 137Cs with Monte


Fig. 9. A comparison of the measured HPGe spectrum for gamma-rays from 60Co with Monte


Fig. 10. A comparison of the measured HPGe spectrum for gamma-rays from 137Cs with Monte


Fig. 11. A comparison of the measured HPGe spectrum for gamma-rays from 60Co with Monte


6

2. Extracting information about shield thickness

2.1. Studied approaches The possibility of use of the Compton part of a spectrum for extraction of the

information on thickness of the container on measurements of a radiation from a source placed inside the container is investigated. In general the problem will consist in a choice of values on which change it is possible to judge change of thickness. The possibility of use of the ratio of areas taken from different windows of a spectrum and use of only part of Compton tail has been investigated. This work is based on study pulse-height spectra from 137Cs and 60Сo sources shielded by lead.

The assumption of a possibility of use of a Compton part of a pulse-height spectrum is based on fact, that the shape of a registered spectrum is influenced by gamma rays undergoes Compton scattering in a material of shield. In a Fig. 12 are presented of Monte Carlo calculated scattered photon fluxes on surface of lead shield for different thicknesses of shield for 662 keV source.

Fig. 12. Monte Carlo calculated scattered photon fluxes on surface of lead shield after

transition of different thicknesses from 662 keV source.

The analysis of these distributions shows that the energy averaged on a spectrum

shape of scattered gamma rays increases with increase in thickness of shield. The ratio scattered and nonscattered gamma rays also increases. Table 3. Energies of gamma rays averaged on spectrum shape versus shield thickness

H, mm E, keV 5 489 10 504 15 512 20 517 25 520 30 523

7

Thus, with increase of shield thickness high-energy gamma rays scattered in shield make a contribution to measured pulse-height spectrum simultaneously with nonscattered gamma rays reached detector. It results the deformation of spectra within Compton continuum. It allows, in general, explaining presence of effects studied in the work. For an illustration of this statement in a Fig. 13 response function of detector NaI(Tl) are presented for various thickness of shield normalized on a maximum of full-absorption peak of 137Cs source.

Fig. 13. Normalized on maximum in full-absorption peak pulse-height spectra of 137Cs

for different thicknesses of lead shield. NaI(Tl) detector.

2.2. Using Peak to part of Compton tail ratio method From the theory of interaction of radiation with matter is known that Compton

scattering gamma rays in a material of the detector results formation of Compton tail. This arises due to incomplete transfer of energy of gamma rays to detector material. This Compton tail on the right is limited to energy determined by equation

221 ⋅= −

+CE Emc

EE E

γ

γγ , (1)

where Eγ - energy of emitted gamma rays and CEE - Compton edge energy.

Table 4. Calculated energies of Compton edges

Eγ , keV CEE , keV662 478 1174 964 1332 1118

General procedure of the analysis of spectra is consist in a choice of two windows to

calculate ratio of the count sums of pulse-height spectra. In Table 5 the sizes of windows in keV for used types of detectors and used gamma-ray sources are presented.

8

Table 5. Width of windows in keV used for peak/Compton ratio analysis

137Cs 60Co I II I II

NaI(Tl) 440–530 600–720 850–1000 1100–1440 Ge 440–478 660–665 850–964 (1172–1177) + (1329–1334) The choice of width of windows can be made arbitrary in some extent, but we were

guided by the following criteria: 1. Do not use a low-energy part of a tail in view of the significant contribution scattered gamma-rays. 2. The right border is defined by Compton edge corresponding to energy of radioactive source. In a case 60Co for HPGe the detector as border energy corresponding to a line with smaller energy 1174 keV is chosen. In case of detector NaI(Tl) the account of Gaussian broadening of a spectrum results in a choice of the right border different from exact value.

Generally procedure should be reduced to summation of the areas of all peaks (region II) and then this sum it is necessary to divide on Compton component (region I). Ratio peak/Compton should be beforehand known for the specific detector and a radioactive source stored in the container. Generally it is possible to calculate with program MCNP pulse-height spectra from sources with separate energies for various thicknesses and then make summation according to emission probabilities. In case of discreet set of spectra it is necessary develop interpolation procedure to obtain spectra corresponding exact value of energy. Described procedure is some kind of mathematical calibration. Using precalculated spectra seems the most preferable and flexible. It is enough to set a radioactive material to receive a required set of peak/Compton ratios for various thicknesses.

In Figs 16–17 comparisons experimental results with Monte Carlo simulations are shown. The received satisfactory agreement of experiment with calculation has allowed us to use of calculated calibration values of peak/Compton dependent of thickness of shield.

Fig. 14. Ratio of counts in the full-energy peak to part of Compton tail in pulse-height spectrum of

137Cs as function of lead shield thickness for NaI(Tl) detector.

Fig. 15. Ratio of counts in the sum of full-energy peaks to part of Compton tail in pulse-height spectrum of 60Co as function of lead shield

thickness for NaI(Tl) detector.

9

In a Figs 16–17 peak/Compton ratios for NaI(Tl) detector are presented. The received dependences are approximated exponential function. The received dependences are calibration characteristics of the detector for measurement of shield thickness.

Fig. 16. Lead shield thickness versus ratio of counts in the full-energy peak to part of Compton tail in pulse-height spectrum of 137Cs for NaI(Tl)

detector.

Fig. 17. Lead shield thickness versus ratio of counts in the full-energy peaks to part of Compton

tail in pulse-height spectrum of 60Co for NaI(Tl) detector.

Similar measurements and calculations have been executed for detector HPGe.

Results of comparison of experiments and calculations are resulted in a Figs 18–19.


137Cs as function of lead shield thickness for HPGe detector.


60Co as function of lead shield thickness for HPGe detector.

In a Figs 20–21 are presented calibration curves for HPGe detector.

10

Fig. 20. Lead shield thickness versus ratio of counts in the full-energy peak to part of Compton tail in pulse-height spectrum of 137Cs for HPGe

detector.

Fig. 21. Lead shield thickness versus ratio of counts in the full-energy peaks to part of Compton

tail in pulse-height spectrum of 60Co for HPGe detector.

On Fig. 22 shown result of zero extrapolation of ratio of counts in the full-energy

peak to part of Compton tail in pulse-height spectrum of 137Cs for HPGe detector. Result obtained allows estimating of minimum of upper limit of thickness of lead shield which can be determined using described method. Generally speaking this is very rough estimation because real dependence is not linear. The general analysis allows asserting, that with increase in thickness of shield the accuracy decreases in view of that the steepness of curve dependence grows.

Fig. 22. Extrapolated lead shield thickness versus ratio of counts in the full-energy peak

to part of Compton tail in pulse-height spectrum of 137Cs for HPGe detector. Estimation of accuracy of the suggested method can be performed by setting initial

conditions. Assume that precision in measurement of the peak/Compton ratio is 2%. Having analytical expressions of dependence of peak/Compton ratio from thickness it is possible to receive easily analytical dependence of accuracy of measurement versus of measured thickness. In a Figs 23–24 dependences for detector NaI(Tl) are shown. The

11

behavior of these curves completely depends on the calculated peak/Compton ratio and can be considered as estimation.

Fig. 23. Estimated standard deviation in determining of shield thickness with 137Cs source

inside container. NaI(Tl) detector.

Fig. 24. Estimated standard deviation in determining of shield thickness with 60Co source

inside container. NaI(Tl) detector.

It is important to estimate time of measurement necessary for achievement of precision of 2%. Having the calculated response functions and setting value of source intensity 3.7·105 cps it is possible to calculate required time. In a Fig. 25 dependences of necessary time of measurement versus of thickness of shield are shown.

From obtained distributions is clear advantage of NaI(Tl) detector. In such measurements efficiency of registration is more important factor for maintenance of comprehensible accuracy in comparison with the energy resolution of the detector.

Fig. 25. Measurement time versus lead shield thickness for reaching 2% precision in determining of ratio of counts in the full-energy peak to part of Compton tail in pulse-

height spectrum of 3.7·105 Bq 137Cs and 60Co inside container.

12

2.3. Method based on analysis Compton part of pulse-height spectra only

In this method for the analysis the part of a spectrum corresponding Compton tail is used by itself. This is preferable approach from our point of view. Powerful methods developed in the theory of recognition of images can be used in this approach. In our case the problem of classification can be formulated as comparison of the chosen quantitative characteristics of the measured distributions with measured or calculated beforehand for known thickness of shield.

The analysis of this region allows asserting, that the shape of a spectrum depends on thickness of a wall of container. In general case it would interest to find of such transformation which convert the space of spectra into space of attributes. Thickness is such attribute in considered problem. In a Figs 26–35 the simulated spectra for HPGe and NaI(Tl) are presented depending on thickness of lead shield. The analysis of these distributions allows asserting, that the form of a spectrum strongly depends on thickness of shield. The general approach will consist in search of a matrix of transformation R which satisfies to equation

SRT ⋅= , (2) where T – identity matrix with dimensions NxN equal to number of discrete values of thickness of shield, S – a matrix made of vectors of the spectra corresponding to discrete thickness with dimension MxN, where M – number of channels of a spectrum. Dimension of a matrix R is NxM. Having a set of spectra it is possible to calculate value of a matrix from matrix equation

1−⋅= STR . (3) Thus the task is reduced to task of finding of Moore-Penrose inverse of a matrix with

columns representing of spectra for various thickness of shields. It is important that dimensions of matrix satisfy condition M > N. At the solving of this problem it is necessary to be limited to reasonable number of channels in a spectrum, i.e. the analyzed spectrum should be convoluted till the reasonable sizes. We are considered 50 groups in interesting energy range. Also it is necessary to scale a spectrum in limits from 0 up to 1. The calculations have shown suitability of such approach.

In work we used method based on calculation of area under normalized Compton tail. On Figs 27–28, 30-31, 32, 34 presented spectra normalized on a maximum of Compton peak and full-absorption peak. Such normalization allows transforming a spectrum to such units when dependence on distance between detector and source disappears and there is only a dependence on thickness of shield. The area of normalized Compton part of a tail depends on thickness of shield thickness.

13

Fig. 26. Monte Carlo simulated pulse-height spectra of 60Co for different thicknesses of

lead shield. NaI(Tl) detector.

Fig. 27. Fragment of Monte Carlo simulated pulse-height spectra of 60Co normalized on maximum in

Compton tail for different thicknesses of lead shield. NaI(Tl) detector.

Fig. 28. Fragment of Monte Carlo simulated pulse-height spectra of 60Co normalized on full-energy

peak for different thicknesses of lead shield. NaI(Tl) detector.

Fig. 29. Monte Carlo simulated pulse-height spectra of 137Co for different thicknesses of

lead shield. NaI(Tl) detector.

14

Fig. 30. Fragment of Monte Carlo simulated pulse-height spectra of 137Cs normalized on maximum in

Compton tail for different thicknesses of lead shield. NaI(Tl) detector.

Fig. 31. Fragment of Monte Carlo simulated pulse-height spectra of 137Cs normalized on maximum in full-energy peak for different thicknesses of lead

shield. NaI(Tl) detector.

Fig. 32. Fragment of Monte Carlo simulated pulse-height spectra of 60Co normalized on maximum in 964 keV Compton peak for different thicknesses of

lead shield. HPGe detector.

Fig. 33. Fragment of Monte Carlo simulated pulse-height spectra of 60Co for different thicknesses of


15

Fig. 34. Fragment of Monte Carlo simulated pulse-height spectra of 137Cs normalized on maximum in Compton peak (478 keV) for different thicknesses

of lead shield. HPGe detector.

Fig. 35. Fragment of Monte Carlo simulated pulse-height spectra of 137Co for different thicknesses of


In Table 6 the sizes of windows in keV for various types of detectors and radioactive sources are presented. Table 6. Size of windows (keV) used for Compton part of spectra analysis

137Cs 60Co NaI(Tl) 440-530 850-1000

Ge 450-650 750-1150

On Figs 36–39 normalized areas are plotted.

Fig. 36. Area under Compton part in pulse-height spectrum of 137Cs normalized on height of full-

energy peak versus lead shield thickness for NaI(Tl) detector.

Fig. 37. Area under Compton part in pulse-height spectrum of 60Co normalized on height of full-energy peak (1174 keV) versus lead shield

thickness for NaI(Tl) detector.

16

Fig. 38. Area under Compton part in pulse-height spectrum of 137Cs normalized on height of full-energy peak versus lead shield thickness for

HPGe detector.

Fig. 39. Area under Compton part in pulse-height spectrum of 60Co normalized on height of full-energy peak (1174 keV) versus lead shield

thickness for HPGe detector. The resulted calculations are in good agreement with experimental results. In a

Figs 40–41 are shown calibration characteristics for HPGe detector. Values of the calculated calibration values for HPGe detector are shown in table 7.

Fig. 40. Lead shield thickness versus area under Compton part in pulse-height spectrum of 137Cs

normalized on height of full-energy peak for HPGe detector.

Fig. 41. Lead shield thickness versus area under Compton part in pulse-height spectrum of 60Co

normalized on height of full-energy peak (1174 keV) for HPGe detector.

17

Table 7. Calibration values for determining shield thickness with HPGe detector using area of Compton part of pulse-height spectrum normalized on height of full-energy peak

60Co 137Cs Thickness,

mm Area, norm.

counts SD Area, norm. counts SD

0.0 16.99 0.38 2.13 0.06 2.5 17.98 0.40 2.49 0.06 5.0 19.00 0.44 2.82 0.07 7.5 20.09 0.49 3.13 0.09 10.0 21.09 0.53 3.43 0.10 12.5 22.12 0.59 3.70 0.12 15.0 23.24 0.65 3.96 0.14 17.5 24.21 0.72 4.23 0.17 20.0 25.26 0.80 4.49 0.20 22.5 26.44 0.88 4.73 0.23 25.0 27.08 0.96 4.97 0.27 27.5 27.79 1.05 5.18 0.32 30.0 28.27 1.14 5.43 0.38 32.5 29.08 1.25 5.68 0.45 35.0 29.71 1.36 5.88 0.53 37.5 30.37 1.49 6.04 0.62 40.0 31.04 1.62 6.21 0.73 42.5 31.70 1.77 6.45 0.86 45.0 32.31 1.93 47.5 32.96 2.11 50.0 33.54 2.27

Year 1 Conclusion

In work is analyzed the opportunity of use Compton parts in a spectrum measured

by the solid-state detectors for determining of thickness of shield by measuring of pulse- height spectra of a sample placed in the container. It is approved, that this problem can be solved with use either NaI(Tl) or HPGe detectors. The simple technique of measurement of thickness with 137Cs and 60Co sources is offered. The accuracy of the method was analyzed. In such formulation proposed method will work in case of other radioactive sources. Proposed method can be effective used in practice of customs services and it can be easily implemented as the software for wide use. The given work is preliminary research and is intended to show an opportunity of extraction of the information on a design of the container on the registered pulse-height spectrum of radioactive source placed inside a container. The suggested approach has been tested on sets of measured and Monte Carlo simulated spectra from radioactive samples are placed behind lead shield of various thicknesses. We shall note, that developing methods of such kind will allow supervising non-authorized transportation of sources.

18

Year 2 Activities The goal of the continuation project was implementation of the developed approach for concrete realization as the software and hardware to estimate of transportation container wall thickness and source activity. This phase of the project were carried in cooperation with ATOMTEX enterprise, Minsk, Belarus (producer of radiometric and gamma-spectrometry equipment) [8]. An algorithm, allowing analyzing spectrums of wide range of shielded radio nuclides with the purpose of their quantitative analysis, is suggested

3. Experimental studies and Monte Carlo simulation for Year 2 Activities

As the basic device for implementation of a method for determination of activity of a radioactive source located inside of the transport container had been took the identifier of radionuclides made by Atomtex enterprise AT-6101 with detector NaI (Tl). Energy resolution of detector NaI(Tl) with the size of the crystal 4x4 is <9 % on 662 keV. The measured and calculated spectra were analyzed by ATAS software. Additional processing of pulse-height spectra (smoothing, subtraction of a background) was made. Parameters of energy calibration and parameters of peak Gaussian broadening were used at processing and as input data at Monte Carlo simulation. Acquired data was stored in 512 channels.

Fig 42. Radionuclide identifier AT-6101. ATOMTEX enterprise offered detailed information on a design of the detecting

module with NaI(Tl) detector. On Fig. 43 section view of the detector is shown.

19

Fig. 43. Sectional view of NaI(Tl) detector. General-purpose Monte Carlo computer code MCNP was used for simulation of

pulse-height spectra. MCNP input file for calculation of necessary calibration data was created. Measurements of pulse-height spectra of wide range of radioactive sources for different distances was performed for creation and test of adequate mathematical model of detector. The model was adjusted to provide difference between experimental efficiencies and calculated within the limits of 5%.

For example comparison of measured and simulated pulse height spectra from 152Eu source is shown on Fig. 44.

20

Fig. 44. Comparison of measured spectra and simulated pulse height tally from

152Eu source. The good agreement of calculation with experiment has allowed performing of

necessary calculations for the spectrometer calibration. Computer Monte Carlo simulations of the detector response functions was

performed for Pb-shield thicknesses varied from 0 up to 50 mm with step of 5 mm. Energy diapason from 50 to 2000 keV was chosen. Used thicknesses are typical for transport containers designed for radioactive sources transportation. Measurements and calculations were carried out in geometry when the detector was on distance of 50 mm from a radioactive source which was placed behind lead shield. Standard gamma-ray point type sources manufactured by VNIIM Russia were used for testing purposes. Calculated response functions are input data for developed software for quantitative measurements of activity and determination of characteristics of shield.

Note that calculated response functions make possible to obtain pulse height spectra for any shielded radioactive source. Simulation of response functions can be considered as mathematical calibration procedure. Simulated response functions where used for calculating all necessary calibration values. The using of precalculated spectra seems the preferable and flexible approach in many assessment tasks.

On the basis of the simulated response functions for various energies and thickness of Pb-shield the table of full-energy-peak efficiency was obtained (Table 8). Calculated data is used in the software for measuring of source activity located in transportation container.

21

Table 8. Dependence of full-energy-peak efficiency as function of energy and Pb shield thickness

Pb shield thickness, cm Energy, keV 0 1 2 3 4 5

50 1.68E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 100 2.41E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 200 2.05E-02 4.34E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 300 1.36E-02 2.26E-04 3.35E-06 4.59E-08 0.00E+00 0.00E+00 400 8.87E-03 9.15E-04 8.43E-05 7.73E-06 6.37E-07 3.75E-08 500 6.56E-03 1.37E-03 2.70E-04 5.22E-05 9.80E-06 1.87E-06 600 5.18E-03 1.57E-03 4.57E-04 1.31E-04 3.57E-05 1.03E-05 700 4.28E-03 1.59E-03 5.75E-04 2.06E-04 7.32E-05 2.56E-05 800 3.63E-03 1.53E-03 6.32E-04 2.60E-04 1.05E-04 4.28E-05 900 3.16E-03 1.46E-03 6.68E-04 3.04E-04 1.35E-04 6.07E-05

1000 2.79E-03 1.38E-03 6.72E-04 3.25E-04 1.57E-04 7.53E-05 1100 2.51E-03 1.31E-03 6.70E-04 3.44E-04 1.76E-04 8.93E-05 1200 2.27E-03 1.23E-03 6.56E-04 3.52E-04 1.86E-04 9.92E-05 1300 2.08E-03 1.15E-03 6.34E-04 3.52E-04 1.93E-04 1.06E-04 1400 1.92E-03 1.09E-03 6.15E-04 3.50E-04 1.96E-04 1.11E-04 1500 1.78E-03 1.04E-03 5.95E-04 3.45E-04 1.97E-04 1.14E-04 1600 1.66E-03 9.70E-04 5.67E-04 3.31E-04 1.93E-04 1.12E-04 1700 1.56E-03 9.22E-04 5.45E-04 3.23E-04 1.91E-04 1.14E-04 1800 1.46E-03 8.75E-04 5.22E-04 3.13E-04 1.88E-04 1.13E-04 1900 1.38E-03 8.31E-04 5.01E-04 3.02E-04 1.83E-04 1.11E-04 2000 1.30E-03 7.91E-04 4.80E-04 2.91E-04 1.77E-04 1.08E-04 On Fig. 45 energy dependence of full-energy-peak efficiencies for different Pb-shield

thickness are shown.

Fig. 45. Full-energy-peak efficiencies for different Pb-shield thickness.

4. Determination of shield thickness The Table 8. contains the information necessary for determination of thickness of a

container wall in case of the radioactive sources emitting of more than one gamma ray

22

with energies enough far apart. On ratio of full-energy-peak areas corrected by branching ratio fraction to find container wall thickness according to equation

)()(

)()(

2

1

12

21

γ

γ

γγ

γγ

εε

EE

IESIES

R ≡⋅

⋅= (4)

where )( γES - full-energy-peak area; γI - branching ratio fraction; )( γε E - full-energy-peak efficiency.

Exact values of efficiencies for the certain energies are gained by interpolation of table data. On Fig. 46 the ratio of the calculated efficiencies for 1000 and 500 keV depending on thickness of a wall of the container is shown. In a semi-log scale it is can be approximated by a straight line.

Fig. 46. Ratio of the full-energy-peak efficiencies for energies 1000 and 500 keV versus

Pb-shield thickness. For analyzing radionuclide’s with single or very close gamma-ray lines in their

emitting spectra approach developed on the first stage of project was utilized.

5. Accounting the distance from source to detector

Calculations of efficiency have been made only for one fixed distance. Nevertheless, these data it is enough to use the calculated values for any distances between a radiation source and the detector. For this purpose we shall take advantage of the concept of effective centre of the detector according to which there is a point inside of the detector for which efficiency of registration of the detector varies in inverse proportion to a square of distance between a radioactive source and this point. As definition this is point located inside the detector with inverse square of distance to source dependence of photo-efficiency.

23

For transition to other distance the formula is used

2

2

)()'()',(),(

hdhddEdE

++

⋅= γγ εε (5)

where ),( dEγε - full-energy-peak efficiency for energy γE and distance source from

detector d ; h – distance from detector face to effective detector centre; )',( dEγε –calculated value; 'd =5 cm.

MCNP simulation has shown relevancy of such approach in correction of full-energy-peak efficiency for various distances. On Fig. 47 the experimental dependence of effective detector centre on energy of gamma-rays is presented.

Fig. 47. Energy dependence of effective detector centre.

6. Software algorithm Obtained data have allowed creating algorithm implemented as addition to ATAS

software allowing to analyze spectrums of gamma rays of radioactive sources located inside of the transport container.

General procedure of operation of algorithm consists in following steps. Before the beginning of measuring it is enough to input the information on distance from a radioactive source to the detector.

24

Fig. 48. Input and output results window.

As a result of the analysis of a spectrum identification of radioactive source located inside of the container is made. Thickness of a wall of the container is determined from the analysis of a spectrum. There is an individual algorithm for particular radionuclide. Additional data necessary for analysis are contained in library of nuclides of ATAS software.

Fig. 49. Nuclide library window.

25

In case of the container is made of a material different from lead software find thickness of lead with the same attenuation ability as well as a wall of the real container. The equivalent thickness of lead gained thus allows choosing from the Table 1 proper value for full-energy-peak efficiency.

On Fig. 50 shown result of measurement. On the MCA interface we can see identified radioactive source, estimated activity and equivalent to lead shielding thickness.

.

Fig. 50. MCA window. 7. Year 2 Conclusions For implementation of the developed algorithm we inserted additional module to the ATAS software which is used by ATOMTEX enterprise for control of electronics and processing of gamma-ray spectra in identifiers of radionuclide’s. Measurements and Monte Carlo simulation of a basis set of pulse-height data for various gamma sources and various shielding compositions and thicknesses was performed.

As input data the information on the declared source activity, placed inside the container, thickness of a container wall and distance from an entrance window of the detector to a measured radionuclide is required. The accounting of distance from source to detector is performed with using of the effective detector centre concept. Thickness of a wall of the container calculated from the analysis of the spectrum transformed by shield in terms of equivalent lead thickness. Thus for determining of activity we need only information about distance from source to detector. Using of thickness entered manually by operator for measurement of activity is provided. At present time the correctness of software was tested for lead, steel and a tungsten shields with thicknesses from 3 to 50 mm for 152Eu, 137Cs, 60Co, 133Ba, 88Y, and 22Na.

Energy dependent attenuation approach was implemented for analysis of wide set of radioactive sources. The accuracy of the method was analyzed.

Software has been tested on sets of measured and Monte Carlo simulated spectra from radioactive samples are placed behind shields of different compositions and thicknesses.

26

8. Project Summary The possibility of using the Compton parts in a spectrum to determine the thickness of possible shielding materials was analyzed. Measurements were made, using 137Cs and 60Co sources, showed that it is possible to do this using data from either NaI(Tl) or HPGe detectors. The accuracy of the method was analyzed. An algorithm based upon this work was developed, and inserted into commercially-available software. At present time the correctness of software was tested for lead, steel and a tungsten shields with thicknesses from 3 to 50 mm for 152Eu, 137Cs, 60Co, 133Ba, 88Y, and 22Na. This developed software can be used in practice of customs services and can be easily implemented for routine procedures.

REFERENCES [1] Model S504 Genie-2000 InSpector Basic Spectroscopy S504-CUS 7/97. [2] Model S501 Genie-2000 Gamma Analysis. S501-CUS 7/97. [3] K. Debertin, R. G. Helmer, Gamma- and X-ray Spectrometry with Semiconductor

Detectors, Elsevier, Amsterdam, 1988. [4] D. Reilly, N. Ensslin, H. Smith, Jr., and S. Kreiner, Passive Nondestructive Assay of

Nuclear Materials, NUREG/CR-5550, LA-UR-90-732, 1991. [5] J.F. Briesmeister (Ed.), MCNP A General Monte Carlo N-Particle Transport Code

Version 4B, LA-12625-M, 1997. [6] J. Rodenas, A. Martinavarro, V. Rius, Validation of the MCNP code for the

simulation of Ge-detector calibration, Nucl. Instr. and Meth. A450 (2000) 88–97. [7] M.A. Ludington, R.G. Helmer, High accuracy measurements and Monte Carlo

calculations of the relative efficiency curve of an HPGe detector from 433 to 2754 keV, Nucl. Instr. and Meth. A446 (2000) 506–521.

[8] www.atomtex.com.

The Vehicle Monitor for the Detection of Radioactive Materials

Qingli Zhang, Jizeng Ma, Minchen Han, Yawen Huang

China Institute for Radiation Protection, Taiyuan, China

Abstract

At this time, almost all RPMs used in China are produced by foreign companies. To

achieve a higher implementation rate of these devises, it is important to develop a

vehicle monitor that has a user-friendly interface to the Chinese language and is more

economically than foreign products. The objectives of this project were to:

1. Make a prototype vehicle monitor with a Chinese man-machine interface & with the minimum performance requirements by IAEA-TECDOC-1312

2. Test this prototype monitor with the international standard procedures 3. Make some improvements to this monitor, to make the monitor to have the

abilities to reduce the NORM alarm rate with multi-energy-window methods. 4. Make a movable RPM for accident & emergency situation

Under this project, the objectives were achieved. Prototypes of a fixed RPM and a

movable RPM were constructed. They achieved the desired specifications and

functions, such as, gamma sensitivity and low false alarm rate, in the tested static

modes. They operated with user-friendly Chinese man-machine interfaces. The

monitors were tested both in the laboratory and at a scrap recycle mill.

It is anticipated that advanced filtering techniques such as an exponential smoothing

filter, a stepwise monitor and the sequential probability ratio test should be adopted in

the near future. Tests of multi-energy-window method should be made to get more

information from this technique, using medical radionuclides and NORM sources.

Dynamic tests still need to be made with moving sources.

1

1. The Objects and Specifications

In 1950s, China began to establish it’s atomic energy industry. With the development of nuclear industry and nuclear technology in our country, radioactive materials, radioactive sources and radioactive devices are widely used in industry, agriculture, medical, science, education, military etc. They have made important contributions to economy development and social advancement. According to the statistical data from the specialized action “check radiation sources to assure the public” in 2004, there are 12412 units which use radioactive sources and 108504 radioactive sources in total. Among these radioactive sources, 137Cs are about 45 percent and 60Co are 27 percent. About 25 thousands radioactive sources are out of service, and there are certain amounts of special nuclear material. It is estimated that there are more than 2000 orphan sources in China. In recent years, average 30 radiation incidents annually in our country and 80 percent are accidents of lost or stolen sources. Some of them were severe accidents and source melting accidents in the past [1]. For example, in 1991, a lost radioactive source 60Co (370GBq), which was out of service for agriculture breeding purposes, killed two persons and severely injure several others in Shanxi province in north of China.

As we know, a Radiation Portal Monitor (RPM) is one of the most effective monitoring device to enforce the management of radioactive materials and prevent such radioactive accidents. In generally, they are installed at checkpoints, such as customs, airport, seaport, toll-gate of highway, steel mill and public venue. But RPMs have not been widely accepted and deployed in the above mentioned necessary site in China. They only have been installed in nuclear power plants (NPPs)& some of customs where the staffs of NPP & Custom have better radiation and foreign language professional training experience than the staffs of other end-users, such as in medium and small steel mill, toll-gate of highway, garbage treatment mill and so on. In recent two years, some big steel companies begin to install RPMs at the requirement of China National Environment Protection Agency. As an example, in one year after RPMs had been deployed, tens of orphan sources had been discovered in scrap in the steel mill of Wuhan Steel Co. Ltd. in the central south of China.

Up to now, almost all of these RPM in China are from foreign companies such as BICRON, EBERLINE, SAIC, EXPLORIUM, MGP, EURISYS, ESM, RADOS etc. The high price and also foreign man-machine interface are the main obstacles to these end-users for more widely use of these monitors. So, for a more popular RPM installation rate in China, beside to fulfill the technical requirements of respective international standards, it is even more important to develop a vehicle monitor that has a more user-friendly interface to Chinese and is more economically than foreign products.

The objectives of this project are: • Make a prototype vehicle monitor with a Chinese man-machine interface & with the

minimum performance requirements by IAEA-TECDOC-1312

2

• Test this prototype monitor with the international standard procedures • Make some improvements to this monitor, to make the monitor to have the abilities to

reduce the NORM alarm rate with multi-energy-window methods. • Make a movable RPM for accident & emergency situation The minimum specifications for these type of the monitor are as following:[2, 3] • Sensitivity to gamma radiation:

At background 0.2µSv·h-1, with increment of 0.1µSv·h-1 for 1 second, the alarm probability of the monitor should be not less than 99.9%.

• False alarm rate: For background less than 0.2µSv·h-1, the false alarm rate should be not more than 1 false in 10000 measurements.

• Environmental conditions: Temperature: –15℃ to +45℃

Humidity: 10–90%

2. Description of Research

2.1. The prototype fixed monitor

The prototype fixed monitor only for gamma channel have been built up in our lab under this contract. The configuration of this monitor is showed in Figure 1. As usual, it is composed of subassemblies as follows:

Two detection assemblies, which sense gamma radiation. They consist of the plastic scintillators, PMTs, electronic circuit boards and other accessories.

One remote PC: it can be used as a process center of the whole system, a graphic man-machine interface, a huge data storage center. It also can be connected to other peripherals.

Controller unit: to perform management of the following accessories, such as occupancy sensors, speed sensors, indicating units, etc.

The photograph of prototype fixed monitor is showed in Figure 2. We have self-developed the signal conditioning unit with preamplifier, amplifier, high voltage adjust unit, discriminator.

3

Figure 1. The configuration of prototype fixed monitor.

Figure 2. The fixed RPM (for gamma channel).

2.1.1. The composition of a fixed monitor

Detector assemblies（for gamma channel) Figure 3 is a photo of the inside of detector panel. A plastic scintillator with the dimensions of 1000 mm×500 mm×51 mm is used in both of the detector assemblies. The model of plastic scintillator is BC408 (made by BICRON); the specifications of BC408 are showed in Table1 and its relative light output versus wavelength is showed in Figure 4. It can be seen that the light attenuation length of BC408 can reach as long as 210 cm. The longer the light attenuation length of the plastic scintillator, the better the plastic scintillator for application of large area usage. The BC408 plastic scintillators are wrapped in reflective foil and black vinyl for light collection and lightproof.

Temp.Adj.

Detector Assembly

Detector Assembly

Signal Condi.

Signal Condi.

Power Supply

Acquisition CPU

I/O I/O

PC

Occup. sensors

Speed sensors,etc. Alarm Light

4

Figure 3. The photo of the inside of detector panel.

Figure 4. The relative light output of BC408.

5

Table 1. The specifications of BC408 plastic scintillators

Base Density Refractive Index

Softening Point

Light Output,%

Anthracene

Rise Time,

ns

Decay Time,

ns

Pulse Width, FWHM,ns

Wavelength of Max.

Emission,nm

Light Attenuation Length, cm

Polyvinyl- toluene 1.032g/cc 1.58 70 64 0.9 2.1 2.5 425 210

We have chosen the photomultiplier tube model CR105, made by Beijing- HAMAMATSU Ltd.; it is a head-on PMT with Sb-Cs photocathode and box-and-grid type. This type of bialkali photocathode PMT can operate with low photomultiplier noise at room temperature. The low photomultiplier noise is essential to good intrinsic detection efficiency in organic scintillators. Its specifications and spectrum response are showed in Table 2 and Figure 5

Table 2. The specifications of CR105 photomultiplier tube

Spectral Response Cathode Sensitivity

Diameter (mm)

Length Max. (mm)

Range (nm)

Peak Wave- length (nm)

Luminous (μA/lm)

Radiantat λp

(mA/W)

Anode to Cathode Supply Voltage (Vdc)

Anode Luminous Sensitivity

(A/lm)

Current Ampli- fication

Typ.

Anode Dark

Current (after

30 Min.) Typ. (nA)

Anode Pulse Rise Time (ns)

Max. Supply Voltage (Vdc)

51 125 300-650 420 >60 >60 1000 2000 106 50 7.0 1500

Figure 5. The spectrum response of CR105. To reduce the measured level of natural background radiation, a lead box with thickness of 30 mm lead shield all faces of plastic scintillator except the one facing to vehicle. A aluminum plate with the thickness of 1 mm is used to cover this face of the detector. Signal conditioning unit is also in the detector assemblies.

6

Signal Conditioning Electronics Unit A signal conditioning unit has been developed by ourselves for this monitor. It consists of a high voltage divider, preamplifier, Its diagram is showed in Figure 6. The electronic circuits of monitor (preamplifier, amplifier) can be checked by test pulses which are produced by an internal pulses generator. A high quality AC/DC module is used to provide DC power supplies for electronic system. A DC/DC converter module is used to provide high voltage supplies for HV divider of photomultiplier. A digital potentiometers is used to adjust the high voltage between 200V and 1200V. Five analog thresholds (Thres.1 to Thres.5) that their values can be adjusted by five potentiometers are used to form four analog windows. We can distinguished the difference of radiation induced pulses amplitude with recognize the nature of radioactive materials which trig the alarm of monitor and to suppress the NORM alarm [4], 241Am, 137Cs, 60Co, 226Ra sources has been used for the test. Up to now, this unit is still tested in our laboratory.

Figure 6. Diagram of Signal Conditioning Unit.

Speedometer A radar speedometer is used to measure the speed of vehicle for speed range from 1.5 to 50 km/h with accuracy 1km/h. The speed measurement data can be transmitted to central computer via RS-232. It also can provide an alarm signal when the vehicle speed is over allowable limit (such as 10km·h-1). A audio/visual alarms light give the indications whether or not the vehicle contains allowable amount of radioactive materials by comparing the measurement results with the alarm threshold. Figure7 is the radar speedometer and a audio/visual alarms light.

7

Figure 7. Radar speedometer and audio/visual alarm light. Occupancy sensor Photoelectric sensor (Beam-Array Banner) and magnetic loop are used to be a vehicle recognize unit, A curtain of light with the height of 120 cm is provided by the Beam-Array Banner to recognize the air gap between vehicle & its trailer, a magnetic loop and its control unit (Sarasota 526B from Peek-Traffic Ltd., UK) are used to be occupancy sensor to trig the monitor from background measurement to vehicle monitoring. Sarasota 526B (Figure 8) is a dual channel boxed detector designed for use in traffic and access control application with magnetic loops. It has a single presence relay output per channel. The microcontroller is programmed with an enhanced detection algorithm to give optimum operational stability and compensates for environmental changes that occur around the detection circuit. It has the features of high crosstalk and noise immunity, loop fault indicator, four switch selectable sensitivity level.

Figure 8. Sarasota 526B for a occupancy sensor.

Power supply unit & a thermostat

8

A module AC/DC power unit provides the DC (+5V, ±12V) for signal processing unit and a DC/DC module which can produce a variable high voltage (range about from 200 V to 1100 V by a digital potentiometer) for voltage divider of PMT. A simple thermostat and a semi-conductor thermoelectronic cooling keeps air temperature around PMT not higher than the set point. AC/DC and DC/DC power modules provide an excellent stable power supplies for whole system. The variations of power almost have no effects on the performance of the monitor. Model 4NIC-X is a linear integrated AC/DC power supply with a low ripple and high reliability & high efficiency. The voltage accuracy is smaller than 1%, voltage regulation is smaller than 0.5%, current regulation is smaller than 1%, ripple is smaller than 1 mV. It is also provided for over-current protection, short-circuit protection, over-heat protection and over-voltage protection.

Central/Control computer A industrial computer (Advantech model IPC 610 with PCL-836 card in Figure 9) is used to be a control center and also provide a friendly man-machine interface for user and record the historical alarm measurement data. It also can make some algorithms as signal processing unit. The PCL-836 is a counter/timer and digital I/O card for PC/AT compatible computers. It provides six 16-bit counter channels, it also includes 16 digital outputs and 16 digital inputs. The PCL-836 also includes a unique digital filter to eliminate noise on the input signal. The frequency can be adjusted to provide more stable output readings. It can recognize the vehicle & its trailer, catch the speed of the vehicle, and also provide diagnostic pulses for electrical self-test of monitor.

Figure 9. The central computer of fixed PRM.

2.1.2. Data Acquisition & Control Software in Chinese

We have developed data acquisition, processing & control software by ourselves. MCGS (Monitor and Control Generated System) is used to build acquisition software and it is developed on object-oriented language.

9

A lot of program have been developed for counting subroutine, timing subroutine, testing subroutine, communication subroutine, self-diagnostic subroutine, reset subroutine, parameter- access subroutine, filter subroutine and data processing subroutine.

The main page of man-machine interface in Chinese is showed in Figure 10. It can monitor four vehicle lanes. We can enter the page of one of these four lanes by click the “Detail” button on main page or by selecting the four lane menus. It is the mode of background monitoring in the left part of Figure 11. and the mode of vehicle monitoring in the right part of this figure. The monitor works in the mode of background monitoring when there is no vehicle entering into the monitoring area. When there is a vehicle passing through the monitoring area, the occupancy sensor is trigged and the monitor turn into the mode of vehicle monitoring. In the mode of background monitoring, the monitor measures the counts of current background and renews the alarm threshold. In the mode of vehicle monitoring, the monitor measures the counts of vehicle monitoring and compares the monitoring results with the alarm threshold and makes the decision if there is a significant amount of radioactive material in the vehicle. If the decision is NO, the monitor will turn back to the mode of background monitoring when the vehicle passing through the monitoring area. If the decision is Yes, the monitor will trig the alarm unit to produce a sound/light signal. The vehicle must be stopped and checked thoroughly.

Figure 10. The main page of man-machine interface.

10

Figure 11. The page for one of four lanes in background & vehicle monitoring.

We also can execute the other commands by pressing the respective button or by selecting the

respective sub-menu of the main menu. The man-machine interface shows measurement data &curve, the buttons for alarm & fault information, the buttons for preset of parameters, HV adjust, the buttons for self-testing, a lots of help information in Chinese, and so on. The operating parameters concern the aspects related to operation and use of the monitor. The different operating parameters are summarized in the Table 3, these parameters can be set and adjusted by the man-machine interface.

Table 3. The system operating parameters

Parameters Description Date & Time Use the clock of computer，this parameter contains the

current date and time of the system. W Moving average coefficient, the number of

background measurement which is recorded to refresh the alarm threshold.

Low Background Threshold The counts of background are not in the normal range because of the malfunction of the system.

High Background Threshold The counts of background are not in the normal range because of the malfunction of the system and other reasons which have to be investigated clearly.

Sampling period of measurement We can select the 100 ms or 200 ms as sampling time to increase the response speed of system.

K Number of σ, standard deviation of background measurement.

Number of Pass Vehicle The count of vehicles what pass the monitoring area.

Number of alarm The count of vehicles what trig the alarm of monitor be confirmed to be contaminated.

We can digitally adjust the value of high voltage for PMTs by the up & down arrow button on

the page of high voltage adjusting. It is showed in the left part of Figure 12 and the right part in this figure is HV fault & the help information for user to deal with the possible HV faults.

11

Figure 12. The pages for HV adjusting & HV fault information.

It also can keep the record of information of monitoring results (the mean value of

measurement value and their standard deviation) and alarm. The pages of alarm information is showed in Figure 13. These information are those as the start & end time of events, the type of alarm, the alarm threshold, the measurement data of alarm event and so on. The historical measurement information are showed in Figure 14.

Figure 13. The pages of alarm information.

12

Figure 14. The pages of historical measurement information. 2.1.3. Description of the Data Processing Method

In order to achieve the possible reliability in the triggering of the alarm signals and the indicating of faulty operation, background noise is steadily analysed and the estimated value undergoes a “smoothing” by the calculation of moving averages. Moving average algorithm is used to reduce the statistics fluctuations of background measurement. When the RPM is in operation, a constant counting is performed and acquisition is done at time intervals of 200 ms, that is to say that BT, BT+1, BT+2, … raw counting rates (background noise) are obtained. In order to get a rapid response, a shorter monitoring time interval of 100 ms or 200 ms is used for the monitoring of background and vehicles. The raw counting data can be smoothed by the following formula to get the smoothed background noise.

Tmm BW

BW

WB •+

+•+

= − 11

1 )1(

where: W: Operator parameter ( we selectively preset W=80 in our test experiment) Bm: Background noise with moving average upon acquisition B(m-1): Background noise with moving average on the previous acquisition at T-1 BT: Background noise on acquisition T

The alarm threshold is: σKBm +

where:

mB : is the above mentioned Background noise

σ : is the standard deviation of background monitoring K : is a factor that it is usually 3 or 4 for different level of false alarm rate

13

When the monitor is trigged to the mode of vehicle monitoring by the occupancy sensor, the system stops the process of background monitoring and keeps the same value of this smoothed background. Then, we can set the alarm threshold with this smoothed background noise and its standard deviation. Also we can detect the faulty operation of the monitor (low background counting rates or high background counting rates).

As soon as the occupancy sensor (magnetic loops) is trigged, the monitor starts its measurements in order to monitor any possible radioactive contamination. Successive measurements for every 100 ms or 200 ms are performed by the software as long as no contamination has been detected. The processing algorithm for these measurement data is to make the sum of ten (for 100 ms cycle time) or five (for 200 ms cycle time) most recent data, then to compare the mean value of the ten or five measurement data with the above mentioned alarm threshold from background monitoring, finally to decide whether or not to trig the alarm unit. In case of contamination alarm, these measurements are interrupted and confirming measurements should be performed.

In the application of RPM, the frequent “innocent alarm” caused by naturally occurring radioactive materials (NORM) and medical radioisotopes administered to patients is a major problem. NORM radionuclides such as 40K, 226Ra, 232Th and 238U are often found at seaports, trains, and truck traffic at land borders when large quantities of materials are transported. On the other hand, medical radionuclides such as 111In,123I, 125I, 129I, 131I are often found at airports or at the checkpoint of public venue. The high “innocent alarm” rate will cause large expense of investigation and decrease the traffic flow cross the checkpoint. A number of methods have been developed to solve this problem, such as RPMs based on the NaI scintillation detector for vehicle and pedestrian monitors. But even these monitors have higher energy resolution for the recognition of radioisotopes, the lower sensitivities and higher price than that of RPM with plastic scintillator limit their wide application. The most popular detectors of RPMs are plastic scintillators. The plastic scintillators which can be used as a large volume detector have high detection sensitivities and worse energy resolution. The multi-energy-window method [7] is a good approach to make plastic scintillators to have the abilities to recognize radioactive materials to be NORM or medical radioisotopes. The principle of the method is to split the gamma spectra into three energy windows, a low energy window for low energy gamma from SNM and medical isotopes, a medium energy window for most industrial radioactive sources and a high energy window for the detection of high energy gamma from NORM. The principle of this method is showed in Figure 15 [7].

MEDICALand SNM

HIGHMEDIUMLOWINDUSTRIALSOURCES NORM MATERIAL

TOTAL channel

TOTAL

LOW

MEDIUM

HI

ANALYSIS SNM/MEDICAL INDUSTRIAL NORM

Figure 15. The principle of multi-energy-windows method.

14

2.2. The prototype of moveable RPM

We have built the moveable RPM for vehicle/equipment/person radiation monitoring in radioactive accident & emergency situation. It consists of two detector units, one portable computer and one control unit in which there are UPS and linear actuator elevation control unit. Figure 16 shows the photograph of the moveable RPM. The main page of man-machine interfaces is showed in Figure 17 and Figure 18 is the pages for settings of ultrasonic sensor and alarm information.

Detector assemblies: Each detector assembly has a 500 mm×500 mm×51mm plastic scintillator with PMT. In order to decrease the weight of the monitor, no shielding lead plates are applied to the plastic scintillators. Signal conditioning unit is also in the assemblies

Ultrasonic sensor: Instead of magnetic loops for fixed RPM, an ultrasonic sensor is used to be occupancy sensor to trig the monitor from background measurement to vehicle monitoring.

Power unit & a thermostat: Power unit provides the DC for Signal processing unit, high voltage for PMT. A simple thermostat adjusts air temperature around scintillator & PMT. A UPS is used for improvement of power qualities in emergency field.

A portable computer provides a user friendly man-machine interface and record the historical measurement data.

The elevation unit: The center of plastic scintillator can be lift up and down rapidly and smoothly from 1000 mm to 1800 mm by a light compact linear actuator. It is useful to monitor the different size of vehicles in the emergency circumstance.

Figure 16. Moveable PRM for accident & emergency situation.

15

Figure 17. The main page for the moveable PRM.

Figure 18. The pages for ultrasonic sensor, alarm information.

3. The measured performance and applications of RPM:

We have made some tests on the performance of the monitor in our lab. We have tested the some effects of parameters setting. We test the false alarm rate, gamma sensitivities at test point, gamma sensitivities at different point in detection region with collimated and non-collimated 241Am, 137Cs and 60Co sources.

We take the moveable RPM to a steel mill to monitor the vehicles which are loaded with recycle scraps. We have tested the moveable RPM at a recycle mill for 30 days. We have monitored the vehicles and also checked them with a portable survey meters. We have got the information about the influence of rain, vehicle screen effects, and so on.

3.1. Setting of parameters

In order to detect the gamma radiation with a more wide energy range, we adjust the amplifier to make the pulse height to be about 200mV for 241Am gamma ray (energy 60keV), about 2V for the 137Cs gamma ray (energy 660keV) and the pulse height to be about 5V for gamma ray (energy 1500keV). The curves of detection efficiency vs high voltage at different bias voltages have been measured with a 137Cs sources and some of these curves are showed in Figure19. We choose to set the high voltage to about 900V~950V for PMTs. In fact, because of the excellent stabilities of the

16

AC/DC and DC/DC modules, we almost can not observe the significant variations of the counts that they are caused by the fluctuations of main power. The result of this test is showed in Table 4.

Thres.50mV

0

1000

2000

3000

4000

5000

400 500 600 700 800 900 1000 1100 1200

High Voltage (V)

c/0.1s

a. The detection efficiency vs high voltage at thres.50mV

Thres.100mV

0

500

1000

1500

2000

400 500 600 700 800 900 1000 1100 1200

High Voltage (V)

c/0.1s

b. The detection efficiency vs high voltage at thres.100mV

Thres.150mV

0

500

1000

1500

2000

400 600 800 1000 1200

High Voltage (V)

c/0.1s

c. The detection efficiency vs high voltage at thres.150mV

Thres.200mV

0

500

1000

1500

2000

400 600 800 1000 1200

High Voltage (V)

c/0.1s

d. The detection efficiency vs high voltage at thres.200mV

Figure 19. The curves of detection efficiency versus high voltage on PMT.

17

Table 4. The variation of counts with the fluctuations of main power from 220V

187V 198V 210V 230V 242V

Detector A

Detector B

Detector A

Detector B

Detector A

Detector B

Detector A

Detector B

Detector A

Detector B

-5.7‰ -4.1‰ -4.7‰ 1.4‰ -0.7‰ -0.8‰ -4.0‰ 1.7‰

-1.4‰ 0.4‰ -4.0‰ -0.3‰ -4.5‰ -0.9‰ -5.8‰ 2.9‰ -4.4‰ 1.3‰

We have adopted the signal-to-noise method to determine the value of bias voltage for discriminator. We measure the counts of the monitor with and without radioactive source at different bias voltage, then determine the ratio of signal-to-noise by the following equation. We select the bias voltage that it is correspondent to a larger R value. In case of our monitor, the bias voltage of discriminator is set to be greater than 30mV.

Where: S : the counts with radioactive source B : the counts of background noise without radioactive source R : the ratio of signal-to-noise

The minimum detection activity (MDA) is determined by the following formula, the lower bias voltage of discriminator is a better selection for more higher detection efficiency. At the different settings of bias voltage, we have tested the MDA of the monitor with 241Am(399.4kBq±1.8%, k=3), 137Cs(114.8kBq±2.8%, k=3), 60Co(35.4kBq±1.9%, k=3). The distances between the surface of detector cover and the radioactive source are 20 cm for 241Am source and 50 cm for 137Cs&60Co sources. Also there is 10 cm gap from the cover to the surface of plastic scintillator. The results of the test are showed in Tables 5-1 and 5-2. Figure 20 is the curve of MDA vs bias voltage for 241Am, 137Cs and 60Co. Among the range of test bias voltage, it can be seen from data that there is a strong dependence between the value of MDA and the bias voltage of discriminator for lower energy gamma rays than for higher energy gamma rays. Among them, MDA can be defined as follow:

ρT

BMDA ××=

22

Where: B : the counts of background noise without radioactive source T : the measurement time

ρ : the detection efficiency MDA : the minimum detection activity at 2σ

BBSR −

=

18

The raw measurement results of monitoring sensitivities with 241Am, 137Cs and 60Co are listed in Tables 6–8.

0

50

100

150

0 100 200 300

Bias Voltage (mV)

MDA (kBq)

241Am

0

5

10

15

20

0 200 400 600

Bias Voltage (mV)

MDA (kBq)

137Cs 60Co

Figure 20. The MDA vs bias voltage for 241Am, 137Cs and 60Co.

Table 5-1. The MDA of the monitor at the settings of bias voltage for 241Am

Source/Distance 50mV 110mV 200mV 230mV 250mV 241Am(20 cm) 31.4kBq 38.3kBq 46.4kBq 69.9kBq 141.5kBq

Table5-2. The MDA (kBq) of the monitor at the settings of bias voltage for 137Cs, 60Co

Source/Distance 50mV 110mV 200mV 250mV 300mV 350mV 400mV 500mV137Cs(50 cm) 7.66 7.73 7.92 7.36 7.63 8.11 9.74 16.9 60Co(50 cm) 4.23 4.39 4.19 3.94 3.39 3.33 3.45 3.45

Table 6. The raw data of count rate vs bias voltage with 241Am at 20 cm

Bias voltage Background 241Am source

1418 1428 1411 1417 1420 2010 2004 2020 2026 2017 50mV

1414 1416 1431 1413 1397 1416

2029 2026 2034 2054 2010 2023

1401 1405 1375 1390 1383 1862 1889 1895 1883 1884 110mV

1390 1383 1390 1385 1395 1390

1875 1876 1858 1870 1881 1877

1269 1259 1256 1262 1267 1666 1668 1673 1648 1655 200mV

1253 1252 1261 1256 1258 1253

1632 1635 1630 1632 1615 1645

1096 1095 1079 1076 1086 1330 1313 1308 1334 1314 230mV

1079 1077 1064 1061 1071 1078

1302 1314 1322 1316 1301 1315

891 868 874 873 876 997 994 995 993 978 250mV

884 868 886 878 872 877

961 980 977 985 967 983

19

Table 7. The raw data of count rate vs bias voltage with 137Cs at 50 cm

Bias voltage Background 137Cs source

1419 1405 1406 1423 1397 2157 2076 2129 2196 2086 50mV

1391 1432 1402 1403 1391 1407

2197 2004 2088 2015 2257 2121

1396 1385 1398 1363 1390 2155 2034 1998 2049 2167 110mV

1380 1378 1370 1374 1391 1382

2130 2127 2139 1990 2017 2081

1203 1220 1208 1199 1205 1856 1874 1835 1873 1866 200mV

1220 1199 1198 1221 1201 1207

1796 1826 1844 1855 1812 1844

840 832 835 832 833 1404 1400 1392 1390 1411 250mV

835 844 835 840 824 835

1433 1421 1407 1402 1390 1405

585 571 578 577 585 1024 1052 1036 1027 993 300mV

577 578 582 570 575 578

1041 1053 1041 1037 1044 1035

443 444 448 450 454 844 823 842 810 832 350mV

457 439 447 448 445 448

845 848 790 815 810 826

373 362 360 352 353 637 632 646 644 627 400mV

362 363 355 360 358 360

631 642 659 670 636 642

286 285 284 277 277 423 427 411 425 418 500mV

275 277 283 272 284 280

417 424 425 432 438 424

Table 8. The raw data of detection efficiency vs bias voltage with 60Co at 50 cm

Bias voltage Background 60Co source

1419 1405 1406 1423 1397 1851 1799 1803 1788 1808 50mV

1391 1432 1402 1403 1391 1407

1901 1666 1876 1794 1762 1805

1396 1385 1398 1363 1390 1756 1682 1802 1852 1768 110mV

1380 1378 1370 1374 1391 1382

1764 1760 1774 1710 1743 1761

1203 1220 1208 1199 1205 1552 1587 1603 1610 1524 200mV

1220 1199 1198 1221 1201 1207

1569 1553 1609 1582 1600 1579

840 832 835 832 833 1189 1154 1151 1176 1199 250mV

835 844 835 840 824 835

1155 1150 1159 1148 1155 1164

585 571 578 577 585 924 918 873 872 913 300mV

577 578 582 570 575 578

893 888 893 911 863 895

443 444 448 450 454 748 732 732 737 737 350mV

457 439 447 448 445 448

733 724 741 721 713 732

373 362 360 352 353 612 584 612 592 605 400mV

362 363 355 360 358 360

609 611 627 611 597 606

286 285 284 277 277 491 496 516 504 493 500mV

275 277 283 272 284 280

508 467 499 502 498 497

20

3.2. The gamma detection efficiency in the detection region of monitor

The detection region of monitor is limited between the two detection assemblies. The distance between the two detection assemblies is 380 cm, the height of central line of the plastic scintillator is 185 cm, the length of plastic scintillators is 100 cm. The gamma sensitivities of the plastic scintillators is measured with collimated and non-collimated 137Cs and 60Co radioactive sources at different positions in the detection region. The detection sensitivities for one detection assembly with the source-detector distance are showed in Figure 21. It can be seen that the less effective detection position is along the central line between the two detection assemblies. The results of gamma detection efficiency with 137Cs radioactive source are showed in Figure 22. The test points on the plastic scintillator with 100 cm×50 cm are showed in Figure 22-a. The PMT is mounted at the center of left side of the plastic scintillator. The gamma sensitivities of the monitor with collimated and non-collimated 137Cs radioactive sources are showed in Figure 22-b and Figure 22-c It can be seen that the most effective detection position is along the central line of the plastic scintillator for non-collimated radioactive sources.

137Cs

0

40000

80000

120000

160000

200000

240000

280000

320000

0 0. 5 1 1. 5 2 2. 5 3 3. 5 4

The sour ce- det ect or di st ance ( m)

count

Figure 21. The sensitivities with the source-detector distance.

21

- 25- 20- 15- 10- 505

10152025

- 50

- 45

- 40

- 35

- 30

- 25

- 20

- 15

- 10

- 5 0 5 10 15 20 25 30 35 40 45 50

a. The test point on the plastic scintillator (unit: cm)

137Cs sour ce wi t h col l i mat or

B

中心点F

A

HGC

DE

00. 10. 20. 30. 40. 50. 60. 70. 80. 9

11. 11. 21. 31. 4

0 1 2 3 4

位置

Response)

cent r al l i ne 20cm above 20cm Bel ow

b. The gamma sensitivities with collimated 137Cs

137Cs sour ce wi t hout col l i mat or

B

中心

F

A

H

G

C

D

E

00. 10. 20. 30. 40. 50. 60. 70. 80. 9

11. 11. 21. 31. 4

0 1 2 3 4

位置

Response

cent r al l i ne 20cm above 20cm bel ow

c. The gamma sensitivities with non-collimated 137Cs

Figure 22. The sensitivities at different position with collimated and non-collimated 137Cs.

3.3. Gamma sensitivities

Gamma sensitivities are tested by putting radioactive source at the position which is 0.5 or 1 meter from the monitor. The radioactive sources are: 241Am: ( 399.4kBq±1.8%, k=3 ) 137Cs: ( 114.8kBq±2.8%, k=3 ) 60Co: ( 35.4kBq±1.9%, k=3 )

We have tested the MDA of the prototype monitors at the bias voltage of 100mV with these sources. MDAs at 0.5 and 1 meter for different sources are listed in Table 9.

22

Table 9. MDAs at 0.5 and 1 meter for fixed monitor with different sources

241Am 137Cs 60Co 0.5m 1.0m 0.5m 1.0m 0.5m 1.0m MDA

(kBq) 57.7 110.9 8.27 22.8 4.83 14.7

3.4. False alarm rate

The false alarm rate was tested at the background radiation environment for the prototype monitors. The dose rate of background radiation is about 0.11 to 0.15μGy/h.

The false alarm rate was tested in the absence of vehicle. The time period of measurement is 200 ms. After 1 second measurement (5 cycle time periods), the monitor turns from vehicle measurement to the background refresh measurement. The monitor is trigged automatically to 1 scond vehicle measurement once in a minute. The average value of the sum of 5 successive measurements is compared to the alarm threshold which is based on moving average method. The test results are that 9 alarms events happened in total 6736 trigs at alarm threshold (2σ) in about 4 days test, the false alarm rate are 1.3‰. No alarms events happened in total 6856 trigs at alarm threshold (3σ) in about 4 days test. 3.5. Multi-Energy-Window method

As mentioned above, the multi-energy-window method is used to decrease the alarm rate induced by naturally occurring radioactive materials (NORM) and medical radioisotopes. We have made some primary tests on the method with 241Am, 137Cs, 60Co, 226Ra. The principle of judgment is showed in right part of Figure 15. The pulse height which is set for low, medium and high energy windows are as following: Low-window: 50mV―1V about equivalent to range of energy 20 keV―300keV Medium-window: 1V―3V about equivalent to range of energy 300 keV―1MeV High-window: 3V― upper limit about equivalent to range of energy 1MeV― Total-window: 50mV―upper limit about equivalent to range of energy 20 keV― 3.6. Environment test

We have tested the moveable RPM in the walk-in environmental test chamber. The moveable RPM passed the test at the following environmental condition:

At low temperature of -20℃, storage 2 hours, then start the computer & monitor, then works half hours.

At high temperature of +45℃, relative humidity 90%, works 24 hours

3.7. The variation of background radiation [4, 5, 6]

Alarm thresholds are set by the background radiation and the variation of its intensity with time. The rapid variations in background radiation intensity, which can be caused by natural background radiation processes, will cause the monitoring sensitivity to decrease. The screen effect of vehicles and the precipitation are the examples of these rapid variations in background radiation intensity. These effects are evaluated by the field tests. Figure 23 is the field test at a recycle mill and the photo in the right-down corner of this figure is to check the vehicle with survey meter.

23

Figure 23. Field test for the moveable PRM at a recycle mill.

3.7.1. The screen effect of vehicles

When a vehicle enters the monitoring region of RPM, the counts of background measurement will decrease because of screen effect of vehicle. This effect will decrease the detection sensitivities when the alarm threshold is set by the background radiation level. So, it is important to know the variation of effect induced by different vehicles. At the recycle mill, we test this screen effect for various vehicles which they are different in type, size, loaded, dense or loose, etc. The small, less and loose loaded vehicles have less screen effect on background radiation level. The screen effects of monitoring results are given in Figure 24 for 20 vehicles. Some of the measurement data are listed in Table 10. It can be seen that the screen effects are very different from vehicles to vehicles. They vary to about 15% for different loaded vehicles in our test.

Table 10. Some monitoring results of vehicles at a recycle mill

Detector A Detector B No. of Vehicle

Count rate Alarm threshold Count rate Alarm threshold

1 3165.2 3669.5 3000.7 3461.0

2 2856.0 3534.0 2315.0 3056.8

3 2872.9 3541.0 2285.3 3025.8

4 2992.4 3615.3 2659.2 2971.0

24

5 2784.8 3448.5 2486.6 2959.8

6 2930.1 3497.0 2566.0 3333.0

7 2807.6 3497.0 2576.9 3333.0

8 2929.0 3547.8 2859.1 3436.4

9 3334.3 3753.6 3106.3 3592.6

10 3451.2 3854.8 3148.0 3729.5

11 3168.3 3888.0 3078.8 3710

12 3418.9 3922.5 3278.3 3770.5

13 3184.8 3518.2 2917.7 3343.8

14 3071.0 3518.2 2889.0 3343.8

15 2979.4 3518.2 2617.4 3343.8

16 2940.4 3530.0 2573.8 3310.0

17 3011.3 3530.0 2864.3 3310.0

18 2953.1 3530.0 2808.0 3310.0

19 2980.0 3530.0 2661.3 3310.0

20 2954.2 3483.2 2757.5 3244.0

21 3034.0 3505.2 2794.6 3185.2

22 2853.8 3445.0 2519.0 3098.8

23 2945.8 3445.0 2429.3 3098.8

24 2971.0 3425.8 2510.9 2998.5

25 3089.9 3425.8 2791.1 2998.5

26 2862.6 3470.2 2549.5 3140.6

27 3074.5 3470.2 2819.7 3140.6

28 3316.9 3727.0 2647.2 3189.8

29 3026.3 3608.0 2416.4 3165.0

30 2984.8 3608.0 2530.3 3165.0

31 3214.0 4046.2 2905.8 3456.5

32 3242.5 3707.0 2680.1 3275.8

25

rear end 3240.1 3701.8 2667.2 3141.8 33

foreside 3277.0 3701.8 2692.7 3141.8

Rear end of trailer 3020.4 3514.3 2598.1 3064.2

foreside of trailer 3118.5 3514.3 2597.5 3064.2

rear end 3121.2 3514.3 2647.6 3064.2 34

foreside 3050.5 3514.3 2597.1 3064.2

rear end of trailer 3061.8 3660.0 2472.0 3050.4


rear end 3010.9 3660.0 2447.1 3050.4

35

foreside 3001.7 3660.0 2519.0 3050.4

rear end of trailer 2922.7 3555.8 2555.6 2954.8


rear end 2802.0 3555.8 2506.0 2954.8

36

foreside 2795.9 3555.8 2533.6 2954.8

37 3372.6 3723.0 2770.8 3149.8

38 3325.5 3746.6 2740.1 3102.0

39 3090.9 3717.6 2491.7 3055.0

40 3123.6 3800.8 2270.4 2821.2

41 3063.5 3823.6 2325.1 2824.0

rear end 3176.6 3829.3 2567.1 3113.3

middle 2957.8 3829.3 2497.1 3113.3

42

foreside 3023.0 3829.3 2509.2 3113.3

43 2854.8 3569.6 2525.8 3078.0

26

Shield effect of different vehicle

0

5

10

15

20

No.of vehicles

Perc

ent f

or d

epre

ss o

fba

ckgr

ound

A detector B detector

A detector 7.95 14 12.1 8.92 13.9 11.8 15.5 11.9 5.91 4.84 12.6 7.46 3.99 7.42 10.2 10.9 8.27 10.5 9.67 9.08

B detector 7.22 17.5 17.9 4 10.9 18.2 17.9 10.7 8.17 9.39 11.4 7.14 6.36 7.28 16 16.1 6.61 8.44 13.2 9.47

汽车1 汽车2 汽车3 汽车4 汽车5 汽车6 汽车7 汽车8 汽车9汽车

10

汽车

11

汽车

12

汽车

13

汽车

14

汽车

15

汽车

16

汽车

17

汽车

18

汽车

19

汽车

20

Figure 24. Screen effect of different loaded vehicles at a recycle mill.

3.7.2. Rainfall on background radiation level

It is the decay of 226Ra in soil, which produce the gaseous daughter, 222Rn, which can escape the soil to decay in atmosphere. These daughter products may attach to dust particles that form condensation center for raindrops. When these raindrops fall to the ground, the level of background radiation will temporarily increase and then slowly return to its normal value after about 2 hours. A brief intense precipitation will make this increase sharply. A medium-small precipitation is recorded during the field test. About 20% increase of background level was observed. The monitoring results are showed in Figure 25.

var i at i on of backgr ound count r at e dur i ng r ai n

2500

3000

3500

4000

Ti me ( mi nut e)

counts(1/s)

Backgr ound A 3379. 3477. 3577. 3667. 3760 3845. 3813. 3791.Backgr ound B 2919. 3035. 3145. 3219 3277 3309. 3274. 3250

09: 24 09: 26 09: 28 09: 30 09: 33 09: 36 09: 40 09: 45

Figure 25. The variation of background radiation during rain.

27

4. The Results of the project:

We have finished preliminary physical and functional and mechanical design and installation of the prototype monitor for gamma channel. The specifications of these monitors are showed below. We have developed the preliminary signal processing unit and central/control unit with user-friendly interface for Chinese. We also have developed a prototype of movable RPM for emergency situation. We have tested some characteristics of these monitors and developed a multi-window signal-processing unit to suppress the NORM alarm.

4.1. The measured performance of the fixed RPM

The dimensions of plastic scintillator: 1000 mm×500 mm×51 mm×2 The thickness of shield lead: 30 mm False alarm: about 1‰ Gamma sensitivities: MDA(1m,1second,2σ):

241Am(111kBq, 3μCi) 137Cs (22.8kBq, 0.62μCi) 60Co (14.7kBq,0.40μCi).

Power: when the main voltage varies from 187V to 242V, there are not significant variations of response that can be observed.

4.2. The measured performance of the movable RPM

The dimensions of plastic scintillator: 500 mm×500 mm×51 mm×2 The weight of one monitor: 75kg Height of the center of plastic scintillator: 1000 mm―1800 mm False alarm: about 4‰ Gamma sensitivities: MDA(1m,1second,2σ):

241Am(2.11MBq,57μCi) 137Cs (39.5kBq, 1.1μCi) 60Co (21.4kBq,0.58μCi). Environmental conditions:

Temperature: -20℃ to +40℃

Humidity: 10-90%

5. Conclusions

The project has made a prototype of a fixed RPM and a movable RPM with reasonable specifications and functions, such as, gamma sensitivity and false alarm rate. friendly Chinese man-machine interfaces are developed for the monitors. The monitors have been tested in laboratory and a recycle mill.

More advanced filtering techniques such as an exponential smoothing filter, a stepwise monitor and the sequential probability ratio test should be adopted in the near future. The tests of multi-energy-window method should be further made to get more information from this technique with more medical radionuclides and NORM. The dynamic tests should be made with moving sources.

We are very grateful to IAEA for technique and financial support to this project.

28

REFERENCES

[1] Ziqiang Pan, Radiation protection at the beginning of 21st century, Proceedings of experience exchange for radiation protection at nuclear facilities. Beijing, Oct, 2002.

[2] IAEA-TECDOC-1312, Detection of Radioactive Materials at Borders. [3] Beck, P., ITRAP Illicit Trafficking Radiation Detection Assessment Program. [4] P.E. Feklau, G.S. Brunson,“Coping with Plastic Scintillators in Nuclear Safeguards”IEEE

Transactions on Nuclear Science NS-30,158(1983). [5] P.E. Feklau, C. Garcia, et al. “Vehicle Monitors for Domestic Perimeter Safeguard”, Los

Alamos National Laboratory report LA-9633-MS(1983). [6] P.E. Feklau, “Perimeter Radiation Monitors”. [7] K.E. Duftschmid, The application of gamma spectrometric techniques with plastic

scintillators for the suppression of “innocent alarms” in border monitoring for nuclear and other radioactive materials.

Development and Demonstration of the New Methods for the Detection of Hidden Highly Enriched Uranium

L. Meskhi

State Military Scientific and Technical Center DELTA, Tbilisi, Georgia

Abstract

At the airports of many countries, in parallel with modern x-ray monitoring systems, there are

passive devices to detect nuclear materials. Unfortunately, it is problematic to detect highly enriched

uranium (HEU) in this way, due to the low energy of emitted gamma-radiation and HEU’s practical

absence of spontaneous fission. Using “active interrogation” techniques, it is possible to detect even

small amounts of shielded HEU. It this research, it was proposed to develop a mechanical

“switchable” source of neutrons that is different from normal “D-T generator” style of pulsed

neutrons.

The type of switched-off neutron source was chosen to be Am-LiF. Extensive calculations were

performed, and a “switchable” source was designed. It was designed to meet the following technical

requirements:

1. Overall dimensions of equipment do not exceed the size of modern x-ray visual devices. 2. During the working regime radiation conditions outside of device correspond to the radiation

safety requirements. 3. The lowest detectable mass of detected substance do not exceed 10 gram that corresponds to

the one pellet of assemblage of heat-generating cell. 4. Time of detection of the lowest mass (with probability 99%) in the luggage compartment

(having dimensions 60 x 60 x 50) does not exceed one second. Three Russian nuclear centers (Khlopin’s Radium Institute in St-Petersburg, IPPE in Obninsck and

the Scientific Research Institute of Nuclear Reactors of the Russian Federation in Dimitrovgrad city)

have express preliminary consent (with juridical and financial reservations) to develop the

experimental models of modern sources. At this time a physical model has not been fabricated.

1

Prevention of illegal movement of nuclear and radiating materials undoubtedly is a task of extreme importance. Last years IAEA has carried out works on ITRAP (Illicit Trafficking Radiation Detection Assessment Program) [6, 7], has developed corresponding TECDOC Technical/Functional Specifications for Border Radiation Monitoring Equipment and conducts Coordinated Research Project: Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and other Radioactive Materials. Our project is a part of these efforts.

At the airports of many countries in parallel with modern systems x-ray monitoring systems of luggage are already widely applied passive detectors of nuclear materials. These detectors allow finding out and identifying the majority of radioisotopes and nuclear materials on neutrons of spontaneous fission and characteristic gamma-radiation [4, 5]. However, till now, problematic there is a detection of highly enriched uranium (HEU). The reason of it is rather low intensity both small energy of gamma-radiation and practical absence of spontaneous fission.

The technique of active detection of hidden in luggage high-enriched uranium developed by us can be used together with known x-ray visual systems and/or as a new part in structure of the conveyor monitoring system of luggage [2, 3, 16].

Modern systems of the x-ray control of luggage (for example, HI SCAN EDS 10050) operate on the basis of a multi level principle of research of subjects of luggage that guarantees high reliability of detection at high throughput. At speed of a tape of the conveyor 50 cm/s it is possible to appear through at one o'clock of 1500 units of luggage in length up to 100 cm.

The first level completely automatically distinguishes the basic part of luggage as safe which are transported further on delivery to the owners. That part of luggage which does not give clear results at the control, transmit on the second level. This level of the control also automatically carries out the additional computer analysis which provides very high degree of the resolution for a various chemical compound materials. Only stayed after that small a share of all luggage it is carefully examined by the safety personnel.

Such organization of the monitoring system quite corresponds also to a task of detection of the hidden nuclear materials.

With the purpose of reduction radiological problems and improvements of operational characteristics of the device, it was provided to develop a "switched off" radioisotope source of the neutron radiation - Switchable Neutron Source (SNS) [10] — on the basis of Americium-241 and Lithium-7 as target. The basic preconditions of our researches in these directions are development the USA portable equipment AWCC-Active Well Coincidence Counter for measurement of parameters HEU-235 [1] and patent USA #4.829.191.

Development of a new "switched off" source of the neutron radiation is conducted together with the IPPE (Obninsk, Russia), Khlopin’s Radium Institute (St. Petersburg, Russia) [2] and with the Scientific-Investigation Institute of Nuclear Reactors of the Russian Federation (city of Dimitrovgrad-10, Ulyanovsk region).

Specific objectives and research directions

1. Development of a technique of active detection of the latent highly enriched uranium for

combined use in the structure of existing conveyor systems of the x-ray control or as the independent control system [8, 9];

2

2. With the purpose of reduction of radiological problems and improvement of operational characteristics, it is planned to develop a switchable radioisotope source of neutron radiation, so called SNS — Switchable Radiation Neutron Source;

3. A choice of source type and development of the SNS design; 4. Development and optimization of the block of the neutron source; 5. Development and optimization of the block of detection of the fission neutrons; 6. Formation of radiating fields in a zone of the control; 7. Minimization of control time - no more than 5 seconds; 8. Achievement of the maximal sensitivity of detection (the minimal found out weight) — no more

than 10–20 gram of 235U. 9. Maintenance of radiating safety outside of the control system 10. Development of data acquisition system and algorithm of decision-making 11. The solution of technical and technological problems of realization of new units of the device.

The general description of the control system

In the general view the circuit of control system (installation) can be presented in the following (Fig. 1):

As a whole the control system consist of the following basic units: top (SU) and bottom (SD)

blocks of the neutron source, working area with the conveyor (Tr), two blocks of neutron detectors (D1, D2), researched object (B). Biological protection in figure is not shown. Besides one of basic elements of installation is preamp electronics and data acquisition system.

D1 D2

SU

SD

B

Tr

Fig. 1.

3

Accordingly, each of the listed units is a subject of optimization of nuclear-physical parameters and demands careful design study. Optimization is directed on achievement of the set characteristics of control system (installation).

The unit of the neutron source should provide of the homogeneous field of slow neutrons in the control area. Density of a flux and a spectrum of neutrons in working area of installation should be such to provide certain detection of the nuclear materials.

Units of detecting D1 and D2 are identical. Own experience and the literary data testify, that proportional tubes with working gas 3He under 4 atmospheres pressure are the most suitable for the fission neutrons detection. The tube sizes (diameter and length) and their arrangement, the geometrical parameters of moderators and the reflectors are a subject of optimization.

The principle of construction of the biological protection for such devices is known. In our case, in conditions of the concrete design of installation it is necessary to minimize the sizes, weight and accordingly the cost of protection.

All calculations were carried out by the Monte Carlo method with use of software package GEANT3 [11] with a set of subroutines for simulation of the slow neutrons interaction with substance. At passage of the neutrons through substance all basic processes of the neutrons interaction with nucleus, including elastic and inelastic scattering, neutron fission of nucleus, radioactive capture, radioactive fission etc, are taken into account [15]. The destiny of neutrons is traced up to the threshold energy Ttr = 10-3 eV. Neutrons with smaller energy are considered lost.

Choice of a Type of Neutron Source

The general requirements to a radioisotope source of neutrons is a low energy of neutrons, absence of fission neutrons and accompanying high energy γ-quantum.

Neutrons of isotope sources are usually produced in reactions (α-n) on nucleus of light elements. At presence in quality of emitter mostly are used americium-241 (a half-life period 432 years, specific activity 1.27 Bq/g) or plutonium-238 (a half-life period of 87 years, specific activity 1.27 Bq/g) more often. Traditional isotope sources of neutrons usually represent tablets from a mix of a α-emitter with a light element in the ratio approximately 1:10, placed in a hermetic capsule. The neutron yield depends on energy of alpha particles, on a kind of a target and can change a little depending on manufacturing techniques [12, 13].

As energy of an outcome of alpha particles of plutonium-238 (5.49 и 5.45 MeV) and americium-241 (5.43 и 5.48 MeV) differs a little, table 1 gives good representation about neutron yield from a source made on the basis of 241Am or 238Pu and a corresponding target [14].

4

Table 1

Nucleus-target Peak energy of neutrons, MeV

Average energy of neutrons, MeV

Neutron yield for 106 alpha-particles

Beryllium-9 13 4,7 80 Boron-11 7,3 3,2 26 Boron-10 8,1 3,4 13 Magnesium-25 10,1 3,6 6,1 Fluorine-19 2,5 1,4 12 Lithium-7 1,2 0,4 2,6

As an alpha emitter 241Am is more acceptable, as in case of 238Pu outcome of spontaneous fission neutrons is rather high — about 2000 neutrons in a second on 1 gram. 241Am at the expense of spontaneous fission lets out all less than one neutron in a second on gram. The simple analysis of preliminary theoretical, experimental researches and the table data gives the unequivocal answer that only Am-Li or Am-LiF sources, having least average energy of neutrons, shall be used. Besides, this energy is much less than energy of fission neutrons (secondary, informative). This factor promotes reliable allocation of informative signals from background.

Optimization of the Source Block

Constructed on the Base of Point Like Source

The purpose of optimization of a design of the source block is achievement of the maximal flux density of slow neutrons in a working zone of installation. Besides for achievement of uniform sensitivity of detection on all volume of a working zone, it is desirable to have maximum homogeneous flux of the neutrons.

At all stages of performance of the project two variants of a design of a source — a standard small size source (industrial manufacturing), conditionally named further point-like source and switchable source were considered.

Development of a design and manufacturing of the experimental model of the switchable source is one of tasks of the given Project. However, construction and technological study of 241Am-Li switchable source has revealed serious technological problems connected with use of lithium as a target. Therefore, as alternative, alongside with Am-Li in pair, a pair Am-LiF was considered. Though, average energy of neutrons from fluorine is three times higher -<Tn> = 1.2 MeV , but the output of neutrons almost is five times higher at rather low value of the maximal energy of neutrons -3.5 MeV. Furthermore, physical and chemical characteristics of LiF make its rather attractive to application. Use of LiF significantly reduces technological problems and essentially simplifies a design of switchable source. As the spectrum of neutrons from these sources is various, the basic calculations were carried out in parallel for Am-Li and for Am-LiF sources.

5

As known, reflectors and moderators forming the block of the source not only return a part of neutrons in working area, but also soften a spectrum of neutrons. These both factors are extremely important for a task of detection of nuclear materials by active methods.

Optimization of a design of the block of the source means selection of such form and thickness of emitter and moderator which will provide maximum high density of a flux of low-energy neutrons in working area of installation. Instead of determining two values - spatial distribution of density of a flux on volume of a working area and spatial dependence of a energy spectrum of neutrons, we have used that, that the method of Monte Carlo allows to fix in a course of performance of calculation events satisfying certain criterion - accordingly to the effect. For this purpose, in the center (on height) of a working zone it is placed thin (0,4 cm) uranium plate (90% 235U + 10% 238U) the area of the equal area of a working zone and we fix the fact of 235U fission and х, у-coordinates of this event. Thus, but also spatial distribution (heterogeneity) of these events is defined not only number of partitions (probability). To an essence it is equivalent to calculation of convolution of section of division on spectral density of a stream of neutrons in points (xi yi).

On picture 2 the elementary scheme of the block of a source — a point-like source (it is designated by an asterisk) and polyethylene reflector RF are shown. The distance between a source and a reflector makes dS = 0.5 cm. The uranium plate (target) having thickness dTG = 0.4 cm is on distance L = 50.0 cm from a source. The sizes of a reflector and a target make 60х60 cm2. As a rule, the number of modeled events - neutrons emitted by a source makes number Ntot = 105.

During simulating the number of neutrons hit in target N_TG and number of nucleus fissions 235U – N_Fiss. was counted up. Number of modified calculations in this simple configuration the are carried out. In absence of a reflector for discrete values of neutrons energies a range of values of energy makes from 1 keV up to 10 MэВ. The relation of the number of neutrons hit in a target to total number of emitted neutrons N_TG/Ntot determines a solid angle of a target (for homogenous source) equal to 0.085 steradian. On picture 3 results — dependence of number of fissions in a target from energy of neutrons are resulted.

*

RF

TG

dRF dS

L

dTG

Fig. 2.

6

This dependence is proportional to section of partition of nucleus through 235U neutrons. The following calculation is carried out for thick (semi-infinite) reflector dRF = 20cm. Energy of neutrons again discretely varied in the same range. Dependence of number of partition in a uranium plate from energy of neutrons is shown on Fig. 4. From figures it is visible, that the number of fission has grown almost ten times. This growth is caused by appreciably smaller energy of the reflected neutrons.

Fig. 3.

Fig. 4.

1 .10 3 0.01 0.1 1 100

2500

5000

7500

1 .104

Neutron kinetic energy, MeV

Num

_Fis

s

7

Calculations with Am-Li and Am-LiF sources for different thickness of a reflector in order to optimize the thickness of a reflector in a same simple geometry are carried out. Dependence of number of neutrons hits in a target upon thickness of a reflector is shown on Fig. 5, and number of fissions in a target on Fig. 6.

Fig. 5. (boxes-Am-Li, circles- Am-LiF).


8

In both figures it is well visible effect of application of a reflector and distinction for Am-Li and Am-LiF sources. At thickness of a reflector 4–6 cm. the number of neutrons hit in a target is doubled (Fig. 5), and number of fission, because of mitigation of a spectrum of the reflected neutrons, increases in 10 times! From figures it is visible also, that optimum value of thickness of a reflector makes 5–6 cm.

Efficiency of a reflector can be increased having applied an additional cylindrical form reflector (Fig. 7). Internal radius of polyethylene cylinder R, external radius (R + dC) and height of the cylinder — hC. Here two parameters — radius of cylinder R and its height hC are subject to optimization. Thickness of a reflector dRF and the cylinder dC are fixed and are accordingly equal 4 cm and 3 cm.

On Fig. 8 shows results of calculations are resulted: dependence of number of fission in a target from height of the cylinder for four different values of internal radius of the cylinder.


*

RF

TG

dRF dS

L

dTG

hC

2R dC

Fig. 7.

9

From these figures it is visible, that dependence upon the internal radius of a cylindrical reflector is weak and optimum makes 3–4 cm. At small values of height of the cylindrical reflector hC dependence seems to be strong, the exit on a plateau poorly depends on radius of the cylinder and occurs at hC ~3–6 cm. In all cases, for a point like source application of an additional reflector having cylindrical form increases number of fissions in a target still twice! From these figures it is visible, character of dependence does not vary for Am-Li and Am-LiF sources, and value of efficiency differ on (15–25)%.

Source of neutrons having the form of a disk

Development of a design and manufacturing of a pre-production model switchable source of neutrons is one of the tasks of the Project. The switchable source of neutrons essentially has the big area - a little hundred square centimeters. Structurally, the switchable source can have the form as of a flat disk and the form of a square or a rectangular as well. More simple design of a switchable source turns out for a source as a disk. The geometry of the block of a source with a thin flat source is shown on Fig. 9. Radius of disk R = 20 cm, thickness of a reflector — dRF and thickness of moderator — dMod. Gaps between a disk, a reflector and moderator make 0.1 cm. At this stage of calculations (the design of a switchable source is not concretized yet) thickness of a disk is zero. A point of neutrons emission is performing uniformly on a surface of a disk.

For optimization of thickness of a reflector modeling in absence of moderator — dMod = 0 is carried out. Dependence of N_FISS from dRF is shown on Fig. 10.

From figure it is visible, that as well as in a case with a point-like source, N_FISS appears on a plateau at thickness of a reflector 4–6 cm. Having fixed thickness of a reflector on size dRF = 4.0 cm dependence on thickness of moderator dMod is performed. On Fig. 11 this dependence is shown, an optimum thickness of moderator makes dMod = 2.0 cm.

L

dMod

dTG

dRF 60x80

0.1 0.1

2R

Fig. 9.

10

It is necessary to note, that despite of so big size of a disk — 2R = 40 cm the absolute

number of fission in a uranium plate is approximately equal to number of fission in a case with a point-like source. As to homogeneity of a field of neutrons created by an switchable source, studying of this question should be postponed before specification of details of a design of asource.

Other important characteristic of the block of a source is homogeneity of density of a flux of

neutrons in working volume of installation. For studying homogeneity of a neutron field modeling with one, with two and with four point-like sources and with a source in the form of a disk has been carried out. The best homogeneity (10–15)%, obviously, is reached in case of four allocated point-like sources and with a source in the form of a disk.

Fig. 11.

Fig. 10.

11

Neutron detection

The basic purpose of the detection block to provide the maximum of the neutrons detection efficiency of the fission 235U and the minimal efficiency for the primary neutrons from Am-Li source.

As detectors of neutrons, the tubes filled by the 3Не gas at the pressure of 4 atmospheres, operated in proportional mode are chosen. This type of neutron detectors possesses high efficiency of registration of slow neutrons. The industry lets releases the wide nomenclature, both tubes, and electronics to them.

In real conditions of installation, neutrons get in the working volume of the tubes both from Am-Li a source and the fission neutrons of 235U. The design of the detector block should be optimized so that to provide high efficiency of the fission neutrons and low efficiency for neutrons from Am-Li or Am-LiF sources. As a rule the detector block, besides detectors of the neutrons proportional 3He tubes, contains moderator and a reflector. In some cases use as well filters of a boron or cadmium.

For definition of moderator optimum thickness and of the reflector in the detector block, calculations for different values of thickness of a moderator and a reflector are carried out. The point-like source in the isotropic regime lets out neutrons according to a fission spectrum. Thickness of 3He detector 2.5 cm, pressure of working gas 4 atmosphere. In the program the number of operations of the detector, i.e. number of events in which energy deposition in the detector exceeds threshold value Eth = 50 keV is counted up. By results of calculations (Fig. 12) optimum values of moderator thickness - dMod = 3–4 cm. and a reflector - dRF = 5–6 cm are chosen.

Fig. 12.

dRF dMod

12

During the work the analysis of expediency of use of the boric filter in the block of detectors is lead. The spectrum of neutrons from a source is essentially softened in comparison with a spectrum of the fission neutrons. The effect of application of the boric filter on model of the detector block was investigated. Dependence of number of registered neutrons from Am-Li and Am-LiF sources and the fission neutrons was investigated depending on thickness of the filter. A material of the filter — boron carbide — B4C. According to the general configuration of the control system, Am-Li/F source were located hardly above blocks of detectors along the center of a working zone, and a source of partition neutrons - in the central area of a working zone.

Results have shown, that for the fission neutrons practically there is no dependence on thickness of the filter dFL and for neutrons from Am-Li and Am-LiF sources sharp dependence is observed. So, even at thickness of the filter dFL = 0.3–0.4 cm the stream of neutrons causing operation of tubes decreases twice. One more issue which is not obvious without simulation, is orientation of proportional tubes along a vertical or a horizontal. Results of modeling did not give distinctions within the limits of statistical errors.

Optimization of Biological Protection

With the purpose of optimization of biological protection corresponding calculations are carried out in conditions of concrete configuration of the monitoring system. Two-layer protection from all sides surrounds installation. Internal (first) layer is made from polythene, an external (second) layer of lead or iron. Hereinafter, the following designations are accepted: first layer - IX, Y, Z, the second — IIX, Y, Z, indexes designate a protection layer along the corresponding axes.

First of all influence of protection along a z-direction (along conveyor) has been investigated. On Fig. 13 dependence of an withdrawal of neutrons and scale of gamma-quantums out of limits of protection is submitted depending on thickness of a polyethylene layer from end faces (in z-direction). In other directions IX,Y = 0 and IIX, Y, Z = 0. Full number of neutrons emitted from sources Ntot = 104.

0 5 100

1500

3000

4500

6000

Thickness of Iz layer

Num

ber o

f Neu

trons

0 5 100

2000

4000

6000

8000

Thickness of Iz layer

Num

ber o

f Pho

tons

Fig.13.

13

From the figure it is visible, that the neutron number almost twice decreases at thickness of a polyethylene layer 5–6 cm. accordingly, the number of gamma quantum’s moved out of limits of installation, equally grows.

On Fig. 14 results of calculations of dependence of a neutron yield and gamma quantum’s from the thickness of the first protective layer in a direction x and y are shown at different IZ values (thickness of protection in a z-direction) — IZ = 4.0 (red squares), IZ = 6.0 (dark blue dots) and IZ = 10.0 cm. (black diamonds).

From figure it is visible, that, number of neutrons, predictably, does not appear on a plateau, therefore thickness of the first protective layer is chosen on the base of absolute value of a neutron yield. At thickness IZ = 6.0 and IX,Y = 6.0 cm. the neutron yield from the whole surface of installation makes about 400 neutrons on 104 neutrons from the sources. This size of a neutron yield is quite acceptable and as it will be visible from the further, provides allowable levels of radiation outside of installation. The gamma quantum’s yield achieves maximum at 5 cm. The further decrease in number of gamma quantum’s is connected to absorption of gamma quantum’s in polythene.

The absorption of the gamma quantum’s moved out of the first protection layer of shield (IX,Y,Z = 6.0 cm.) should be ensured by the second protective layer made from the heavy material — iron or lead. In Figure 15 results of calculations of withdrawal of a neutron yield and gamma quantum’s out of limits of the second protective layer depending on thickness of the second iron (red squares) or lead (dark blue dots) layer are shown

0 5 1010

100

1 .103

1 .104

Thickness of Ix,y layer

Num

ber o

f Neu

trons

0 5 105000

6250

7500

8750

1 .104

Thickness of Ix,y layer

Num

ber o

f Pho

tons

Fig. 14.

14

Simulations show, that, in concrete geometry of installation 2 cm of lead provide quadruple

decrease in gamma quantum’s yield, hardly less — 4 cm of iron. It is obvious, that increasing thickness of the second layer it is possible to lower gamma quantum’s yield. Thickness of the first and second protective layer should be such to provide an optimum ratio between a neutron and gamma quantum’s yield namely,

Noutcomeγ/Noutcomen ~ КК where the КК-factor means radiation weighting factor.

The total surface of installation, external surface of shield, makes 44.103 cm2. This is minimal value and after real design study, in view of constructive elements, the sizes of installation will increase.

The limits of shield leave, basically, the slow neutrons preliminary exposed to several acts of dispersion and gamma quantum’s having energy 2.2 MeV initiated by 1H (n, γ)2H reaction.

Table 2

Am-Li Am-LiF Fe

4.0 сm. Pb

2.0 сm. Fe

4.0 сm. Pb

2.0 сm.

NВЫХn 1267 3206 2809 5119 NВЫХγ 41300 22231 40355 21704

According to the table 2 the output of neutrons and gamma-quantum’s out of limits of shield makes accordingly about (1–5)103 and (2–3)104 on 105 neutrons from sources. These figures give area density of a flux for slow neutrons about 0.1 neutrons/cm2 sec and for gamma about 0.5 gamma/cm2 sec. These levels completely meet sanitary norms of safety.

0 2 40

150

300

450

600

Thickness of IIx,y,z layer

Num

ber o

f Neu

trons

0 2 40

2500

5000

7500

1 .104

Thickness of IIx,y,z layer

Num

ber o

f Pho

tons

Fig. 15.

15

Calculation of Control System Efficiency

General configuration of installation without biological shield it is shown in Figure 16. The sizes and composition both the block of a source, and the detection blocks have been specified by results of simulation. The 3He tubes are located in the plane xy. The tubes diameter are 2.5 cm, active length 50 cm, thickness of aluminum wall of 0.5 mm. Total number of tubes 48. The tubes are located in chessboard. It is supposed, that the tubes are filled with working gas 3He at pressure 4 atm. The tubes external plane is closed by the polyethylene reflectors RF, from the inside — preliminary moderator PMd, boron carbide (B4C) filter-FL and the polyethylene basic moderator-Md. The target (90% 235U + 10% 238U) is located in the center of working area and is model of the hidden sharing material in luggage. The sizes of a target are 2х2х0.1 cm3. Weight of a sample about 8 g.

Fig. 16.

*

*

FL

PMd

Md

dRF

RF

RF

RF

TG

16

Calculations were carried out both with Am-Li, and with Am-LiF point-like sources. The total number of released neutrons made 5х105. A signal of detection of a sharing material is simultaneous registration of two (double coincidence) or three (triple coincidence) neutrons by proportional tubes.

The threshold of operation of a tube is chosen at a level 50 keV energy deposition in 3He. This value of a threshold is much less than in 3He (n, p)3H cases (kinetic energy of a proton 0.57 MeV) and is more than level of noise of tubes and electronics.

In case of 105 neutrons in a second the rate of the count of an each tube will not exceed 100–200 hertz. Dependence of number of single response of tubes has been investigated depending on a threshold for Am-Li sources in absence of a uranium target (there are no fission neutrons). For three values of a threshold 50, 100 and 200 keV, dependences on a threshold in these limits is not observed

For reception of more authentic values simulation with the big statistics - 5 х 105 neutrons from sources has been carried out. An average of double and triple coincidence (number of events of effect) have made, accordingly for Am-Li and Am-LiF sources 14.3 ± 1.2 and 10.8 ± 1.1. Without a target the same values are equal 0.20 ± 0.14 and 0.3 ± 0.17 correspondingly. Such high values of effect mean that the flux of neutrons from source can be reduced by 20–30% having kept high reliability of detection (above 99%).

Development of the design of a switchable source

One of the primary goals of the project was development of the switchable source which

prototype is U.S. Patent 4,829,191. For the purpose of use of the switchable source in the luggage monitoring system researches on the following directions have been lead:

• A theoretical assessment of the parameters of SNS on the basis of Am-Li, • Development of technical requirements to a source of neutrons, • An experimental assessment of nuclear physical characteristics of a source,

development of the technological proposal on manufacture of a source and an assessment of its approximate cost.

The principal scheme of the neutron switchable source is shown in Fig. 17 [8, 9]. The sources consist of several parallel mobile and three motionless disks. Both surfaces (half of every surfaces) of mobile disks are covered with 241Am and corresponding surfaces of motionless disks — Li or LiF. Permanent magnets provide reliable fixing rotating disks in the switched off condition. Initial turning on and final turning off is carried out mechanically. Each disk is divided into segments that match those of the adjacent plates; alternate plates are in a fixed position. Those disks between the fixed plates can be rotated into the opposition so that the Am surfaces are directly opposite the Li surfaces and α can strike the Li, or alternately, into the "off' position in which α from Am surfaces cannot strike Li.

The neutron switchable source has a non-magnetic stainless steel capsule. The capsule can be evacuated, or alternatively, can contain a non-chemically contaminating atmosphere.

17

Fig. 17.

18

Fig. 18.

19

Fig. 19.

The sketch of the variants SNS design are shown in Fig. 18 and 19 [8, 9]. The choice of a variant

of a design will be possible after of much testing of experimental model of the sources. The major technology requirement to switchable isotope source is: to result in working position

in contact substances of α-emitter and targets so that at switching them off it will be possible to divide them again, not having polluted one another.

As have shown experimental researches optimum thickness of an α-emitter layer makes about 0,1 microns, i.e. approximately 10% from the maximal run of α-particles in plutonium or in americium. This thickness allows use effectively about half of emitted α-particles. As for exception of auto absorption of α-particles in a material of a source it is necessary to make a layer of a α-emitter thin, at their limited specific activity, a neutron yield from sources will be proportional to the area of covering of substance of a target and an alpha emitter. So, for a source on the basis of plutonium-238 and beryllium with an output 105 n/s the area of an active surface of a metal plutonium layer should be equal to 3,2 cm2, for is those a source on the basis of americium-241 — about 15 cm2, and for source Am241-Li7 the covering area will turn out already about 500 cm2. Activity of Americium-241 is about 2 curie, about 0,7 gram.

20

Especially it is necessary to note presence at plutonium-238 of neutrons of spontaneous fission with intensity about 2 х 103 n/s per gram. Hence, as for such source with an output 105 n/s it is required about 20 mg of plutonium-238, which in the switchable regime will let out 40 neutrons in a second. Americium-241 at the expense of spontaneous fission lets out all less than one neutron in a second on gram.

The assessment of size of the covering area of α-sources and targets was given earlier. It is evident, that dimensions of "switchable" neutron sources will be far from the traditional "point" sizes and their design will depend on tasks of application and special rigid technical and operational requirements to sources of radiation. As much as possible taking into account necessary requirements, we have developed some variants of a design of a switchable neutron source of updating.

It is similarly possible to design a source of the rectangular form, and make inclusion trough shift of alternating plates for width of strips of alpha-emitter and target.

SNS case provides necessary tightness and radiating protection against accompanying soft gamma-radiation caused by Americium-241.

The last phase of the CRP the experimental study of the switchable neutron source has not been executed. As stated above, all three organizations not only have shown interest to creation of a source, but also have carried out the certain work on development of manufacturing techniques Аm and Аm-Li components of SNS. However, these organizations belong to a military-industrial complex, therefore to receive from them real products without intermediary IAEA is not possible.

Therefore the relations of this kind do not give IAEA and Delta the right to use the practical results, e.g., we have no right to include in this report some important results of preliminary investigations and discussions. In this regard the memorandum IAEA confirming interests of the Agency in the given subject, i.e. interest of the Agency in the future development of the themes and receipt of the working model of SNS for testing, will be helpful. At the same time authority of the Agency will be the key factor during adjustment of the legal relations.

Conclusions

The main technical requirements to the Project are solvable and available from practical point of view. Suggested technical decisions is sufficient for realization of developing method of detection of High Enriched Uranium in luggage:

1. Overall dimensions of equipment do not exceed the size of modern x-ray visual devices. 2. During the working regime radiation conditions outside of device correspond to the radiation

safety requirements. 3. The range of neutron output (3–5) x 105 is sufficient for switched-off neutron sources. 4. The neutron source of given activity ensures dynamic control of luggage within the system of

existing X-Ray visual control devices. 5. The lowest detectable mass of detected substance do not exceed 10 gram that corresponds to the

one pellet of assemblage of heat-generating cell. 6. Time of detection of the lowest mass (with probability 99%) in the luggage compartment

(having dimensions 60 x 60 x 50) does not exceed one second. 7. Technical parameters of serial electronic system of analysis in time domain of coincidence of

secondary partition neutrons are quite sufficient.

21

8. Extension of activity of neutron source as well as area of detection could accordingly increase the speed of detection.

9. The basic parameters of emanation and detection units were optimized in order to form the neutron fields of primary and converted secondary partition neutrons.

10. The type of switched-off neutron source is chosen: Am-LiF. 11. The model of frame for switched-off neutron source is developed. 12. The technical issues of safe plotting of components (alpha-emitter and target) of switched-off

neutron sources on layers are studied. 13. Three Russian nuclear centers (Khlopin’s Radium Institute in St-Petersburg, IPPE in Obninsck

and the Scientific Research Institute of Nuclear Reactors of the Russian Federation in Dimitrovgrad city) express preliminary consent (with juridical and financial reservations) to develop the experimental models of modern sources. Support of IAEA is strongly needed to support organizational efforts due to construction, testing and delivery of switched-off sources to the customers.

Recommended Future Action by Agency

It is recommended to maintain activities for the further development of methods the active

detection of hidden highly quality nuclear materials. Specifically, it is recommended to initiate a CRP in a direction of “The comparative analysis of application of radioisotope sources and neutron generators for the active detection of hidden highly quality nuclear materials”.

REFERENCES [1] Description and Operation Manual for the Active Well Coincidence Counter. Los Alamos,

LA-7823-M. [2] V. Smirnov, “Analysis of Performance of a System for Explosives Detection in Airline

Baggag”, Pros. of NATO Advanced Research Workshop “Detection of Explosives and Land Mines: Methods and Field Experience”, 2001, St. Petersburg, Russia.

[3] L. Meskhi, Methodology and Field Equipment for Detection Explosives, Drugs and Other Substances of Organic Origin, Pros. of NATO Advanced Research Workshop “Detection of Explosives and Land Mines: Methods and Field Experience”, 2001, St. Petersburg, Russia.

[4] Beck P., Schmitzer C., Duftschmid K.E., Arlt R., ITRAP- International Laboratory and Field Test Site Exercise for Radiation Detection Instruments and Monitoring Systems at Border Crossings, Proceedings International Conference on Security of Material, Stockholm, May 2001, IAEA-CN-86 (2001).

[5] Gunnink R., and Arlt R., “Illicit Trafficking Isotope Identification: Software Problems and Solutions”, Unpublished report presented at the Technical Co-ordination meeting on Hand-Held Device Isotope Identifiers, January 8–11, 2002, IAEA, Vienna. Copies available on request.

[6] P. Beck, Austrian Research Center Seibersdorf, “ITRAP Project — Illicit Trafficking Radiation Assessment Programme”; www.arcs.ac.at.

[7] F. Gabriel, A. Wolf, D. Proehl, R. Jainsch, K.W. Leege, R. Arlt, P. Schwalbach, B. Richter, “Mini MCA — Evaluation, Field Test and Commercialization”. Proceedings of the 19th ESARDA Annual Symposium, Montpellier, France, May 1997, EUR 17665 EN, ESARDA 28, pp. 419; www.gbs-elektronik.de.

22

[8] L. Meskhi, L. Kurdadze, G. Cirekidze, “Development of the equipment to reveal clandestine highly-enriched uranium”. International conference on Nuclear Material Protection, Control and Accounting. May 16–20, 2005, Obninsk.

[9] L. Meskhi, L. Kurdadze, Working Material RCM on the CRP “Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials”. IAEA, SGTS/NSNS. 24–28 April 2006, Vienna, Austria.

[10] E.A. Rhodes, D.L. Bowers, R.E. Boyar, C.E. Dickerman, “Advanced concept proof-of-principle demonstration: Switchable radioactive neutron source”, ANL/ACTV-95/2, Technical Report. OSTI ID: 197135; DE96004828.

[11] GEANT, Detector Description and Simulation Tools. CERN, Geneva, Switzerland. [12] Safeguards Techniques and Equipment. International Nuclear Verification Series #1, IAEA,

2003 Edition. [13] T. Gozani, Active Nondestructive Assay of Nuclear Materials: Principles and Applications,

1981. [14] Donald R. Rogers, Handbook of Nuclear Safeguards Measurement Methods, 1983. [15] W.M. Bowen and Carl A. Bonnett, Statistical Methods for Nuclear Management, 1988. [16] Passive Nondestructive Assay of Nuclear Materials/Edited by D. Reilly, N. Ensslin, H. Smith

and S. Kreiner, 1991.

Comparative Study of New Scintillation Materials in Application to the Border Monitoring Equipment

M. Moszyński

The Andrzej Soltan Institute for Nuclear Studies, Poland

Abstract

At present there is a continuous need for highly efficient and of good energy resolution

(selectivity) detectors to use them in border monitoring hand-held instrumentation to fight

illicit trafficking of nuclear materials. There are at least two major technical obstacles of the

instrumentation: sensitivity and selectivity. The sensitivity is of great importance parameter,

which allows finding nuclear material, while the selectivity is required to avoid a

misinterpretation of the benign radioactivity as being a threat.

An extensive set of measurements was performed; some of the results are listed below: 1. A high energy resolution was measured for the LaCl3:Ce scintillator above 100 keV

gamma rays. For lower energies, NaI(Tl) crystal is still superior because of the excess of the light yield at theses energies.

2. The superior performance of the LaBr3 crystal in the whole range of gamma-rays energy was reflected in its high selectivity and good detection efficiency, better than that of NaI(Tl) crystal. A high linearity of the new crystal response was observed, which should simplify isotope identification.

3. A good energy resolution and high detection efficiency was measured for gamma rays of small CWO and CaWO crystals. However, a larger crystal exhibited a degradation of the light output and energy resolution. In the case of both crystals further efforts are necessary to get large volume detectors of comparable performances.

4. A high detection sensitivity for thermal neutron detection of 6LiI(Eu) crystal was observed, as well as its a high selectivity against gamma ray background. As such, 6LiI(Eu) could be used in compact handheld neutron monitors to replace He-3 detectors.

5. The comparative study of LaBr3 and CZT detectors of comparable size showed a better energy resolution of LaBr3. A poor charge collection in a large CZT limits still obtainable energy resolution. More efforts are necessary for a further development of larger volume CZT detectors with an energy resolution similar to that measured with small detectors.

6. It was recognized that further work is necessary to select and to optimize photomultipliers for LaBr3, assuring particularly a good linearity of the PMT response. To utilize full capabilities of LaBr3 detector, photomultipliers with a reduced number of linear-focused dynodes (to 7 or 8 stages) and characterized by high quantum efficiency of about 35% is required.

1. General description of the research

At present there is a continuous need for highly efficient and of good energy resolution (selectivity) detectors to use them in border monitoring hand-held instrumentation to fight illicit trafficking of nuclear materials [1]. There are at least two major technical obstacles of the instrumentation: sensitivity and selectivity. The sensitivity is of great importance parameter, which allows finding nuclear material, while the selectivity is required to avoid a misinterpretation of the benign radioactivity as being a threat. As of now, the small hand-held instrumentations utilize scintillation detector NaI(Tl) and semiconductor detector CZT. The former one is characterized by a very high detection efficiency but poor energy resolution, while the latter one is of better selectivity but lower efficiency due to its small size.

New scintillation detectors are needed to significantly improve current border monitoring instrumentation. Particularly, new scintillators as LaCl3, LaBr3, CdWO4 (CWO), CaWO4 and LiI(Eu) seem to be good candidates and their advantages and drawbacks should be evaluated. A high energy resolution (selectivity) of LaCl3 and LaBr3, a high detection efficiency of CWO and CaWO4 for high-energy gamma rays due to their density and high Z number can improve the border monitoring. In the case of LiI(Eu), its high sensitivity to thermal neutrons, as well as, its excellent selectivity against γ-ray background have to be pointed out.

The scientific program of the contract was addressed to evaluation of the performance of new scintillation detectors for the border equipment characterized by both a high selectivity in one case and a high sensitivity in the other. Besides of new scintillators, the new photomultipliers necessary for the efficient work with LaBr3 crystal was studied. The study covered tests of:

• Scintillators of a high selectivity as LaCl3 [2] and LaBr3 [3] in comparison to

commonly used NaI(Tl) and CZT detectors, • Scintillators with a high sensitivity as CdWO4 (CWO) [4] and CaWO4 (CaWO) [5] in

comparison to BGO, • The program was extended by the tests of LiI(Eu) crystal as a very efficient neutron

detector • Tests of new photomultipliers for efficient work with LaBr3 crystal, • First tests of new photodetectors, as Avalanche photodiodes (APDs) and Silicon Drift

Detectors (SDD). The studies were summarised in the following papers, attached at the end of the report: [I] M. Moszyński, M. Balcerzyk, W. Czarnacki, M. Kapusta, W. Klamra, A. Syntfeld,

M. Szawlowski, „Intrinsic Energy Resolution and Light Yield Non-proportionality of BGO”, IEEE Trans. Nucl. Sci., 51 (2004) 1074.

[II] M. Balcerzyk, M. Moszyński, M. Kapusta,

„Comparison of LaCl3:Ce and NaI(Tl) scintillators in γ-ray spectrometry”, Nucl. Instrum. Meth., A537 (2005) 50.

[III] M. Moszyński, M. Balcerzyk, M. Kapusta, A. Syntfeld, D. Wolski, G. Pausch, J. Stein, P. Schotanus, “CdWO4 Crystal in Gamma-ray Spectrometry”, IEEE Trans. Nucl. Sci. 52 (2005) 3124.

[IV] A. Syntfeld, M. Moszyński, M. Balcerzyk, M. Kapusta, D. Wolski, M. Majorov and

P. Schotanus, “6LiI(Eu) in Neutron and γ–ray Spectrometry – A High Sensitive Thermal Neutron Detection”, IEEE Trans. Nucl. Sci. 52 (2005) 3151.

[V] M. Swoboda, R. Arlt, V. Gostilo, A. Loupilov, M. Majorov, M. Moszynski,

A. Syntfeld, „Spectral Gamma Detectors for Hand-held Radioisotope Identification Devices (RIDs) for Nuclear Security Applications“, IEEE Trans. Nucl. Sci. 52 (2005) 3111.

[VI] M. Moszyński, M. Balcerzyk, W. Czarnacki, A. Nassalski, T. Szczesniak, H. Krus,

V.B. Mikhalik, I.M. Solskii, „Characterization of CaWO4 scintillator at room and liquid nitrogen temperature”, Nucl. Intstr. Meth., A553 (2005) 578.

[VII] A. Syntfeld, R. Arlt, V. Gostilo, A. Loupilov, M. Moszynski, A. Nassalski,

M. Swoboda, D. Wolski “Comparison of a Large Volume CdZnTe Detector with a LaBr-3 Scintillation Detector”, presented at 2005 IEEE NSS/MIC Conference, submitted to IEEE Trans. Nucl. Sci.

2. Results 2.1. High energy resolution scintillators - comparison to NaI(Tl) scintillator and CZT

detector

Recently new chloride and bromide compounds, as LaCl3:Ce [2] and LaBr3:Ce [3], showed both a very high light output and good energy resolution [26]. The value of 2.9% measured for 662 keV γ-rays from a 137Cs source with LaBr3 scintillator is unequalled [3]. A good density of both crystals equal to 3.86 g/cm3 for LaCl3 and to 5.3 g/cm3 for LaBr3 assure a comparable or better sensitivity than that of NaI(Tl). Fast light pulses with the decay time constant of about 18 ns allow for a high counting rate measurements. 2.1.1. LaCl3:Ce crystal

The LaCl3:Ce crystal was proposed first by van Loef et al [2]. In the study [II], the commercial sample of LaCl3:(9±1)%Ce with the size of ∅25 × 25 mm was compared to NaI(Tl). A good light output of 9400±100 photoelectrons per MeV and energy resolution of 4.2 ± 0.2% for 662 keV γ-rays were measured with the LaCl3 crystal coupled to the XP3212 photomultiplier with bialkali photocathode. Below 122 keV, energy resolution of LaCl3 was unexpectedly worse than that of NaI(Tl). Estimated photofraction of 17.9% at 662 keV is comparable to 21.4% for NaI(Tl) of ∅25 × 31 mm size. The radioactive background of the natural 138La radioactive isotope was observed in LaCl3, estimated to be about 28 c/s in the crystal. Moreover, the contamination of the crystal by α emitting isotopes of uranium series was discovered. The LaCl3 showed a good proportionality of the light yield versus energy within 3% down to 20 keV. The total and intrinsic energy resolutions are discussed.

The new LaCl3:Ce crystal delivered by Saint-Gobain showed a reduced alpha contamination at least by one order of magnitude in comparison to the first tested crystal.

Fig. 1 presents the comparison of energy spectra of 662 keV γ-rays from a 137Cs source, as measured with LaCl3 and NaI(Tl) crystals at 3 μs shaping time constant. Note a very good energy resolution of LaCl3 of 4.2% in comparison to 6.5% observed with NaI(Tl).

0 1000 2000 3000 40000

10000

20000

300000 1000 2000 3000 4000

0

1000

2000

3000

4000

5000

FWHM 4.2%

LaCl3:Ce

Cou

nts

Channel

FWHM 6.7%

NaI(Tl)

Cou

nts

Fig. 1. The comparison of energy spectra of 662 keV γ-rays from a 137Cs source, as measured with LaCl3 and NaI(Tl) crystals under the same gain of spectroscopy amplifier.

The non-proportionality curves for LaCl3 and NaI(Tl) are shown in Fig. 2. The non-proportionality is defined here as the phe number yield measured at specific γ-ray energy relative to the phe number yield at 662 keV γ-peak. LaCl3 is clearly superior to NaI(Tl) in terms of non-proportionality.

90%

100%

110%

120%

10 100 1000 10000Energy, keV

Non

prop

ortio

nalit

y LaCl3 NaI(Tl)

Fig. 2. Non-proportionality of the light yield of LaCl3 and NaI(Tl). The curve error is 3%.

0%

10%

20%

30%

10 100 1000 10000Energy, keV

Ener

gy re

solu

tion NaI(Tl) Sample 3

LaCl3

Fig. 3. Energy resolution of LaCl3 and NaI(Tl) versus energy.

The advantages and properties for LaCl3 and NaI(Tl) are summarized in Table 1.

Table 1. Comparison of properties of LaCl3 and NaI(Tl)

Property NaI(Tl) LaCl3:Ce

Light yield 12750 phe/MeV 9400 phe/MeV

Afterglow Large Substantial

Radioactivity None 138La, 2.34 decays/cm3

Energy resolution at 662 keV

6.76±0.15% 4.2±0.13%

Proportionality +15% at 20 keV ±3% above 20 keV

α/γ 0.64 0.33±0.01

Density 3.67 g/cm3 3.86 g/cm3

Zeff 50 59.5

Photofraction at 662 keV

21.4% 17.9%

No doubt that the main advantage of LaCl3 over NaI(Tl) is superior energy resolution at energies above 120 keV. This fact and a comparable detection efficiency and photofraction will make it, in the near future, a crystal of choice for a precise γ-ray spectrometry. A much faster light pulse allows for a high counting rate measurements, several times larger than that with NaI(Tl). A high speed of the fast component of light pulse and high light output assure also fast timing capabilities of the new crystal.

For more details, please look to the enclosed copy of the paper.

2.1.2. LaBr3:Ce crystal

The performance of hand-held radioisotope identification devices (RIDs) is still hampered by the performance of the NaI(Tl) detectors, which are commonly used in such instruments. In the paper [VII] we continue the search for better detector options. One of the largest ever made single elements, coplanar CdZnTe detector (30 × 15 × 12.1 mm3, volume 5.45 cm3, designed by University of Michigan) is compared with a commercially available LaBr3 detector (∅1” × 1”; volume 12.9 cm3). Parameters, relevant to the performance of isotope identification devices, such like resolution and efficiency as function of the γ-ray energy, temperature shift, linearity and others are measured and compared. According to first measurement results it seems to be very likely that for this application LaBr3 detectors are an alternative to CdZnTe detectors. This, even more, if one bears in mind that LaBr3 came only recently commercially available and detectors with larger volumes are likely to appear in the nearest future.

The overall energy resolution for LaBr3 crystal connected to PM tube was measured as 3.2% at 662 keV (see Fig. 4) at the shaping time being 3 μs. In the same figure the spectrum of 137Cs measured with CP CZT is included and compared to LaBr3. An energy resolution of 5.1% was measured for the CP CZT detector. The energy resolution of LaBr3 is much better than that obtained for the CP CZT detector. On the other hand, all large volume (2.25 cm3) CZT detectors presented in [8] had energy resolutions less than 2% on the energy 662 keV. For the measured 137Cs spectra the peak-to-Compton and peak-to-valley ratios for LaBr3 and CP CZT were calculated and presented below in Table 2.

0 100 200 300 400 500 600 700 800 900 10000

200

400

600

800

1000

1200

1400

1600

CP CZT

LaBr3

5.1% 3.2%

661.6 keV1 ch = 1keV

Cou

nts

Channel number

Fig. 4. The γ-ray spectra from a 137Cs source measured with LaBr3 and CP CZT.

Figure 5 shows an overall energy resolution as a function of γ-ray energy measured for LaBr3, CP CZT and NaI(Tl) detectors. CP CZT exhibits poor energy resolution in the whole energy region (18.7% and 5.1% at 81 and 662 keV, respectively.) The energy resolution measured for LaBr3 is extremely good for γ-ray energies above 100 keV. The deterioration of energy resolution below 100 keV follows probably the light yield decrease observed for the LaBr3 crystals (see Fig. 6). The increasing tendency of NaI(Tl) light output while the γ-ray energy is going down makes NaI(Tl) still competitive amongst scintillators in the low energy region.

10 100 10001

10

Ene

rgy

reso

lutio

n(F

WH

M, %

)

Energy (keV)

NaI(Tl) LaBr

3

CP CZT

Fig. 5. The overall energy resolution measured for LaBr3 and CP CZT (compared to ∅25 x 31 mm NaI(Tl)).

The light yield as a function of γ and X-ray energy relative to the yield at 662 keV is

presented in Fig. 6. The non-proportional response of LaBr3 is compared to that of NaI(Tl). For the energies above 200 keV both scintillator responses are quite well proportional while in the low energy region the LaBr3 detector is more proportional than NaI(Tl). In contrary to NaI(Tl), the LaBr3 non-proportionality curve is bending down when energy decreases and the light production is much lower compared to NaI(Tl) (see Fig. 6) The non-proportionality curve presented in Fig.6 for the LaBr3 detector is in agreement with the non-proportionality reported by Dorenbos et al. [10]. However, one should be careful in determination of light yield in the high energy region (above ~1 MeV) owing to a possible non-linearity of PM tube.

10 100 10000.80

0.85

0.90

0.95

1.00

1.05

1.10

1.15

1.20

Non

-pro

potio

nalit

y(n

orm

aliz

ed to

662

keV

)

Energy (keV)

NaI(Tl) LaBr

3

Fig. 6. A comparison of the non-proportionality curves measured for the LaBr3 and NaI(Tl) (∅25 × 31 mm) crystals, respectively.

Figure 7 shows γ spectra for the enriched 235U. The lower spectrum presents the 235U source behind fertilizer, 14 cm in thickness, taken with the LaBr3 detector. Since 235U emits γ-radiation mainly in the low energy region, a good energy resolution is crucial to γ-ray separation and identification in this energy region. The 186 keV peak is still observed even if the 235U sample is masked by the natural radioactive shield.

0 500 1000 1500 2000 2500 30001

10

100

1k

10k

1

10

100

1k

10k 1 ch = 1 keV

40K

Cou

nts

Channel number

235U

40K (background)

235U behind 14 cm fertilizer

235U

Fig. 7. The 235U spectra taken for the LaBr3 detector.

Table 2 summarizes a comparison of performance of the LaBr3 scintillation detector, a large volume coplanar grid CdZnTe (CP CZT) detector and those of NaI(Tl) crystal.

Table 2. Properties of LaBr3, CP CZT and NaI(Tl) detectors

Unfortunately, poor performance of the tested large volume (5.45 cm3) CP CZT indicates bad quality of this crystal that is a consequence of the still immature fabrication of such big single slab crystals. Previously measured smaller volume (up to 2.25 cm3) CdZnTe detectors were characterized by a small temperature drift, the highest energy resolution and the best linearity of the energy scale. More efforts are necessary for a further development of larger volume CdZnTe detectors with an energy resolution similar to that of smaller ones.

Quite new LaBr3 scintillation crystals occurred to be a compromise between NaI(Tl) and CdZnTe. Recent samples of a volume of about 13 cm3 can register γ radiation with an energy resolution as high as 3.2% at 662 keV compared to about 2% for CP CZT (volume 2.25 cm3) and 6.5% for the selected NaI(Tl) samples. In the low energy region (<100 keV) LaBr3 has

poorer energy resolution and comparable to NaI(Tl). On the other hand, LaBr3 is more linear in the whole energy region. High detection efficiency of LaBr3 scintillators is due to their high density and recently available larger size of the crystals.

2.2. Scintillators with a high sensitivity — comparison to BGO and NaI(Tl) scintillators

New high-Z scintillator detectors are needed to significantly improve current border monitoring instrumentation utilizing plastic detectors, as well as, for the development of highly efficient handheld instrumentation. BGO, widely used high-Z scintillator suffers because of a low light yield and in consequence a poor energy resolution. CdWO4 (CWO) [4] and CaWO4 (CaWO) [5] are candidate to be used for this purpose. An application of CWO and CaWO in the gamma spectrometry is limited because of a long main decay time constant of the light pulses, which does not allow measurements at high counting rates. Both crystals belong to the group of “very slow” scintillators with a long decay time constant of the light pulse of 14 µs for CWO [4] and 9 µs for CaWO [5]. However, in the border monitoring equipment, one is looking for traces of gamma rays emitted by well-shielded radioactive sources. Thus, high rate capabilities of the detectors are less important. 2.2.1. CdWO4 (CWO)

In the paper [III], the properties of CdWO4 (CWO) crystals in gamma spectrometry were studied. Several small samples of 10x10x3 mm size, typically used in CT X-ray detectors, were tested and then compared to performance of a larger crystal of 20 mm in diameter and 20 mm high. A light output, energy resolution and non-proportionality of the CWO response versus gamma-ray energy were measured and compared to those of small BGO to discuss further the origin of the intrinsic resolution of pure scintillating crystals. A high light output of 6500 ± 200 phe/MeV and a good energy resolution of 6.6 ± 0.2% were measured for 662 keV gamma rays from a 137Cs source coupling the small samples to XP3212 photomultiplier. Common non-proportionality curves and consequently common intrinsic resolutions of small CWO and BGO suggest that they represent fundamental characteristics of the scintillating material themselves, for heavy oxide crystals.

Fig. 8 shows the energy spectra of 662 keV γ-rays from a 137Cs source measured with a small CWO crystal and with a 25 mm in diameter and 30 mm high NaI(Tl).

Note the high energy resolution of 6.6 ± 0.2% for the 662 keV peak obtained with the CWO, which is comparable to that of NaI(Tl). Note also the comparable photofractions of 26% and 23% in both spectra for the CWO and the NaI(Tl) crystals, respectively, while the volume of the NaI(Tl) crystal is 50 times larger. A shift down of 32 keV KX-ray peak in CWO spectrum in relation to that of NaI(Tl) is seen. It reflects an excess of light at low energies in NaI(Tl) observed in the non-proportionality curve [5–8] and a reduced light yield in the case of CWO, see Fig. 10, below.

Fig. 9 presents the spectra of gamma rays from a 207Bi source and that of the background of the laboratory measured with a larger CWO crystal of 20 mm in diameter and 20 mm in height. High photofractions are observed for the high energy gamma peaks of 1063 keV and 1770 keV. In the background spectrum, well defined peaks of 1460 keV for 40K, 1764 keV due to Radium series and 2614 keV due to Thorium series, are present. All of them are typically observed in background spectra.

0 500 1000 1500 2000 2500 30000.0

5.0x104

1.0x105

1.5x105

0 500 1000 1500 2000 2500 3000 3500 40000.0

2.0x103

4.0x103

6.0x103

8.0x103

FWHM 6.6%

137Cs12 μs shaping

Num

ber o

f cou

nts

Channel #

CWO 1L10×10×3 mm

FWHM 6.5%

137Cs3 μs shaping

Num

ber o

f cou

nts

NaI(Tl)∅25×30 mm

Fig. 8. Energy spectra of 662 keV γ-rays from a 137Cs source, as recorded with the 10 x 10 x 3 mm3 CWO

(bottom panel) and with the ∅25 mm x 30 mm NaI(Tl) (upper panel). An escape peak of K X-rays of Tungstate is seen at the 662 keV peak in the CWO spectrum.

Fig. 9. Energy spectra of γ-rays from a 207Bi source (top trace) and that of the laboratory background (bottom

trace), as recorded with a large CWO crystal at 12 μs shaping time constant in a spectroscopy amplifier.

Fig. 10 presents the non-proportionality characteristics of the small CWO in comparison to that of a BGO crystal, according to Ref. [I]. The non-proportionality is defined here as the ratio of the photoelectron yield measured for photopeaks at a specific γ-ray energy relative to the yield at 662 keV γ-peak [5, 7]. Note a comparable shape of the measured curve for CWO to that reported in [I].

10 100 100060

70

80

90

100

110

CWO BGO

Ligh

t yie

ld [%

of6

62 k

eV]

Energy [keV] Fig. 10. A comparison of the non-proportionality curves measured for the small CWO and BGO crystals. The

curve for BGO follows that of [I].

In the whole range of energies, the curves are well matched. Since BGO showed the same non-proportionality at room and at LN2 temperatures, it was postulated in [I] that its non-proportionality is a fundamental characteristic of BGO material. This may imply a further conclusion that the curve presented in Fig. 10 also represents fundamental characteristics of the scintillating materials themselves, i.e. characteristic of heavy oxide crystals. Note that one can expect a high purity of CWO crystals, developed for CT scanners. Both BGO and CWO crystals belong to the scintillators with the lowest afterglow.

Fig. 11 presents the energy resolution versus the energy of gamma rays obtained with the CWO 1L crystal in comparison to that measured in [I] with BGO crystals of 4 mm thickness.

10 100 1000

10

100

CWO BGO

Ener

gy re

solu

tion

[FW

HM

, %]

Energy [keV] Fig. 11. Energy resolution of CWO and BGO crystals versus gamma rays energy. Error bars are within the size

of the points. Data for BGO are taken from [I].

Note a much better energy resolution of the CWO crystal compared to BGO in entire range of energy. There is no doubt that this is due to about 3.5 times higher photoelectron number of the CWO.

The study of the small CWO crystals showed a high light output of 27300 ± 2700 ph/MeV and an energy resolution of 6.6 ± 0.2% for the 662 keV γ-rays from a 137Cs source, both placing the CWO crystal within bright scintillators with a good energy resolution.

The high photofraction estimated for the CWO crystals studied here and particularly that of 47% for the 662 keV peak, estimated for a 20 mm in diameter and 20 mm in height CWO crystal, confirmed the high efficiency for gamma rays detection and the high potential of CWO crystals for application in the border monitoring equipment. A further improvement of the quality of larger crystals is required to improve their energy resolution.

2.2.2. CaWO4

The properties of CaWO4 (CaWO) crystals in gamma spectrometry were studied at room and liquid nitrogen temperatures. Small samples of 10 × 10 × 4 mm3 and 10 × 10 × 8 mm3 size were tested, coupled to a Photonis XP3212 photomultiplier at room temperature and a large area avalanche photodiode at liquid nitrogen temperature. Light pulse shape and light output at room and liquid nitrogen temperatures were measured. Energy resolution and non-proportionality of the CaWO response versus gamma-ray energy were studied and compared with those of small BGO and CdWO4 crystals to discuss further the origin of the intrinsic resolution of pure undoped scintillating crystals. A high light output of 4800 ± 200 phe/MeV and a good energy resolution of 6.6 ± 0.2% for 662 keV gamma rays from a 137Cs source were measured for the small samples coupled to the XP3212 photomultiplier.

Fig. 12 shows the energy spectra of 662 keV γ-rays from a 137Cs source as measured with the CaWO1 crystal at room and LN2 temperatures. Both spectra exhibit a comparable energy resolution of the full energy peak.

Fig. 12. Energy spectra of 662 keV γ-rays from a 137Cs source recorded with a CaWO crystal at room and LN2

temperatures.

Worthwhile noting is the clearly observed escape peak of tungsten K X-rays on the low energy side of the peaks and a high photofraction of 29% in the small volume crystal (0.4 cm3). This parameter is a factor of two greater in comparison to the 10 mm in diameter and 10 mm high NaI(Tl) sample having about twice the volume (0.8 cm3). The measured energy resolution is superior, as compared to those reported in [5], but it was achieved with a small-size crystal.

Fig. 13 presents the energy resolution versus energy of gamma rays obtained with the small CaWO1 crystal at room and LN2 temperatures measured with the PMT and LAAPD readout, respectively. A straight line in the double logarithmic plot shows that the energy resolution of CaWO in the whole range of measured energies is independent of temperature.

Fig. 13. Energy resolution measured with CaWO at room and LN2 temperature versus energy of gamma rays.

Fig. 14 presents the non-proportionality characteristics of CaWO crystals determined at both room and LN2 temperatures. The non-proportionality is defined here as the ratio of the light yield measured at specific γ-ray energies relative to the light yield at the 662 keV γ-peak [5, 6].

The curves exhibit a good match and there is no temperature dependence. This finding is consistent with earlier measurements for BGO crystals [I]. This implies that in contrast to halide crystals [5–7], the non-proportionality of BGO and CaWO seems to be a fundamental characteristic of the scintillating materials. A characteristic dip of the non-proportionality curve around the K-absorption edge of tungsten can be noticed and that was also observed for other scintillators [5]. It may suggest also that the nonlinearity of CaWO is independent of temperature in a typical range required for border monitoring instrumentation.

Fig. 14. Non-proportionality characteristics of a CaWO crystals determined at both room and LN2 temperatures.

The study of small CaWO crystals in gamma spectrometry showed a high light output of 15800 ± 1600 ph/MeV and an energy resolution of 6.6 ± 0.2% for the 662 keV γ-rays from a 137Cs source, as measured at both room and LN2 temperatures. It places CaWO crystals within the range of bright scintillators with good energy resolution. The light pulse shape studied by means of a single photon method at room temperature showed that it is represented well by a single main decay time constant of 8.2 ± 0.1 µs.

The high photofraction determined for the CaWO crystals studied here confirmed the high efficiency for gamma ray detection and the high potential of CaWO crystals for application in border control equipment.

The study of energy resolution and non-proportionality of the light-yield of CaWO as function of gamma ray energy showed common characteristics at both room and LN2 temperatures and confirmed the good correlation of the intrinsic resolution of the crystal with the non-proportionality curves. It is shown that the non-proportionality of the light yield and the intrinsic resolution of CaWO are superior to those of other heavy scintillators, namely BGO and CWO.

2.3. LiI(Eu) crystal as a very efficient neutron detector

In the paper [II], Europium activated 6LiI crystal (enriched to 96% 6Li) has been

extensively studied in neutron and γ-ray spectrometry with a special attention laid to its sensitivity to thermal neutrons and selectivity against the γ-ray background. A sample of

Ø50 mm × 5 mm size was tested using a light guide between the crystal protective window and the calibrated Photonis XP5200 photomultiplier window. The response to neutrons emitted from the Pu-Be source, shielded with paraffin blocks, has been investigated and the thermal neutron peak has been found to be located at Gamma Equivalent Energy (GEE) of about 3.5 MeV with the very good peak-to-background. The high sensitivity of the 6LiI(Eu) crystal is demonstrated by observing the neutron peak at very low neutron rate of the order 10-2 counts per second. Moreover, a high selectivity against the γ-ray background is measured as the apparent peak of neutrons originating in the atmosphere is located beyond the highest-energy gamma peak of the background radiation. Apart from neutron spectra a light output, energy resolution and non-proportionality of the 6LiI(Eu) response versus γ-ray energy have been measured. The light yield obtained was 1.5·104 ph/MeV (about 40% of NaI(Tl)) and the energy resolution at 662 keV (Cs-137) of 7.5 ± 0.2%. The non-proportionality curve has been measured to have more “proportional” character as compared to NaI(Tl) crystal although an upward bending is still observed for energies below 200 keV.

An example of a spectrum measured for the shielded Pu-Be source is shown in Fig. 15. Note a well defined Gaussian neutron peak characterized by a high energy resolution of 3.9%. A continuous spectrum on the left from the neutron peak corresponds to the laboratory γ-ray background with a characteristic peak of 1460 keV from 40K.

Fig. 15. The spectrum of a shielded Pu-Be source (0.6 Ci). Thermal neutron peak appears at about 3.5 MeV with a very good peak-to-background ratio.

To determine a sensitivity of 6LiI(Eu) crystal to thermal neutrons, a neutron monitor

Nuclear Enterprises NM2B1 was used to control a dose equivalent (DE) expressed in both µSv/h units and counts of neutrons per second registered in the monitor. Since the 6Li nuclei have a large cross-section for thermal neutron capture, the strong Pu-Be source (0.6 Ci) was surrounded with paraffin blocks to slow down the fast neutrons. Moreover, several 5 cm thick lead bricks were used in front of the source to absorb 4.4 MeV γ-rays following (α,n) reaction in the Pu-Be source. The source was placed several meters far from the 6LiI(Eu) detector and the neutron monitor. Thus, in fact, instead of the neutron flux, detectors responded to a homogeneous field of diffused thermal neutrons.

Provided the neutron field was homogenous, the high sensitivity of the detector is illustrated in Fig. 16 where almost linear dependence of the number of detected neutrons by

1 NMB2 monitor based on boron trifluoride proportional counter surrounded by a combined modulator/attenuator assembly to produce the

correct dose corresponding to the human.

6LiI(Eu) on the dose equivalent goes down to very low DE values. 6LiI(Eu) is able to detect neutrons at the dose equivalent as low as 0.05 μSv/h registered in the NMB2 neutron monitor.

Fig. 16. Measured linear dependence between the number of neutron detected in 6LiI(Eu) (left Y-axis) vs. dose equivalent (bottom X-axis) and counts of neutrons (top X-axis), both registered by the neutron monitor NM2B.

To test the 6LiI(Eu) response to gamma rays, the energy spectra were measured using

different radioactive source covering the energy range from 16 keV to 1.33 MeV. Fig. 17 shows the energy spectrum of 662 keV γ-rays from a 137Cs source as measured with 6LiI(Eu) (L). The energy resolution of 7.5 ± 0.1% was measured to be much better than a value of 8.8 ± 0.1% obtained for the small one. These both values are inferior to that obtained for NaI(Tl).

Fig. 17. The γ-ray spectrum from a 137Cs source measured with 6LiI(Eu) (L) with a light guide placed between

the crystal and the PMT window.

Fig. 18 presents the non-proportionality characteristic of 6LiI(Eu) in comparison to that of a small NaI(Tl) [8]. The non-proportionality is defined here as the ratio of photoelectron yield measured for photopeaks at specific γ-ray energy relative to the yield at 662 keV γ-peak [5, 8]. The non-proportionality of 6LiI(Eu) shows a characteristic shape for other halide crystals with an excess of light at low energies [5–7]. It points out that this shape is independent of a doping agent and correlates with the structure of the halide crystal. Light yield non-proportionality obtained in our measurements, presented in Fig. 18, has much better proportional character of the curve as compared to the NaI(Tl) crystal. On the other hand,

more proportional scintillators have the downward bending of proportionality curve at few tens of keV [5, 8].

Fig. 18. A comparison of the non-proportionality curves measured for the 6LiI(Eu) (L) and NaI(Tl)

(∅10 × 10 mm) crystals, respectively.

Fig. 19 shows radiation of the low and high burnup samples of Pu shielded with a 30 mm thick Pb brick, registered in the 6LiI(Eu) crystal. The brick was placed between the sample and the detector and significantly attenuated intense low energy photons. No neutron moderator was used in both measurements. The sample of low burnup 239Pu was measured within 25 minutes while the other for 7 minutes. In both cases, the peak of neutrons which were thermalized by surrounding materials is distinct from the γ-rays registered by the detector at the same time. Finally, neutron count rates were measured as 45 n/s/kg and 240 n/s/kg for the low and high burnup 239Pu samples, respectively.

Fig. 19. The γ-ray and neutron spectra of the low burnup Pu (bottom) and high burnup Pu (top) samples

measured with 6LiI(Eu). The 30 mm Pb shield was placed between the samples and the detector.

A very high sensitivity to thermal and slow neutrons was demonstrated for 6LiI(Eu) crystal, as well as, a high selectivity against γ-ray background was observed as the neutron peak was distinct from the highest energy background peaks. Due to the high neutron detection efficiency of 6LiI(Eu), the crystal detects slowed down neutrons emitted from plutonium samples. The high sensitivity and efficiency in detection of thermal and slow neutrons are major merits of the 6LiI(Eu) scintillator and can play an important role in the selection of the best crystal for the border monitoring equipment.

2.4. Tests of new photomultipliers for the efficient work with LaBr3 crystal,

A further improvement of energy resolution of scintillation detectors can be achieved selecting new photodetectors. Particularly important is a development of new photomultipliers for the LaBr3 crystals. Independently, for small volume crystals, silicon drift detectors [12] and avalanche photodiodes [13] are of a great interest. A large number of primary e-h pairs has been measured with all these devices due to a high quantum efficiency of silicone.

The most of the classical PMTs do not fit to LaBr3 crystals because of a lack of a large dynamic range of anode signal. A fast light pulse of LaBr3 with 18 ns decay time constant and its high intensity, above 60000 ph/MeV, requires more than one order of magnitude larger linear range of the output current of the PMT, than that for NaI(Tl).

Independently, a large progress of the quantum efficiency of the most modern photomultipliers up to 35% for the current production and above 40% for the experimental tubes is achieved at Photonis. It is of importance for the development of scintillation detectors with a good energy resolution particularly for low energy gamma rays.

Within the carried out study numerous different photomultipliers were tested, listed in Table 3.

Table 3.

Tested photomultipliers

PMT type Blue sensitivity [μA/lm blue]

Diameter Number of dynodes/types

XP5200 10.7 51 mm 8/LF

XP5212 12.2 51 mm 9/LF

XP3422 16.3, QE = 35% Hexagonal, 60 mm 8/LF

R3998 9.2 28 mm 9/ Box and LF

XP2930 11.2 29 mm 11/ LF

XP3142 9.2 25 mm 8/LF

XP2920 11.2 29 mm 11/LF

XP2950 11.3 29 mm 11/LF

XP2960 12.5 29 mm 8/LF

Photomultipliers XP5200 and XP5212 were tested mainly with LaBr3 crystal. Results for

the XP5212 are summarized in sect. 2.1.2 presenting properties of the LaBr3 crystal. A very good linearity up to 2.5 MeV gamma rays was obtained in the test with 226Ra source. A precise measurement of the non-proportionality characteristic of LaBr3 presented in Fig. 6 confirmed a good performance of the XP5212 PMT. A high energy resolution of 3.2% was measured for 662 keV gamma rays from 137Cs source.

Fig. 20 presents the progress in the development of new photocathodes achieved recently by Photonis. The best XP3422 are equipped with the photocathode characterized by QE of about 40%.

Fig. 20. A progress in the development of photocathodes at Photonis. Courtesy of Photonis.

An excellent performance of the XP3422 PMT, associated with its high quantum efficiency is reflected in Fig. 21. It shows a comparison of energy resolution measured with NaI(Tl) crystal coupled to XP3422 and to the XP5200 PMT with a standard bialkali photocathode.

0

5

10

15

20

25

30

10 100 1000 10000

Energy [keV]

Ener

gy re

solu

tion

[%]

QE = 25%

QE = 35%

Fig. 21. Energy resolution measured with a ∅25 mm × 30 mm NaI(Tl) crystal coupled to the XP5200 of a standard QE and to the XP3422 with QE of 35%.

Note a significant improvement of energy resolution in the low energy region due to a high QE of the XP3422 PMT. It is reflected in the resolution of 9.3% for 59.6 keV gamma rays from 241Am source.

Fig. 22 presents the energy spectrum of 662 keV gamma rays from a 137Cs source measured with LaBr3 crystal coupled to the XP3422 PMT with quantum efficiency of 42%. Note an excellent energy resolution of 2.8% measured with the crystal from the first batch of Saint Gobain production.

0 500 1000 1500 2000 2500 3000

0

500

1000

1500

2000

2500

3000

3500

Num

ber o

f cou

nts

Channel number

PMT XP3422, QE - 42% 137Cs LaBr3

2.8%

Fig. 22. The energy spectrum of 662 keV gamma rays from a 137Cs source measured with LaBr3 crystal coupled

to the XP3422 PMT with quantum efficiency of 42%.

Tests of the small PMTs were addressed mainly to the linearity problem in gamma spectrometry with LaBr3 crystal. The most of the tests were done observing a detector response to 511 keV and 1274.5 keV gamma lines from a 22Na source. In each case a deviation of the measured 1274.5 keV peak position was presented. All photomultipliers were tested with the tapered voltage dividers to increase the linearity range of the PMTs and their gain was adjusted by HV to get the same 511 keV peak position. Table 4 summarizes collected data.

Table 4. Number of photoelectrons and non-linearity of tested small PMTs as measured with Ø25 mm x 25 mm LaBr3

PMT type Number and type of

dynodes HV [V]

Phe number [phe/MeV]

Non-linearity [%]

R3998 9/ Box and LF 600 12450 1.74

XP2930 11/ LF 670 14000 0.6

XP3142 8/LF 950 10400 0.03

XP2920 11/LF 490 13100 0.35

XP2960 8/LF 890 10430 0.27

Table 4 shows a better performance of the photomultipliers with the linear-focused dynode structure. The R3998 PMT, with first three dynodes built as the box and grid, showed the poorest non-linearity, close to 2%. The XP3142, with 8 stages, exhibits an excellent linearity of the response. It is related to the highest voltage used for this PMT, close to 1000 V, which assures sufficiently high inter-dynode voltage for a good collection of electrons.

The importance of the tapered voltage divider has to be pointed out, not shown in the Table 4. With the linear voltage divider, the non-linearity of the most of PMTs is much poorer because of the space charge effect in the last stages. Moreover the importance of the reduced number of dynodes is easily observed.

Based on the above study, one can formulate the requirements addressed to PMTs for LaBr3 scintillators. To get a good linearity in the large dynamic range of energy, the linear focused PMTs with a reduced number of dynodes down to 8 or even 7 stages is required, working with the tapered voltage divider. It will increase the inter-stage voltage reducing the space charge effect in the PMTs.

A further improvement of energy resolution can be achieved using PMTs with high quantum efficiency of order of 35%. It is generally important to use PMT with a larger diameter of the photocathode than the external diameter of the crystal. 2.5. Tests of avalanche photodiodes in scintillation detection

The results of the first tests of Large Area Avalanche Photodiodes are presented in Figs. 23 and 24. Measurements were done with ∅9 mm x 9 mm CsI(Tl) and ∅10 mm x 10 mm NaI(Tl) crystals coupled to the 16 mm in diameter LAAPD from Advanced Photonix, Inc. A high number of electron-hole pairs produced by scintillation light in APD allowed getting a high energy resolution not only for 662 keV gamma rays but also for low energy gammas. Fig. 22 presents the energy spectrum of 662 keV gamma rays from a 137Cs source measured with CsI(Tl) crystal. Note a high energy resolution of 4.8%.

Fig. 23. Energy spectrum of 662 keV gamma rays from a 137Cs source measured with CsI(Tl) crystal coupled to

the LAAPD.

Fig. 24 presents the spectra of low energy gamma rays measured with NaI(Tl) crystal coupled to the LAAPD.

Fig. 24. Energy spectra from 241Am and 57Co sources measured with NaI(Tl) crystal coupled to a 16 mm LAAPD.

The best energy resolutions of 11.3% and 8.4%, respectively, were obtained at the

LAAPD gain of 100 and 0.5 µs shaping time constant in the spectroscopy amplifier. The energy threshold of the collected spectra is below 10 keV. These results are comparable to those measured with the best scintillators coupled to a photomultiplier. 3. Conclusions

The performed studies were addressed to characterize new scintillators for the border monitoring equipment. It has covered:

• A study of new LaCl3:Ce and LaBr3:Ce scintillators showing a superior energy resolution in gamma spectrometry.

• A study of new heavy CWO and CaWO scintillators with the efficiency of gamma ray detection comparable to that of BGO scintillator.

• A study of 6LiI(Eu) crystal in thermal neutrons and gamma rays detection using a modern detection systems.

• The complementary study of photomultipliers for future LaBr3 detectors with the highest performance.

The study showed: • A high energy resolution of LaCl3:Ce scintillator above 100 keV gamma rays. For

lower energies NaI(Tl) crystal is still superior because of the excess of the light yield at theses energies. A high linearity of the new crystal response should be pointed out.

• The superior performance of LaBr3 crystal in the whole range of gamma-rays energy reflected in its high selectivity and good detection efficiency, better than that of NaI(Tl) crystal. A high linearity of the new crystal response should be pointed out too, which makes simpler isotope identification.

• A good energy resolution and high detection efficiency for gamma rays of small CWO

and CaWO crystals. However, a larger crystal of 20 mm x 20 mm exhibited a degradation of the light output and energy resolution. In the case of both crystals further efforts are necessary to get large volume detectors of comparable performances.

• A high detection sensitivity for thermal neutron detection of 6LiI(Eu) crystal and its a

high selectivity against gamma ray background. No doubt that it can be used in the compact handheld neutron monitors replacing He-3 detectors.

• The comparative study of different detectors confirmed the superior energy resolution

of LaCl3:Ce and LaBr3:Ce scintillators and the superior detection efficiency of CWO and CaWO crystals. However, in all the cases further efforts are needed to increase their volume and to reduce price.

• Particularly, the comparative study of LaBr3 and CZT detectors of comparable size

showed a better energy resolution of LaBr3. A poor charge collection in a large CZT limits still obtainable energy resolution. More efforts are necessary for a further development of larger volume CZT detectors with an energy resolution similar to that measured with small detectors.

• However, a further work is necessary to select and to optimize photomultipliers for

LaBr3, assuring particularly a good linearity of the PMT response.

• To utilize full capabilities of LaBr3 detector, the photomultipliers with a reduced number of linear-focused dynodes to 7 or 8 stages and characterize by high quantum efficiency of about 35% is required.

REFERENCES [1] R. Arlt, J. Brutscher, R. Gunnik, V. Ivanov, K. Parnham, S.A. Soldner, and J. Stein,

"Use of CdZnTe Detectors in Hand-Held and Portable Isotope Identifiers to Detect Illicit Trafficking of Nuclear Material and Radioactive Source," Nuclear Science Symposium Conference Record, 15–20 Oct. 2000, IEEE vol. 1, pp. 4/18–4/23.

[2] E.V.D. van Loef, P. Dorenbos, C.W.E. van Eijk, “High-energy resolution scintillator Ce3+ activated LaCl3”, Appl. Phys. Lett., 77, 2000, 1467–1468.

[3] E.V.D. van Loef, P. Dorenbos, C.W.E. van Eijk, K. Kramer, H.U. Gudel, “High-energy-resolution scintillator: Ce3+ activated LaBr3”, Appl. Phys. Lett., 79, 2001, 1573–1575.

[4] C.L. Melcher, R.A. Manente, J.S. Schweitzer, “Applicability of Barium Fluoride and Cadmium Tungstate Scintillators for Well Logging”, IEEE Trans. Nucl. Sci., Vol. 36, No. 1, pp. 1188–1192, Febr. 1989.

[5] Yu.G. Zdesenko, et al, Nucl. Instr. Meth. A538 (2005) 657.

[6] M. Moszyński, “Inorganic scintillation detectors in γ-ray spectrometry”, Nucl. Instr. Meth., A505, 2003, 101–110.

[7] M. Moszyński, "Energy resolution of scintillation detectors," in Proceedings of SPIE Hard X-Ray and Gamma-Ray Detector Physics VII (SPIE, San Diego, California, 2005) Vol. 5922, edited by R.B. James, L.A. Franks, and A. Burger, 592205-1.

[8] M. Moszyński, J. Zalipska, M. Balcerzyk, M. Kapusta, W. Mengeshe, J.D. Valentine, “Intrinsic energy resolution of NaI(Tl)”, Nucl. Instr. Meth., A484, 2002, 259–269.

[9] W. Mengesha, T.D. Taulbee, B.D. Rooney and J.D. Valentine, “Light Yield Nonproportionality of CsI(Tl), CsI(Na), and YAP”, IEEE Trans. Nucl. Sci., 45, 1998, 456–460.

[10] B.W. Sturm, Z. He, E. Rhodes, T.H. Zurbuchen, and O.L. Koehn, "Coplanar grid CdZnTe detectors for space science applications," in Proceedings of SPIE Hard X-Ray and Gamma-Ray Detector Physics VI (SPIE, Bellingham, WA, 2004) Vol. 5540, edited by A. Burger, R.B. James, and L.A. Franks, 0277-786X/04$15.

[11] P. Dorenbos, J.T.M. de Haas, and C.W.E. van Eijk, "Gamma Ray Spectroscopy with a ∅19 × 19 mm3 LaBr3:0.5%Ce3+ Scintillator," IEEE Trans. Nucl. Sci., vol. 51, no. 3, pp. 1289–1296, June 2004.

[12] C. Fiorini, A. Longoni, F. Peroni, C. Labanti, P. Lechner, L. Struder, “Gamma Ray Spectroscopy with CsI(Tl) scintillator coupled to Silicon Drift Chamber”, IEEE Trns. Nucl. Sci., 44, 1997, 2553–2560.

[13] M. Moszynski, M. Szawlowski, M. Kapusta, M. Balcerzyk, “Large Area Avalanche Photodiodes in scintillation and X-ray detection”, Nucl. Instr. Meth., A485 (2002) 504.

1

Analysis of Procedures of Testing the Technical Means for Detection, Localization and Identification

of Nuclear Materials and Radioactive Substances

G. Yakovlev Russian Research Center Kurchatov Institute, Moscow

Abstract The existing procedures of testing and performance monitoring of fixed radiation portal monitors

were considered. From analysis of the failure statistics in the conditions of the real custom

control, the least fault-tolerant (crucial) elements/units were revealed for all (pedestrian,

vehicular, and railway) types of the devices. The averaged mean lifetime of elements/units of the

fixed radiation portal monitors and the causes of failure were determined. The possibility of

increasing the total lifetime of the devices by preventive replacement or duplication of items was

studied. Based on the above examinations, the recommendations on how to increase the total

lifetime and reliability of the fixed radiation portal monitors were worked out.

The procedures of testing the hand-held devices and technical characteristics of the device s

(including ANSI, IEC, ISO, Russian specifications/test procedures, IAEA Specifications and

Test Procedures, IRAP, US Customs — RADTAP, etc.) were analysed. On the basis of the

Russian and foreign standards, a Unified Test Procedure was developed for testing hand-held

devices that are used for qualitative evaluation of parameters of radioactive materials during the

customs control of the fissile materials and radioactive substances. The specific features of the

customs control procedure for nuclear and radioactive materials crossing the border, effective in

the system of customs agencies of the RF were taken into account, as well as the current practice

of using portable instruments in the customs control technology.

2

SUMMARY ....................................................................................................................................4 1. MODERN STRATEGIES IN CUSTOMS CONTROL OF RISK GOODS..............................6 2. PROCEDURE FOR CUSTOMS CONTROL, AIMED AT PREVENTION OF ILLICIT

TRAFFICKING OF NUCLEAR MATERIALS AND RADIOACTIVE SUBSTANCES ACROSS THE BORDER, EFFECTIVE WITHIN THE CUSTOMS SYSTEM| OF RUSSIA ................................................................................................................................7

3. ANALYSIS OF THE FAILURE STATISTICS OF FIXED RADIATION PORTAL MONITORS. RECOMMENDATIONS ON INCREASING THE TOTAL LIFETIME AND RELIABILITY OF THE FIXED RADIATION PORTAL MONITORS.........................9 3.1. General requirements to installation of stationary radiation portal monitors

at various border checkpoints for customs control purposes ..............................................9 3.2. Overview of Stationary Portal Monitors ...........................................................................11

3.2.1. YANTAR STATIONARY CUSTOMS FISSILE AND RADIOACTIVE MATERIAL DETECTION SYSTEM (Aspect Research and Production Center, Russia, Dubna) ............................................................................................11

3.2.2. URK-RM5000 STATIONARY RADIATION MONITORING SYSTEM (Polimaster Joint Venture, Minsk, Belarus).............................................................12

3.2.3. СDM 002 STATIONARY RADIATION MONITORING SYSTEM (SAPHIMO, France)................................................................................................13

3.2.4. PUMA-TM STATIONARY RADIATION MONITORING SYSTEM (USA, NucSafe) .......................................................................................................13

3.2.5. STATIONARY RADIATION MONITORING BICRON APM GN (BICRON, USA, BICRON).....................................................................................14

3.2.6. KSAR 1U.041 STATIONARY RADIATION MONITORING SYSTEM (Nuclear and Technical Research Center Scientific and Technical Center, Saint-Petersburg, Russia).........................................................................................14

3.2.7. RIG-08 N STATIONARY RADIATION MONITORING SYSTEM (SNIIP-KOMVEL Research and Engineering Center, Moscow, Russia) ...............15

3.2.8. SPEKTR RADIATION MONITORING & METAL DETECTION STATIONARY COMBINED SYSTEM (GUP Dedal, Moskovskaya oblast, Russia) .....................................................................................................................17

3.2.9. VM – 250 STATATIONARY RADIATION MONITORING SYSTEM (TSA Systems Ltd, USA.) .......................................................................................17

3.2.10. JPM-32A STATIONARY RADIATION MONITORING SYSTEM (Canberra Industries, Inc., USA) .............................................................................18

3.3. Basic technical characteristics and general design............................................................18 3.4. Basic Principles of the FRM Operation ............................................................................21 3.5. Main Modes of Operation of a Typical FRM System.......................................................23 3.6. Research of FRM detection equipment reliability ............................................................23

3.6.1. Systemic approach to the research of reliability indicators to analyze functionality and performance features of stationary FRM detection equipment.................................................................................................23

3.6.2. Stationary detection systems failure statistics in the Russian Federation................24 3.7. Testing of stationary portal monitors aimed at the identification of the most

critical elements (units) .....................................................................................................27 3.7.1. Development of reliability test plan and methodology............................................27 3.7.2. Reliability proof tests of standard FRM detection systems. Duration of

reliability test ...........................................................................................................30 3.7.3. Outcomes of reliability proof testing of FRM detection systems............................31 3.7.4. Failures of individual units of the FRM over the regions during the period

of 2004–2005 ...........................................................................................................31 3.7.5. Probable FRM faults and remedies..........................................................................33

3

3.8. Recommendations on improvement of radiation monitor reliability through implementation of group-shared kit of spare parts and accessories..................................34

3.8.1. General strategies in development of Spare Parts and Accessories Kit (SPAK).....34 3.8.2. Assessment of required number of units to be uncluded to the group

shared SPAK............................................................................................................35 3.9. CONCLUSIONS...............................................................................................................36

4. DEVELOPMENT OF UNIFIED TEST-PROCEDURE FOR TESTING THE HAND-HELD DEVICES ................................................................................................37 4.1. Current practice of using portable equipment in the customs control procedure..............37 4.2. Analysis of standards and specifications for testing the hand-held devices .....................38 4.3. Unified procedure for testing the hand-held devices ........................................................50

4.3.1. Evaluation Category 1: functional evaluation of an instrument ..............................50 4.3.2. Evaluation Category 2: False identification (to be performed first of all prior to

detection tests) .........................................................................................................53 4.3.3. Evaluation Category 3: Possible detection of objects, irradiating gamma-quanta ..54 4.3.4. Evaluation Category 4: Possible detection of objects, irradiating neutrons ............54 4.3.5. Evaluation Category 5: Possible detection of surface radioactive alpha- and beta-

contamination...........................................................................................................55 4.3.6. Evaluation Category 6: Requirements on isotope library........................................55 4.3.7. Evaluation Category 7: Possible identification of one isotope (shielded) ...............55 4.3.8. Evaluation Category 8: Possible identification of more than one isotope

simultaneously .........................................................................................................56 4.3.9. Evaluation Category 9: Possible identification of an isotope under the conditions

of increased radiation gamma-background..............................................................56 4.3.10. Evaluation Category 10: Ergonomical properties of the instrument face panel. .....57 4.3.11. Topic 1: Evaluation of control elements..................................................................57 4.3.12. Topic 2: Evaluation of display.................................................................................57 4.3.13. Topic 3: Evaluation of work ....................................................................................57

4.4. Adoptation of the Unified Procedure for testing the hand-held devices ...........................58 4.5. CONCLUSION .................................................................................................................61

APPENDIX 1. Format, rules and procedure for filling out the register to track the time of operation, failures and damages to equipment items during testing.........................................62

APPENDIX 2. The RADTAP Program (USA)............................................................................65 APPENDIX 3. Protocol for the evaluation of portable isotope identification instruments

at Los Alamos National Laboratory on May 17–19, 2004 .......................................................67

4

SUMMARY

The project “Analysis of procedures of testing the technical means for detection, localization and

identification of nuclear materials and radioactive substances” was implemented in the

framework of IAEA Research Contract # 12598\Nuclear Security Multi-donors Fund by Russian

Research Center “Kurchatov Institute” from December 01, 2004 till May 31, 2006. The project

manager is Guenrikh Yakovlev.

In the framework of the project the existing procedures of testing and performance monitoring of

fixed radiation portal monitors were considered. From analysis of the failure statistics in the

conditions of the real custom control, the least fault-tolerant (crucial) elements/units were

revealed for all (pedestrian, vehicular, and railway) types of the devices. The averaged mean

lifetime of elements/units of the fixed radiation portal monitors and the causes of failure were

determined. The possibility of increasing the total lifetime of the devices by preventive

replacement or duplication of items was studied. Based on the above examinations, the

recommendations on how to increase the total lifetime and reliability of the fixed radiation portal

monitors were worked out.

The procedures of testing the hand-held devices and technical characteristics of the device s

(including ANSI, IEC, ISO, Russian specifications/test procedures, IAEA Specifications and

Test Procedures, IRAP, US Customs — RADTAP, etc.) were analysed. On the basis of the

Russian and foreign standards, the Unified Test Procedure was developed for testing hand-held

devices that are used for qualitative evaluation of parameters of radioactive materials during the

customs control of the fissile materials and radioactive substances. The specific features of the

customs control procedure for nuclear and radioactive materials crossing the border, effective in

the system of customs agencies of the RF were taken into account, as well as the current practice

of using portable instruments in the customs control technology.

5

A number of proposals on improvement of the test procedures for portable instruments, captured

in the Unified Test Procedure, were implemented in the Project IAEA-TECDOC-XXXX

Technical/Functional Specifications for Border Radiation Monitoring Equipment.

The first version of the Unified Test Procedure was approved during the experiments in the

Russian research center “Kurchatov Institute” in 2003. After the corrections, the release version

of the Unified Test Procedure was approved during the experiments in Los Alamos National

Laboratory.

The results were reported on the «Research Co-ordination Meeting of the International Atomic

Energy Agency’s Coordinated Research Project on Improvement of Technical Measures to

Detect and Respond to Illicit Trafficking of Nuclear and radioactive Materials» held in Vienna in

2003 and 2006, and in Dagomys (Russia) in 2004.

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1. MODERN STRATEGIES IN CUSTOMS CONTROL OF RISK GOODS

Customs services of many countries are going through a new stage of their upgrade, where primary focus is shifted towards such policy goals and objectives as development of customer-oriented services, promotion of foreign trade and improvement of customs control procedures.

Globalization of the international economy, integration of national economies into the global economic environment determine an increasing role and significance of customs regulation as a management tool in foreign trade.

At the same time, one of the most important responsibilities of the customs remains to be the reduction of potential threats related to trafficking of the prohibited and restricted goods over the border.

Customs control of the goods with inherent qualities allowing their further use for production of mass destruction weapons or for nuclear and radiological terrorist attacks, shall remain the most efficient instrument in prevention of illicit trafficking of such materials, and provides the most immediate impact on the support of national security and compliance of a country with its international obligations. Escalation of international terrorism necessitates improvement of customs control effectiveness through creation of organizational, technical and legal mechanisms aimed at minimization of the above threats.

The Council of the World Customs Organization (WCO) approved the Framework of Standards to Secure and Facilitate Global Trade on June 23, 2005 at is annual session in Brussels. The above document allows moving to the new principles of the safe world trade and signifies the beginning of a radically new approach in organization of activities and interaction between customs administrations and business circles.

The WCO’s Framework of Standards starts with the following introduction: “International trade is an essential driver for economic prosperity. The global trading system is vulnerable to terrorist exploitation that would severely damage the entire global economy. As government organizations that control and administer the international movement of goods, Customs administrations are in a unique position to provide increased security to the global supply chain and to contribute to socio-economic development through revenue collection and trade facilitation. … Customs administrations have important powers that exist nowhere else in the government — the authority to inspect cargo and goods shipped into, through and out of a country. Customs also have the authority to refuse entry or exit …”

In compliance with Standard 3 of the Framework of Standards to Secure and Facilitate Global Trade, equipment for non-intrusive inspection and radiation detection shall be available and used wherever available to conduct inspections in accordance with risk assessments. Such equipment is necessary for operational search of containers and higher risk goods without interruption of the legitimate trade processes.

Physical inspection (search) of every shipment, involving opening of cargo compartments and unloading of goods would create unacceptable and unnecessary impediments.

Thorough inspection of each shipment would halt the world trade and inflict additional hardships for the foreign trade participants, as it would result in increase of time for customs clearance and downtime of transportation means, additional expenses for cargo handling operations and consignment of goods.

The current customs control technologies are based on the following key principles:

1. Facilitation of the world trade and creation of conditions for unimpeded transfer of goods and transportation means over the border by the law compliant participants of the foreign economic activities; and

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2. Customs control based on the risk management system and random inspection of goods (switch from dispersed shooting to pointed strikes).

Accordingly, criteria or conditions for selection of goods for inspection should be in place to serve as grounds for customs officials to identify the goods (vehicles, or persons) to be subject to physical inspection in the form of customs examination and/or search.

The above criteria shall be spelled out in appropriate documents describing the customs control technologies.

The radioactive and nuclear materials are known to be qualified as risk goods, because their illicit trafficking over the border impacts the national security and implies considerable political, economic and environmental consequences for a state.

Consequently, physical customs inspection involving opening of a certain object (container), search and identification of a location of radiation source in it (i.e. customs visual examination and search of an object) is a necessary technique in preventing the illicit trafficking of risk goods over the border.

Justification for such customs physical inspection is an alarm generated by the radiation monitor, i.e. identification of objects with ionizing radiation higher than the natural background in the entire flow of goods, vehicles and persons crossing the border.

This is the way the general technology of customs control is implemented when aimed at the prevention of illicit trafficking of nuclear and radioactive materials via the border:

- initial detection of NRM (identification of a suspicious object) with the help of radiation monitor;

- follow-up radiation monitoring and in-depth radiation investigation to search, locate and identify the NRM in the object.

As it is seen from the above, such technology fully complies with the principles of selective customs control and supports the risk management system, where radiation monitors are trusted with the role of main instruments to ensure the selectiveness of physical inspection.

2. PROCEDURE FOR CUSTOMS CONTROL, AIMED AT PREVENTION OF ILLICIT TRAFFICKING OF NUCLEAR MATERIALS AND RADIOACTIVE

SUBSTANCES ACROSS THE BORDER, EFFECTIVE WITHIN THE CUSTOMS SYSTEM OF RUSSIA

In the process of the customs control, aimed at prevention of illicit trafficking of nuclear materials and radioactive substances across the customs border of the country, the authorized customs personnel successively executes the following steps:

• - The customs observation with the help of technical means (primary radiation monitoring);

• - The customs external inspection with the help of technical means (additional radiation monitoring of an object without opening it);

• - The customs inspection with the help of technical means (additional radiation monitoring and close radiation investigation).

The purpose of the customs observation with the help of technical means (primary radiation monitoring) is to detect objects with an increased level of ionizing radiation, as compared with the natural radioactive background, in the overall traffic of goods and vehicles subject to the customs control.

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The continuous customs observation with the help of technical means is performed at the border crossing points, at the goods delivery areas, as well as during the actual control of goods shipped as international post items.

The customs observation (primary radiation monitoring) is performed by authorized customs personnel with the help of stationary radiation monitoring equipment, or, in case of its absence or failure, with the help of portable radiation monitoring detectors

A stable non-false operation of a detector, confirmed by a repeated measurement, serves as a criterion for classification of an object subject to control (a transportation vehicle, a storage room, a package containing goods, luggage, etc.) as an object with an increased level of ionizing radiation

The following serves as a justification for performing additional radiation monitoring of an object without opening it:

• - The customs observation results (primary radiation monitoring); • - Obtaining of indirect indications as to the presence of nuclear and radioactive materials

(radiation danger signs on goods packages, specific protective containers, massive lead structures, etc.);

• - Obtaining of rapid information, results of checking of transportation and/or commercial documentation.

In the process of additional radiation monitoring of an object without opening it, an authorized customs officer:

• - Determines points (areas) with the maximum intensity of ionizing radiation on an object surface;

• - Measures radiation characteristics of an object, including the levels of surface contamination with alpha- and beta-radiating radionuclides;

• - Determines the degree of radiation danger of an object. During the customs inspection with the opening of goods package or vehicle cargo areas or bays, containers or any other areas where the cargo items are or may be located, an authorized customs officer executes:

a) additional radiation monitoring in order to search for and localize an ionization radiation source (IRS) within the cargo item, measurement of its radiation characteristics and evaluation of the degree of radiation danger;

b) close radiation investigation in order to achieve maximum possible localization, initial identification of the IRS, and preliminary classification of the IRS within one of the following groups:

• Nuclear materials or products made of them; • Radioactive substances or products made of them; • Radioactive waste materials; • Goods with an increased content of radionuclides (scrap metal, mineral raw materials,

civil engineering materials, etc.). The additional radiation monitoring and close radiation investigation are performed by authorized customs personnel, certified to handle ionization radiation sources.

If, in the process of the customs observation, an event of stationary equipment operation in the neutron channel is recorded, the additional radiation monitoring should be performed with the help of portable equipment, provided with neutron radiation detectors.

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The goods and transportation vehicles, with an increased level of ionization radiation detected in the process of the customs control, should be located at a certain section of the customs control zone, specified by the customs, taking all the measures necessary to ensure radiation safety.

The opening of goods packages or cargo bays of vehicles or vessels, containers or any other areas where cargo items could be located, as well as any other activities on the search for and localization of a source of ionization radiation should be performed with the use of individual protection means. Before the beginning of the above-mentioned work, levels of surface contamination by alpha- and beta-radiating radionuclides should be measured on the surface of each object under investigation.

The opening of detected sources of ionization radiation, resembling containers for transportation of nuclear and radioactive materials by their external characteristics (in the form of cylinders, hermetically sealed vessels, tubes, flasks, cans, etc.) is not allowed in the customs control zone. The above activities can be implemented only in the process of expert investigation at specialized enterprises.

In order to make a decision about classifying the detected source of ionization radiation as nuclear or radioactive materials, an expert investigation is assigned.

3. ANALYSIS OF THE FAILURE STATISTICS OF FIXED RADIATION

PORTAL MONITORS. RECOMMENDATIONS ON INCREASING THE TOTAL LIFETIME AND RELIABILITY OF THE FIXED RADIATION PORTAL

MONITORS

3.1. General requirements to installation of stationary radiation portal monitors at various border checkpoints for customs control purposes

Type of radiation monitor Purpose of use Place of installation

VEHICLE CHECK POINT (VEHICLE BORDER ENTRY) Pedestrian type with gamma

and neutron detectors. Initial radiation monitoring of drivers,

passengers, and their luggage. Customs passengers control zones (customs clearance for

persons). Vehicle type with gamma

and neutron detectors. Initial radiation monitoring of vehicles

and goods they transport. Vehicle entrance to restricted

access area of checkpoints (usually behind a drop bar). The number depends on lane

width and availability of individual lanes for trucks, cars,

and buses. INTERNATIONAL AIRPORT CHECK POINT

Pedestrian type with gamma and neutron radiation

detectors

1. Initial radiation monitoring of airplane crews, passengers, and their

luggage. 2. Initial radiation monitoring of

international mail.

Customs passengers control zones (customs clearance for persons), VIP halls, customs

clearing of aircraft crews, international mail exchange

Vehicle type with gamma and neutron detectors.

Initial radiation monitoring of goods liable to customs control and

transported to and from airport cargo terminals by vehicles.

Entrances to/exits from airport cargo terminals.

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SEA CHECK POINT (SEA PORT OPEN FOR INTERNATIONAL PASSENGER AND CARGO TRAFFIC)

Pedestrian type with gamma and neutron radiation

detectors

Initial monitoring of ship crews, passengers, and luggage.

Seaport zones for customs control over passages and

luggage. Vehicle type with gamma

and neutron radiation detectors.


transported to and from airport cargo and container terminals by vehicles.

Vehicle entrance/exits to cargo and container terminals of

seaports

Railway type with gamma and neutron detectors


transported to and from airport cargo and container terminals by vehicles.

Train entrance/exits to cargo and container terminal of seaports

RAILWAY CHECK POINT ( BORDER RAILWAY STATION) Pedestrian type with gamma


Initial radiation monitoring of train crew, and passengers and their

luggage.

In areas for customs inspection of passengers (customs

clearance). Railway type with gamma


Initial radiation monitoring of trains as well as transported cargo.

At a railway entrance to restricted access areas of check

points.

In addition to monitors the system will include data acquisition and processing equipment (including control unit, video surveillance system, and data and control server, operator’s workstation) to implement the following tasks:

1. Integration of stationary radiation monitors into a single response system.

2. Recording, storage and display of information on alarms generated by radiation monitors.

3. Recording, storage, and display of video data on events that triggered radiation monitors.

The typical control point and the possible location of the fixed portal monitors are displayed in Fig. 1.

Fig.1. Typical control point and the possible location of the fixed portal monitors.

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3.2. Overview of Stationary Portal Monitors

3.2.1. YANTAR STATIONARY CUSTOMS FISSILE AND RADIOACTIVE MATERIAL DETECTION SYSTEM (Aspect Research and Production Center, Russia, Dubna)

Yantrar system is designed to detect movement of fissile and radioactive materials (FRM) over the border. The system works 7 days a week 24 hours a day in automatic mode and generates light and sound alarms if FRMs are detected in the system coverage.

Yantar family products include several versions of systems such as a pedestrian system (Yantar -1P and Yantar -2P), a vehicular system (Yantar -1A and Yantar -2A), and a railway system (Yantar -1G and Yantar 2G).

Yantar systems vary by the number of elements included in each version (e.g. number of pillars), the types of detected radiation, size of detection zones, minimal quantities of detected materials, and by other features and specifications (Table 1).

Differences in Yantar Systems

Model ID Application Detection Channels The maximum width of

monitored lane (in meters)

Yantar – 1P Pedestrian Gammas and neutrons 1.5

Yantar – 2P Pedestrian Gammas and neutrons 1.5 Yantar – 1A Vehicle (trucks) Gammas and neutrons 8 Yantar – 2A Vehicle (trucks) Gammas and neutrons 4 Yantar – 1G Railway Gammas and neutrons 10 Yantar – 2G Railway Gammas and neutrons 5

Minimal mass values of various nuclear materials (in grams) detectible by Yantar system

System type Plutonium-239 Uranium-235 Uranium-238 Plutonium -239 shielded (4 cm lead)

Yantar – 1P 1 10 100 40 Yantar – 2P 1 10 160 30

Yantar – 1G» 20 2500 7000 400 Yantar – 1A 10 1000 6000 100

Minimal activity values of various nuclear materials detectible by Yantar systems, in kBq (µCu)

System type Cs137 Co60 Ra226 Th232 K40 Yantar – 1С» 420(12) 118(3,2) 177(4,7) 180(5,1) 2,360(64) Yantar – 1P 40 (1.1) 56(1.5) 43(1.2) 43(1.2) 1,133(31) Yantar – 2P 11(0.3) 7.4(0.2) 7.4(0.2) 7.4(0.2) 444(12) Yantar – 1G 900(24) 300(8) 400(11) 450(12) 5,400(146) Yantar – 1A 300(8.1) 91(2.5) 136(3.6) 143(3.9) 1,800(4.9)

Composition:

- gamma-channel: plastic scintillating detectors, each with a sensitive volume of 4,000 cm3; and

- neutron channel: 3 helium-based neutron detectors, each with a sensitive volume of 800 cm3

False alarms do not exceed 1/1000;

Operating mode –24 hours a day 7 days a week;

Data display unit can be positioned at 2,000 m maximum from the detectors;

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Operating temperature range: -50 - +50 С0;

Power supply: 220V, 50 Hz;

Power consumption – does not exceed 60 W;

Time of continued operation with use of a built-in chargeable battery is 10 hours at least;

Weight:

• Yantar 1A: 700 kg • Yantar 1P: 200 kg; • Yantar 2 P: 500 kg; • Yantar 1G: 1,200 kg;

Dimensions: • Yantar 1A: 370 мм х 600mm х 2,987 mm; • Yantar 1P 250 mm х 520 mm х 1,670 mm; • Yantar 2P 610 mm х 1,190 mm х 2,340 mm; • Yantar 1G 500 mm х 2,220 mm х 2,400 mm;

Warranty: 24 months;

Service life: 12 years.

3.2.2. URK-RM5000 STATIONARY RADIATION MONITORING SYSTEM (Polimaster Joint Venture, Minsk, Belarus)

URK-RM5000 Radiation Monitoring System is designed to monitor movement of radioactive and nuclear materials over the borders of secure areas. The system consists of independent units to assemble eight types (models) of radiation monitors. Gamma detection unit – an organic plastic scintillating detector with a sensitive volume of ∼4,500 cm3 (optionally per client’s request – 8,500 cm3). Neutron detection unit: Не-3 proportional meters with a sensitive volume of ∼3,200 cm3. Sensitivity of detection unit: - Gamma radiation:

150 (pulse/sec)/(µR/h) – for 241Am;

125 (pulse/sec)/(µR/h) – for 137Cs.

- Neutron radiation:

Up to 500 pulse cm2/neutron for Pu-α-Be,

2,000 pulse cm2/neutron for Cf.

False alarms level: no more than 1 alarm per 1,000 movements of an object crossing the area in the system coverage.

Operating temperature: -30 - +50 °С.

Operation mode: continued, 24 hours.

Power supply: Commercial power 95–250 V/AC, 50/60 Hz or built-in chargeable battery, 12 V during eight hours (in case of a blackout).

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Power consumption: shall not exceed 50 VA.

RS-232 port to connect to a personal computer.

Minimal detectible quantities of radioactive and nuclear materials Models

UPK-RM5000

Max. speed,

km/h

Max.lane

width, m

241Am,µCu

137Cs, µCu

60Co, µCu

238U, g

235U, g

239Pu,g

239Pu, g

TIGR.412151.001 10 6 150 10 4 2700 300 4,3 50.3 RM 5000 20 6 - - - - - - -

TIGR.412151.001-02 10 6 200 12 5,5 2000 250 4,2 50.3 RM 5000-02 20 6 - - - - - - -

TIGR.412151.001-03 5 1,5* 80 3 2,5 1000 35 1,2 - RM 5000-03 5 3* - - - - - - 50.3

10 3* - - - - - - - TIGR.412151.001-11 10 6 - - - - - - 50.3

RM 5000-11 20 6 - - - - - - TIGR.412151.001-14 10 3* - - - - - - -

RM 5000-14 5 3* - - - - - - 50.3

3.2.3. СDM 002 STATIONARY RADIATION MONITORING SYSTEM (SAPHIMO, France)

СDM 002 is designed to monitor movement of radioactive and nuclear materials at various checkpoints.

Composition:

- Gamma/neutron plastic scintillation detector (used as a neutron moderator

- Three He-3 neutron detectors (1000 mm)

- Data processing unit.

System Specifications:

- Thresholds of gamma radiation detection at 1 meter from detectors (exposure of 1 sec with the background 70 nGr/hour, 5 sigma of standard error ) – 30 kBq Cs-137;

- Neutron sensitivity 450 (units/sec/cm2);

- Operating temperature: – 15 - + 50 С0;

- Dimensions of gamma detector: 1,000 mm х 250 mm х 50 mm;

- Overall dimensions: 2,000 mm х 350 mm х 440 mm;

- Weight: 170 kg.

3.2.4. PUMA-TM STATIONARY RADIATION MONITORING SYSTEM (USA, NucSafe)

PUMA-TM is designed to monitor movement of radioactive and nuclear materials at various checkpoints.

Composition:

Gamma scintillating detector (NaI(Tl));

Neutron detector based on scintillating glass;

Data processing unit.

System specifications:

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- Operating temperature range: – 15 + 50 С0;

- Dimensions of gamma detector: 50 mm х 100 mm х 400 mm;

- Dimensions of neutron detector: 5,000 cm2;

- Overall dimensions: 2,600 mm х 1,430 mm х 310 mm;

- Weight: 300 kg.

3.2.5. STATIONARY RADIATION MONITORING BICRON APM GN (BICRON, USA, BICRON)

BICRON APM is designed to monitor movement of radioactive and nuclear materials at various check points.

Composition:

- Gamma scintillation detector (NaI(Tl));

- Не-3 neutron detector;

- Data processing unit.

Specifications:

- Operating temperature range: -35 - + 40 С°;

- Gamma detector dimentions: 75 mm x 75 mm;

- Neutron detector dimensions 100 cm;

- Overall dimensions: 610 mm х 630 mm х 280 mm;

- Weight: 115 kg.

Minimal gamma dose rate detected:

For 241Am – 0.1 µSv/h;

For 241Co – 0.1 µSv/h;

For 137Cs – 0.1 µSv/h with natural background gamma radiation of 0.2 µSv/h, with radiation source movement speed of 15 kmph and false alarm rate of 1/10,000;

Minimal mass of weapons-grade 239Pu detected at 2 meters from detector is 300 g where exposure time is 5 sec and false alarm rate is 1/10,000.

3.2.6. KSAR 1U.041 STATIONARY RADIATION MONITORING SYSTEM (Nuclear and Technical Research Center Scientific and Technical Center, Saint-Petersburg, Russia)

RUBEZH KSAR 1U.041 Vehicle stationary radiation monitoring system KSAR 1U.041 Vehicle System is designed to monitor movement of radioactive and nuclear materials at various checkpoints.

COMPOSITION:

- Two detecting pillars: 3,100 mm х 710 mm х 210 mm;

- Signal processing unit: 1,000 mm х 470 mm х 404 mm.

- Each detecting pillar includes:

- Eight polystyrene gamma detectors, 26,400 cm3 each

- Eight He-3 neutron detectors.

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Specifications:

System detection threshold:

- For gamma radiation: 400 g of 235U and 5 g of 239Pu;

- For neutron radiation: 100 g of 239Pu with 5 meters between pillars, with vehicle speed of 8 kmph and false alarm rate of 1/4,000;

Due to a detector heat setting unit, operating temperature ranges from -50 up to 50 Cº.

The system is certified by Gosstandart Agency of the Russian Federation, and has ASTM USA conformance certificate.

Dozor KSAR 1U.031 pedestrian stationary radiation system KSAR 1U.031 Pedestrian System is designed to monitor movement of radioactive and nuclear materials at various checkpoints.

COMPOSITION:

Two detecting pillars: 2,210 mm х 740 mm х 196 mm;

Detection zone: 2,000 mm х 800 mm.

Each detecting pillar includes:

Three polystyrene gamma detectors, 2,640 cm3 each

Two He-3 neutron detectors.

Specifications:

System detection threshold:

- For gamma radiation: 10 g of 235U and 0,3 g of 239Pu;

- For neutron radiation: 50 g of 239Pu with 0.8 meters between pillars, 0.8 sec of measurement time at mean speed of 5 kmph and 1/1,000 false alarm rate;

Operating temperatures ranges from -50 up to 50 Cº due to a detector heat setting unit.

The system is certified by Gosstandart of the Russian Federation and has certificate of conformance to requirements of ASTM USA.

3.2.7. RIG-08 N STATIONARY RADIATION MONITORING SYSTEM (SNIIP-KOMVEL Research and Engineering Center, Moscow, Russia)

RIG-08N Radiation Monitoring Systems (hereinafter referred to as systems) are designed for continuous monitoring of gamma and neutron radiation levels and for generation of alarms if the relative value of a radiation background exceeds a certain threshold. Photon ionizing radiation level is monitored by value of measured effective dose rate.

Systems can be used to monitor (detect) illicit trafficking of radioactive and fissile materials by persons and vehicles crossing a detection area.

The system is produced in two versions: RIG-08N-1 system equipped with one pillar, and RIG-08N-2 with two pillars.

Options to use:

RIG-08N-1 active at a distance of 1.5 m when speed of persons carrying target materials is 4,0 ± 0,4 kmph.

RIG-08N-2 with 0.8 m between pillars when speed of persons with target materials is 4.0 ± 0.4 kmph.

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RIG-08N-2T with 4 meters between pillars when speed of vehicles transporting target materials is 10 ± 1 kmph.

Scintillating plastic is a detector unit.

The measurement range of photon effective dose rate shall be from 0.05 µSv/h up to 3µSv/h. The allowable basic error shall be ±30 % where the confidence level is 0.95.

RIG-08 sensitivity

Fissile material, radionuclide

Unit of threshold detection

Maximal thresholds detection where the confidence level is 0.95, for various options

Option a Option b Option c Weapons-grade Plutonium-239

g 0.2 0.03 2

Weapons-grade Uranium-235

g 30 5 300

137Cs kBq 480 55 - 60Co kBq 450 50 -

False alarm rate shall not exceed one false alarm per 8 hours of continuous monitoring when radiation background does not exceed 25 µR/h.

Time of measurement is 0.125 sec.

The least period of continuous operation is 24 hours, reading instability does not exceed ±10 %.

Power consumption: does not exceed 10VA/AC

Dimensions:

Pillar: 150 mm х 1,005 mm х 75 mm;

Control unit: 225 mm х 170 mm х 85 mm.

Weight:

Pillar – 7 kg;

Control unit – 1 kg:

Service conditions:

- Ambient temperature: from -30 to +50 °С;

- Relative humidity: 95 % at +35 °С and at lower temperature, without moisture condensation

The set of equipment included in the system is as follows:

Description Quantity Detecting pillar 2

Detection control unit 1 The RIG-08H has an interface for a printer port, personal computer or inspection data reader (time of equipment switch on/off, time of alarm events, and the current background radiation when the system was triggered, monitoring of detecting unit failure or shutdown).

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3.2.8. SPEKTR RADIATION MONITORING & METAL DETECTION STATIONARY COMBINED SYSTEM (GUP Dedal, Moskovskaya oblast, Russia)

SPEKTR is designed to detect gamma radiation of fissile and radioactive materials including those shielded in various containers made from ferromagnetic and non-ferromagnetic materials. The SPEKTR is used at sites handling radioactive materials. Composition:

Two stand-alone pillars containing:

-Two plastic scintillating detectors

Distance between pillars: 0.76 m

Pillars height: 2.25 m

Pillars width: 0.61 m

Specifications:

Permissible speed of detected target movement between pillars ranges from 0.3 to 5 m/sec

Detection probability:

Metal items with weight exceeding 100 g is 0.95 at least

Minimal weight of detectible gamma sources:

Pu-239: 0.3 g

U-235: 10 g

Ambient temperature: +5 - +50 Со.

3.2.9. VM – 250 STATATIONARY RADIATION MONITORING SYSTEM (TSA Systems Ltd, USA.)

VM – 250 TSA monitors nuclear and radioactive materials transported by railway and vehicle. The system is used at site handling radioactive materials.

Composition:

Two plastic scintillating detectors with the following dimensions: 150 mm х 760 mm х 38 mm; of 17.6 liters volume; and data processing unit

Specifications:

Distance between pillars: from 3 up to 10 meters

Operating temperature range: from -35 up to +50 oC;

Dimensions of each pillar: 250 mm х 250 mm х 3,050 mm;

Weight of each pillar: 135 kg

- Stand-alone power supply provides 12 hours of operation if the system is disconnected from commercial power;

- The system has the following sensitivity parameters when detecting materials in a vehicle moving with the speed of 8 kmph at 3 m from detectors. The false alarm rate in this case will be 1/1,000 if the background radiation of 0.20 µSv/h:

- 10 g for low enriched Pu-239

- 1000 g for U-238 oxide;

- 18 µCu for Cs-137;

- 9 µCu for Co-60 .

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3.2.10 JPM-32A STATIONARY RADIATION MONITORING SYSTEM (Canberra Industries, Inc., USA)

JPM-32A Canberra monitors traffic of nuclear and radioactive materials transported by vehicles. The system is used at site handling radioactive materials. Composition:

Two stand-alone pillars containing:

-Four plastic scintillating detectors, their volume being 4,572 cm3

-Eight He-3 neutron proportional counters of 1,830mm length and 55mm diameter;

-Data processing unit

Specifications:

- Distance between pillars: from 3 to 6 meters

- Operating temperature: from -40 to +40 C

- Stand-alone power supply provides 12 hours of operation if the system is disconnected from commercial power

- The system has the following sensitivity parameters when detecting materials in a vehicle moving with the speed of 8 kmph at 3 m from detectors. The false alarm rate in this case will be 1/4,000 if the background radiation of 0.20 µSv/h:

-157 g for weapons-grade Pu-239

- 30 g for reactor-grade Pu-239.

3.3. Basic technical characteristics and general design The Fixed Radiation Monitors (FRM) are manufactured by different companies in different modifications. They differ in their purposes (pedestrians, vehicle, and railway), types of registration channels (gamma-, neutron or gamma-neutron), number of measurement racks (one or two), design, control zone dimensions, minimum quantities of detected materials and some other technical data and characteristics.

The FRM operates continuously, 24-hour per day in an automatic mode and ensures visual and tone alarms, in case the radiation level exceeds the background, the recording of the time and duration of the radiation level excess, storage and display of information about exceeding the operation threshold.

A typical SCSD NMF consists of the main and additional equipment. The main equipment consists of the following:

• A rack with detectors and electronic units of the UVK type (Fig. 2); • A console of the PVTs-01 type (Fig. 3).

The UVK-type typical rack consists of the following:

A scintillation detection unit of the BDS-G type

A neutron counter of the SN type

A module of counters power supply MPS

A neutron channel amplifier UNK

A module of the system monitoring of the MSK type

A module of charging and power supply of the MZP type

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A module of power supply and processing of the MPO type

A power supply and processing unit of the BPO type

Tone indicator of the SPEK-5 type

A presence sensor of the PARADOR type

An infrared radiator of the RS type

An infrared receiver of the RS type

A console for information processing and analysis PVTs-01

Fig. 2. A typical layout of the units in the FRM: 1) – the storage battery; 2) the neutron detector; 3) the gamma-radiation detector; 4) the power supply units and information processing unit.

Depending on the tasks to be resolved with the help of the FRM, it can be complete with additional equipment, which may include the following:

• A computer, a printer, a device for mating the channels between the USK-01 racks; • A matching unit of the BX-01 type; • Annunciators of additional visual and tone signal indication of the BOP type; • An alarm button; • Check sources of gamma- (Cs-137) and neutron (Cf-252 or Cu-244) radiation; • A system of visual display and recording of information.

From the viewpoint of its design, the FRM is manufactured in one rack (of the USK type) or in two racks (of the USK type). To ensure maximum sensitivity, the racks should be installed at the minimum distance from the controlled objects.

Currently, automated complexes of the FRM are installed at the customs control areas. In addition to the monitor racks of the UVK type with the TV observation system, a server computer (a control module) and automated working stations (ARM) (Fig. 4) are included into

1

2

3

4

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the software/hardware part of this complex. One server of the complex can support up to a few dozens of the FRM systems and transmit information to a few automated working stations (ARMs). The server/ARMs interface is implemented via a local computer network. The ARM computer ensures the work of the standard Internet Explorer program; there is no need to install any specialized software. The computer is equipped with a tone adapter and columns to ensure the alarm tone indication; the printer is connected to it to print out the protocol.

Fig. 3. The control and information processing unit of the FRM (an option without the use of the computer).

General Features for All the FRM:

• Visual and tone alarm signal indication; • Automatic adaptation to the natural background variation; • Archiving of information about the event: date, time, count rate of detectors, channel type

(gamma- or neutron); if complete with the video-recording system, can additionally record a video-clip of the alarm object;

• A gamma-detector is an organic plastic scintillator (or NaI-based scintillator); • A neutron detector is He-3-proportional counter; • Lightning protection of power lines and data transmission lines; • Possibility for combining of systems of different types into a unified information network

without additional software/hardware expenditures; • A developed system of self-diagnostics.

21

Fig. 4. The server computer (the control module) and automated work station (ARM) of the FRM.

3.4. Basic Principles of the FRM Operation The basic principle for all the FRM operation is always the same, despite the existing differences in the level of completeness, in particular, in the number of registration channels, availability or absence of a video-camera and a computer, etc.

The registration and analysis of the ionization radiation in the FRM are performed by the method of comparison of the controlled object radiation with the background radiation level at the place of installation of the monitor without direct measurement of the numerical value at a certain rated accuracy.

The detection of the nuclear materials and radioactive substances by the system is generated in the form of signals, containing no values of physical units, but confirming that the radiation of the controlled objects exceeds the levels of a certain threshold with respect to the external radiation background, processed in accordance with slightly different but still close mathematical models.

The equipment for storage and processing of information of the FRM ensures the monitoring, procession of information from all the detector units, its delivery via the main consecutive channel to the information procession unit PVTs and external devices (the computer and printer, when connected), as well as the initiation of the tone and visual alarm signals in case the established threshold is exceeded.

• The maximum length of the communication line to the console is not more than 2000m (when connecting the matching-type unit BX, up to 4000 m).

• It is possible to connect a video system to the FRM to record the transportation vehicle that initiated the system triggering on a video tape. The video system is controlled by forming a “dry-contact”-type signal with the parameters:

• Mах commutated voltage is 50 V; • Mах commutated current is 250 mA. • The system is powered from the AC mains with the voltage from 187 to 242 V 50±1 Hz

or a storage battery with the voltage from 14 to 10 V. The time of the system operation from the storage battery during the mains shutdown is, as a rule, not more than 10 hours.

• The following can be connected to the PVTs-type control console: • A computer to perform automated collection and processing of information about the

system operation;

22

• A printer to print out the data about the system triggering; • A visual-type annunciator of the BOP type, i.e. a lamp, operating in the “flash” mode,

when the FRM system is triggered. A block diagram of the UVK-type rack is given in Fig. 5.

The rack consists of 2 separate detection channels (gamma- and neutron), units for processing signals from the gamma- and neutron detectors, a controller for procession of data on the detectors status, sensors detecting the presence of an object in the controlled zone (between the racks), elements of information and signal indication.

From the viewpoint of their design, the UVK racks are made in the form of metal cabinets, housing the electronic units and protecting them from the external effects. Protective shields, used to decrease the effect of the external background on the detectors and to increase their sensitivity, are installed inside the rack around the detecting units, except for their front part. As a rule, the shield for the gamma-detectors is made of lead sheets with the thickness of about 10 mm. For neutron detectors, provision is made for shields made of borated polyethylene with the thickness up to 50–70 mm.

Fig. 5. Block diagram of the UVK-type rack.

The exchange of information between the console and racks is performed via the main consecutive channel, satisfying the requirements and recommendations of the RS-485 interface. The information is supplied by a consecutive digital code via the information transmission line (ITL) in the form of messages, using the “command-response” principle. Each message contains an address byte and the data bytes. The ITL are made in the form of a cable with a shielded twisted pair of conductors in a protective envelope.

The system has a flexible structure, with possibilities of extending the number of information channels and connection of additional peripheral devices. When operating a few racks, to automate the information processing and visualization processes, provision is made for connection of a computer and a printer.

In the absence of the PVTs, it is possible to use the device for mating of the channels RS-485 и RS-232 (USK) to control the system, using the computer keyboard, and to send the visual information to be displayed on the monitor screen.

23

3.5. Main Modes of Operation of a Typical FRM System The operation of the FRM system is based on evaluation of statistical parameters of the natural radiation background in the absence of an object in the monitored zone and their subsequent comparison with the radiation parameters, obtained in the presence of the monitored object.

The count rates of the number of pulses from the gamma- and neutron detectors are the evaluated radiation parameters. Each measurement of the count rate consists of a few intervals; the time of one measurement is defined as:

Т = m*t, Where m is the number of intervals;

t is the exposure time (duration of one measurement interval), ms.

A number of measurements of the count rate is performed in the mode of measurement of the natural background to get the sampling, on the basis of which the count rate average value is evaluated. The sampling size is defined as:

TT

M b= , where

Tb is the background measurement time.

For the sampling the following is calculated:

Evaluation of the count rate average value:

M

NNav

M

ii∑

== 1 ,

Evaluation of dispersion:

1

)(1

2

2

−

−=

∑=

M

NNS

M

icpi

, where Ni is the sampling element.

The sampling refreshment and calculation of evaluations take place upon completion of the next measurement; and the result of this measurement is added into the sampling with a simultaneous removal of the oldest in time measurement from it. As a result, provision is made for adaptation of the FRM system to the natural background variations.

3.6. Research of FRM detection equipment reliability

3.6.1. Systemic approach to the research of reliability indicators to analyze functionality and performance features of stationary FRM detection equipment

The analysis is based on the main assumption of random failure process of the multi-unit stationary FRM detection system. Random processes are best described by the probability density function: one-dimensional, two-dimensional, etc. However, it is very difficult to operate with such multi-dimensional functions for most cases, therefore, in this situation the problem was approximated, and the random failure process of the FRM detection equipment can be described with characteristics and parameters of principles describing the random process in part, rather than in full. The most important among them are mathematical expectation and dispersion.

Measurement of parameters and characteristics of random processes is simplified significantly if they are stationary (the number of hardware units in FRM detection equipment stays the same, etc.) and ergodic (parameters and specifications of hardware units in FRM detection equipment do not change, etc.).

24

Stationary processes include random processes, developing homogeneously over time, with specific functions varying around some average function with a steady magnitude. The properties of stationary processes are characterized by the following conditions:

а) mathematical expectation of a stationary random process is a constant, i.e. mx(t)=mx=const. This requirement, although, is not essential for the multi-unit systems, as one can always switch from a random X(t0 function to the centered one. The latter will have a zero level of mathematical expectation (centered random process). Therefore, if a random process of FRM detection units is not stationary only due to the mathematical expectation of a unit failure fluctuating in time, then its centering would convert it into a stationary process anyway.

b) for stationary random process dispersion by sections is a constant value, i.e. Dx(t) =Dx=const.

Practically all real electric switching and detecting random processes in FRM detection equipment refer to stationary ones. The overwhelming majority of stationary random processes possess ergodicity as their property, allowing averaging-out by the ensemble of functions with averaging-out by time of one function within an infinitely long interval Тx.

Measurement of probability of stationary ergodic processes is of interest in most cases. The main numeric characteristics of these processes would include:

• mathematical expectation (mean value) of a random process is calculated by averaging-out the values in a given function.

• dispersion of a random process determines the magnitude of fluctuations D=σ2x; Parameter σx = √Dx – is a standard error

Real characteristics of such random processes as failures of FRM detection equipment, specified as infinite limit integrals, can be found on the basis of infinitely large number of their possible functions, or, in case of ergodicity — on the basis of a function of infinite duration. In reality, there are only limited samples of failures of FRM detection equipment, i.e. an ultimate ensemble of functions or a function of a finite duration. Therefore, estimation of reliability is a task involving evaluation of characteristics of a random process of FRM detection equipment failures on the basis of finite samples.

The examples of actual samples for analysis of FRM detection equipment failures as used in this report, are samples of failures of FRM detection equipment in some regions of the Russian Federation (impact of climate conditions) and samples of failures of individual units of the FRM detection equipment systems, taken for a period of time (analysis during 4 years of observation of FRM detection equipment in operation)

3.6.2. Stationary detection systems failure statistics in the Russian Federation

Sample of FRM detection equipment failures broken down by the regions of the Russian Federation was conducted over the period of two years of observation of these systems. Reliability of the FRM detection systems was evaluated based on the four key assumptions:

• FRM detection systems are installed in customs by manufacturer's representatives, and the systems maintenance is performed by authorized trained professionals. These measures allow avoiding a human factor impact on failure statistics of FRM detection equipment.

• Some processing operations are disabled and protected by passwords as the misuse of these operations may cause the failure of FRM detection equipment during the maintenance activities. The above processes are performed by the specialists additionally trained and properly certified.

• To improve reliability of FRM detection systems and to reduce the impact of human factor on failures, all customs officers authorized to work with FRM detection systems as per their duties shall be trained at the special training courses in branches of Russian

25

Customs Academy. The courses shall include both practice and theory issues on FRM detection system functioning.

• The procedures of maintenance and operation of FRM detection systems were the same in various regions of the Russian Federation.

The results of the analysis of failure statistics of FRM Detection Systems Failure in the Russian Federation Regions over Two-Year Period from 2002 through 2005 are given in table 1.

Fig. 6 displays the total FRM equipment by regions. Total statistics of failed FRM detection systems repaired with/without replacements are given in Diagrams (Figs 7–8).

Fig. 6. FRM equipment by regions.

Table 1.

Region of activities

Type and number of FRM detection systems

Repair with replacement

of units

Repair without

replacement of units*

Ratio of repaired systems to the

total number of FRM detection

systems (with/without replacements)

North-Western region

Vehicle FRM detection systems (101) Railway FRM detection systems (41) Pedestrian FRM detection systems (143)

69 29 80

12 12 13

0.684/0.11 0.71/0.29 0.56/0.09

Western region


76 6

21

24 9

11

1.17/0.37 1/1.5

0.5/0.26 Southern

region Vehicle FRM detection systems (53) Railway FRM detection systems (32) Pedestrian FRM detection systems (49)

60 36 80

21 20 34

0.69/0.24 1.1/0.6

0.87/0.37 Urals region Vehicle FRM detection systems (32)

Railway FRM detection systems (-) 60 -

14 -

2.2/0.51 -/-

26

Pedestrian FRM detection systems (32) 28 12 1/0.42 Siberian region


99 15 12

48 4 4

1.11/0.54 2.14/0.57

0.6/0.2 Far-Eastern

region Vehicle FRM detection systems (29) Railway FRM detection systems (16) Pedestrian FRM detection systems (38)

49 27 22

27 18 13

1.1/0.6 1.8/1.2

0.4/0.24 Central region

(Moscow, Moskovskaya

oblast)

Vehicle FRM detection systems (8) Railway FRM detection systems (-) Pedestrian FRM detection systems (119)

8 -

41

6 - 9

0.6/0.46 -/-

0.31/0.07

Notes * - Replacement of rechargeable batteries, control lamps, fuses, cables, and software are not considered as replacements of units.

Fig.7. Statistics of failed FRM with replacements.

27

Fig.8. Statistics of failed FRM without replacements.

3.7. Testing of stationary portal monitors aimed at the identification of the most

critical elements (units) A special test plan was used to identify the most critical elements (units) in typical stationary portal monitors. One sample of each type of FRM detection systems (pedestrian, vehicular, and railway) was subject to proof testing.

3.7.1. Development of reliability test plan and methodology

Test plan and methodology were developed to perform reliability proof testing of stationary portal monitors.Three various versions of monitors are tested. All main components of monitor shall be subject to burn-in process in conformity with the equipment specifications.

The aim of proof testing is to evaluate conformity of equipment with requirements as specified in design documentation and to make conclusion on whether this equipment meets these requirements. Nature of failures occurred during testing shall be reviewed to develop recommendations on prevention of future failures. The test shall result in the development of recommendations in the following areas:

• Improvement of equipment reliability; • Improvement of equipment repair, maintenance and supply with spare parts; and • Improvement of design documentation and user's manual quality.

Period of Test Performance:

- In accordance with a diagram on Figure 1 but within tус=22,400 hours;

- If no failures occur, the test will last tmin=7,420 hours (310 days).

Successfully tested equipment will be shipped after replacement of the components that expired more than 75% of their service life, while appropriate records confirming replacement and actual hours in operation are made in the equipment certificate.

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The proof test shall result in evaluation of reliability level. The mean time before failure rate shall not be less than the level specified in operating manuals (e.g. 14,000hours);

The monitor is considered failed when it does not meet the following requirements:

• When test radiation source moves through the detection zone, the system shall provide detection with 0.5 probability where the confidence level is 0.95;

• The system shall provide recording, storage and transmitting of data from all system sensors and detecting units;

• When triggered, the system shall generate visual and audible alarms. Methodology of reliability proof testing

Given the limits in duration and number of items to be tested, the test program was selected to truncate the test duration and number of tested items, while retaining the risks of manufacturer and customer.

The input data for planning of reliability testing are assumed as follows:

- Number of tested items ……………………………………. N = 3 pcs;

- Acceptable mean time to failure (MTTF) .………………… Тα = 16,000 hours;

- Rejected MTFF …… ….. …………………… Тβ = 8000 ч;

- Risk of vendor (manufacturer) ………………………………α = 0,2;

- Risk of customer (user)…………………………………….. β = 0,2.

The exponential law was selected to distribute MTTF.

Notes:

During the test, one system can be replaced with its analog if required. When such replacement occurs, the sum of MTTFs is used.

Units must be recovered after a failure.

For this plan, the profile of sequential testing method was developed as shown on Fig. 9.

Figure 9. Profile of sequential reliability testing.

a) 1 – inconsistency lines;

b) 2 – consistency lines;

29

c) tус – the maximum possible non-failure operating time to consider the item consistent with reliability requirements;

d) t min – minimal time of testing if failures do not occur

(t min = 1.39 t∑/Tα);

e) rtr – truncation parameter as per number of failures (rtr =7);

f) t∑ - Total time (hours) of operation before failure since the beginning of a test, as calculated by equation:

∑t = ∑ tі, і=1

Where tі - total non-failure operating time of i-item during the test;

The system is considered consistent with reliability requirements if the graph of sequential reliability testing intersects one of the consistency lines (Line 2). The system is considered inconsistent if the graph intersects the inconsistency line (Line 1).

When reliability indicators are evaluated, all of the failures identified in testing are broken into accountable and unaccountable.

The following failures are not taken into account:

• Dependent; • Caused by external factors not envisaged by equipment specifications; • Caused by violation of operating requirements by operating/maintenance staff; • planned for elimination during design improvement process, if their efficiency is obvious

or demonstrated by further reliability test or another additional test; • Damages: when failed equipment remains functional; • Malfunctions: transient failures or once-occurring failures repaired by minor involvement

of an operator; and • Fault-tolerant status: when system can operate if some of its elements fail.

Reliability proof tests are performed in real operating conditions during all four seasons.

Safety requirements shall be met during testing procedures.

Routine maintenance shall be conducted in accordance with requirements of operating manuals.

Personnel conducting tests during working hours shall consist of two persons experienced and qualified as specified in operating manuals.

Before and after tests, and every 720 hours, each item shall be subject to functionality tests in accordance with failure criterion, as specified in this test plan.

Operating time before failure and and the time taken for equipment recovery after failures shall be recorded by authorized testing personnel in computerized Reliability Proof Testing Log where the following information is entered:

Day and time when the equipment was switched on/off;

Day and time of when failures were detected; and

Time taken to recover after each failure;

Environment factors such as temperature and humidity: the accuracy of measured temperature will be ±3 оС, humidity will be measured until the temperature exceeds +5 оС with accuracy of ±3%, with no restriction to the location of measurement instruments installation.

The log form is attached in Appendix 1.

30

All alarms shall be recorded in computer log in format as described in Appendix 1.

In the course of testing activities, the total operating hours before each failure will be calculated. All information will be shown on the sequential testing profile, as specified on Fig. 1.

If the test results achieve Line #2, the tested items are considered to meet the reliability requirements; and testing will be stopped.

If the test results achieve Line #1, the tested items are considered inconsistent with reliability requirements; and testing will be stopped.

During the test, responsible persons involved in testing maintain records in Reliability Test Logs on a computer.

Test results will be reported and approved in accordance with the established procedure, and the test protocol (report) will include the following:

• Name of equipment (system); • Name of manufacturer; • References to test plan and methodology; • List of failures and their grouping; • Causes of identified failures; • Processed test results; • Conclusions on whether the tested items meet requirements as specified; • Recommendations on equipment design modifications in order to improve or achieve the

specified reliability, and to review design documentation; and • Test profile (graph) shall be attached to the report/certificate.

The following recommendations shall be developed during the test procedure: • Improvement of equipment reliability • Improvement of equipment repair, maintenance and supply with spare parts; and • Improvement of design documentation and user's manual quality

3.7.2. Reliability proof tests of standard FRM detection systems. Duration of reliability test

Types of systems

Non-failure operating hours during testing

Total non-failure operating hours of all tested systems

Minimal test duration hours

(r=0)

Actual test duration

Pedestrian Vehicle Railway.

14,040 14,040 14,040

42,120

7,420

14,040

Functionality of each item (system) was checked before and after testing procedure, and every 720 hours, to check the failure criterion consistency. Test logs evaluation results:

Type of system Number of false alarms Number of failures and damages

Number of detected targets

1 2 3

23 18 21

20 18 19

39,923 39,916 39,903

Total 62 57 119,742

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3.7.3. Outcomes of reliability proof testing of FRM detection systems

The equipment was tested for non-failure operation during 42,120 hours. The total number of 57 failures and damages were recorded. Results of evaluation of failure causes are specified in Log for Tracking Failures and Damages during Testing Procedures. All of them can be ignored. The failures were mostly caused by equipment mishandling and significant changes in natural radiation background in the area of test performance.

3.7.4. Failures of individual units of the FRM over the regions during the period of 2004–2005

Table 2.

Region of activity in the Russian

Federation Type of FRM units

Number of repairs of

units North-West region BDS-G (the gamma-radiation detection unit)

SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

15 7 4 4 5 6 9 0 1 3 2 0 1

Western region BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

11 11 20 5 14 4 6 13 0 0 4 3 3

Southern region BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

21 4 6 1 1 12 1 0 0 1 4 0 2

32

Urals region BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

4 2 2 2 4 2 2 2 1 0 1 1 1

Siberian region BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

8 4 3 3 6 1 2 2 1 0 2 1 1

Far East region BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

5 3 2 1 2 3 1 3 0 1 1 1 1

Central region of activity (Moscow, the Moscow region)

BDS-G (the gamma-radiation detection unit) SN (the neutron radiation detection unit) MPO (the power supply and processing module) MSK (the system monitoring module) UNK (the neutron channel amplifier) MZP (the charging and power supply module) BPO (the power supply and processing unit) The storage batteries The tone annunciator The presence sensor The matching unit BX The “Alarm” lamp PVTs (the console for information processing and analysis)

7 3 4 3 1 5 1 1 0 2 1 1 1

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3.7.5. Probable FRM faults and remedies

The most likely faults and their recommended effective remedies are given in Table 3.

Table 3. Probable FRM Faults and Remedies

Fault manifestation Probable cause Remedy method

With the mains power on, the MAINS indicator fails to glow on the face panel of the MZP-type module.

а) no power supply of 220 V to the UVK rack; b) the fuse of the MZP-type module is burnt; c) the MZP-type module failed.

а) provide for the power supply; b) replace the fuse; c) replace the MZP-type module.

No signal from the presence sensor as an object drives into the monitored area.

а) the presence sensor is dirty and covered with ice; b) the PARADOR-type presence sensor is faulty.

а) wipe the optical parts of the presence sensor; б) replace the presence sensor.

With the mains power off, there is no change over to the power supply from the storage batteries.

а) the storage batteries voltage is lower than permissible; b) the MPZ-type module failed.

а) charge storage batteries; b) replace the MZP-type module.

The LOW BACKGROUND message is induced.

а) the parameters of the system operation do not comply with the required values; b) the BDS-G-type unit failed.

а) establish the required values of parameters; b) replace the BDS-G-type unit.

The HIGH BACKGROUND message is induced.

а) the parameters of the system operation do not comply with the required values; b) the BDS-G or SN unit failed.

а) establish the required values of parameters; b) replace the faulty unit.

The frequency of false triggerings is higher than the permissible one.

а) the parameters of the system operation do not comply with the required values; b) there is a change in the natural background value; c) the FRM system grounding is violated; d) the presence sensor is faulty; e) the BDS-G-type or SN-type unit failed.

а) establish the required values of parameters; b) repeat the check at a constant value of the natural background; c) restore the grounding; d) reveal and remedy the fault; e) replace the faulty unit.

The NO COMMUNICATION message is indicated

а) the communication cable is broken; b) the MPO-type power supply and processing module is faulty.

а) restore the connection of the PVTs console; b) replace the MPO-type faulty module.

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3.8. Recommendations on improvement of radiation monitor reliability through implementation of group-shared kit of spare parts and accessories

3.8.1. General strategies in development of Spare Parts and Accessories Kit (SPAK)

The required level of reliability in the process of operation of stationary portal monitors within the expected lifetime determines the need for development and delivery of SPAKs to the operational divisions. The main properties of SPAKs are as follows:

• SPAK types • List of items within a SPAK • Methods for replenishment of spare parts in SPAKs • Indicators of sufficiency of items in a SPAK • Optimal number of spare parts in SPAKs • Cost, weight and volume of SPAKs.

A group shared SPAK is meant for support of routine preventive and corrective maintenance of a suite of stationary portal monitors of the same and varying types, to be stored at the centralized facility in the region of monitor operation (the location for SPAK storage can be either a warehouse of the regional customs office, or a warehouse of the organization in charge of the routine preventive and corrective maintenance of monitors as per the agreement with a regional customs agency). A group shared SPAK should be used and spent in accordance with its declared objectives following the criteria in the operational manuals. A standard group shared SPAK shall include:

• Spare parts, • Tools, • Accessories and supplies, • Documentation, and • Packaging materials.

Spare parts are understood primarily as the following units of the stationary portal monitors:

Units Purpose Gamma radiation detector unit (GRDU) Recording of gamma rays in the controlled space and

generation of a pulse with peak value proportional to the gamma radiation energy Processing of the received data and its transmission to the power supply and processing unit

Neutron radiation detector unit (NRDU) Recording of neutron flux in the controlled space Processing of the received data and its transmission to the power supply and processing unit

Power supply and processing unit (PSPU) Data accumulation and processing Information exchange with a recording and control console Storage of the following information in the nonvolatile memory: Number and type of devices in the system System parameters Archive of events recorded during the system operation.

Power and charge modules DC power supply for the system components Switch of the system to the power supply from the chargeable batteries if the power network is lost, or of the power parameters go beyond acceptable parameters Charging of batteries

35

Presence sensors Annunciation of the movement of a controlled object to the controlled area and its departure from the controlled area

Recording and control console Specification of the operation modes and display of the system status

3.8.2. Assessment of required number of units to be uncluded to the group shared SPAK

The baseline data to be used for estimation of the required number shall include the following: • - total number of portal monitors in operation by the regional customs’ structure • - number of systems with a unit of a given type in operation by the regional customs’

structure – F; • - number of units of a given type per a specific monitor version – pedestrian, vehicular or

railway version - Kin. • - failure rate of each unit type in relation to the total number of these units in operation

for 2 years – Kfailure. In addition, the following factors shall be taken into account:

• - expected increase of the number of monitors in this region– Kgrouth • - peculiar features (specifically, the climatic conditions) of the region – Kregion. • - elevating factor to take into account the previous experience in operation of these units–

Kexp • - Knorm, to determine the needs over the future 3 years (it is 1.5). • Knew– the factor to distinguish the old and new versions of the monitors. It is Кnew=0

for the old versions, and Knew=1 for the new ones. The estimated value of the SPAK number Ncalc, required to support the operation of all previously installed monitors, as well as monitors to be installed in the next year over a 3 year period shall be determined by the formula

Ncalc = N x Kin x Kfailure/100 x Kregion х Kexp х Knorm x (1+(Kgrouth-Knew) х Knew)

Number of units Ntot, to be included in the group-shared SPAK is arrived at through rounding of Ncalc to the whole number.

Review of statistics on portal monitor units failure over the two years of operation produced the following results:

Unit % of unit repairs of the total number of these units operated as part of the portal monitors

Gamma radiation detector unit (GRDU) 2.3 Neutron radiation detector unit (NRDU) 1

Power supply and processing unit (PSPU) 1.2 Power and charge modules (PCM) 5.2

Presence sensors (PS) 3.8 Recording and control console (RCC) 8

The procedure for estimation of the SPAK contents for a specific region (one regional customs structure) over a three-year period:

36

Baseline Data

Unit Kfailure Kin Kexp Knew Knorm GRDU 2.3 4 1 1 NRDU 1 4 1 1 PSPU 1.2 1 1.2 1 PCM 5.2 2 1.2 1 PS 3.8 2 1 1

RCC 8 1 1 1

1.5

For the described region of operation: Kgrouth =1, Kregion = 1

Unit F Ncalc Ntot GRDU 7 0.97 1 NRDU 7 0.42 1 PSPU 7 0.15 1 PCM 7 1.31 1 PS 7 0.79 1 RCC 7 0.84 1

3.9. CONCLUSIONS • It is evident from the analysis of the failure statistics on the equipment subject to the

corrective maintenance where units were not replaced as a whole (minor failures are mainly meant here, where interface failures, or safety fuses, or oxidation processes are mostly involved), that the major share of such failures happen in the regions with unstable climatic conditions (Far East, Western, and Siberian regions).

• The analysis of the failure statistics on the equipment subject to the corrective maintenance where whole units have to be replaced demonstrates that the major share of the failures happen to the equipment for detection of fissile radioactive materials installed at the locations with changing climatic conditions. (Statistics on pedestrian detection systems installed in more or less comfortable locations with positive temperature gradient and low relative air humidity is lower than in vehicular and railway systems by the order of magnitude).

• Statistics of the systems subject to repairs involving replacement of the whole units reveals that the failures are caused by the stationary random processes (failures) of some equipment units.

• Sustainability of the detection systems can be supported through the creation of SPAKs of standard units and assemblies.

• It is recommended to perform preliminary diagnostics of the SRM system in order to reveal initial indications of failures of individual units (assemblies) of the SRM system in the process of periodical maintenance of the SRM system, performed by certified diagnostics centers.

• Taking into account the performed analysis of failures of the SRM system individual typical unit, it is recommended that the designers of these devices should develop a set of measures to increase reliability of those units that fail most often. It is proposed to ensure redundancy of the most important individual low-current units as one of the most efficient measures.

• It is recommended to implement systematic analysis of failures of the SRM systems operated in the IAEA participant countries in order to reveal repeated systematic failures induced by insufficient reliability of the component units and devices.

37

4. DEVELOPMENT OF UNIFIED TEST-PROCEDURE FOR TESTING THE HAND-HELD DEVICES

4.1. Current practice of using portable equipment in the customs control procedure Currently, taking into account the existing technologies of the customs control for nuclear fissile materials (NFM) in the RF, and the equipment for detection, recording and identification of NFM, developed and available for the customs, the following classification of the use of the customs control equipment can be presented:

Customs external inspection/inspection on results of Customs observation

Customs

observation Additional radiation monitoring

Close radiation investigation

Search instrument for radiation monitoring with gamma- and neutron-radiation detectors

+

Universal radiometer-spectrometer + +

Individual dosimeter +

Individual dosimeter of X-Ray and gamma-radiation

+

Universal dosimeter of X-Ray and gamma-radiation

+

Description of Customs

Operation (Forms of Customs Control)

Main Goals and Tasks Resolved Portable Instrument Used

Primary radiation monitoring (Customs observation)

Detection of objects with increased level of ionization radiation, as compared with natural radiation background, in the overall traffic of goods and transportation vehicles

Search instrument of radiation monitoring with gamma- and neutron-radiation detectors

Additional radiation monitoring of an object without its opening (within the framework of customs inspection) Additional radiation monitoring of an object with its opening (within the framework of customs inspection)

1. Search for and localization of an ionization radiation source (IRS) within an object.

2. Measurement of IRS radiation characteristics.

3. Evaluation of degree of IRS radiation danger.

- Search instrument of radiation monitoring with gamma- and neutron-radiation detectors - Dosimeters - Radiometer-spectrometer

Close radiation investigation of detected source of ionization radiation (within the framework of customs inspection)

1. IRS initial identification. 2. Preliminarily classification of IRS within one of the following groups: - nuclear materials or products made of them; - radioactive substances or products made of them; - radioactive waste products; - Goods with increased content of radionuclides (scrap metal, mineral raw products, civil engineering materials, etc.).

Radiometer-spectrometer

38

The current practice of using this equipment on the whole makes it possible to draw the following conclusions:

• The use of the radiation monitoring equipment, different in its purposes, specific applications, not always complying with the specific features of the customs control of NFM, often resulted in insufficient use of such equipment, decreasing the overall efficiency of the customs control of the NFM.

• In order to evaluate the technical characteristics and capabilities of using the above equipment in the procedures of the customs control, this equipment was subjected to different tests. At that, in view of the diverse technical capabilities of the equipment, the testing procedures also displayed considerable differences

• Problems also arose in connection with the recognition of the test results by the ministries and industries of the Russian Federation, because some of them (the “MinAtom” Ministry of Atomic Energy of the RF, the “GosSanEpidNadzor” State Sanitary Epidemic Supervision Inspection of the RF, etc.) had their own procedures for testing of this class of equipment at their disposal

• The customs personnel directly engaged in performing the customs control of goods and transportation vehicles, crossing the customs border of the Russian Federation, resolve a broad number of tasks on inspection and control of goods, the task of the customs control of the NFM being one of them. Therefore, it is not feasible to use a few types of equipment for control of the NFM in performing the NFM customs control.

By combining possible tasks to be resolved by a customs specialist in connection with the NFM, the following main tasks can be specified:

• Search and localization of the IRS within an object. • Measurement of radiation characteristics of the IRS. • Evaluation of the IRS radiation danger degree. • Implementation of preliminary identification of the IRS.

4.2. Analysis of standards and specifications for testing the hand-held devices In view of the above-mentioned tasks, currently, unified equipment for the customs control of the NFM is being designed, which will be subjected to verification and testing. In order to develop a unified procedure for testing of this class of equipment, the data captured in the following documents were analyzed in the process of implementation of the present Contract:

• Russian Specifications and Test Procedures- Monitors of NM Radiation. GOST R 51635-2000

• Standard sources of NM for radiation monitors. GOST R 52118-2003 • IAEA Specifications. • IAEA TECDOC-XXXX. • Illicit Trafficking Radiation Assessment Program (ITRAP) • ASTM Standards C 1112-93, C 1237-93 • ANSI Standards ANSI N42.32-2003, N42.33-2003, N42.34-2003. • Radiation Detection Technology Assessment Program (RADTAP) • IEC 61526. Direct reading personal dose equivalent meters. • ISO 22188 Standards. • Technical Requirement Specifications of hand-held devices

39

• Procedures for testing the technical means for detection, localization, and identification of NM and RS. Testing in RRC “Kurchtov Institute” (1995)

The summary materials of this analysis are given in Table 3. The RADTAP program differed considerably from the major part of the test procedures in its evaluation criteria; therefore, the information about this program is given separately in Appendix 3.

40

Tabl

e 3.

Id

entif

icat

ion

Rad

iois

otop

e so

urce

s

Typ

e of

in

stru

men

t T

ype

of r

ecor

ded

radi

atio

n an

d se

nsiti

vity

(d

etec

tion

and

its

prob

abili

ty)

Nuclear materials

industrial

medical

of natural origin

Climatic conditions

Power supply

Probability characteristics (false

operation)

Notes

Rep

ort o

f the

I.V

. Kur

chat

ov In

stitu

te o

f Ato

mic

Ene

rgy,

199

5 1.

Spec

trom

eter

-rad

iom

eter

of

gam

ma-

radi

atio

n ba

sed

on R

KG

-09N

(‘

KO

RA

D’)

in

stru

men

t 2.

Porta

ble

gam

ma-

spec

trom

eter

PG

S

1 D

ensi

ty o

f 137 C

s ac

tivity

in th

e ra

nge

of 0

,5-4

00 µ

Cu/

m2

2. O

n γ

in th

e ra

nge

of e

nerg

ies o

f 50

keV

– 3

.0 M

eV

235 U

, 23

8 U, P

u,

252 C

f (s

hiel

ded

by

4-cm

of

Pb),

24

1 Am

, Th

137 C

s, 60

Co,

60

Co+

13

7 Cs

Te

sts w

ith th

e go

al o

f col

lect

ing

expe

rimen

tal d

ata

to p

erfo

rm e

valu

atio

n ca

lcul

atio

n on

se

nsiti

vity

. Acc

umul

atio

n of

the

back

grou

nd d

ata,

mea

sure

men

t of s

ourc

es o

f diff

eren

t m

asse

s, in

a st

atio

nary

pos

ition

and

in m

ovin

g w

ith re

spec

t to

the

inst

rum

ents

. Pl

us th

e co

mpo

sitio

n of

137 C

s + 60

Co.

Th

e sc

ient

ific

and

engi

neer

ing

subs

tant

iatio

n of

typi

cal p

rogr

ams a

nd p

roce

dure

s for

ac

cept

ance

test

s of t

he T

KD

RM

sear

ch e

quip

men

t is a

vaila

ble.

1.

NaI

(Tl) ф

50×5

0 , e

xpos

ure

of 0

.5-5

s in

a p

ower

ful P

b sh

ield

(5 c

m),

the

tem

pera

ture

st

abili

zatio

n of

the

duct

is a

bsen

t, th

e γ-

radi

atio

n sp

ectru

m is

mea

sure

d in

128

cha

nnel

s. 2.

NaI

(Tl) ф

63×6

3 ca

n be

use

d on

a st

and-

alon

e ba

sis,

128

chan

nels

.

Tes

t Res

ults

of H

and-

Hel

d Id

entif

ier

in IA

EA

, 200

2 Sp

ectro

met

er:

GR

-135

–

2014

NaI

de

tect

or,

GR

-135

–

2580

NaI

de

tect

or,

GR

-135

–

2580

CZT

de

tect

or,

GR

-135

–

2014

NaI

de

tect

or,

MKС

-02A

- 96

/01,

fie

ldSP

EC-

CZT

-138

2,

field

SPEC

- N

aI -

1122

, fie

ldSP

EC-

NaI

-150

7

Iden

tific

atio

n in

γ 1)

reco

rded

val

ue

(thre

shol

d) o

f 0.3

4 w

ith b

ackg

roun

d of

0.

13 µ

Sv/h

2)

1.5

at 0

.13

3) 3

.7 a

t 0.1

3 4)

0.4

at 0

.15

5) 0

.5 a

t 0.1

5

Pu

Ba-

133

+ Pu

-239

1) w

arhe

ad P

u (m

ass:≈

6.6g

,93

% P

u-23

9; 6

.3 %

Pu-

240)

в P

b co

ntai

ner w

ith th

e w

all

thic

knes

s: 1

0mm

. G

R-1

35 (N

aI),

MK

S-02

and

fiel

dSPE

C-N

aI a

nd C

ZT is

/was

cap

able

of i

dent

ifyin

g Pu

-239

un

der t

hese

con

ditio

ns (s

tate

s). R

ecom

men

d to

mea

sure

tim

e: 3

00 s.

2)

reac

tor P

u (m

ass:≈

6.6

g, 6

2.5%

Pu-

239;

25.

4 %

Pu-

240)

в P

b co

ntai

ner w

ith w

all

thic

knes

s: 1

0 m

m

GR

-135

(NaI

), M

KS-

02 a

nd fi

eldS

PEC

-NaI

and

CZT

is/w

as c

apab

le o

f ide

ntify

ing

Pu-2

39

unde

r the

se c

ondi

tions

(sta

tes)

. Rec

omm

end

to m

easu

re ti

me:

100

s.

3) re

acto

r Pu

free

in th

e ai

r. 4)

reac

tor P

u fr

ee in

the

air w

ith 1

.2 m

m fi

lter o

f Cd.

5)

a m

ixtu

re o

f Ba-

133

and

Pu-2

39 (9

3 %

), en

clos

ed (c

over

ed) b

y 1.

2 m

m fi

lter o

f Cd

6) id

entif

icat

ion

time

depe

ndin

g on

radi

oact

ive

mat

eria

l and

inst

rum

ent:

30,6

0,12

0,24

0,30

0 an

d 60

0 s.

41

GR

-135

–

2014

NaI

de

tect

or,

GR

-135

–

2580

, M

KС

-02A

- 96

/01,

fie

ldSP

EC-

NaI

-112

2,

field

SPEC

-N

aI-1

507

137 C

s

1) T

he so

urce

act

ivity

is 3

70 M

Bk

and

it is

in a

con

tain

er w

ith u

nide

ntifi

ed c

hara

cter

istic

s. 2)

Mea

sure

men

ts w

ere

take

n at

dis

tanc

es o

f 40

and

25 c

m fr

om th

e so

urce

cen

tre.

3). A

ll th

e sp

ecifi

ed st

anda

rds (

rate

s) o

f the

dos

e fr

om th

ree

diff

eren

t typ

es o

f ins

trum

ents

are

in

per

fect

agr

eem

ent.

Che

ck w

ith B

AB

YLI

NE

by sh

owin

g th

e in

stru

men

t with

the

max

imum

di

ffer

ence

by

5 %

. 4)

At t

he h

igh

dose

est

imat

es th

e de

ad ti

me

of G

R-1

35, a

nd th

e M

KS

inst

rum

ent i

s so

high

th

at th

ere

are

prob

lem

s with

iden

tific

atio

n. P

ropo

sal:

1. Im

plem

enta

tion

in te

stin

g of

the

stan

dard

(rat

e) o

f the

dos

e is

pos

ition

ed to

ens

ure

poss

ibili

ty o

f ide

ntifi

catio

n.

2. T

he p

ropo

sed

uppe

r lim

it is

at l

east

100

µSv

/h.

K.E

. Duf

tsch

mid

200

3-11

-15

IAE

A T

echn

ical

/Fun

ctio

nal S

peci

ficat

ions

for

Bor

der

Rad

iatio

n M

onito

ring

Equ

ipm

ent M

inim

um R

equi

rem

ents

for

Tes

t L

abor

ator

ies D

raft

15

Nov

embe

r 20

03

Indi

cato

rs

Spec

trom

eter

s

in γ

: fr

om 1

0 nS

v/h

to

10 S

v/h

< ±

5 %

, in

n:

10 n

Sv/h

to 1

mSv

/h

< ±

30 %

For

indi

cato

rs

241 A

m

20 M

Bq

252 C

f 0.

2 M

Bq

(2–2

.104

n/s)

ur

aniu

m

oxid

e in

po

wde

r (>

90%

23

5 U)

100g

ox

ide

(6

% 24

0 Pu)

10g

For

spec

tro-

met

ers:

24

1 Am

; 10

, 104

235 U

; 1;

238 U

; 1

239 Pu

; 1;

233 U

23

2 Th

10;

HEU

ox

ide

pow

der

(> 9

0 %

23

5U)

100g

; D

U 1

00g;

For

indi

cato

rs57

Co

0.5

– 1;

137 C

s 0.

2 –

1,

100;

60

Co

0,25

,100

,10

5 M

Bq;

13

1 I 10;

22

6 Ra

103

for

spec

tro-

met

ers:

13

7 Cs;

1,

100

, 10

00;

60C

o;

0.25

, 10

, 100

, 10

5 ; 111 In

; 1;

99mTc

; 10

; 201 Tl

; 10

; 67G

a;

1; 13

3 Xe;

10

; 125 I;

100;

123 I;

1; 13

1 I; 1;

18

F; 1

; 57

Co;

1;

237 N

p;

100;

1;

133 B

a; 1

; 40

K; 1

;

- 20

° C to

+

50 °

C a

nd

a re

lativ

e hu

mid

ity o

f 90

%

Min

imum

requ

irem

ent f

or a

test

lab:

1.

A ro

om n

o le

ss th

an 1

00m

2.

2. H

igh

mea

ns o

f illu

min

atio

n of

the

stan

dard

(rat

e) o

f the

gam

ma

dose

, the

stan

dard

(rat

e) o

f th

e do

se ~

1 S

v/h

60C

o в

> 0.

5 m

. · A

clim

atic

boa

rd fo

r PR

D a

nd R

IID

test

s, co

verin

g (e

nclo

sing

) - 2

0°C

to +

50°

C w

ith >

re

lativ

e hu

mid

ity b

y 90

%.

· A ra

dar o

r IR

spee

dom

eter

for e

xact

mea

sure

men

t of a

tran

spor

tatio

n ve

hicl

e or

a p

edes

trian

sp

eed

(< ±

10

%).

· An

auto

mat

ed sy

stem

of a

trol

ley

for t

rans

fer o

f a so

urce

from

is a

ccel

erat

ed to

15

km\h

on

a di

stan

ce o

f 20

m.

· A sy

stem

of p

ositi

onin

g a

sour

ce fo

r det

erm

inin

g (th

e in

tent

) of t

he a

rea

(reg

ion)

of t

he

sear

ch, t

he v

ertic

al ra

nge

up to

4 m

., w

ith th

e le

ad o

f thi

ckne

ss c

ollim

ator

s by

5 m

m.

· A m

icro

phon

e-ba

sed

tone

ala

rm, s

ensi

tive

to th

e de

vice

for a

utom

ated

reco

rdin

g of

diff

eren

t si

gnal

s fed

by

PRD

s and

RII

Ds.

· IC

RU

PM

MA

pha

ntom

30

x 30

x 1

5 cm

. · A

rota

ry d

isk

prin

ts a

uni

t of t

estin

g of

the

busy

sens

or.

42

NU

100

g;

LEY

(<

20

%

235U

) 10

0g;

WG

U

(30-

90 %

23

5U)

100g

; W

G P

u ox

ide

ball

(< 9

0 %

23

9Pu)

10

g;

RG

Pu

(~65

%

239P

u)

10g

226 R

a; 1

; 19

2 Ir; 1

; 90

Sr/Y

; 10

; 252 C

f; 0.

2

(2.1

04 n/

s);

AN

SI №

42.3

2-20

03 a

ppro

ved

as o

f 23.

12.2

003

Pocket-size instruments

Personal dosimeter

Ener

gy ra

nge

from

20

0keV

to

1.5

MeV

. Exp

osur

e tim

e is

2 se

c. T

he

alar

m sh

ould

be

initi

ated

by

an

exce

ss o

f the

sign

al

thre

shol

d by

50

µr/h

for m

ore

than

0.5

sec.

Rad

ionu

clid

es a

re n

ot id

entif

ied.

Te

mpe

ratu

re

rang

e fr

om -

20о С

to +

50о С

.

Rec

harg

eabl

e ba

tterie

s. U

nder

ala

rm

cond

ition

s the

in

stru

men

t sh

ould

ope

rate

co

ntin

uous

ly

durin

g 30

min

utes

.

Prob

abili

ty o

f si

gnifi

cant

op

erat

ion

is

95%

. For

in

stru

men

ts

with

a d

igita

l in

dica

tion

of

the

dose

ra

te, t

he

mea

sure

men

t er

ror +

30%

, if

calib

rate

d in

Cs

-137.

The

alar

m si

gnal

type

s: to

ne, v

isua

l, vi

brat

ion.

A

cces

s to

chan

ge th

e in

stru

men

t fun

ctio

ns (c

alib

ratio

n pa

ram

eter

s, si

gnal

thre

shol

d le

vel)

shou

ld b

e cl

assi

fied

(req

uire

d to

ent

er a

pas

swor

d).

The

inst

rum

ent s

houl

d re

tain

wor

king

cha

ract

eris

tics a

fter m

echa

nica

l im

pact

s (dr

ops)

and

th

e ef

fect

of v

ibra

tion

with

in th

e ra

nge

from

10H

z to

500

Hz.

If th

e in

stru

men

t mea

sure

s the

le

vel o

f rad

iatio

n in

diff

eren

t uni

ts o

f mea

sure

men

t (µr

/h, µ

Sv/h

, µG

r/h),

then

the

dose

rate

va

lue

shou

ld b

e di

spla

yed

on th

e sc

reen

with

the

indi

catio

n of

the

units

of m

easu

rem

ent.

The

batte

ry in

dica

tor s

houl

d in

dica

te th

at th

e tim

e of

ope

ratio

n fo

r the

inst

rum

ent i

s lim

ited

to

4 ho

urs i

n th

e ab

senc

e of

ala

rms.

43

AN

SI №

42.3

2-20

03 a

ppro

ved

as o

f 23.

12.2

003

Pocket-size instruments

Instruments for evaluation of radiation danger

R

adio

nucl

ides

are

not

iden

tifie

d.

Tem

pera

ture

ra

nge

from

-20

о С to

+50

о С.

Rel

ativ

e hu

mid

ity u

p to

93

%.

The

inst

rum

ent

shou

ld re

mai

n fu

lly

oper

atio

nal a

fter

bein

g su

bjec

ted

to w

ater

sp

lash

es (s

ea

wat

er in

clud

ed).

The

inst

rum

ent s

houl

d im

med

iate

ly re

spon

d to

the

exce

ss o

f the

thre

shol

d le

vel.

Indi

catio

ns

shou

ld b

e di

spla

yed

in th

e sc

reen

with

illu

min

atio

n (p

ossi

ble

oper

atio

n un

der l

ow il

lum

inat

ion

cond

ition

s).

The

inst

rum

ent s

ize

shou

ld n

ot e

xcee

d 20сm

х10с

mх5сm

(if i

t is n

ot a

par

t of a

noth

er

inst

rum

ent).

The

inst

rum

ent w

eigh

t sho

uld

not e

xcee

d 40

0 gr

am.

The

inst

rum

ent s

houl

d be

pro

vide

d w

ith a

ring

or a

grip

allo

win

g it

to b

e re

liabl

y fix

ed.

The

inst

rum

ent s

houl

d be

mar

ked

allo

win

g th

e fr

ont a

nd th

e ba

ck p

anel

to b

e di

stin

guis

hed,

as

wel

l as t

he d

etec

tor l

ocat

ion

indi

cate

d.

The

inst

rum

ent c

asin

g sh

ould

be

mad

e of

wat

er-r

esis

tant

eas

y-to

-dec

onta

min

ate

mat

eria

ls.

Req

uire

men

ts o

n m

echa

nica

l im

pact

: the

inst

rum

ent s

houl

d re

tain

its w

orki

ng c

hara

cter

istic

s af

ter a

dro

p fr

om a

hei

ght o

f 1.5

m o

n a

conc

rete

floo

r with

any

side

. Th

e to

ne a

larm

freq

uenc

y is

100

0-40

00 H

z. In

cas

e of

an

unst

able

ala

rm, t

he si

gnal

inte

rval

sh

ould

not

exc

eed

2 se

c. T

he to

ne si

gnal

leve

l sho

uld

be n

ot le

ss th

an 8

5 dB

but

not

mor

e th

an

100

dB a

t a d

ista

nce

of 3

0 сm

from

the

inst

rum

ent.

For “

quie

t” in

dica

tion,

use

is m

ade

of

visu

al, v

ibra

tions

and

oth

er ty

pes o

f ala

rms.

Prov

isio

n sh

ould

be

mad

e fo

r a sp

ecia

l sig

nal i

ndic

atin

g ab

out t

he in

stru

men

t sta

tus,

unde

r w

hich

its n

orm

al o

pera

tion

is im

poss

ible

(the

bat

tery

low

leve

l, th

e se

nsor

failu

re, e

tc.).

Th

e ef

fect

of r

adio

freq

uenc

ies (

from

20

mH

z to

100

0 m

Hz)

shou

ld b

e el

imin

ated

.

AN

SI №

42.3

3-20

03 a

ppro

ved

as o

f 23.

12.2

003

Portable (hand-held) instruments

Instrument of type 1

Det

ectio

n an

d al

arm

in

itiat

ion

at th

e le

vel

of th

e do

se ra

te fr

om

0nнS

y/h

to

10 µ

Sy/h

. Ene

rgy

rang

e fr

om 6

0keV

to

1.33

МeV

.

Rad

ionu

clid

es a

re n

ot id

entif

ied.

Te

mpe

ratu

re

rang

e fr

om -

20о С

to +

50о С

, at

the

rela

tive

hum

idity

from

40

to 9

3%.

New

bat

terie

s:

unin

terr

upte

d w

ork

durin

g 24

hou

rs.

Und

er th

e co

nditi

ons o

f co

ntin

uous

al

arm

sign

al,

oper

atio

n du

ring

15 m

inut

es. F

or

char

ged

stor

age

batte

ries:

12

hou

rs o

f no

n-st

op w

ork

and

15 m

inut

es

unde

r the

co

nditi

on o

f tri

gger

ing

of a

ll th

e al

arm

type

s.

Th

e in

stru

men

t wei

ght s

houl

d be

less

than

4.5

5 kg

. Th

e in

stru

men

t sho

uld

be p

rovi

ded

with

mak

ing

show

ing

the

loca

tion

of d

etec

tors

insi

de th

e ca

sing

. Th

e al

arm

shou

ld b

e a

com

bina

tion

of to

ne, v

isua

l and

vib

ratio

n si

gnal

s.

The

sign

al th

resh

old

is e

stab

lishe

d w

ith th

e ac

coun

t tak

en o

f the

nat

ural

bac

kgro

und.

Th

e al

arm

is in

itiat

ed in

the

inst

rum

ent b

y th

e pr

esen

ce o

f a ra

dioa

ctiv

e m

ater

ial (

on th

e ba

sis

of th

e do

se ra

te in

crea

se).

The

inst

rum

ent s

houl

d w

ithst

and

the

win

d ve

loci

ty u

p to

5 in

ches

per

hou

r, sh

ould

be

stab

le

unde

r the

eff

ect o

f dus

t par

ticle

s at t

he sp

eed

of 1

750

ft pe

r min

. and

atm

osph

eric

pre

ssur

e fr

om 5

20 to

720

mm

Hg.

It

is fo

rbid

den

to su

bjec

t the

inst

rum

ent t

o th

e ef

fect

of w

aves

with

the

freq

uenc

y fr

om

20 M

Hz

to 1

000

MH

z.

The

disp

lay

scre

en o

f the

inst

rum

ent s

houl

d be

illu

min

ated

in su

ch a

way

that

it w

ould

po

ssib

le to

use

und

er d

iffer

ent i

llum

inat

ion

cond

ition

s. Th

e M

enu

stru

ctur

e sh

ould

be

sim

ple

and

conv

enie

nt.

Th

e in

stru

men

t sho

uld

with

stan

d a

drop

on

a co

ncre

te fl

oor f

rom

a h

eigh

t not

less

than

1m

(w

ith a

ny su

rfac

e) a

nd v

ibra

tion

with

a fr

eque

ncy

of 1

0-30

Hz

durin

g 15

min

utes

.

44

AN

SI №

42.3

3-20

03 a

ppro

ved

as o

f 23.

12.2

003

Portable (hand-held) instruments

Instrument of type 2

Det

ectio

n an

d al

arm

in

itiat

ion

at th

e le

vel

of th

e do

se ra

te fr

om

1µSy

/h to

1 S

y/h.

Th

e en

ergy

rang

e fr

om 6

0keV

to

1.33

MeV

.

Rad

ionu

clid

es a

re n

ot id

entif

ied.

Te

mpe

ratu

re

rang

e fr

om -

20о С

to +

50о С

, at

the

rela

tive

hum

idity

from

40

to 9

3%.

New

bat

terie

s:

unin

terr

upte

d w

ork

durin

g 24

hou

rs.

Und

er th

e co

nditi

ons o

f co

ntin

uous

al

arm

sign

al,

oper

atio

n du

ring

15 m

inut

es. F

or

char

ged

batte

ries:

12

hou

rs o

f no

n-st

op w

ork

and

15 m

inut

es

unde

r the

co

nditi

on o

f ac

tuat

ion

of a

ll th

e al

arm

type

s.

Th

e in

stru

men

t wei

ght s

houl

d be

not

mor

e th

an 2

.7 k

g.

The

inst

rum

ent s

houl

d be

pro

vide

d w

ith m

akin

g sh

owin

g th

e lo

catio

n of

det

ecto

rs in

side

the

casi

ng.

The

alar

m sh

ould

be

a co

mbi

natio

n of

tone

, vis

ual a

nd v

ibra

tion

sign

als.

The

sign

al th

resh

old

is e

stab

lishe

d in

depe

nden

t of t

he n

atur

al b

ackg

roun

d.

The

inst

rum

ent i

s use

d to

eva

luat

e ra

dioa

ctiv

ely

dang

erou

s and

pot

entia

lly d

ange

rous

car

go

item

s und

er th

e co

nditi

ons o

f an

incr

ease

d le

vel o

f ion

izat

ion

radi

atio

n.

The

tone

ala

rm fr

eque

ncy

from

100

0 to

400

0 H

z, in

the

even

t of a

n un

stab

le a

larm

sign

al, t

he

inte

rval

shou

ld n

ot e

xcee

d 2

sec.

Th

e le

vel o

f the

tone

sign

al in

the

ear p

hone

s is a

bout

75

dB, w

hile

at a

dis

tanc

e of

30

cm fr

om

the

inst

rum

ent,

no le

ss th

an 8

5 dB

. Th

e in

stru

men

t sho

uld

be p

rovi

ded

with

the

batte

ry st

ate

indi

cato

r. Th

e in

stru

men

t sho

uld

with

stan

d th

e w

ind

velo

city

up

to 5

inch

es p

er h

our,

be st

able

und

er th

e ef

fect

of d

ust p

artic

les a

t the

spee

d of

175

0 ft

per m

in. a

nd a

tmos

pher

ic p

ress

ure

from

520

to

720

mm

Hg.

It

is fo

rbid

den

to su

bjec

t the

inst

rum

ent t

o th

e ef

fect

of w

aves

with

the

freq

uenc

y

from

20

MH

z to

100

0 M

Hz.

Th

e in

stru

men

t sho

uld

with

stan

d a

drop

on

a co

ncre

te fl

oor f

rom

a h

eigh

t not

less

than

1m

(w

ith a

ny su

rfac

e) a

nd v

ibra

tion

with

a fr

eque

ncy

of 1

0-30

Hz

durin

g 15

min

utes

.

AN

SI №

42.3

4-20

03 a

ppro

ved

as o

f 23.

12.2

003

Pock

et-s

ize

inst

rum

ent f

or

dete

ctio

n an

d id

entif

icat

ion

The

ener

gy ra

nge

of

the

gam

ma

spec

trum

fr

om

25 k

eV to

3

MeV

. Th

e ac

tuat

ion

in th

e ne

utro

n ch

anne

l with

th

e C

f-25

2 so

urce

w

ith a

wei

ght o

f 0.

01 µ

g lo

cate

d at

a

dist

ance

of 2

5 cm

(in

this

cas

e th

e ne

utro

n ra

diat

ion

dose

rate

am

ount

s to

3 µS

y/h)

, th

e ex

posu

re ti

me

is

2 с

233 U

, 23

5 U,

237 N

p,

Pu

137 C

s 57

Co,

60

Co,

133 B

a,

192 Ir

, 20

4 Tl,

226 R

a,

241 A

m

67G

a,

51C

r, 75

Se,

99mTc

, 10

3 Pd,

111 In

, 12

3 I, 12

5 I, 13

1 I, 20

1 Tl,

133 X

e

40K

, 22

6 Ra,

23

2 Th

238 U

The

tem

pera

ture

ra

nge

from

-2

0°C

to

+50°

C, w

ith

the

rela

tive

hum

idity

up

to

93 %

.

Uni

nter

rupt

ed

wor

k po

wer

ed

from

a n

ew

batte

ry is

not

le

ss th

an

2 ho

urs.

The

batte

ry st

ate

indi

cato

r sho

uld

be a

vaila

ble

on

the

disp

lay

scre

en. I

t sho

uld

not b

e di

ffic

ult

to re

plac

e ba

tterie

s und

er

field

con

ditio

ns.

The

inst

rum

ent

defin

itely

id

entif

ies

radi

onuc

lides

in

8 o

ut o

f 10

subs

eque

nt

test

s.

The

inst

rum

ent d

ispl

ay sc

reen

shou

ld b

e ill

umin

ated

in su

ch a

way

that

it w

ould

pos

sibl

e to

us

e un

der d

iffer

ent i

llum

inat

ion

cond

ition

s. Th

e M

enu

stru

ctur

e sh

ould

be

sim

ple

and

conv

enie

nt. A

cces

s to

the

radi

onuc

lides

libr

ary

shou

ld b

e lim

ited.

Th

e in

stru

men

t sho

uld

have

two

mod

es o

f ope

ratio

n: a

utom

atic

(con

veni

ent)

and

clas

sifie

d (f

or e

xper

ts).

Pr

ovis

ion

shou

ld b

e m

ade

for a

mea

ns o

f dat

a tra

nsfe

r fro

m th

e in

stru

men

t to

an e

xter

nal

devi

ce. P

ossi

bilit

y to

con

nect

the

pock

et c

ompu

ter t

o an

ext

erna

l pow

er su

pply

sour

ce (f

rom

12

Vol

ts).

The

inst

rum

ent s

houl

d be

cap

able

to st

ore

up to

50

unpr

oces

sed

spec

tra, c

onta

inin

g in

form

atio

n ab

out t

he d

ate

and

time,

iden

tifie

d ra

dion

uclid

e, ti

me

of sp

ectru

m a

ccum

ulat

ion,

le

vel o

f gam

ma

and

neut

ron

radi

atio

n. If

the

radi

onuc

lide

is n

ot id

entif

ied,

ther

e sh

ould

be

a gl

owin

g in

scrip

tion

info

rmin

g ab

out i

t.

It sh

ould

be

poss

ible

to h

andl

e th

e in

stru

men

t wea

ring

glov

es a

nd p

uttin

g it

into

a

poly

ethy

lene

bag

to p

rote

ct it

from

con

tam

inat

ion.

Th

e in

stru

men

t sho

uld

be c

apab

le o

f ide

ntify

ing

the

follo

win

g ra

dion

uclid

es:

- W

ithou

t enc

losu

re: 4

0К, 1

37C

s, 57Со,

60C

o, 9

9mTc

, 201

Tl, 6

7Ga,

125

I, 13

1I, 1

92Ir

, 13

3Ва,

226

Ra,

232

Th, 2

33U

, 235

U, 2

38U

, Pu

(>6%

240

Pu),

241А

m.

- B

ehin

d a

stee

l enc

losu

re b

y 5

mm

: 40К

, 137

Cs,

57Со,

60C

o, 9

9mTc

, 201

Tl, 6

7Ga,

125

I, 13

1I, 1

92Ir

, 133Ва,

226

Ra,

232

Th, 2

33U

, 235

U, 2

38U

, Pu

(>6%

240

Pu),

241А

m.

The

inst

rum

ent s

houl

d be

cap

able

to d

istin

guis

h at

leas

t tw

o ra

dion

uclid

es si

mul

tane

ousl

y.

The

inst

rum

ent s

houl

d be

cap

able

to id

entif

y ra

dion

uclid

es u

nder

the

cond

ition

s of i

ncre

ased

ga

mm

a-ra

diat

ion

by n

atur

al T

h or

incr

ease

d be

ta-r

adia

tion.

Th

e in

stru

men

t sho

uld

not i

dent

ify a

n un

avai

labl

e ra

dion

uclid

e.

The

inst

rum

ent s

houl

d be

abl

e to

with

stan

d vi

brat

ion

durin

g 15

min

. with

in th

e fr

eque

ncy

rang

e of

10-

33 H

z.

45ISO

221

88

Pock

et-s

ize

inst

rum

ents

Tr

igge

ring

of th

e ga

mm

a-ch

anne

l co

uld

be in

itiat

ed

by e

xces

s of t

he

dose

rate

val

ue o

f 1

µSy/

h at

the

time

of e

xpos

ure

from

2

sec.

Tr

igge

ring

of th

e ne

utro

n ch

anne

l ca

n be

initi

ated

by

loca

tion

of th

e in

stru

men

t at a

di

stan

ce o

f 0.2

5 m

fr

om th

e C

f-25

2 so

urce

with

a

wei

ght o

f 0.0

1 m

icro

gram

(in

this

ca

se, t

he n

eutro

n ra

diat

ion

dose

rate

am

ount

s to

3 µS

y/h)

with

the

time

of e

xpos

ure

of 1

0 se

c.

(i.e.

abo

ut

2000

0 ne

utro

ns

per s

econ

d).

Rad

ionu

clid

es a

re n

ot id

entif

ied

The

inst

rum

ent

oper

atin

g co

nditi

ons:

the

inst

rum

ent

tem

pera

ture

ra

nge

from

-1

50 С to

+4

50 С, a

t the

re

lativ

e hu

mid

ity o

f 95

%.

Stan

d-al

one

oper

atio

n of

the

inst

rum

ent

pow

ered

from

th

e st

orag

e ba

tterie

s is n

ot

less

than

12

hou

rs. T

he

batte

ry c

harg

e le

vel s

houl

d be

di

spla

yed

in

the

info

rmat

ion

pane

l.

The

inst

rum

ent

oper

atio

n pr

obab

ility

is

99%

, i.e

. on

the

aver

age,

2

failu

res p

er

days

. W

ith th

e ne

utro

n de

tect

or

subj

ecte

d to

ga

mm

a-ra

diat

ion

of

not m

ore

than

1%

of t

he to

tal

radi

atio

n, th

e co

nfid

ence

of

ala

rm

trigg

erin

g sh

ould

am

ount

to

50%

, i.e

. 6

failu

res p

er

1 ho

ur.

The

inst

rum

ent s

ize

shou

ld a

llow

it to

be

carr

ied

in a

side

poc

ket o

r on

a be

lt.

The

inst

rum

ent s

yste

m sh

ould

pro

vide

for a

djus

tmen

t of t

he a

larm

thre

shol

d le

vel.

Th

e in

stru

men

t sho

uld

not f

ail a

t a c

onsi

dera

ble

incr

ease

of t

he io

niza

tion

radi

atio

n do

se ra

te.

Prov

isio

n sh

ould

be

mad

e in

the

inst

rum

ent f

or th

ree

type

s of a

larm

sign

als:

vis

ual,

tone

, and

vib

ratio

n.

The

tone

ala

rm sh

ould

be

no le

ss th

an 3

0 dB

at a

dis

tanc

e of

30

cm fr

om th

e in

stru

men

t.

Prov

isio

n sh

ould

be

mad

e fo

r “qu

ietly

” in

form

ing

the

oper

ator

abo

ut th

e pr

esen

ce

of ra

dioa

ctiv

e m

ater

ials

. In

ord

er to

war

n th

e op

erat

or a

bout

the

dang

er, a

s wel

l as t

o lo

caliz

e th

e so

urce

, the

to

ne a

larm

sign

al sh

ould

cha

nge

prop

ortio

nally

to th

e in

crea

se o

f the

ioni

zatio

n ra

diat

ion

dose

rate

. Fa

lse

alar

ms s

houl

d no

t exc

eed

one

oper

atio

n pe

r day

(in

view

of t

he p

rese

nce

of a

n in

sign

ifica

nt a

mou

nt o

f nat

ural

Th

and

U in

the

good

s).

The

inst

rum

ent s

houl

d be

subj

ecte

d to

che

ck te

sts b

y a

spec

ializ

ed a

genc

y on

ce a

ye

ar.

ISO

221

88

Porta

ble

(han

d-he

ld)

inst

rum

ents

Trig

gerin

g of

the

gam

ma-

chan

nel

coul

d be

initi

ated

by

an

aver

age

valu

e of

the

dos

rate

of 0

.4 µ

Sy/h

at

the

time

of

expo

sure

of m

ore

than

3 se

c.

Trig

gerin

g of

the

neut

ron

chan

nel

can

be in

itiat

ed b

y lo

catin

g a

sour

ce

Rad

ionu

clid

es a

re n

ot id

entif

ied.

Th

e in

stru

men

t op

erat

ing

cond

ition

s:

the

inst

rum

ent

tem

pera

ture

ra

nge

from

-1

50 С to

+4

50 С, w

ith

the

rela

tive

hum

idity

of

95%

.

Stan

d-al

one

oper

atio

n of

the

inst

rum

ent

pow

ered

fr

om th

e st

orag

e ba

tterie

s is

not l

ess t

han

12 h

ours

. Th

e ba

ttery

ch

arge

leve

l sh

ould

be

The

oper

atio

n pr

obab

ility

st

anda

rd is

m

ore

than

90

%, i

.e.

6 fa

ilure

s pe

r hou

r. W

ith th

e ne

utro

n de

tect

or

subj

ecte

d to

ga

mm

a-

The

inst

rum

ent s

houl

d be

pro

vide

d w

ith a

tone

indi

catin

g th

e ex

cess

of t

he d

ose

stan

dard

.

The

tone

sign

al sh

ould

be

no le

ss th

an 8

5 dB

at a

dis

tanc

e of

30

cm fr

om th

e in

stru

men

t.

It is

des

irabl

e to

hav

e a

diff

eren

t ala

rm to

ne le

vels

in c

ase

of o

pera

tion

of th

e ga

mm

a- a

nd n

eutro

n de

tect

ors.

To

be

used

in th

e se

arch

mod

e, th

e in

stru

men

t sho

uld

not w

eigh

mor

e th

an 2

kg

and

be p

rovi

ded

with

a c

onve

nien

t han

dle

for c

arry

ing.

Th

e in

stru

men

t sho

uld

allo

w m

easu

rem

ents

to b

e ta

ken

durin

g a

shor

t per

iod

of

time

(less

than

1 s)

, in

orde

r to

quic

kly

insp

ect s

urfa

ces,

bags

, ped

estri

ans a

nd

trans

porta

tion

vehi

cles

. In

cas

e of

det

ectio

n of

the

radi

atio

n so

urce

loca

tion,

the

alar

m si

gnal

shou

ld e

ither

be

repe

ated

aut

omat

ical

ly o

r cha

nge

the

sign

al to

ne fr

eque

ncy,

whi

ch sh

ould

46

of C

f-25

2 w

ith a

w

eigh

t of 0

.01

mic

rogr

am (i

n th

is

case

the

neut

ron

radi

atio

n do

se

rate

am

ount

s to

3 µS

y/h)

with

the

time

of e

xpos

ure

of 1

0 se

c.

(i.e.

abo

ut

2000

0 ne

utro

ns

per s

econ

d).

disp

laye

d in

th

e in

form

atio

n pa

nel.

radi

atio

n of

no

t mor

e th

an 1

% o

f th

e to

tal

radi

atio

n, th

e co

nfid

ence

of

alar

m

trigg

erin

g sh

ould

am

ount

to

50%

, i.e

. 6

failu

res

per 1

hou

r.

incr

ease

with

the

dose

rate

incr

ease

. Pr

ovis

ion

shou

ld b

e m

ade

in th

e in

stru

men

t for

adj

ustm

ent o

f the

ala

rm th

resh

old

leve

l.

The

inst

rum

ent s

houl

d be

eas

y to

han

dle.

Th

e in

stru

men

t sho

uld

be su

bjec

ted

to c

heck

test

s by

a sp

ecia

lized

age

ncy

once

a

year

. Th

e in

stru

men

t sho

uld

oper

ate

adeq

uate

ly e

noug

h at

a su

ffic

ient

ly h

igh

leve

l of

ioni

zatio

n ra

diat

ion

dose

rate

. The

hig

her t

he d

ose

rate

, the

qui

cker

the

inst

rum

ent

shou

ld tr

igge

r.

ISO

221

88

Inst

rum

ents

fo

r id

entif

icat

ion

of

radi

onuc

lides

Det

ectio

n of

ga

mm

a-ra

diat

ion

spec

tra in

the

ener

gy ra

nge

from

60

keV

to 1

.5 M

eV.

Afte

r cal

ibra

tion,

th

e ra

dion

uclid

es

(Со-

57, С

о-60

, К

-40,

Cs-

137,

A

m-2

41),

givi

ng

the

dose

rate

of

gam

ma-

radi

atio

n in

the

dete

ctor

by

0.5

µSy/

h hi

gher

th

an th

e ba

ckgr

ound

leve

l, sh

ould

be

iden

tifie

d.

If p

rovi

sion

is

mad

e fo

r det

ectio

n of

gam

ma-

radi

atio

n, th

e al

arm

sh

ould

be

initi

ated

at

the

valu

e of

0.

4 µS

y/h

of th

e do

se ra

te w

ith th

e ex

posu

re ti

me

of

3 s.

If p

rovi

sion

is

mad

e fo

r det

ectio

n

233 U

|

235 U

23

9 Pu

241 Pu

60C

o 13

7 Cs

192 Ir

22

6 Ra

241 A

m

238 Pu

18F

57C

o 67

Ga

99m

Tc

201 Tl

12

3 I 12

5 I 13

1 I 11

1 In

192 Ir

40K

22

6 Ra

232 U

23

8 U

No

requ

irem

ents

on

pow

er

supp

ly.

The

inst

rum

ent

oper

atin

g co

nditi

ons:

th

e in

stru

men

t te

mpe

ratu

re

rang

e fr

om -

150 С

to

+450 С

, with

th

e re

lativ

e hu

mid

ity o

f 95

%.

Prob

abili

ty o

f co

nfid

ent

oper

atio

n at

an

incr

ease

d le

vel o

f ga

mm

a-ra

diat

ion

is

90%

, i.e

. 6

failu

res

Poss

ible

iden

tific

atio

n of

radi

onuc

lides

with

the

help

of t

he in

stru

men

t by

non-

spec

ialis

ts.

The

inst

rum

ent s

houl

d be

cal

ibra

ted

on a

stan

d-al

one

basi

s, pr

efer

ably

, on

the

basi

s of

an

inte

rnal

sour

ce.

The

libra

ry o

f rad

ionu

clid

es sh

ould

be

prof

ound

ly st

udie

d by

the

desi

gner

, as w

ell

as e

dite

d, o

ptim

ized

and

test

ed.

47

of n

eutro

n ra

diat

ion,

the

alar

m

shou

ld b

e in

itiat

ed

in lo

catio

n of

the

Cf-

252

sour

ce

with

a m

ass o

f 0.

01 m

icro

gram

at

a d

ista

nce

of

0.25

m fr

om th

e in

stru

men

t (th

e ne

utro

n ra

diat

ion

dose

rate

in th

is

case

am

ount

s to

3

µSy/

h) w

ith th

e ex

posu

re ti

me

of

10 se

c. (i

.e. a

bout

20

000

neut

rons

pe

r sec

ond)

.

per 1

hou

r. W

ith th

e ne

utro

n de

tect

or

subj

ecte

d to

ga

mm

a-ra

diat

ion

of

not m

ore

than

1%

of t

he

tota

l rad

iatio

n,

the

conf

iden

ce

of a

larm

tri

gger

ing

shou

ld

amou

nt to

50

%, i

.e.

6 fa

ilure

s per

1

hour

.

48

G

OST

Р51

635-

2000

Rad

iatio

n M

onito

rs fo

r N

ucle

ar M

ater

ials

(G

ener

al S

pecs

for

Use

by

the

Cus

tom

s)

Porta

ble

(han

d-he

ld)

sear

ch

inst

rum

ents

fo

r rad

iatio

n m

onito

ring

1. T

rigge

ring

of

the

gam

ma-

chan

nel c

an b

e in

itiat

ed b

y tra

nsfe

r of a

C

O sa

mpl

e of

ur

aniu

m w

ith a

m

ass o

f 10

gram

or

plu

toni

um w

ith

a m

ass o

f 0.3

gra

m

to a

dis

tanc

e of

20

cm

with

the

trans

fer r

ate

of

0.5

m/s

ec.

2. T

rigge

ring

of

the

gam

ma-

chan

nel c

an b

e in

itiat

ed b

y tra

nsfe

r of a

C

O sa

mpl

e of

ur

aniu

m w

ith a

m

ass o

f 250

gra

m

or p

luto

nium

with

a

mas

s of 3

gra

m

to a

dis

tanc

e of

10

0 cm

with

the

trans

fer r

ate

of

0.5

m/s

ec .

1.

Trig

gerin

g of

th

e ne

utro

n ch

anne

l can

be

initi

ated

by

trans

fer o

f a

СО

sam

ple

of

plut

oniu

m w

ith a

m

ass o

f 10

gram

, sh

ield

ed b

y le

d w

ith a

thic

knes

s of

0.03

-0.0

5 m

to a

Rad

ionu

clid

es a

re n

ot id

entif

ied.

Th

e tim

e of

un

inte

rrup

ted

oper

atio

n w

ith a

stan

d-al

one

pow

er

supp

ly b

y st

orag

e ba

tterie

s is

not l

ess t

han

6 ho

urs.

Tem

pera

ture

ra

nge

from

-30

to +

50 С

at

the

hum

idity

of

95%

The

fals

e tri

gger

ing

freq

uenc

y is

no

t mor

e th

an 1

fals

e tri

gger

ing

per

10 m

in. o

f co

ntin

uous

op

erat

ion

of th

e in

stru

men

t. A

t tha

t, at

le

ast 1

5 tri

gger

ing

of

the

inst

rum

ent

out o

f 20

cas

es o

f th

e so

urce

ca

rryi

ng

(det

ectio

n pr

obab

ility

is

0.5

at a

co

nfid

ence

|o

f no

less

th

an 9

5%)

Dur

ing

anal

ysis

of p

relim

inar

y te

sts,

it is

allo

wed

to u

se a

ltern

ativ

e so

urce

s of

gam

ma-

radi

atio

n B

a-13

3 (7

50 k

Bk)

, neu

tron

sour

ces C

f-25

2 or

Cm

-244

(flu

x of

1.

5*10Е4

) to

cont

rol t

he d

etec

tion

thre

shol

d of

inst

rum

ents

for t

he g

amm

a-ch

anne

l.

The

uran

ium

stan

dard

: the

con

tent

of t

he m

ass f

ract

ion

of u

rani

um sh

ould

be

no

less

than

99.

75%

, oth

er im

purit

ies o

f not

mor

e th

an 0

.25%

, at t

hat,

the

cont

ent o

f ur

aniu

m-2

35 sh

ould

be

not l

ess t

han

89%

. Th

e pl

uton

ium

stan

dard

: the

con

tent

of t

he m

ass f

ract

ion

of p

luto

nium

shou

ld b

e no

less

than

98%

, oth

er im

purit

ies o

f not

mor

e 2%

, at t

hat,

the

cont

ent o

f pl

uton

ium

-239

shou

ld b

e no

t les

s tha

n 93

.5%

, plu

toni

um-2

40 –

not

mor

e th

an

6.5%

. In

add

ition

, the

Fun

ctio

nal t

est a

t the

max

imum

per

mis

sibl

e ba

ckgr

ound

of

0.25

µSy

/h fo

llow

ing

is v

erifi

ed:

49

dist

ance

of 2

0 cm

w

ith th

e tra

nsfe

r ra

te o

f 0.5

m/s

ec.

2. T

rigge

ring

of

the

neut

ron

chan

nel c

an b

e in

itiat

ed b

y tra

nsfe

r of a

СО

sam

ple

of

plut

oniu

m w

ith a

m

ass o

f 350

gra

m,

shie

lded

by

led

with

a th

ickn

ess o

f 0.

03-0

.05

m to

a

dist

ance

of 1

00 c

m

with

the

trans

fer

rate

of 0

.5 m

/sec

.

50

4.3. Unified procedure for testing the hand-held devices The logic of adopting a particular testing procedure is specified by the following:

• Specific features of the customs control procedure for nuclear and radioactive materials crossing the border, effective in the customs system of the Russian Federation (refer to Section 1).

• The current practice of the use of portable instruments within the customs control technology (refer to Section 4.1).

It follows from the analysis of the above-mentioned procedures for testing of technical means for detection, localization and identification of nuclear materials and radioactive substances that the procedures, on the whole, make it possible to perform evaluation of equipment belonging to the class of indicators (with the dosimeter functionality) and radiometers-identifiers. Using the existing experience gained in testing, the testing procedures can be subdivided into 10 sections:

1. Functional evaluation of an instrument.

2. False identification evaluation.

3. Possibility to detect objects irradiating gamma-quanta.

4. Possibility to detect objects irradiating neutrons.

5. Possibility to detect surface radioactive alpha- and beta-contamination.

6. Requirements on the isotope library.

7. Possible identification of one isotope (shielded).

8. Possible identification of more than one isotope.

9. Possible identification of an isotope under the conditions of an increased gamma-radiation background.

10. Ergonomic properties of the instrument front panel.

4.3.1. Evaluation Category 1: functional evaluation of an instrument

The instrument is analyzed for the presence of the following properties, being desirable but not mandatory.

Characteristic features of instrument:

• To be operated by one person (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004)

• It allows gamma- and neutron radiation, as well as the alpha- and beta-contamination, to be simultaneously detected with the help of an additional detector (the Use of Technical Means of Radiation Monitoring in the RF, the Rules for Transportation, IAEA, ST-1).

• The “timer/counter” mode for neutron and gamma-channels in order to improve the detection efficiency (ISO, TECDOC IAEA, the US Standard, Los-Alamos Test Program, 2004)

• Mechanically strong, intended for outdoor use within a broad range of temperatures and humidity (the instrument and its component parts working capacity should be retained at the temperature from -20 to +40 С). (TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF)

51

• An illuminated display screen to be used during the dark time, its readings can also be taken under a bright sun light (ISO, TECDOC IAEA, US Standard, RF Standard, Los Alamos Test Program, 2004)

• The battery service life is sufficient for at least 8 hours of operation (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004)

• An automatic (for a wide range of users) and a manual (for experts) modes of operation, provided with a protected toggle switch.

• The set critical parameters can be reset to the manufacturer’s settings by default; they are protected by classified access; the parameters selected by the user are preserved in case of the system is switched off in the mode for experts.

Properties of the detector equipment:

• Detectors are installed for simultaneous measurement of gamma- and neutron radiation, the detectors are integrated within one casing of the instrument (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004)

• An external detector for measurement of alpha- and beta-radiation (the Use of Technical Means of Radiation Monitoring in the RF).

• A neutron detector functional under the conditions of gamma-radiation background (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004).

Properties of the display screen and visual indication:

• An adequate size of the display; the display is reliable under the field conditions and at extreme temperatures; the time during which the display is in the “on” position can be adjusted with the help of the “always on” option; automatic adjustment of the degree of contrast, depending on temperature; limited manual adjustment, if required (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004).

• The number of pulses (count rate) per second (counts/s) for gamma-quanta and neutrons.

• Automatic scaling for graphical representation of the count rate (for example, histogram, belt-type diagram for search).

• After activation the date and time are indicated, as well as the hardware and software versions.

• Indicates the excess of the range for the dose rate > 1 µSy/h (It is proposed to initiate a tone (picked) signal in excess of the unit dose rate by more than 1 µSy/h (transportation category 1 in compliance with the “Rules for Safe Transportation of Radioactive Materials (ST-1), IAEA, Vienna (1996)».

• Indicates the residual capacity of the battery (storage battery) when using recommended batteries (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004).

Display indication in isotope identification:

• The spectrum display with the cursor functions (only in the mode for specialists).

• Indication of the dead time and maximum range of the count rate. Indication of the detector optimal count rate on the graphical display (information: “move the instrument closer or away. The instrument is installed at the optimal distance”) (the Use of Technical Means of Radiation Monitoring in the RF).

52

• The remaining time until the spectra procession is complete (back count), if used in the fixed time mode.

• Neutron count per second; total count during a selected time (only in the mode for specialists). (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004).

• Indication that the instrument is busy processing the spectrum. (ISO, TECDOC IAEA, the US Standard, Los Alamos Test Program, 2004).

• Indication of results: an isotope and category, or “an unknown isotope”, or “a weak signal”, in case identification failed. It is recommended to have indication of the identification degree of confidence. In all the cases it is possible to continue measurements without restarting the instrument.

Acoustic signals:

• Different tone search signals for gamma-quanta and neutron count. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• An increase of the frequency rate (not the tone level) of the tone signal is proportional to the count rate of gamma-quanta (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF)

• Single tone signals to indicate the neutron count.

• An acoustic alarm signal in excess of preset level of the dose rate. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF.)

• Visual and acoustic warning about the battery low voltage. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF)

• Possibility of a toneless and audible warning signal (vibration or earphones); information of the display screen if the signal is toneless. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

Characteristics of the amplifier/multichannel analizer (MCA) of the spectrometer:

• At least 1000 каналов of the analog-to-digit transformer (ADT). (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• Stabilization of energy dependence and a repeated calibration mode with the help of the K-40 isotope). (The Use of Technical Means of Radiation Monitoring in the RF).

• Maximum nonlinearity of the energy scale (after the linearization): < ± 0.25% below 200 keV, < ± 0.5% above 200 keV.

• Linearization depending on the energy range for scintillation detectors.

• Calibration correction in energies with the help of a radioactive source; preferably, a natural isotope, which is not subject to the Rules due to its activity. (The Use of Technical Means of Radiation Monitoring in the RF).

53

• A memory to store at least 30 spectra, 1000 channels per each. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• A response to a strong temperature variation: the amplifier should withstand temperature variations from the room temperature to +50/-20 degrees C (one repeated stabilization is permitted, if the instrument is used for more than 8 hours). (TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• Nonobligatory automatic saving of spectra in the automatic (for a wide range of users) mode.

In-built memory and a PC Interface:

• A format of spectral data in the ASCII.

• A PC interface with the help of a standard connection with communication means (for example, RS232, USB, IR or WIFI, the port is protected against splashes of water). (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• Standardized information saved in a file, including all the settings and diagnostics, as well as the measurement results.

• User-friendly software for the PC support in the mode for experts. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

Power supply source:

• A standard-size storage battery (rechargeable) or non-rechargeable (to be used if the internal battery is discharged completely or the charging device is faulty). (ISO, TECDOC IAEA ,the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• The preset parameters, data and time settings should not be lost, if the main battery is discharged completely or replaced.

• An automatic voltage reading, an adapter/charging device of AC, used throughout the world (not mandatory), automobile battery adapter.

• Change-over to the mains charging device, if the system is powered from the mains.

• The indication of the states: “charging” and “discharged completely”. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

• It is possible to charge during operation.

• The time of charge (up to reaching 90%) is less than the time of the battery discharge in the switched off instrument.

4.3.2. Evaluation Category 2: False identification (to be performed first of all prior to detection tests)

• Requirement: The instrument should not identify an unavailable radionuclide during operation in a stable mode with a low-radiation background of the environment. (ISO, TECDOC IAEA, the US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

54

The Method of testing: Perform identification of a radionuclide, using the instrument under the conditions of a stable background with the intensity of not more than 0.1µSy/h in the absence of any radiation source (a shielded box or shielding cover can be required to perform the test). Under these conditions not a single unavailable radionuclide should be identified. The test consists of 10 subsequent attempts, and the characteristics of the instrument are appropriate if the instrument has not identified any unavailable radionuclide in 9 cases out of 10 subsequent attempts. If a natural radionuclide, for example, К-40 has been identified, measures should be taken in order to decrease or eliminate this source before the performance of the test. If the presence of a radionuclide is anticipated, and it cannot be eliminated, the test result will be acceptable, if the anticipated radionuclide is identified.

4.3.3. Evaluation Category 3: Possible detection of objects, irradiating gamma-quanta

• Requirements: Provision should be made for correlation between the instrument sensitivity and the portal monitors sensitivity, since these portable instruments are used to localize the radioactive sources of gamma-radiation, detected by the portal monitors. The instrument should comply with the standards of ASTM C1237 of the US and GOST Р51635 of the RF on the minimum detectable quantities of nuclear materials and alternative radioactive sources. (The US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF).

The method of testing: The test should be performed with the following:

• Highly enriched uranium (HEU) in the form of a sphere with a mass of 10 g;

• Cs-137 source with the intensity of 1 µCurie;

• Co-60 source with the intensity of 2 µCurie;

• Ва-133 source with the intensity of 21 µCurie.

Each source should move along the front surface of the detector at a speed of more than 0.5 m/s; and, at the closest approach, the distance from the front surface of the detector to the source centre should be not less than 0.25 m. It is required to perform at least 20 tests for each source and the number of successful measurements should be recorded. The goal is to achieve the detection probability of at least 50% at the level of confidence of 95% (i.e., at least, 15 attempts out of 20 should be successful).

4.3.4. Evaluation Category 4: Possible detection of objects, irradiating neutrons

Requirements: Provision should made for correlation between the instrument sensitivity and the portal monitors sensitivity, since these portable instruments are used to localize radioactive sources of neutron irradiation, detected by the portal monitors. The instrument should generate an acoustic alarm signal, when located in the field of neutrons with the intensity higher than the threshold value.

• It should be taken into account that the criterion for detection of neutron radiation is an important factor, indicating that a fissile material or a neutron source has been detected. In this case, the criterion of minimum detectable value of the neutron flux is a very important value. (the RF Standard, the Use of Technical Means of Radiation Monitoring in the RF, ISO, TECDOC IAEA).

The method of testing: 1. – The instrument is carried in the neutron radiation field of the Cf-252 source, equivalent

to the flux of (15000 n/s ± 30%), generated by the Cf-252 source without a moderator

55

(i.e. a sample with a weight of about 0.1 µg), at a distance of about 100 cm from the instrument at a speed of about 0.5 m/s.

2. – Perform 20 tests and record how many times an acoustic signal was generated. The goal is to achieve the detection probability of at least 50% at the level of confidence of 95% (i.e., at least, 15 attempts out of 20 should be successful). (The US Standard, the RF Standard, Los Alamos Test Program, 2004, the Use of Technical Means of Radiation Monitoring in the RF, ISO, TECDOC IAEA).

Notes: 1. A source of 252Cf with a mass of 0.01 µg, located at a distance of about 25 cm from the

instrument, generates about 3 µSy/h (0.3 mР/hour).

2. The neutron detection sensitivity of a manual instrument strongly depends on the neutron moderator. Therefore, the test should be performed under the conditions of the absence of reflection, for example, on a thin steel plate. Use is made only of the neutron moderator being a component part of the instrument.

4.3.5. Evaluation Category 5: Possible detection of surface radioactive alpha- and beta-contamination

A criterion of permissible level of surface radioactive contamination:

• Alpha-particles of not more than 1.0 particle/сm2min;

• Beta-particles of not more than 100 particle/сm2min. (ST-1, IAEA)

4.3.6. Evaluation Category 6: Requirements on isotope library

Requirements: The instrument should have the following isotopes in its library: K-40, Co-57, Co-60, Ga-67, Tc-99m, I-125, I-131, Ba-133, Cs-137, Ir-192, Tl-201, Ra-226, Th-232, U-233, U-235, U-238, Pu [reactor-grade plutonium (> 6% Pu-240)], Am-241. At that, it is required to be taken into account that, in view of the danger of nuclear material illegal trafficking, the spectra of U-233, U-235, U-238, Pu (> 6% Pu-240), Am-241 should have a priority value from the viewpoint of appearance of possible multiplets with the rest of the nuclide library; in order to prevent that, the lines of other radionuclides located closely to the lines of nuclear materials should not be processed.

4.3.7. Evaluation Category 7: Possible identification of one isotope (shielded)

Requirements: The instrument should be capable to identify the following radionuclides, shielded by a steel shield, within 5 minutes: Co-57, Co-60, Ba-133, Cs-137, Ra-226, Th-232, U-233, U-235, U-238, Pu [reactor-grade plutonium (> 6% Pu-240), Am-241. (ISO, TECDOC IAEA, Los Alamos Test Program, 2004).

The method of testing:

1. The instrument measures each of the above-mentioned radionuclides (one by one). The radionuclide should have a nominal activity of 2 µCurie. The instrument is spaced 10 cm apart from the source (in the cases when a source with the 2 µCurie activity is not available, the detector location should be selected proceeding from the assumption that the dependence on distance of 1/r2 is implemented). A steel shield of a sufficient thickness is then located between the source and the detector until the measured intensity decreases by a factor of 2.

2. 10 tests should be performed; in the middle of test sequence it is required to recalibrate the instrument. The test is treated as successful, when the instrument correctly identifies

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an isotope, irrespective of its other readings. The goal is achieved when the instrument correctly identifies a radionuclide in 9 out 10 successive attempts.

As a rule, actual deliveries of radioactive materials are performed in protective containers made of lead. Therefore, it is extremely important to check the identification of a number of radioactive sources in the lead containers. To perform identification, the 137Cs, 226Ra, 60Co sources are to be put into lead containers with such a thickness that the measured intensity will not decrease by more than a factor of 10 (possible correct identification in case of appearance of “single” and “double” leaks) (ISO, TECDOC IAEA, Los Alamos Test Program, 2004).

4.3.8. Evaluation Category 8: Possible identification of more than one isotope simultaneously

Requirements: The instrument should be capable of identifying, at least, two radionuclides simultaneously. (ISO, TECDOC IAEA, Los Alamos Test Program, 2004).

The method of testing: 1. The instrument simultaneously measures isotopes Ba-133 and Pu (of a reactor grade),

each with the activity of about ~ 1 µCurie. The isotopes are spaced 10 cm apart from the instrument. The testing consists of 10 subsequent tests, the duration of each being not more than 5 minutes. The goal is for the instrument to correctly and simultaneously identify both the tested radionuclides in 9 out of 10 subsequent tests. Similarly as in the previous category, the instrument should be recalibrated in the middle of the testing sequence.

2. The instrument simultaneously measures isotopes Co-57 with the activity of about ~ 1 µCurie and a sphere made of HEU with a mass of 10 g. The isotopes are spaced 10 cm apart from the instrument. The testing should consist of 10 subsequent tests, the duration of each being not more than 5 minutes. The goal is for the instrument to correctly and simultaneously identify both the tested radionuclides in 9 out of 10 subsequent tests. Similarly as in the previous category, the instrument should be recalibrated in the middle of the testing sequence.

3. The instrument simultaneously measures isotopes Th-232 and Ra-226 with the activity of about ~ 1 µCurie. The isotopes are spaced 10 cm apart from the instrument. The test should consist of 10 subsequent tests, the duration of each being not more than 5 minutes. The goal is for the instrument to correctly and simultaneously identify both the tested radionuclides in 9 out of 10 subsequent tests. Similarly as in the previous category, the instrument should be recalibrated in the middle of the test sequence.

4.3.9 Evaluation Category 9: Possible identification of an isotope under the conditions of increased radiation gamma-background

Requirement: The instrument should be capable of identifying radionuclides of interest under the conditions of an increased background of gamma-radiation generated by natural thorium. (ISO, TECDOC IAEA, Los Alamos Test Program, 2004).

The method of testing: 1. The instrument is irradiated by natural thorium, the dose rate of which, measured by the

detector, amounts to 50 µRoentgen/h.

2. Put the 241Am source in such a place where the intensity measured by the detector can be increased by 0.5 µSy/h (50 µRoentgen/hour). The goal is that the detector can correctly identify 241Am in 8 out of 10 subsequent attempts; at that, none should last for more than 1 minute.

3. Put the 60Co source in such a place where the intensity measured by the detector can be increased by 0.5 µSy/h (50 µRoentgen/hour). The goal is that the detector can correctly

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identify 60Co in 8 out of 10 subsequent attempts; at that, none should last for more than 1 minute.

4.3.10. Evaluation Category 10: Ergonomical properties of the instrument face panel.

At the end of the testing program, each expert in evaluation should answer the questions in the evaluation review that have to do with the three “easy-to-handle” topics :

4.3.11. Topic 1: Evaluation of control elements

1. Is it easy to find the “On/Off” switch?

2. Are all the control elements marked?

3. Are all the marked control elements easy to read/interpret?

4. Is it easy to handle all the control elements without wearing gloves?

5. Is it possible to handle all the control elements wearing gloves?

6. Evaluate the control elements using a ten-point scale (10 – the best, 0 – unacceptable).

7. Is it possible to adjust the brightness/contrast, manually or automatically, so as to compensate for the brightness levels?

4.3.12. Topic 2: Evaluation of display

8. Is it possible to read everything with poor illumination?

9. Is it possible to read everything at high levels of illumination?

10. Is it possible to read the display readings, when it is located in a polyethylene bag?

11. Did the display contain abbreviations or icons (if not, omit the next question).

12. Is it easy to interpret or understand the abbreviations or icons?

13. Were the date and the time displayed on the screen?

14. Evaluate the control elements using a ten-point scale (10 – the best, 0 – unacceptable).

4.3.13. Topic 3: Evaluation of work

15. Did the instrument inform you about its technical status during initiation (for example, about the battery service life, if the detector is available, if it is activated, the scope of memory available, the mode of operation)?

16. Did you have to turn to the instructions more than once during the execution of the testing?

17. Was the Menu structure simple and intuitively clear?

18. Did the instrument invite you to act at any time during the testing?

19. Did the instrument generate any caution and warning messages? (If not, go over to Question 21).

20. Did the instrument submit you information about the character of the caution and warning messages and the respective order of actions to be taken?

21. Evaluate the instrument from the viewpoint of work convenience, using the ten-point scale (10 – the best, 0 – unacceptable).

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4.4. Adoptation of the Unified Procedure for testing the hand-held devices The first version of the Unified Procedure was preliminarily approved during the testing at the Kurchatov Institute in 2003 (see Figs 10). The following portable instruments were used:

1. Spectrometer “Gamma-1С/NB1”

The spectrometer energy resolution in the gamma-radiation line with the energy of 662 keV (Cs-137) is 8%. The range of energies for recorded γ-quanta is 50 to 3000 keV. The time for establishing the working mode is not more than 30 minutes. The range of working temperatures is from -20 to 50 С. The spectrometer mass is not more than 20 kg.

2. Universal Radiometer-Spectrometer RSU-01 “Signal-M”

The relative energy resolution of the spectrometer in the gamma-radiation line with the energy of 662 keV (Cs-137) is not more than 35%. The range of recorded energies of the gamma-spectrum is 200 to 3000 keV. The time for establishing a working mode is not more than 15 minutes. The range of working temperatures is from -20 to +40 С. The overall dimensions (of the electronic panel) are 180*140*75 mm. The mass (of the electronic panel) is 1.5 kg.

3. Universal Radiometer-Spectrometer MKS-А02

The relative energy resolution of the spectrometer in the gamma-radiation line with the energy of 662 keV (Cs-137) is not more than 8%. The range of recorded energies of the gamma-spectrum is 50 to 3000 keV. The time for establishing a working mode is the following: - not more than 30 minutes in the identification mode; - not more than 2 minutes in the other modes. The range of working temperatures is from -20 to +50 С. The overall dimensions are 290*160*135 mm. The mass is 3.6 kg.

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4. Search Radiometer-Spectrometer RM-1401К

The relative energy resolution of the spectrometer in the gamma-radiation line with the energy of 662 keV (Cs-137) is not more than 7%. The range of recorded energies of the gamma-spectrum is 15 to 3000 keV. The time for establishing a working mode is the following: - not more than 30 minutes in the identification mode; - not more than 2 minutes in the other modes. The range of working temperatures is from -20 to +50 С. The overall dimensions are 290*160*135 mm. The mass is 0.65 kg.

Fig. 10. The testing at the Kurchatov Institute in 2003 in Category 7: Possible Identification of One Isotope (Shielded).

After preliminary coordination and correction by the RF/US testing participants, the working version of the Unified Procedure was approved in the course of testing in the Los Alamos national laboratory executed from May 16 to May 19, 2004 (see Figs 11 and 12). The following portable instruments were used:

1. Search Radiometer-Spectrometer RM-1401К.

2. Universal Radiometer-Spectrometer MKS-А02.

3. Universal Radiometer-Spectrometer MKS-А03.

4. Radiometer-Spectrometer FieldSpec (the Target Company, Germany).

The Protocol compiled on the basis of the instrumentation testing results is given in Appendix 3.

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Fig. 11. Testing in Evaluation Category 3: Possible Detection of Objects Irradiating Gamma-Quanta.

Fig. 12. Testing in Evaluation Category 7: Possible Identification of One Isotope (Shielded).

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4.5. CONCLUSION The developed Unified Test Procedure is used for portable instruments, performing qualitative evaluation of parameters of radioactive materials during the customs control of the fissile materials and radioactive substances.

The Unified Test Procedure is developed on the basis of the Russian and foreign standards for the test procedures and technical characteristics of portable instruments.

In developing the Unified Test Procedure, specific features of the customs control procedure for nuclear and radioactive materials crossing the border, effective in the system of customs agencies of the RF were taken into account, as well as the current practice of using portable instruments in the customs control technology.

A number of proposals on improvement of the test procedures for portable instruments, captured in the Unified Test Procedure, were implemented in the Project IAEA-TECDOC-XXXX Technical/Functional Specifications for Border Radiation Monitoring Equipment, July, 2004.

The experimental work on defining qualitative characteristics of the isotopes of nuclear materials and radioactive substances confirmed the applicability of the developed Unified Test Procedure and necessity to use it in the process of the customs control.

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APPENDIX 1. Format, rules and procedure for filling out the register to track the time of operation, failures and damages to equipment items during testing.

(Tables А.1 and А.2) А.1. Format, Rules and Procedure for Filling Out the Table to Track the Time of Operation А.1.1. The table for tracking the time of operation during the test is filled out by a person responsible for the test, its sample and recommended form are provided in Table А.1. А.1.2. The Date and Time column shall have time when the item was switched on and off (hour, minutes, day, month, year) at the beginning of the test and after the scheduled or forced switching off. А.1.3. The Impact Factor Set shall have a daily record of temperature and humidity in the room and outside, with relative humidity outside measured at the temperature above +5 оС. А.1.4. The time of item operation after switching it on shall include the time of item operation in hours from the moment when it was last switched on till it is switched off due to any reason. By the time the item is switched off the total time is recorded in hours. А.1.5. The total Non-failure Operating Time column shall contain the total operating time (in hours) from the moment the test started till the latest switch-off. А.1.6. The Reason for Switching the Item off column shall specify the reason for switching the item off. А.2. Format, Rules and Procedure for Filling Out the Table to Track Failures and Damages А.2.1. The table for tracking the failures and damages during the test is filled out by a person responsible for the test, its sample and recommended form are provided in Table А.2. А.2.2. The Date and Time of Failure (Damage) Detection column shall have the date and time for failure manifestation (day, month, year, hour, minutes), А.2.3. The Time for Item Operation from the Moment of Switching it on before Failure (Damage) Column shall specify the time of item operation (in hours) from the moment of its scheduled switching off. А.2.4. The Manifestation of the Failure and Terms under which the Failure (Damage) was Detected during Operation or Examination Column shall include the character of a failure (what was seen at the moment of a failure, what happened), and the conditions under which it was detected – during a performance test or during scheduled functional testing. А.2.5. The Recovery Time column shall cover the time (in hours and minutes) from the moment the failure was detected to the time the item was switched back on. А.2.6. The Commission’s Opinion on the Reasons of Failure (Damage) column shall have the opinion of a commission which analyzed the reasons of a failure. Based on the commission’s findings the person responsible for the test shall prepare a report of findings in any form А.2.7. The Measures Taken column shall specify what measures were taken to eliminate and prevent similar failures in the future. The data shall be entered based on the report of findings. А.3. Sample of a filled out register:

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Register to Track the Time of Operation, Failures and Damages to Equipment Items during Testing (Format) Table А.1 – Tracking the time of equipment operation during testing

Date and Time Impacting Factors as a Set (temperature,

humidity)

Operation of Items since

Switching on

Total Non-failure

Operation Time

Reason of Switching

off the Item

Name of a person

responsible for the test

On Off Inside Outside 08.08.2005

17 Hrs 00 Min

24 оС, 65 %

15 оС 80 %

24.12.2005

12 h 00 m 3313 h 30 m 3313 h 30 m Scheduled

activities A.S. Ivnov

Further the

time is given as total

Table А.2 – The table for tracking the failures and damages during the test

Date and Time of Failure

(Damage) Detection

Time for Item

Operation from the

Moment of Switching it

on before Failure

(Damage)

Manifestation of the Failure

and Terms under which the Failure

(Damage) was Detected during

Operation or Examination

Recovery Time after Failure or Damage

Commission’s Opinion on the

Reasons of Failure

(Damage)

Measures Taken to eliminate and

prevent similar failures in the

future

20.09.2005 13 h 00 m

1,024h 30m. Malfunction of internal timer

during information exchange

3 h Software imperfection

Replacement of software. Creation of a pool of chips

with software at the consumers’ quarters

Format, Rules and procedure for filling out the register for tracking the alarms B.1 The register to track down the alarms generated by the items shall be maintained by the person responsible for the test, with recommended form and sample of a filled out register attached in Table B.1. B.1.1 The Date and Time of Alarm column shall include the date and time of the alarm В графе «Дата и время срабатывания» указать дату и время срабатывания (day, month, year, hour, minutes).

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B.1.2 The Reason of the Alarm column shall specify the reason of the item alarm – whether it was an authorized or unauthorized detection or a false alarm (reason unidentified) B.1.3 The Code of the Responding Item shall be used to record the ID number of the piece of equipment which responded to the event. B.1.4 Comment Column shall be filled as required if an operator decides that something should be further explained or clarified. Table B.1 – Register for Tracking the Alarms

Date and Time of the Alarm

Reason of the Alarm ID Number of the Responding Item, its Serial Number

Comment Name of the Operator on

Duty 15.09.2005 г.

13 h 55 m

23.11.2005 г. 03 h 10 m

Attempt to transport through the guarded area

Reason unidentified

Serial № 141-01

Serial № 008-00

Trespasser detained

False alarm

I. Petrov

I.P. Vetrov

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APPENDIX 2. The RADTAP Program (USA)

The main purpose is to evaluate the functional capabilities, sensitivity, and reliability of the existing industrial and pilot samples of the radiation control equipment.

Evaluation of possible use of this equipment for the customs purposes

Evaluated Equipment

- Personal portable instruments (pagers, etc.)

- The so-called “manual” (hand-held) instruments

- Small portal systems

- Spectrometers

Methodology

Initially, within the framework of the check tests, the minimum sensitivity was checked, as well as the sensitivity to the detection of a certain quantity of radioactive sources with the energy range 60 keV to 2.6 MeV.

After the successful implementation of the check tests, each instrument was evaluated in view of the possible detection of gamma-radiation sources within the framework of the developed scenarios. The scenarios were developed taking into consideration the real-life conditions of the customs control implementation at different check points (vehicle check points, airport, cargo terminal, a conveyer belt with post packages). The user (operational) characteristics of the equipment were evaluated in parallel.

1. QUANTITATIVE EVALUATION

1.1. Check Tests

1.1.1. Checking the Minimum Sensitivity

All the instruments (except the spectrometers) were checked for compliance with the requirements on the detection of the Cs-137 source (20 µr/hour) at a standard distance.

1.1.2. Determination of Sensitivity to Sources with Different Energy (targeted reaction)

Personal portable instruments and “manual” (hand-held) instruments were checked. The maximum distance for detection of various isotopes was defined.

Source Am-241

Co-57 Ba-133 Cs-137 Y-88 Th-232 I-131 LEU (U-235, 4.3%)

HEU (U-235,93%)

Pu-239

Activity 50 µC 212 µC 195 µC 68.2 µC

170 µC 59,7 µC

104,7 µC

2.2 mC 590 µC 63 µC

Distance

1.2. Evaluation of technical means of detection in the process of verification under real conditions of the customs control

Scenarios Adapted to Real-Life Conditions of the Customs Control:

1. Vehicle Check Point

- Passing-by inspection of automobile vehicles. With the use of personal portable and “manual” (hand-held) instruments, the detection and localization of sources hidden in a vehicle is performed.

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Sources: Co-60, LEU (4.3%), I-131, Ga-67, Tb-160, Pu-239, Cs-137, HEU (93%)

- Driving-by inspection of automobile vehicles (2-5 miles per hour). With the use of portal, “manual” and personal portable instruments the detection of sources is performed. The portal systems are installed on the motion line. The “manual” and personal portable instruments are located in the inspector place of location (a booth), which the automobile vehicles drive by.

Sources: Co-60, LEU (4.3%), I-131, Ga-67, Tb-160, Pu-239, Cs-137, HEU (93%)

2. The Airport

2.1. The Airport (Passengers and Hand-Carried Baggage)

With the use of portal, “manual” and personal portable instruments, the detection of sources carried in the passengers’ baggage is performed.

Sources: I-131, Tl-201, Co-60, Tb-160, Am-241, Ga-67, Th-232, LEU (4.3%), Pu-239

2.2. The Airport (Unattended Baggage and Cargo Items)

- With the use of portal, “manual” and personal portable instruments, the detection of sources, passing by the detectors along the conveyer belt, as a part of the passengers’ baggage or small-sized cargo items is performed.

Sources: I-131, Tl-201, Co-60, Tb-160, Am-241, Ga-67, Th-232, LEU (4.3%), Pu-239

- Detection of sources contained in the cargo items (packages, containers, etc.), located at the customs cargo terminal, during passing-by along the territory. Use is made of “manual” and personal portable instruments.

Sources: Co-57, LEU (4.3%)

Evaluation Criterion: Detection/Failed Detection

1.3. Evaluation of spectrometers in the process of verification under real conditions of customs control

Sources: LEU (4.3%), I-131, Co-60, Pu-239, Ga-67, Tb-160, Cs-137, HEU (93%), Th-232

The sources were located in different objects (baggage, a container, transportation vehicle).

A situation was developed in the procedure when the spectrometer was used at the second stage of control (after the detection and localization of a source within the object).

Evaluation Criterion: Identified/Failed Identification

2. QUALITATIVE EVALUATION

Using a 5-grade scale, the following parameters were evaluated:

Display

- Readability of information

- Readability under solar illumination

- Information clarity

Operating Characteristics

- Easy to use

- Mass and overall characteristics

- Easy to replace power supply sources

- Display of information about triggering

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APPENDIX 3. Protocol for the evaluation of portable isotope identification instruments at Los Alamos National Laboratory on May 17-19, 2004

Representatives of the Russian delegation arrived at Los Alamos National Laboratory, LANL, on May 17, 2004. We met and discussed the testing procedure for the evaluation of the hand-held isotope identification instruments.

An side-by-side evaluation of the ASPECT model MKC-A02-1M RIID, the ASPECT model MKC-A03-1 RIID, and the Polymaster model PM-1401KEKE RIID was performed at Technical Area 35 of Los Alamos National Laboratory on May 17-19, 2004. The evaluation procedure was designed to evaluate the search and identification capabilities of these detectors. For comparison purposes, the Bicron FieldSpec-N RIID was evaluated in the same manor. In this report, we will summarize:(i) the central characteristics of each detector, (ii) the results of the search tests, (iii) the results of the gamma isotope identification tests (with and without shielding), and (iv) the results of the multiple source tests. Also included in this report are general observations of "Human Interface" issues: ease of use, controls and display evaluations, etc.

1. Central Characteristics. The main components of each evaluated detector are summarized below.

a. The ASPECT model MKC-A02-1M RIID is the older of the two ASPECT RIIDs that were evaluated. It contains a 3.4 cm dia X 4.7 cm long Nal gamma detector, and two 1.8 cm 14 y cm 8-atm Helium 3 detectors for neutron identification.

b. The ASPECT model MKC-A03-1 RIID is the newer of the two ASPECT RIIDs that were evaluated. It contains a 4 cm dia X 4 cm long Nal gamma detector, three 1.5 cm X 14 cm 8-atm Helium 3 detectors for neutron identification, and a G-M tube for dose-rate measurements.

c. The Polymaster PM-1401KEKE RIID contains two gamma detectors: a 2.5 cm dia. X 2.5 cm long CsI crystal (seeded with Csl37 for calibration purposes)with a photodiode, and a Silicon photodiode as a semiconductor detector. For neutron detection, one 1.5 cm dia. X 12.5 cm long 3He tube detector is used.

d. The Bicron FieldSpec contains a 1" dia. X 2" long Nal detetctor (also seeded with Csl37), and a x" x y" N aim He-3 detector.

2. Search Mode However, it was determined prior to testing that many of the RIIDs would not be able to'successfully indicate the presence of a continuous moving source at this distance. Each detector was evaluated at an initial distance; if 5 or more unsuccessful trials were performed, the distance was reduced by 5cm and the trials repeated. This process was iterated until the detector could alarm on at least 16 of 20 consecutive trials.

Initially, the plan was to only run this conveyer belt continuously, which would lead to the source passing in front of the detector every 3 seconds. A modification to the plan was agreed upon by all parties that, in addition to this procedure, additional tests were performed where the drive motor for the conveyor belt was manually switched on & off so that the source passed in front of the detector with a frequency of no more than once every 10 seconds. In the fast-moving or continuous mode it was very difficult to distinguish independent alarm signals. The ASPECT A03 exceeded the minimum performance requirements in both the fast and slow test modes. The Bicron Fieldspec met the performance criteria for 60Co in both fast and slow testing but failed for HEU and 137Cs. The ASPECT A02 passed in both modes of testing for 60Co and HEU but failed for 137Cs. The Polymaster

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140 IKE met the minimum requirements for all source and testing procedures except the fast test for 137Cs where its performance was only slightly below the requirement.

3. Unshielded Isotope ID: All of the instruments failed to identify the unshielded reactor grade plutonium. The ASPECT A02 and A03 instruments identified all of the isotopes except Pu. The Polymaster 140 IKE could not identify Pu, 131Ba and 57Co; the CsIl detector-silicon diodes combination seems to have limitations at low energies. The Bicron Fieldspec identified all the isotopes except Pu but had several instances of identifying isotopes which were not present with a lower confidence level.

4. Shielded Isotope ID: The ASPECT A03 and the Bicron Fieldspec instrument were able to identify the shielded reactor grade Pu source as 239Pu when placed at approximately a 10 cm spacing. The other instruments could not. The ASPECT A03 met all of the other identification requirements for shielded sources. The ASPECT A02 met all the other requirements except for 232Th. The Bicron Fieldspec met all of other requirements. The PolyMaster 1401KEKE met all of the other requirements except for 57Co and 133Ba. It also had several instances of identifying isotopes which were not present.

5. Neutron detection: We extended the neutron detection testing to evaluate the detection of the 252Cf source(21K neutrons per second) at 25 cm, 40cm and 50 cm. The PolyMaster 140 IKE met the detection-criteria at 25 cm and 40 cm but not at 50 cm. The ASPECTA02 met the detection criteria at 25 cm, 40 cm, and 50 cm. The ASPECTA03 met the detection criteria at at 25 cm and 40 cm but not at 50 cm. The difference between the A02 and A03 is attributed to the A02 being set at a false alarm rate of 3 per hour and A03 being set at a false alarm rate of 1 per hour. The Bicron Fieldspec did not meet the criteria at 25 cm.

6. Multi-Source Isotope ID: The ASPECT A02 identified the 57Co and HEU combination correctly but could not identify 226Ra and 232Th in combination. The ASPECT A03 identified the 57Co and HEU combination correctly and the 226Ra and 232Th in combination correctly. The Bicron Fieldspec identified the 57Co and HEU combination correctly and the 226Ra and 232Th in combination correctly. The PolyMaster 140 IKE identified the 226Ra and 232Th combination correctly but could not identify 57Co and HEU combination

7. Human Interface: The controls and human interface components of the ASPECT A02 were rated v.-ry highly. The operators found it easy to use and the control and information displays were easy to understand. The ASPECT A03 was found to be further improved in this area. It had additional control and display features that made it even easier to use. The ergonomic improvements make this device considerably easier to handle. The Bicron Fieldspec was rate only fair in its control features but it is by far the easiest to handle. The display were simple but adequate. The operation was somewhat difficult because of discrepancies between the instrument and the manual. One feature which made it difficult was that the acquire time was not self adjusting. The PolyMaster 140 IKE requires experience in the operation of a PDA. For the reason its interface features were considered to be less obvious than the other instruments. The data displays are also less intuitive. Further more the PDA screen could not be viewed by more than one person at a time. This instrument would require more training to be use effectively.

Conclusion: The overall performance of the ASPECT A03 instrument was equal to or better than any of the isotope identification instruments that have evaluated in the past. It met all of the testing requirements except the identification of unshielded reactor grade plutonium but we have

69

not tested any instruments that have reliably met this requirement We would recommend that this instrument be used for identification of the sources of radiation alarms in the Second Line of Defense Program. We would also recommend that a field performance acceptance procedure be developed.

Investigation of the Alternative Gamma-Spectropic Detectors for Quantitative Determination of Activity Isotopes "TYK"

(Standard Shipping Containers)

S. Ulin, V. Dmitrenko, K. Vlasik, Z. Uteshev, N. Ivanova, A. Ischenko, R. Ibragimova Moscow Engineering Physics Institute (the State University) (MEPhI)

N. Kravchenko, I. Bannyh, I. Chirkina, Y. Popkov, A. Korotkov

State Customs Committee of the Russian Federation

A. Dorin Green Star Company

Abstract

Description of the xenon gamma-ray detector and results of its main parameters investigations are

presented. It is shown that the detector provides good energy resolution (∼2% for gamma-line

662 keV) and has average gamma-ray efficiency about 5% in the energy range 0,05–5000 keV.

Higher sensitivity xenon gamma-ray detectors technique for nuclear and radioactive materials

detection is considered. It is shown that using of variable scanning energy intervals for data

processing gamma-ray spectrum increases xenon detector’s sensitivity almost tenfold. Experimental

tests of xenon gamma-ray detector sensitivity were carried out. It was shown that gamma-source with

activity ∼75 kBq placed at a distance of about one meter was detected by a xenon detector for less

than one second. This exposition is 7–8 times shorter in comparison with standard radioactive

sources detecting methods.

Test results of xenon gamma-ray detector installed in a portal monitor, which was developed at All-

Russian Scientific Research Institute of Electro mechanics in 2005 for the purpose of comprehensive

check-up of passengers and their luggage while passing through a checkpoint, are presented. It was

determined that the gamma-ray source 226 Ra with activity about 50 kBq was detected at a distance

of one meter in less then one second and exposure time for identification of this radioactive nuclide

was only several seconds.

2

Using the xenon gamma-ray detector at customs terminal at the “Domodedovo” Airport is

considered. It is confirmed that the gamma-detector can be efficiently applied in order to detect and

identify radioactive materials and to control and investigate radio-nuclides in shipping containers as

well.

Prospects of xenon gamma-ray spectrometer development due to using thin-walled cylindrical

ionization chambers with a composite cover are considered. In this case total mass of xenon detector

can almost be halved, an energy range of measurement gamma-rays will be enlarged to the end of

low energies and gamma-quanta scattering in the walls of working chamber will be essentially

smaller.

Results of gamma-ray spectrum processing by the software, where iterative methods of measurement

spectrum reconstruction were used, are presented. It is shown that, in comparison with ordinary

methods, iterative methods provide more detection efficiency of gamma-lines in measured spectra

and raises nuclide identification reliability.

3

LIST OF EXECUTORS

№ First name, last name

Appointment held Place of employment

1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13.

Sergey Ulin Valery Dmitrenko Konstantin Vlasik Ziyaetdin Uteshev Nikolay Kravchenko Igor Bannyh Iraida Chirkina Yuri Popkov Andrey Korotkov Andrey Dorin Nadezhda Ivanova Artyom Ischenko Rumiya Ibragimova

Professor Head of scientific investigation sector Senior Scientific Researcher Senior Scientific Researcher Deputy Chief of department of Special Equipment and Automatization for Custom Technologies Head customs inspector Head customs inspector Head of customs department Deputy Head of customs department President Post graduate student Engineer Engineer

MEPhI MEPhI MEPhI MEPhI State Custom Committee of RF State Custom Committee of RF State Customs Committee of RF Domodedovo Customs Domodedovo Customs Company “Green Star” MEPhI MEPhI MEPhI

4

CONTENTS

Introduction……………………………………………………………………….. 51. Preparation of test equipment…..……………………………………………... 6 1.1. Brief description of xenon gamma-ray detector…....……..………………. 6 1.2. Determination of main spectroscopic characteristics of xenon gamma-ray detector……………….……………………………………….. 72. Operation of xenon gamma-ray detector in portal monitors….……………….. 10 2.1. Development of higher sensitivity xenon gamma-ray detectors technique for nuclear and radioactive materials detection and identification….…..… 10 2.2. Experimental test of the technique developed.…………….……………... 12 2.3. Description of the portal monitor “ВНИИЭМ-ПМ”…………………….. 13 2.4. Test results of xenon gamma-ray detector comprised in a portal monitor “ВНИИЭМ-ПМ”…...…………………………….……………. 153. Test of xenon gamma-ray detector at customs terminal of the “Domodedovo” airport……………………………………………………………….………… 16 3.1. Measurement conditions………………………………………………….. 16

3.2. Detection and identification of nuclear and radioactive materials withdrawn from passengers…………………………….………………… 173.3. Investigation of medical radioactive materials in shipping containers………………………………………………………………….. 19

4. Prospect of more perfected xenon gamma-ray spectrometers and software construction……………………………………………………….…..…...…. 21

4.1. Development of thin-walled xenon gamma-ray detectors …….………… 21 4.2. Development of reconstruction technique for measured gamma-ray spectra……………………………………………………………………. 22Conclusion..………………………………………………………………………. 25

5

Introduction At present the tasks of development and construction of spectroscopic portal

monitors used for control of nuclear and radioactive materials transportation are becoming extremely relevant. These problems can be solved by means of gamma-ray spectrometers providing high efficiency and energy resolution.

Application of this kind of equipment will permit detection and identification of radioactive materials as early as at the first stage of customs control which will appreciably simplify the task of competent decision-making on further transportation of passengers, their luggage, as well as various loads in shipping containers at their traversing customs terminals.

Gamma-ray spectrometers based on xenon gamma-ray detectors (XeGD) are highly promising for being used as part of portal monitors equipment as well as for carrying out the usual customs control measurements.

Main purpose of the third stage of Contract 12599 was to study the possibility of using the XeGDs developed at Moscow Engineering Physics Institute (the State University) (MEPhI) for customs control of nuclear and radioactive materials transportation.

General tasks of the third stage of Contract 12599 are as follows: – equipment preparation for XeGD testing under real conditions of nuclear and

radioactive materials customs control; – development and experimental testing of effective nuclear and radioactive

materials detecting technique by means of xenon gamma-ray spectrometer; – testing of xenon gamma-ray spectrometer as part of portal monitor “ВНИИЭМ-ПМ”;

– testing of xenon gamma-ray spectrometer at customs terminal at “Domodedovo” airport;

– development of recommendations for construction of new xenon gamma-ray spectrometer models, taking into consideration the experience and test results at Domodedovo customs;

– development of recommendations for new software development for the purpose of further improvement of spectroscopic data processing efficiency.

Works implied by the Contract 12599 were carried out at Radiation Laboratory

(MEPhI), “Green Star” company, All-Russian Scientific Research Institute of Electromechanics (ВНИИЭМ), and customs terminal at “Domodedovo” airport.

6

1. Preparation of test equipment

1.1. Brief description of xenon gamma-ray detector

According to the work program of Contract № 12599 XeGD was made ready at MEPhI. The gamma-detector (XeGD) is based on a cylindrical ionization chamber filled with compressed xenon and supplied with a shielding grid. The chamber operates in pulse mode. Its general arrangement and view of XeGD are given in Fig. 1 and Fig. 2.

Fig. 1. General arrangement of XeGD. Fig. 2. View of XeGD.

1 – charge sensitive amplifier; 2– valve; 3 – high voltage power supply; 4 – metal-ceramic feedthrough; 5 – cylindrical ionization chamber; 6 – anode; 7 – shielding grid; 8 – thermal insulation; 9 – metal case.

XeGD is connected to a personal computer (РС), where a board with shaping

time amplifier and multi-channel analyzer is installed. Spectrometric information is collected and stored in PC and processed by means of special software, developed by “Green Star” company. XeGD shown in Fig. 2 has sensitive volume of two liters. General technical parameters of the detector are presented in Table 1.

Table 1. Technical parameters of XeGD

Energy range of registered γ-quanta Xenon density Sensitive volume End and flank surface (area) Mass Dimensions Supply voltage Power consumption Warranty period

0,05–5 MeV 0,3 g/cm3

2000 cm3

100 cm2 и 200 cm2

5 kg diameter 120 mm, length 320 mm +24 V at most 15 W at least 10 years

1 2 3 4 5 6 7 8 9

7

1.2. Determination of main spectroscopic characteristics of xenon gamma-ray detector.

Major task of preparing the equipment for the test was to check spectrometric

characteristics of XeGD after its continuous service in the course of fulfillment the second stage of the Contract in 2003–2004. Besides, it was necessary to make calibration measurements with the use of standard set of radioactive sources.

Diagram of spectrometric characte.ristics measurements and XeGD calibration is shown in Fig. 3.

The detector’s working substance (Xe gas) purity check was carried out by means

of the system for purification and preparation xenon-hydrogen mixture, which is installed at Radiation Laboratory of MEPhI. Gas control was fulfilled by testing small amounts of working substance from the detector. As a result of analysis of the portion of Xenon gas it was determined that its purity had not changed despite the fact that more than three years had elapsed since the moment of filling the detector with working substance. Life time of electrons in the gas taken from the detector at density of 0.3 g/cm3 was more than 5 μs which corresponds to the value measured at initial filling of the newly made gamma-detector.

РС

220 VPS

1 2

3

4

Gamma-source

XeGD

Lead housing

Fig. 3. General arrangement of measurements and XeGD connection to power supply and PC. XeGD – xenon gamma-detector, PC – personal computer, PS – power supply,

1 – electric signal cable from XeGD, 2 – 24 V cable for the charge sensitive amplifier, 3 – blocking signal cable, 4 – 24 V cable for high voltage signal.

8

Linear characteristic of XeGD. Full absorption peak position versus registered gamma-quanta energy was obtained with the help of nine gamma-sources: 137Cs, 60Co, 57Со, 22Na, 141Am, 54Mn, 139Се, 65Zn, and 113Sn. Typical spectra for 137 Cs and 133Ba gamma-sources measured with the XeGD are given in Figs 4 and 5.

0 200 400 600 8000

2000

4000

6000

8000

10000

12000

14000

Cou

nts

Energy, keV

Cs-137FWHM=15.15keV

0 100 200 300 400 500

0

10000

20000

30000

40000

50000

Cou

nts

Energy, keV

Ba-133

Fig. 4. Gamma-spectrum of 137Cs source

(Eg = 662keV) measured by XeGD. Fig. 5. Gamma-spectrum of 133Ва source (Eg = 80.9; 160.6; 223.1; 276.4; 302.9, 356.0; 393 keV) measured by XeGD.

Measurement results of xenon detector’s linear characteristic are presented in

Fig. 6. Deviation scope of gamma-peak positions from linear dependence was evaluated by criterion χ2, which was 0.18248 at twelve degrees of freedom. In other words, deviation scope from linear dependence of gamma-peak positions on gamma-quanta energy was not more than 0.5%.

0 20 0 400 60 0 80 0 10 00 12 00 14 000

2 00

4 00

6 00

8 00

10 00

12 00

14 00

N

E n e rgy, keV Fig. 6. Gamma-peak position versus gamma-ray energy.

Energy resolution measurement. To measure energy resolution (ΔЕ/Е) a set of

standard gamma-sources “ОСГИ” was used. The results of measurements concerned are given in Fig. 7. As the gamma-quanta energy grows the detector’s energy resolution gets better for gamma-lines 662 keV ΔЕ/Е ≈ 2%, and 1332 keV ΔЕ/Е ≈ 1.5% respectively.

9

0 2 0 0 4 0 0 6 0 0 8 0 0 1 0 0 0 1 2 0 0 1 4 0 01

1 0

Ener

gy R

esol

utio

n (%

)

E , k e V Fig. 7. Curve of energy resolution versus gamma-quanta energy.

Registration efficiency. Efficiency of gamma-quanta registration by means of

xenon detector was defined by gamma-peak area measurement for the gamma-sources from “ОСГИ” set. The distance between the detector and a gamma-source was 50 cm. Fig. 8 represents the results of the measurements.

0 200 400 600 800 1000 1200 1400

1

10

Effic

ienc

y, %

Fig. 8. Gamma-quanta registration efficiency versus its energy.

Registration efficiency of the detector in the energy range from 50 keV up to

1.5 MeV does not exceed 12%, which corresponds to general requirements of the gamma-spectrometric equipment normally used for radionuclides control and identification.

Xenon detectors efficiency is defined by operating substance (xenon) density. Basically, if there is a necessity to increase the detector’s sensitivity it is sufficient to merely enlarge the mass of the operating substance in the detector. In this case main spectrometric properties of the detector will not practically change. There is a possibility for the detector to extend energy range of registered gamma-quanta into the region of low energies (up to 30 keV). This can be done by making the wall of the sensitive volume thinner, as thin as 1mm (normal thickness is 3 mm), as well as strengthening of the thin-walled case at the expense of composite synthetic fiber coating. Sensitivity of the xenon detector, especially for gamma-quanta of low energies, will also increase in this case.

General result of the measurement is that spectroscopic characteristics of the XeGD have not practically changed since the time of manufacture in 2003.

10

2. Operation of xenon gamma-ray detector in portal monitors. 2.1. Development of higher sensitivity xenon gamma-ray detectors technique for

nuclear and radioactive materials detection and identification.

Conventional portal monitors usually use gamma-detectors based on plastic scintillates by means of which excess of registered gamma-quanta over natural background is defined. In fact, those detectors are used as simple counters of the registered particles.

It is easy to show that lack of spectrometric capabilities of those detectors reduces their sensitivity of radioactive materials. Especially visibly this becomes apparent at registration of those radioactive materials whose gamma-radiation is close to background level. As a rule, weak gamma-sources, whose radiation does not exceed background statistical fluctuations (within the limits of 1–2 sigma), altogether cannot be detected by portal monitors in short (1–2 sec) measurement time.

Application of high energy resolution gamma-spectrometers in portal monitors opens fundamentally new vistas for detection as well as identification of various gamma-sources while controlling radioactive materials transportation.

Significant feature of spectrometric detectors is the fact that they permit analysis of specific peaks of gamma-quanta full absorption by measured spectra as well as comparing the number of corresponding gamma-quanta to background radiation within the energy range of measurements.

As a criterion of sensitivity К’(x) of the spectrometric detector the ratio can be chosen between total number of registered gamma-quanta of a particular peak in the measured spectrum and background roof-mean-square deviation in the same energy range. If full absorption peak is described as Gaussian distribution, and background as exponent, then sensitivity К’(x) of the detector can be put down as:

The integrals of the formula can not be calculated analytically, they can only be

integrated by numerical methods. (К’x) values were obtained for various magnitudes of parameters х — energy range of integration and ΔE — full width of half maximum (FWHM) absorption peak.

It was presumed that total magnitude of gamma-quanta in full absorption peak would not exceed the background value by 0.5 sigma in all of the energy range between 100–1000 keV. Calculated results are shown in Figs 9 and 10.

( )

20 0

2

0

0

0

/ 2 ( )0 2

2/ 2

1/ 2 2

/ 2

2'

E x E EE

E x

E xE

E x

N e dEEK x

e dEβ

π

α

+ −−

Δ

−

+−

−

×Δ

=⎡ ⎤

×⎢ ⎥⎢ ⎥⎣ ⎦

∫

∫

11

Fig. 9. Gamma-quanta distributions in backgroundand full adsorption peak.

Fig. 10. Gamma ray detector sensitivity versus scanning intervals for several values

of energy resolution. Registered gamma-quanta distributions in full absorption peak (662 keV) and

natural background that were used in calculations are shown in Fig. 9. In Fig. 10 one can see sensitivity К’(x) of the gamma-spectrometer versus the

scanning interval – х at various values of FWHM of the peak concerned, i.e. energy resolution of the gamma-detector.

It follows from Fig. 10: 1. The given functions have characteristic maximum at scanning interval values х

equal to energy resolution of the detector. 2. The higher energy resolution of the gamma-ray detector the higher its sensitivity

to gamma-radiation. The calculations involved show that sensitivity of XeGD with energy resolution

about 2% can be increased nearly tenfold as compared to simple count of registered events. This high sensitivity of XeGD is provided due to the method of scanning the measured spectrum by energy intervals that correspond to detector’s energy resolution for particular value of registered gamma-quanta.

The calculations in question are of demonstration character. They were carried out with the view of revealing advantages of spectrometric gamma-ray detectors over simple gamma-counters normally used for radioactive gamma-sources detection.

630 640 650 660 670 680 690 7000,0

0,2

0,4

0,6

0,8

N

E keV

Gauss Background Background+Gauss

0 10 20 30 40 50 60 70 80 90 1002

4

6

8

10

12

14

K(x)

x keV

FWHM 14 keV 16 keV 20 keV 24 keV 28 keV 36 keV 46 keV

12

2.2. Experimental test of the technique developed In order to check the results of the calculations concerning evaluation of XeGD

sensitivity to gamma-ray sources series of special measurements was conducted. As a gamma-ray source 137Cs (with activity of 75 kBq) was used. That was placed at various distances from a two-liter XeGD. Exposure time varied from 0.5 to some tens seconds. During the measurements minimal exposure time was defined when sensitivity of the XeGD К’(x) exceeded the background by three sigma. Measurement results are shown in Fig. 11.

Fig. 11. Minimal exposure time of gamma-ray source detection versus distance between the gamma-ray detector and a gamma-ray source for various scanning intervals.

It can be seen from the figure that sensitivity of XeGD to the gamma-ray source

increases with the decrease of scanning intervals width. The best sensitivity of XeGD is achieved at scanning intervals close by value to energy resolution of the XeGD.

At short distances between the gamma-ray source and XeGD the measured curves come nearer to one another since in this case intensity of gamma-radiation flux entering the working volume of the detector increases drastically. At moving the source away from the XeGD the curves in Fig. 11 noticeably diverge. At a distance about 75 cm, at processing the data by scanning intervals procedure (the scanning interval is equal to 15 keV), time of gamma-ray source detection was 7 times shorter than at spectrum processing by simple counting of registered gamma-quanta (counting mode of a conventional counter).

The measurement conducted confirms the earlier computations and shows the possibility of XeGD sensitivity extension owing to application of scanning intervals procedure while analyzing the measured gamma-spectra.

0 25 50 75 100 125 1500

2

4

6

8

10

Tim

e, s

ec

X, cm

1000 keV 100 keV 20 keV

13

2.3. Description of the portal monitor “ВНИИЭМ-ПМ”

A pedestrian portal monitor “ВНИИЭМ-ПМ” was developed at All-Russian Scientific Research Institute of Electromechanics (ВНИИЭМ) in 2005 for the purpose of many-sided check-up of passengers and their luggage at passing through a checkpoint. A passenger will have to walk inside the portal to have measurement taken. Gross time of the procedure will not exceed ten seconds. Duration of time exposure in the portal will be determined above all by the procedure of vocal diagnostics of the passengers as well as the technique of drugs and explosives detection.

By means of portal monitor it is planned to identify passengers and to detect various metal, explosive, narcotic, and radioactive substances. The portal includes following equipment:

Digital camera; Vocal analyzer; Metal finder; Infrared imager; Device for detection of narcotic and explosive materials (with forced air-

circulation system); Xenon gamma-ray spectrometer.

Photos of the portal monitor “ВНИИЭМ-ПМ” и XeGD are shown in Fig. 12.

Fig 12. Portal monitor “ВНИИЭМ-ПМ” и XeGD aimed for radioactive materials detection and identification.

14

Measuring equipment is located on the top of portal monitor except for infrared imager and digital camera which are installed at a height of about 1.5 meters for getting better images of a passenger under examination.

As a multichanal analyzer of the XeGD plate SBS-75 developed by “Green Star” company was used. Software for this plate was also developed by that company.

To operate the XeGD special software was developed by MEPhI employees which was included in the software bundle of the portal monitor.

Portal monitor in the course of assembling and tuning is shown in Fig. 13.

FIg. 13. XeGD adjustment and demonstration test of portal monitor “ВНИИЭМ-ПМ”.

15

2 0 0 4 0 0 6 0 0 8 0 0 1 0 0 0 1 2 0 0 1 4 0 0 1 6 0 0 1 8 0 0 2 0 0 0

0 , 1

1

1 0

1 0 0

Cou

nts,

1/s

ec

E n e r g y , k e V

2 9 5 ( 1 9 , 3 % )2 2 6 R a ( 1 6 0 0 y e a r s )

3 5 1 ( 3 7 , 6 % )

6 0 9 ( 4 6 , 1 % )

7 6 8 ( 4 , 9 % )1 1 2 0 ( 1 5 , 1 % )

1 2 3 8 ( 5 , 8 % )

1 3 7 7 ( 4 % )

1 7 6 4 ( 1 5 , 4 % )

2.4. Test results of xenon gamma-ray detector comprised in a portal monitor “ВНИИЭМ-ПМ” During the test an operator carrying a plastic bag with a gamma-ray source (226Ra

with activity about 50 kBq) (Fig. 10) walked into the portal. The gamma-ray source was being detected for one second, practically at the moment of approaching the portal monitor at a distance of about two meters. This can be accounted for the fact that lead screening of the gamma-ray spectrometer for lessening the viewing angle had not yet been provided for. The measurement taken showed that the XeGD had effectively detected the 226Ra source with activity of 50 kBq at a two-meter distance during the exposure time of one second. The result obtained fully meets the requirements set for modern portal monitors using conventional plastic scintillators for these purposes.

Besides radioactive source detecting its identification was simultaneously accomplished. It took several seconds to obtain a spectrum having necessary statistics.

Spectrum of gamma-ray source 226Ra (50 kBq) measured by XeGD during five seconds is shown in Fig. 14.

Рис. 14. Gamma-ray 226Ra source (50 kBq) spectrum measured by XeGD.

Identification of gamma-ray source 226Ra was carried out by means of software that was developed by “Green star” company to solve the tasks of customs control.

Thus, the test results have shown that XeGD can be used in portal monitor for detection and identification of gamma-sources.

It should be noted that in these measurements XeGD had comparably small volume (2 liters), nevertheless its sensitivity was quite sufficient for effective work of the portal monitor.

Of necessity sensitivity of XeGD can be essentially increased by means of making working volume bigger.

16

3. Test of xenon gamma-ray detector at customs terminal of the “Domodedovo” Airport

3.1. Measurement conditions

Tests of XeGD were carried out at customs radiation control zone of “Domodedovo” airport (Fig. 15) in a special room (Fig. 16) equipped with everything necessary for making measurements of this kind (laboratory tables, light, electricity etc).

The XeGD and a personal computer were delivered from MEPhI to “Domodedovo” airport. All measuring equipment was installed on a laboratory table (Fig. 16). Radioactive materials destined for investigation in shipping containers or without them were also placed on the table if their activity was not too strong.

Powerful gamma-ray sources were placed in a special cart for the time of measurement with the aide of which they could be moved from the XeGD at a distance of 2–5 meters to provide acceptable load of the detector.

Fig. 15. “Domodedovo” airport.

Fig. 16. Customs radiation control zone.

17

200 400 600 800 1000 1200 1400 1600 1800 2000

0,01

0,1

1

10

Cou

nts,

1/s

ec

Energy, keV

295 (19,3%)226Ra (1600 years)

351 (37,6%)

609 (46,1%)

768 (4,9%)1120 (15,1%)

1238 (5,8%)

1377 (4%)1764 (15,4%)

3.2. Detection and identification of nuclear and radioactive materials withdrawn from passengers

To investigate the possibility of using XeGD at the customs terminal of

“Domodedovo” airport gamma-active samples were examined. They had been detected and then withdrawn from passengers who had passed through portal monitors of “Aspect” company that have already been working at the airport for several years.

Major task of the measurements was to clarify the question whether the XeGD involved had sufficient sensitivity for detection and identification of the gamma-ray sources made mention of.

Several objects withdrawn from passengers and their corresponding gamma-ray spectra measured by XeGD at a distance of one meter are shown in Figs 17–20.

Fig. 17. A clock.

Fig. 18. An altimeter.

200 400 600 800 1000 1200 1400 1600 1800 2000

0,1

1

10

100

Cou

nts,

1/s

ec

Energy, keV

295 (19,3%)226Ra (1600 years)

351 (37,6%)

609 (46,1%)

768 (4,9%)1120 (15,1%)

1238 (5,8%)

1377 (4%)

1764 (15,4%)

18

Fig. 19. Grids for gas-rings.

Fig. 20. Camera lens.

Objects introduced in Figs 17–20 contain various radio nuclides, mainly 226Ra and 232Th. To define minimal duration of detecting these gamma-ray sources additional measurements were taken. As a result of measurements it was revealed that it had taken less than one second for the XeGD to detect them at a distance of one meter, though the activity of each gamma-ray source did not exceed 50 kBq. It is worth mentioning that the duration of these objects detection is not longer than the time exposure estimated for portal monitors made by “Aspect” company currently functioning at “Domodedovo” airport.

At the instance of customs control officers gamma-ray spectra of the samples mentioned above were measured within 2–5 minutes. The spectra are shown in Figs 17–20. Based on these spectra 226Ra and 232Th radioactive nuclides were identified in the objects under study.

It is significant that spectra obtained by means of XeGD in several seconds are sufficient for radioactive nuclides identification.

Thus, the results of the research surely show the possibility of XeGD effective application in customs control portal monitors.

200 400 600 800 1000 1200 1400 1600

0,01

0,1

1

10

Cou

nts,

1/s

ec

Energy, keV

238 (43,3%)232Th (1,4*1010years)

510 (22,6%)

583 (84,5%)

727 (10,3%)968

911

2614 - 1022 (100%)

200 400 600 800 1000 1200 1400 1600 1800

0,1

1

10

Cou

nts,

1/s

ec

Energy, keV

238 (43,3%) 232Th (1,4*1010years)

510 (22,6%)583 (84,5%)

727 (10,3%)

968911

2614 - 1022 (100%)

19

3.3. Investigation of medical radioactive materials in shipping containers At the radiation customs terminal of Domodedovo airport investigation of

possibility of XeGD application for detection and identification of various radioactive sources placed in shipping containers were also carried out. For this purpose radioactive sources assigned for medical application were used. More than twenty containers were examined, and only a small fraction of the measurements is submitted in this report.

Apart from radioactive materials detection and identification the investigation in question was also to reveal probable presence of any other additional radioactive sources not included in declaration accompanying forms.

Spectrum shown in Fig. 21 was measured within three minutes from a load destined for a town in Russian Federation.

Fig. 21. Medical radioactive isotope 131I spectrum measured with XeGD. At analyzing the results of measurements it was established that the load was 131I

radioactive nuclide (with activity 203 ± 5 МBq). Metal protecting container underneath packing was absent. Accompanying forms confirmed that the load concerned was in fact a medication 131I with activity of 200 МBq placed in cardboard boxes without additional lead protection. Presence of any other supplementary radioactive isotopes was not revealed. Measurement was taken at a distance of 2.5 m because of high activity of the load.

100 200 300 400 500 600 700 8000

2

4

6

8

10

12

722 (2,3%)

636 (9,1%)

284 (7,6%)

Cou

nts,

1/s

ec

Energy, keV

364 (100%)

131I (8 days)

20

Similar measurement was taken from another load placed in a special container. A photo of it as well as measured gamma-spectrum are shown I Fig. 22.

Fig. 22. Shipping container and measured gamma-ray spectrum of its gamma-radiation.

Measurement presented in Fig. 22 was taken for three minutes with a distance of three meters between the XeGD and container. At this distance measuring equipment dead time did nit exceed 10%.

In the course of analysis of the spectrum measured it was established that the load was 99Mo with activity of (405 ± 10) kBq, which actually was parent radioisotope for 99*Tc. Thus technetium generator widely in use in medical applications was found in the container. It was also revealed that the container had lead protection 2 cm thick. Presence of other supplementary radioactive nuclides was not revealed.

The results of measurement of that load were in good agreement with accompanying forms.

The results in question of various radioactive loads investigations have shown that XeGD can be effectively used for detection and identification of loads alike. At the same time it should be pointed out that during the use of XeGD under real conditions of customs terminal no problems concerning its functioning occurred. As for XeGD reliability and time stability check, it was made and confirmed at the Second Stage of the Contract.

200 400 600 800 1000 1200 14000,01

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nts,

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ec

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99Mo (69 hours)

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21

4. Prospect of more perfected xenon gamma-ray spectrometers and software construction

4.1. Development of thin-walled xenon gamma-ray detectors

XeGD development prospects are determined first of all by perfection of their manufacturing technology and application of new methods for making thin-walled and ultrastrong vessels assigned for working substance of these detectors. At present this kind of work is under way. A pre-production model of XeGD having a thin-walled vessel with composite covering which provides its high strength and reliability has been made. As a result safety of these devices should get better, efficiency should increase and energy range should widen towards especially low gamma-ray energy. However the main result is that the total mass of XeGD becomes three times as less.

A photo of the pre-production model of the XeGD having a thin-walled vessel with composite covering is shown in Fig. 20.

As a result of testing of the two-liter XeGD with composite covering, it was shown that it can withstand inner pressure of more than 400 atm without residual deformation of the vessel.

It should be noted that working pressure inside the detector is not more then 45 atm.

As it was mentioned above safety and reliability of high level are required for wide use of XeGD.

Application of new technologies, namely covering of vessels by composite materials based on ultrastrong synthetic fibers provides high operational characteristics of these detectors.

Fig .23. XeGD thin-walled vessel with composite coverage. Currently, there are other methods of production ultrastrong thin-walled vessels.

For example, using ultrastrong sorts of steel for making the vessel for XeGD can in principle decrease its wall thickness from 3 mm to 0.5 mm. However, using of this kind of steel will essentially increase the cost of XeGD.

Regardless of the XeGD design it is necessary to provide for a safety valve on the vessel for the case if any outside extraordinary impacts occur.

For the final choice of the thin-walled vessel production technique, it is important to carry out additional investigations which will clarify the optimal solution to this question.

22

4.2. Development of reconstruction technique for measured gamma-ray spectra Along with the improvement of XeGD design, development of new data

processing methods is also very important for effective application of this kind of equipment.

Currently, work on development of new software permitting reconstruction of gamma-spectra measured by XeGD has been started at Radiation laboratory MEPhI.

General algorithms to solve tasks of this kind are well known. However to realize these algorithms availability of good software as well as fast

personal computers is essential. The latter has only recently become possible when high clock frequency personal

computers came into being. The gamma-ray spectra measured can be described by the Fredholm integral

equation.

,

where: Z(y) – incident gamma-radiation spectrum; К(x,y) – core of the integral equation, i.e. instrument functions aggregate; f(x) – gamma-ray spectrum measured by XeGD. х and y – gamma-ray energy in our case.

The task of solving this integral equation is to find a function Z(y). As is well

known, problems of this kind are ill-conditioned ones. Special iteration methods are used to solve them. Various stabilizing functionals are usually applied to search for stable solutions.

In this particular case the integral equation can be presented as a matrix equation:

Coefficients of this matrix are determined by calibration measurements and their further approximation for all the energy range of registered gamma-rays. The matrix dimension can be more then 2000∗2000. Matrix equations like these are usually solved by iteration methods.

In Fig. 23 first results of the matrix equation solution with the use of full set of instrument functions of the XeGD are presented.

For convenience the gamma-ray spectra measured by XeGD are shown on the left- hand pictures and the reconstruction gamma-ray spectra (results of calculations) — on the right-hand ones.

)(0

)(),( xfdyyZyxK =∫∞

⋅

⎪⎪⎪

⎩

⎪⎪⎪

⎨

⎧

=⋅

=⋅

∑

∑

=

=

n

ninni

n

nnn

Cxa

Cxa

1

111

...

0≥ika 0≥iс

11

=∑=

ik

n

i

a 11

=∑=

i

n

iс

23

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137Cs

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22Na

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Fig. 24. Measured and reconstructed gamma-ray spectra of 137Cs, 133Ba и 22Na gamma sources.

24

From these pictures it is clear that the radioactive source main gamma-lines in the reconstructed spectra are shown more distinctly. It may be noted, that after several iterations only real gamma-lines are left in the reconstructed gamma-spectra and Compton substrate practically disappears. This circumstance essentially simplifies the procedure of radionuclide identification in many cases.

A very important peculiarity of the software package being developed is the fact that in the course of spectra reconstruction one manages as if to improve energy resolution of the gamma-ray detector. As a result it is possible to detect gamma-lines if a distance between them is smaller than energy resolution of the gamma-detector.

The results of calculations made look very promising. In this connection it is seems very important to continue the development of the software involved. This concerns, first of all, increasing of a spectrum reconstruction speed as well as more precise description of the XeGD instrument functions, and connection of this software to the standard data processing software.

It is also necessary to carry out the developed software comparative research together with the other ones which do not use spectrum reconstruction algorithms dosen’t used.

25

Conclusion In the process of fulfilment of the Contract # 12599 the possibility of XeGD

application for nuclear and radioactive materials traffic control was studied and following results were obtained:

1. General spectroscopic characteristics of XeGD were measured and it was shown that they had not practically changed since the time of manufacture (2003). In particular, the energy resolution for gamma-lines 662 keV remained the same (about 2%).

2. The method of increasing the XEGD sensitivity to gamma-ray sources was developed. It was shown that its sensitivity can be extended almost tenfold if the energy scanning interval is chosen correctly for analysis of gamma-ray spectrum both experimentally and theoretically.

3. Tests of the two-liter XeGD installed in the portal monitor “ВНИИЭМ-ПМ” were carried out. Results of these tests showed that the gamma-ray source 226Ra with activity about 50 kBq was detected at a distance of one meter in less then 1 second and exposure time for identification of this radioactive nuclide was only several seconds.

4. Tests of the two-liter XeGD were carried out in real conditions of customs radioactive control terminal at Domodedovo airport. As result of testing it was determined that the xenon gamma-ray spectrometer can be effectively used both for detection and identification of radioactive materials placed in shipping containers and outside them.

5. General directions of XeGDs development for their further applications as well as for heightening the safety of their use. A first pre-production model of XeGD thin-walled vessel of higher strength was made. The vessel withstands inner pressure of more than 400 atm., though working pressure is at least ten times smaller. The total mass of the XeGD with a thin-walled vessel will decrease to one third.

6. First test results of software developed at MEPhI for reconstruction of gamma-ray spectra were obtained. It was shown that by means of this software it is possible to essentially simplify the process of radioactive materials identification.

Summing up the results of Contract # 12599 fulfillment, it is necessary to note that

further development of xenon gamma-ray spectrometers will promote creating the spectroscopic portal monitors which are very important for the progress of qualified radioactive materials traffic control.

Including xenon gamma-ray spectrometers in the portal monitors will undoubtedly be of aide for solving the tasks of radioactive nuclides detection and identification as early as at the first stage of control.

As follows from the report first steps in this direction have already been made. New models of XeGDs can become regular equipment for various radioactive materials traffic control checkpoints and above all at customs terminals.

26

Taking into account the results obtained in the course of fulfillment Contract № 12599 general directions of further work connected with the development of XeGD for radioactive materials traffic control (detection and identification) passing customs terminals can be formulated.

1. Development of an improved XeGD design due to the use of a new technology producing thin-walled and ultrastrong vessels for working substance.

2. Development of new XeGD samples to be used in pedestrian and other types of portal monitors.

3. Development of new techniques for spectroscopic information processing to increase the efficiency of radionuclide detection and identification.

4. Caring out tests of new types of xenon gamma-ray spectrometers and developed special software application packages in real conditions of the traffic radionuclide control.

In conclusion participants of Contract № 12599 would like to express deep gratitude to the employees of IAEA for their cooperation in the Contract realization.

Using Associated Particle Technique (APT) for the Detection of Shielded Nuclear Materials

A.V. Kuznetsov, D. Vakhtin V.G.Khlopin Radium Institute, St.Petersburg, Russia

Abstract

One of the weak points of all existing instruments used to detect and characterize radioactive materials, including

fixed automated portals, is their inability to detect fissioning materials that are shielded by substances that absorb

their spontaneous gamma and neutron radiation. While spontaneous radiation emitted by some of the most

dangerous fissioning materials cannot be used for their detection and identification, these materials can still be

detected by the so-called “active” methods. These methods use external radiation to induce fission reactions within

the irradiated volume and then measure such products of these reactions as neutrons and high-energy γ-rays, which

have a very high penetrating ability and can penetrate thick layers of shielding material. The main goal of the

Project was to apply active methods to the detection of shielded fissioning materials. The following summarizes

the work performed:

1. Three neutron detectors were made on the basis of fast plastic scintillator and a photo multiplier. These detectors used measurement channels of the data acquisition system, which was been built specially for experiments with the neutron generator.

2. Experiments with an APT neutron generator, one neutron detector, and depleted U sample were carried out.

3. First experiments with shielded and unshielded uranium samples were carried out.

4. Full mathematical model of the experimental setup was created on the basis of MCNP5 Monte-Carlo code.

It was concluded that the characteristics of the created detectors and electronics are adequate for the task of using

them in a system for detection of shielded nuclear materials. Additional experimental work using three neutrons

detectors for detection of triple (n-n-α) and quadruple (n-n-n-α) coincidences between neutrons from induced

fission of shielded fissioning materials is currently under way.

1. DESCRIPTION OF RESEARCH CARRIED OUT

The first year of the Project was devoted to refining the Associated Particle Technique (APT) and assessing its applicability to detection of shielded nuclear materials.

The main idea of the proposed active detection method is to irradiate the inspected volume with 14 MeV neutrons from a miniature DT neutron generator with built-in detector of associated α−particles, and to detect fission neutrons and γ−rays in coincidence with α−particles, that accompany emission of primary neutrons in the d+t n+α reaction. Detection of associated α−particles allows one to determine emission time and flight direction of each primary 14 MeV neutron. Measurement of secondary particles (neutrons and γ−rays originating from induced fission of nuclear materials) in coincidences with these “tagged” neutrons in very narrow (tens of nanoseconds) time “windows” allows one to suppress the background, which is associated with primary 14 MeV neutrons and their (n,2n) and (n,γ) reactions on surrounding materials [1].

The following activities planned for the first year of the Project have been carried out.

1. Applying neutron generators for detection of shielded fissioning materials:

• investigation of the required intensity of the primary flux of 14MeV neutron;

• estimation of the size of the beam spot on the target, which influences the spatial resolution of the device;

• ensuring close to 100% intrinsic efficiency of the α−detector and low noise levels;

• tuning fast electronics to get sub-nanosecond time resolution and fast determination of the number of the detector that was hit by an α−particle;

• determination of the dose of radiation and measures to fulfill the radiation safety rules.

2. Mathematical modeling of the detection process:

• tuning the MCNP4C2 code to describe the geometry of the measurements using available types of detectors;

• estimating count rates of neutron and γ−ray detectors;

• formulating recommendations for the measurement channel of the device.

The second year of the Project was devoted to creation of the neutron detection channel and conducting first test experiments with real samples of fissioning materials.

At the first stage three neutron detectors have been made on the basis of fast plastic scintillator and a photo multiplier. These detectors use measurement channels of the data acquisition system, which has been built specially for experiments with neutron generator. This data acquisition system can handle extremely high counting rates in both neutron channel and “tagging ” α−particle channel.

First experiments with the whole system have been conducted.

The third year of the Project was devoted to experimental detection of real fissioning materials in different shielding conditions.

2. RESULTS OBTAINED

2.1. Applying neutron generators for detection of shielded fissioning materials

2.1.1. Investigation of the required intensity of the primary flux of 14MeV neutron

The optimal intensity of the primary flux of 14 MeV neutrons is chosen as a result of a compromise between the following requirements:

1. The higher is the neutron flux, the shorter is the lifetime of the vacuum neutron tube of the neutron generator. Lifetime of portable APT neutron generators depends for the most part on the stability of the target. Existing prototype APT NG, working at intensities up to 108 n/s into 4π, have typical lifetime of about 1000 hours. With typical time needed for one inspection about 1 minute, this lifetime is equivalent to more that 50,000 inspection cycles before the change of the vacuum tube of the neutron generator is needed.

2. The associated alpha-particle detector, that is built into the neutron generator, should be capable of detecting all the incident alpha-particles. The nine-segment semiconductor detector of associated alpha-particles and associated fast electronics produced by APSTEC Ltd., Russia, (see Figure 2-1) have been shown to work at count rates up to 107 n/s and more. The fast pre-amplifier coupled to the detector produces 20 ns-wide pulses with amplitude about 80 mV, which are transformed into logical levels and are fed into the programmable logic microchip, which produces time reference signal for use in coincidence measurements, as well as the coded number of the hit segment of the detector.

Figure 2-1. Left: nine-segment detector of associated alpha-particles. Right: fast electronics.

3. The counting rate of neutron detectors coming from the primary neutrons should not be too high not to produce too many false coincidences. The typical signal length from fast scintillator-based detectors is of the order of 100 ns or less. Thus, in order to have the accidental coincidence count rate under 1%, the count rate of each of the two neutron detectors should be kept below 105 counts per second.

4. The time needed for detection of shielded fissioning materials is inverse proportional to the intensity of the neutron generator.

On the basis of the above considerations, the optimal intensity of the APT neutron generator was estimated to be between 107 n/s and 108 n/s.

2.1.2. Estimation of the size of the beam spot on the target, which influences the spatial resolution of the device

The size of the deuteron beam spot on the target of the APT neutron generator influences the position resolution of the device. Existing nine-segment alpha-detector is a 3×3 matrix of 1 cm×1 cm semiconductor detectors located at 6 cm from the target of the NG. If the deuteron beam spot on the target were point-like, the in-plane position resolution of the device at 60 cm from the target would be 10 cm×10 cm — the size of the area corresponding to each segment of the alpha-detector. However, a point-like beam spot would lead to a fast burn-out of the target, and would reduce the lifetime of the neutron tube.

0 1 2 3 4 5 6 7 8 9

10

20

30

40

50

60

~10% of neutronsfall into adjacent

areas

half area size: 5 cm

center of area

experiment

Monte-Carlo simulation: point-like target spot diameter = 2 mm spot diameter = 4 mm

C

oinc

iden

ce c

ount

rate

[arb

. uni

ts]

X coordinate [cm]

Figure 2-2. Experimental determination of the deuteron beam spot size on the target.

The beam spot size was experimentally measured using coincidences between alpha-particles measured by the central segment of the 3×3 associated particle detector and corresponding neutrons. Neutrons were detected by a 1 cm×1 cm p-i-n diode covered with a polyethylene converter foil and located at 60 cm from the NG target. The diode was moved with 1 cm step, and coincidence count rate between it and the associated-particle detector was measured for each position (Figure 2-2). Experimental curve was compared to Monte-Carlo calculations, that took into account real geometry, beam profile and reaction kinematics. Comparison shows that the diameter of the beam spot on the target is about 4 mm, which translates into about 10% of neutrons that coincide with alpha-particles detected in the central segment of the alpha-detector “missing” the corresponding area at 60 cm distance from the target. Such “neutron beam defocusing” ratio is totally acceptable, and the beam spot diameter 4 mm is also big enough not to burn the target too quickly.

The beam spot diameter of 4 mm will be the target for all future APT NG prototypes.

2.1.3. Ensuring close to 100% intrinsic efficiency of the alpha-detector and low noise levels

The detector of associated alpha-particles that is built into an APT neutron generator should have close to 100% intrinsic efficiency of detection of ~3.5 MeV alpha-particles. The detector should also remain operational for the whole lifetime of the vacuum neutron tube, during which it should detect about 1012 alpha-particles per cm2.

The nine-segment alpha-detector was tested using a large NG-400 neutron generator at Radium Institute. The total number of incident alpha-particles was 1013, and the signal-to-noise ratio of the detector was still acceptable after such irradiation.

Dependence of the count rate of the central segment of the 3×3 alpha-detector on the detection threshold is shown at Figure 2-3.

-90 -80 -70 -60 -50 -40 -30 -20

5000

10000

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25000

Cou

nt ra

te [s

-1]

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NG intensity: 1.12 x 107 n/s

Figure 2-3. Dependence of the alpha-particle count rate on the threshold.

Presence of the horizontal region between -20 mV and -70 mV indicate, that all the signals from incident alpha-particles are correctly processed by the fast electronics, and the detector has close to 100% efficiency to alpha-particles from d+t n+α reaction.

2.1.4. Tuning fast electronics to get sub-nanosecond time resolution and fast determination of the number of the detector that was hit by an alpha-particle

The first set of fast electronics for servicing the nine-segment detector of associated alpha-particles, produced by APSTEC Ltd., Russia, has been tested (see right part of Figure 2-1). It contains fast pre-amplifiers and low-level discriminators for each of the nine channels, a programmable logic microchip for formation of the timing signal and on-line determination of the number of hit segment, and an RS-485 interface with a controlling PC. Thresholds can be adjusted for each channel independently using an on-board multi-channel DAC. The dead time of the electronics depends only on the signal width, and is typically about 20 ns. The internal delay of the electronics is 3.5 ns.

The fast electronics was installed on a prototype APT neutron generator with built-in nine-segment detector of associated alpha-particles.

Example of signals after pre-amplifier from all nine segments of the 3×3 alpha-detector is shown at Figure 2-4.

Figure 2-4. Left: signals after pre-amplifier (top) and logical signals (bottom) from nine 1 cm×1 cm segments

of semiconductor alpha-detector. Right: noise after pre-amplifier.

The rise time of the analogue signal from any segment of the alpha-detector is about 5 ns, which ensures sub-nanosecond time resolution of the detector.

2.1.5. Determination of the dose of radiation and measures to fulfill the radiation safety rules

The dose, which the operator working with the neutron generator would receive within a year, was calculated with the assumption of 36 hour-long working week and 50 working weeks per year. Examples of MCNP5 dose calculations for a neutron generator with intensity 5×107 n/s are presented on Figure 2-5.

When no additional shielding is used, the operator should be at distances about 10 meters from the working neutron generator. By using appropriate shielding one can reduce the safe distance down to the dimensions of the device.

It must be noted, however, that the relevance of these calculations to real work with the device strongly depends on the measurement scenario. The inspection cycle of the planned portable device for detection of shielded nuclear materials (more precisely, the time during which the NG will be switched on) is expected to be about 1 minute, during which the dose received by the operator will be much less than the allowed values.

However, if the device is used as a stationary constantly-working installation, then the above estimations are correct, and measure should be taken to ensure the compliance with the existing safety standards.

Experimental dose measurements, as well as studies of the role of residual activation of the device’ components will be carried out during the third year of the Project, when the detection device is fully assembled.

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Figure 2-5. Dependence of the dose on distance from the working neutron generator (36 hour-long working week, 50 working weeks per year).

2.2. Mathematical modelling of the detection process

2.2.1. Tuning the MCNP4C2 code to describe the geometry of the measurements using available types of detectors

In order to select the measurement methodology, mathematical modeling of experiments with neutron generator were carried out for different types of neutron and gamma-ray detectors, using MCNP4C2 code [2]. Investigated materials — highly enriched uranium (HEU), weapon-grade uranium (WU), reactor plutonium (RPu), weapon-grade plutonium (WPu) and natural lead (Pb) — were assumed to have spherical shapes (diameters 4.62 cm, masses about 1 kg, 0.58 kg for lead).

The isotopic composition of the samples (see Table 1) was taken from (http://nuclear-weapons.nm.ru/theory).

Isotope content (%) Isotope content (%)

235U 238U 239Рu 240Рu

WU 94 6 WPu 93 6

HEU 20 80 RPu 80 17

Table 1. Isotopic composition of uranium and plutonium samples used in calculations.

Calculations were carried out for two cases: when the sample was surrounded by 5 cm-thick lead shielding, and without any shielding (Figure 2-62.6). In both cases samples were located at 30 cm from the isotropic source of 14 MeV neutrons with intensity 108 n/s into 4π. Two (or three) neutron detectors had dimensions ∅6×6 cm and were located at distances of 15 cm on both sides of the sample.

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Stilben detector 6 x 6 cm

Figure 2-6. Model geometry without lead shielding (left) and with lead shielding (right).

As a result of calculations for two geometries shown at Figure 2-6, the following characteristics were obtained:

1. number of neutrons and gamma-rays forming as a result of spontaneous and induced fission in each sample and then reaching the surface of the sample (see Table 2);

2. number of neutrons per second:

a. primary 14 MeV neutrons incident on the sample;

b. going through the sample without interaction;

c. leaving the sample without causing secondary neutrons;

d. causing (n,2n) reactions;

e. causing fission;

1

3. number of fission neutrons leaving the sample per one incident neutron;

4. average multiplicity of secondary neutrons in events, when an incident neutron has caused fission.

WU HEU WPu RPu

Neutrons from induced fission 2.1×105 1.4×105 3.3×105 3.2×105

Neutrons from spontaneous fission — — 1.2×105 3.3×105

Table 2. Number of neutrons from spontaneous and induced fission leaving the sample for the case without lead shielding.

In the same calculations the background due to accidental coincidences and detector cross-talk were calculated for different neutron detection thresholds and neutron and gamma-ray detection efficiencies.

2.2.2. Estimating count rates of neutron and gamma-ray detectors

The method based on detection of coincidences between fission neutrons [3] was selected for detection of shielded fissioning materials. The main idea of the traditional method, as well as its proposed APT modification, is listed below.

Traditional method: no APT.

The inspected volume is irradiated with fast neutrons. The first neutron (or gamma-quantum) detected in one of the detectors opens the window for detection in another detector. The width of this window is determined by the measurement geometry; in geometries shown at Figure 2-6 this window is 30÷40 ns wide. Detection of coincidences in narrow time windows allows one to reduce the background associated with the primary neutron beam from the neutron generator. Calculated number of coincidences within 40 ns time window caused by neutrons from induced fission in 1 kg sample of weapon-grade uranium is 9 s-1, which is few compared to accidental coincidences between primary neutrons and neutrons, scattered on the lead shielding: 200 s-1. Apart from that, calculation show that 20 kg of lead would produce the same effect due to (n,2n) reactions on lead. Thus, existing active method based on double neutron-neutron coincidences can not detect uranium shielded with large amounts of lead (or other heavy metal). Plutonium in lead shield can be detected by a passive method by neutrons from spontaneous fission of 240Pu.

Using triple coincidences between neutron detectors and subsequent determination of neutron multiplicity by Rossi method [4] allows one to significantly reduce the neutron (and gamma-ray) background originating from (n,2n) reaction on lead, inelastic scattering of neutrons, etc. However, the count rate of triple coincidences effect is small (1÷2 s-1), even when all three detectors are as close as 15 cm from the 1 kg sample.

Modified method: with associated particle technique (APT).

Introduction of additional coincidence with alpha-particles, which accompany neutron emission from the neutron generator (t(d,n)α reaction), allows one to substantially (by two orders of magnitude) improve the signal-to-background ratio compared to the traditional method. Due to strict correlation between directions of neutron and the associated alpha-particle, APT selects only those neutrons, that fly on the direction determined by the geometry of the associated particle detector. Time of arrival of secondary neutrons into neutron detectors (see Figure 2-6) relative to the emission time of the primary neutron (which is determined from detection time of the associated alpha-particle) should fall into a short (10÷40 ns-wide) time interval (Figure 2).

Triple coincidences between two neutrons and an alpha-particle allows one to suppress contributions to the background from (n,2n) reactions on the surrounding shielding materials located outside the spatial area, which corresponds to a given segment of the alpha-particle detector.

Figure 2-7 shows results of calculations of the dependence of coincidence count rate on time difference between two neutron detectors (T1-T2) for different samples: WU, RPu, WPu and natPb. The integral effect (sum of events within 40 ns coincidence window) varies for different samples from 9 s-1 (WU) to 30 s-1 (WPu), while the background due accidental coincidences is just 3÷4 s-1 (Figure 2-8).

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Figure 2-7. Time dependence of coincidence count rate of two neutron detectors caused by induced fission of weapon-grade uranium (WU) – for the geometry with lead shielding (Figure 2-6, right). Time is counted

from the moment of emission of the primary 14 MeV neutron from the neutron generator.

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Figure 2-7. Dependence of coincidence count rate on time difference between two neutron detectors (T1-T2) for different samples: WU, RPu, WPu and natPb in the geometry with lead shielding. Neutron detection

threshold was assumed to be 0.5 MeV.

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Figure 2-8. Dependence of coincidence count rate on time difference between two neutron detectors (T1-T2) in the geometry with lead shielding for WU sample and for accidental coincidence with and without APT.

Neutron detection threshold was assumed to be 0.5 MeV.

Integral count rate of the background due to accidental coincidences is 211 s-1 without APT (straight line with squares on top of Figure 2-8), and 3 s-1 with APT (straight line with circles on the bottom of Figure 2-8), i.e. 70 times less. More detailed results for different samples are summarized in Table 3.

Table 3. Number of coincidences in neutron detectors within 40 ns time window for the geometry with lead shielding with and without APT.

1 NG “ON”: induced fission, plus spontaneous fission. 2 NG “OFF”: only spontaneous fission.

WU WPu RPu Pb sample

NG “ON”1

NG “OFF”2

NG “ON”

NG “OFF”

NG “ON”

NG “OFF”

NG “ON”

NG “OFF”

without associated particles detection (no APT)

background (с-1) 211 — 211 — 211 — 211 —

effect (с-1) 9.3 — 39.7 11 54.3 30.2 0.48 —

with associated particles detection (APT)

background (с-1) 3.37 — 3.37 — 3.37 — 3.37 —

effect (с-1) 9.3 — 28.7 11 24.1 30.2 0.48 —

spontaneous fission in APT window (с-1) 0.18 0.48

total effect (с-1) 28.88 24.6

4

Calculations for the geometry with lead shielding (Figure 2-6) showed, that contribution of γ-n and γ-γ coincidences to both effect and background does not exceed 0.3÷0.4%.

2.2.3. Formulating recommendations for the measurement channel of the device

The detailed studies of the measurement channels of the device — neutron and gamma-ray detectors — will be conducted during the second year of the Project.

Preliminary estimations, arising from the expected detection time and detector counting rates, suggest that neutron detection should be based on two or three large-area plastic scintillators with photo multipliers, while gamma-rays will be detected with BGO or NaI(Tl)-based detectors.

The major issues, that would be addressed during the second year of the project will be:

1. Neutron detection sub-system.

a. number and dimensions of the detectors;

b. detector type (with or without n-γ separation properties;

c. shielding type for neutron detectors.

2. Gamma-detection sub-system.

a. number and type of the detectors

b. type of shielding for gamma-ray detectors

3. Data acquisition system.

2.3. Neutron detectors

Based on results of mathematical modeling, a new concept of neutron detection for the proposed device for detection of shielded nuclear materials has been explored.

The main difficulty associated with extraction of coincidences from fission neutrons is connected with very high background in neutron detectors due to detection of primary 14 MeV neutrons from the neutron generator. If neutron detectors are placed outside the “pseudo beam” of tagged neutrons (i.e. those neutrons, for which the moment of their production in the neutron generator and flight directions are known from the detected associated α−particle), time of arrival of these background primary neutrons is uncorrelated with the detection time of α−particles. On the contrary, if neutron detectors are placed within the “pseudo beam” of tagged neutrons (see Figure 2-9), then each detected primary 14 MeV neutron must have its tagging α−particle detected by the associated particle detector located inside the neutron generator. Moreover, the time of arrival of such background primary neutrons to the neutron detector is fixed by the distance from that detector to the target of the neutron generator. Thus, the background becomes correlated in time with the detection of the associated particles, and can be suppressed by performing time-of-flight analysis of the detected neutrons relative to tagging α−particles.

5

2 1

6

5

3

4 Figure 2-9. Possible layout of the detection device with three neutron detectors. 1 – position of the target in

the neutron generator; 2 – extension with an associated particle detector and electronics; 3 – neutron detectors; 4 – “pseudo beam” of tagged neutrons; 5, 6 – electronics and computer.

Three neutron detectors based on plastic scintillator with dimensions 7×7×21 cm3 were produced (see photo on Figure 2-10). Detectors can be stacked together to form a square with external dimensions about 25×25 cm2 and 8 cm thick, which fits into the “pseudo beam” of tagged 14 MeV neutron at distance about 55 cm from the target of the neutron generator. Thus, any primary neutron detected by this assembly should in principle have a counterpart α−particle detected at one of the nine segments of the associated particle detector. Thus, for each such “background” neutron its time-of-flight is known: they all should arrive to neutron detectors about 11 ns after being emitted from the NG target (they all have ~5 cm/ns velocity; 5cm/ns × 11ns=55cm flight path).

It was estimated, that about 90% of primary neutrons pass through the assembly of neutron detectors unaffected, so such detector placement does not significantly reduce the intensity of neutrons incident on the inspected object, which can be located very close to (e.g. right behind) neutron detectors.

Figure 2-10. Neutron detector based on plastic scintillator and a photo multiplier.

2.4. Experiments with uranium samples

2.4.1. Geometry

First experiments were carried out using four aluminum-coated cylinders (∅2×7 cm3) of metallic depleted uranium. These samples were placed at different locations relative to one neutron detector, as shown on Figure 2-11 and Figure 2-12.

6

neutron detec tor(7x7x21 cm )3

NG target

58 cm

SIDE VIEW FRONT VIEW

Figure 2-11. Measurement geometry. Rectangular neutron detector was placed at 58 cm from the target of the neutron generator (NG) so, that if fit within the “voxels” corresponding to three segments of the

associated particle detector (“pseudo beam” of tagged neutrons).

58 cm

238U sample

28 cm

A

58 cm

238U sample

20 cm

B

58 cm

238U sampleC

Figure 2-12. Three geometries, in which measurements with the U sample were carried out: A – sample between the NG target and the neutron detector; B – sample behind neutron detector; C – sample above

neutron detector. The neutron detector was inside the “pseudo beam” of tagged neutrons (see Figure 2-11).

2.4.2. Results

One of the most important problems in detection of a weak flux of the secondary neutrons emitted in the induced fission of hidden fissile materials is reduction of the neutron and gamma backgrounds. Considerable part of this background comes from primary 14 MeV neutrons that are directly detected by the neutron detector.

In order to investigate this problem the background measurements for two locations of the neutron detector with respect to the “pseudo beam” of neutrons tagged by α−particles detected in different segments of the associated α−particle detector have been performed.

In the first series of experiments a neutron detector was placed outside the “pseudo beam” of tagged neutrons, so that α−particles, which accompanied 14 MeV neutrons hitting this detector, were not detected in the associated α−particle detector. In this case all the accidental coincidences between events in the neutron detector and α−particles were uncorrelated in time.

In the second series of experiments the neutron detector was placed inside the “pseudo beam” of tagged neutrons (see examples of these geometries at Figure 2-12). In this case most of the primary 14 MeV neutrons hitting the neutron detector had their accompanying α−particle detected in one of the segments of the associated α−particle detector, so they produced a sharp peak in time-of-flight distribution. If the time resolution of the whole system

7

allows one to separate between the peak of primary 14 MeV neutrons from the γ−rays and neutrons from induced fission, this geometry is preferable. In this case the neutron detectors are located at the closest possible point to the inspected object, and this may overweight the reduction of the 14 MeV neutron beam on the inspected object by about 10% due to reactions in the neutron detector.

-60 -40 -20 0 20 401

10

100

1000

10000

γ-rays from NG target

Yiel

d

Time [ns]

detector inside tagged neutron "pseudo beam" detector outside tagged neutron "pseudo beam"

14 MeV neutrons

8ns

Figure 2-13. Background time distribution measured by means of neutron detector situated inside (dashed

line) and outside (solid line) the “pseudo beam” of tagged neutron.

Figure 2-13 demonstrates the difference in the time spectrum for different locations of the neutron detector. The background level estimated over the part of the spectrum corresponding to pure accidental coincidences (to the left from the “0” time – moment of emission of the 14 MeV neutron from the NG) and normalized to the total number of detected associated α−particles is approximately 10% higher in the case of the detector located outside the “pseudo beam” of tagged neutrons. This is due to the fact, that primary 14 MeV neutrons, that would have otherwise contributed to this uniformly-distributed background, have their accompanying α-particle detected in the associated particle detector inside the neutron generator, and are “collected” within a narrow peak corresponding to their time-of-flight from the NG target to the neutron detector. On the other hand, the background level at the right side of the spectrum (to the right from the 14 MeV neutron peak) is higher in the case of neutron detector inside the “pseudo beam” due to a “tail” of scattered neutrons. Thus, the optimal choice of the detector geometry depends on the relative positions of the NG target, neutron detector and the inspected object. This question will be addressed in further studies.

The goal of the next set of experiments was to determine time resolution of the α−detector – neutron-detector pair, and to measure time-of-flight spectra of different spectral components. Measurement results are summarized on Figure 2-14.

The top time-of-flight spectrum corresponds to measurement without sample. Only events with α−particle detected in the central segment of the nine-segment α−detector were selected. No neutron/gamma discrimination was used. The main contribution to the spectrum comes from primary 14 MeV neutrons hitting the neutron detector (its central part corresponds to the central segment of the α−detector, see Figure 2-11). The width of this peak is FWHM = 1.55 ns, which includes time resolution of the α-n detector pair, as well as contribution from final dimensions of the α−detector segment neutron detector (width 7 cm). Thus, intrinsic time resolution of the α-n detector pair (plus electronics) is about 1 ns, which is enough for time-of-flight analysis.

Another feature of the top spectrum is a small peak coming from γ−rays produced in the material of the target of neutron generator. These γ−rays reach the neutron detector in about 2 ns, compared to ~11 ns needed for that for a 14 MeV primary neutron. So, the peak corresponding to these γ−rays is located about 10 ns left from the main 14 MeV neutron peak.

Another small peak ~8 ns after the 14 MeV peak comes from γ−rays produced by fast neutrons in the construction materials of the experimental setup.

8

-5 0 5 10 15 20 25 30 35 4010

100

1000

1000010

100

1000

1000010

100

1000

1000010

100

1000

10000

effect (n) no sample geometry C

U above detector

Cou

nts

[rela

tive

units

]

Time [ns]

no sample geometry B

U behind detector

shielding by U

γ-rays from NG target

no sample geometry A

U in front of detector

B ###

14MeV neutronsfrom NG target

effect (γ)

no sample

effect (n)

10ns

Figure 2-14. Time-of-flight spectra of events recorded in the neutron detector relative to associated α−particles detected in the built-in α−detector. Spectra obtained without sample for each case are shown

with thin lines.

U sample in front of neutron detector (geometry A).

When four U cylinders were placed between the NG target and the neutron detector (second spectrum from top on Figure 2-14), the following effects can be seen:

1. A small peak of γ−rays produced in NG target is reduced compared to the case without U sample due to shielding of the central part of the neutron detector by U cylinders.

2. A peak of prompt fission γ−rays appears ~5 ns after the γ−rays from the NG target and ~5 ns before the main 14 MeV neutron peak. These 5 ns represent extra time needed for 14 MeV neutrons to reach the U sample (flight path 28 cm).

3. Fission neutron are not visible, since they fall right in the huge main peak of 14 MeV neutrons.

U sample behind neutron detector (geometry B).

In geometry B effect from fission neutron can be seen as a wide distribution around TOF = 25 ns. Out of this time, ~16 ns are needed by 14 MeV neutrons to reach the U sample (a total flight path of ~80 cm), and the remaining 9 ns are needed by slower fission neutrons (which have average velocity about 2 cm/ns) to fly back to the neutron detector (flight path ~20 cm). Γ−rays from fission in this case are not visible, since they fall right into the huge peak of 14 MeV primary neutrons. Though the time distribution of fission neutrons is rather broad, the integral number of events in it is consistent with the geometry of the experiment.

U sample above neutron detector (geometry C).

This experiment was carried out in order to demonstrate the use of position sensitivity of the associated-particle detector built into the neutron generator. The uranium sample was located in the area above the neutron detector corresponding to segment #6 of the associated-particle detector. So, any effect from this sample was expected in coincidence with segment #6, not with the central segment #9, as in the previous experiments. Time-of-flight spectrum from neutron detector relative to α−particles detected in segment #6 of the associated particle detector is shown on the bottom panel of Figure 2-14. The spectrum was not corrected for the time shift, which is due to the

9

geometry of the associated particle detector, and the effect of scattering of α−particles on the NG target, so the “zero” time on the time axis in this case does not coincide with the “0” label of the axis.

The main peak of primary 14 MeV neutrons in this experiment is a factor of ten lower than in the previous cases. This reflects the fact, that only a small part of the neutron detector falls into the “tail” of the area corresponding to segment #6 of the tagging α−detector. The peak of γ−rays from the NG target is well pronounced, since neutrons coinciding with α−particles detected in segment #6 have to pass through thicker layer of the target holder than those coinciding with α−particles from segment #9.

A rather broad and well-pronounced peak about 5 ns after the main peak from 14 MeV neutrons corresponds to fission neutrons from the U sample. These extra 5 ns are needed by neutrons to travel ~10 cm gap separating the U sample and the neutron detector. Fission γ−rays coincide in time with the maximum of the main peak from primary 14 MeV neutrons.

Role of amplitude distribution.

Another task was to investigate the role of amplitude distribution measured by the neutron detector for better discrimination of fission neutrons. Over 80% of fission neutrons have energies below 5 MeV, while most of the detected 14 MeV neutrons leave more than 5 MeV in the detector. In Figure 2-15 an example of a two-dimensional plot measured in the geometry shown in Figure 2-12C is presented. It can be seen that at amplitudes above the region, where fission neutrons are expected, there are some background events (most probably scattered incident neutrons), which can be removed by introducing of an upper amplitude threshold.

20 25 30 35 40100

200

300

Time [ns]

Ener

gy [c

hann

el]

2.0002.8604.0905.8488.36311.9617.1024.4534.9750.00U

region of interest

Figure 2-15. Two-dimensional amplitude-time distribution measured with U sample in the geometry shown

in Figure 2-12C.

In Figure 2-16 a time-axis projection of the distribution shown at Figure 2-15 is given with and without pulse-height discrimination in the amplitude channel (events with energies over 145 ch were rejected). The spectra were normalized to the maximum of the peak at –15 ns corresponding to simultaneous arrival of prompt γ-rays created by 14 MeV neutrons in the NG target. One can note, that the fission neutron distribution in the vicinity of –4ns is more pronounced in the discriminated case, and the peak corresponding to the incident 14 MeV neutrons is somewhat suppressed.

10

-20 -10 0 10 20 30 40

0.01

0.1

1

Yiel

d [a

rb. u

nits

]

Time [ns]

all amplitudes only lower amplitudesU

Figure 2-16. Time distribution corresponding to two-dimensional distribution shown in Figure 2-15.

Black line – spectrum with the pulse height discriminated below channel 145 (see Figure 2-15).

2.5. Experiments with shielded uranium

First measurements with samples of natural uranium in lead container have been carried out (see Figure 2-17). Neutrons and gamma-rays were measured in coincidence with the associated alpha-particle detector. Double (n-α), triple (n-n-α) and quadruple (n-n-n-α) coincidences have been collected.

Experiments were carried out, using equipment developed for the NATO Science for Peace project.

Data analysis is currently under way.

Figure 2-17. Measurements with natural uranium surrounded by lead.

3. CONCLUSIONS DRAWN

Application of the associated particles technique (APT) to detection of shielded fissioning materials offers a unique possibility to build an all-in-one portable system for detection of both neutron- and γ−ray emitters and materials with no significant spontaneous radiation.

11

Results of the advanced Monte-Carlo calculations and first experiments show, that introduction of the APT technique into the traditional method of detection triple neutron-neutron-alpha coincidences allows one to:

1. detect ~1 kg of fissioning materials (FM) shielded with 5 cm of lead in less than one second;

2. identify the detected FM by spontaneous-to-induced fission ratio and other fission characteristics;

3. significantly improve the effect-to-background ratio and the detection time, relative to the traditional method without APT;

4. localize the detected object with precision determined by the geometry of the device and type of the associated particle detector.

When the 5 cm-thick lead shielding is complimented with a massive shielding against neutrons (e.g., polyethylene or water), the neutron coincidence count rate will drop. However, presence of such additional shielding can be determined by using an additional detector of γ−rays to detect secondary γ−radiation from inelastic scattering of primary (14 MeV) neutrons. Presence of most light chemical elements (carbon, oxygen, etc.) can be detected by characteristic peaks in the spectrum of this γ−radiation, measured using APT technique.

First experiments with one neutron detector and real uranium samples have shown, that observable effect from fission neutrons can be obtained even without use of neutron-neutron coincidences. The possible role of energy discrimination on time-of-flight spectra has been investigated. Introduction of additional neutron detectors and measurement of triple n-n-α and quadruple n-n-n-α coincidences within the time windows determined in the first experiments would allow one to account for different sources of the background in neutron detectors.

Experimental work using three neutrons detectors for detection of triple (n-n-α) and quadruple (n-n-n-α) coincidences between neutrons from induced fission of shielded fissioning materials is currently under way.

4. REFERENCES

[1] Kuznetsov, A.V., Using of the associated particle technique for the detection of shielded nuclear material. Proc. of Research Coordination and Technical Meeting (M2-RC-927), December 1–5, 2003 IAEA Headquarters, Vienna

[2] Vakhtin, D.N., “Using Associated Particle Technique for Detection of Shielded Nuclear Materials”, Report at the Second RCM of the IAEA CPR “Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials”, 4–8 October 2004, Sochi, Russia

[3] Kuznetsov, A.V., et al., “Portable multi-sensor for detection and identification of explosive substances”. //Proc. of the International Conference EUDEM2-SCOT: “Requirements and Technologies for Demining and the Removal/Neutralization of Unexploded Ordnance”, 15–18 September, 2003, Vrije Universiteit Brussel, Bruxelles, Belgium. H.Sahli, A.M. Bottoms, J.Cornelis (Eds.), pp. 633–637

12

EXECUTIVE SUMMARY

1. Contract Number: #12600/Nuclear Security Multi-Donors Fund

2. Title of Project:

“Using associated particle technique for detection of shielded nuclear materials”

3. Institute where research is being carried out: V.G.Khlopin Radium Institute

2-j Murinski pr., 28 194021 St.Petersburg RUSSIA Tel. (812)-2475737 FAX: (812)-2478095 E-mail: [email protected]

4. Chief Scientific Investigator: A.V. Kuznetsov, Director of Physics Department, Ph.D.

5. Time period covered: 15th September, 2003–31th May, 2006

6. Scientific Background and Scope of Project

One of the weak points of all existing instruments used to detect and characterize radioactive materials, including fixed automated portals, is their inability to detect fissioning materials that are shielded by substances that absorb their spontaneous gamma and neutron radiation. While spontaneous radiation emitted by some of the most dangerous fissioning materials cannot be used for their detection and identification, these materials can still be detected by the so-called “active” methods. These methods use external radiation to induce fission reactions within the irradiated volume and then measure such products of these reactions as neutrons and high-energy γ−rays, which have a very high penetrating ability and can penetrate thick layers of shielding material.

The main goal of the Project is to apply active methods to the detection of shielded fissioning materials.

7. Experimental Method

The proposed device will be based on the Nanosecond Neutron Analysis (NNA) method, which rely on measurement of correlations between 14 MeV neutrons produced by a portable neutron generator, and secondary radiation (neutrons and energetic γ−rays), that is produced by these neutrons in the material of the investigated object. The secondary radiation will be detected in very narrow (few nanoseconds) time intervals counted from the time of emission of each neutron from the neutron generator. This time, in turn, will be determined by detecting α−particles, that accompany neutron emission in the reaction d+t n+a, by a position-sensitive or segmented semiconductor detector built into the sealed vacuum of the neutron generator. By detecting secondary radiation in coincidence with α−particles, one would be able to select the region of space, in which this secondary radiation has been produces by the primary neutron, thus achieving substantial suppression of the background arising from reactions of the primary neutrons on surrounding materials.

8. Results obtained

Experimental work.

• Three neutron detectors have been made on the basis of fast plastic scintillator and a photo multiplier. These detectors use measurement channels of the data acquisition system, which has been built specially for experiments with the neutron generator. This data acquisition system can handle extremely high counting rates in both neutron channel and “tagging ” α−particle channel.

• Experiments with an APT neutron generator, one neutron detector, and depleted U sample were carried out. Time-of-flight spectra of neutrons measured in coincidence with associated α−particles were analyzed. Measurements with different neutron detector locations have been carried out. The possible role of energy discrimination on time-of-flight spectra has been investigated.

• First experiments with shielded and unshielded uranium samples have been carried out. Data analysis is currently under way.

13

Mathematical modeling.

• Full mathematical model of the experimental setup has been created on the basis of MCNP5 Monte-Carlo code.

9. Conclusions drawn

The characteristics of the created detectors and electronics are adequate for the task of using them in a system for detection of shielded nuclear materials by NNA technique.

Experimental work using three neutrons detectors for detection of triple (n-n-α) and quadruple (n-n-n-α) coincidences between neutrons from induced fission of shielded fissioning materials is currently under way.

10. Papers Published on Work Done under the Contract

[1] Kuznetsov, A.V., Using of the associated particle technique for the detection of shielded nuclear material. Proc. of Research Coordination and Technical Meeting (M2-RC-927), December 1–5, 2003 IAEA Headquarters, Vienna.

[2] Vakhtin, D.N., “Using Associated Particle Technique for Detection of Shielded Nuclear Materials”, Report at the Second RCM of the IAEA CPR “Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials”, 4–8 October 2004, Sochi, Russia.

[3] Kuznetsov, A.V., et al., “Portable multi-sensor for detection and identification of explosive substances”. //Proc. of the International Conference EUDEM2-SCOT: “Requirements and Technologies for Demining and the Removal/Neutralization of Unexploded Ordnance”, 15–18 September, 2003, Vrije Universiteit Brussel, Bruxelles, Belgium. H.Sahli, A.M. Bottoms, J.Cornelis (Eds.), pp. 633–637.

Development of Methodical Recommendations on Detection of FRM in Objects of the Customs Control Using Stationary Customs System

of Detection of FRM «Yantar», Search Instruments РМ1703М, РМ1703GN, PM1401GN and Universal Radiometers-Spectrometers

MKS-A03 and IdentiFINDER-NGH

A.V. Borisenko, V.N. Kustov, L.G. Eliseenko, V.V. Temchenko, O.G. Alehina, B.A. Semerkov, N.E. Kravchenko, I.N. Bannyh, D.J. Danko

The Russian Customs Academy, the Vladivostok Branch, Vladivostok, Federal Customs Service of Russia, Moscow

Abstract

This research is related to the development of organizational-technical procedures of application of

means of the radiation control and is aimed at increase of efficiency of measures on preclusion of

illegal FRM movement across the border. The research results are based on the authors' great

experience in organizing FRM customs control in the Russian Federation The procedures of

application (methodical recommendations) include:

1. Procedures for preparing and checking the working capacity of instruments; 2. The procedural order of using these instruments 3. The order and sequence of actions when carrying out of the radiation control of passengers,

luggage, goods and vehicles at customs pedestrian and automobile check points.

Specific methodical recommendations are developed for the following instruments:

1. The stationary customs systems of detecting FRM («Yantar» Radiation Portal Monitors) 2. Radioisotope search instruments РМ1401, РМ1703 (and their modifications) 3. Radiometers-spectrometers MKS -А02, MKS-A03 (and their modifications) 4. Radiometer-spectrometer IdentiFINDER and its modifications.

Разработка методических рекомендаций по обнаружению ДРМ в объектах таможенного контроля с использованием стационарной таможенной системы обнаружения ДРМ «Янтарь», поисковых

измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH

А.В. Борисенко1, В.Н. Кустов1, Л.Г. Елисеенко1, В.В. Темченко1, О.Г. Алехина1, Б.А. Семерков1, Н.Э. Кравченко2, И.Н. Банных2, Д.Ю. Данько2

(Российская таможенная академия, Владивостокский филиал, г. Владивосток, Федеральная таможенная служба России, Москва)

РЕЗЮМЕ

Работа представляет собой краткое изложение итогового отчета о результатах,

полученных при выполнении научно-исследовательского контракта МАГАТЭ №12601\Фонд ядерной безопасности «Совершенствование методик применения технических средств таможенного контроля за незаконным перемещением делящихся и радиоактивных материалов (ДРМ) через таможенную границу».

Исследования связаны с разработкой организационно-технических процедур применения технических средств радиационного контроля и преследуют цель повышение эффективности мер по пресечению попыток незаконного перемещения ДРМ через границу.

Процедуры применения (методические рекомендации) включают в себя: − процедуры подготовки к работе и проверки работоспособности технических

средств; − порядок эксплуатации технических средств; − порядок и последовательность действий при проведении радиационного контроля

пассажиров, багажа, товаров и транспортных средств на пешеходном и автомобильном пунктах пропуска через таможенную границу. Методические рекомендации разработаны для следующих технических средств:

− различные модификации стационарной таможенной системы обнаружения (СТСО) ДРМ «Янтарь»;

− поисковые измерители РМ1401, РМ1703 и их модификации; − радиометры-спектрометры МКС-А02, МКС-А03 и их модификации; − радиометр-спектрометр IdentiFINDER и его модификации. Эти технические средства рекомендованы МАГАТЭ для применения при организации

национальных систем контроля за ДРМ и предназначены для обнаружения ДРМ, измерения их радиационных характеристик и испускаемого ими ионизирующего излучения, а также идентификации объектов контроля. В основе приведенных в работе результатов исследования - большой практический опыт полученный авторами при организации таможенного контроля за ДРМ в Российской Федерации.

1 Владивостокский филиал Российской таможенной академии, г. Владивосток. 2 Федеральная таможенная служба, г. Москва.

Разработка методических рекомендаций по обнаружению ДРМ в объектах таможенного контроля с использованием стационарной

таможенной системы обнаружения ДРМ «Янтарь», поисковых измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03

и IdentiFINDER-NGH

А.В. Борисенко1, В.Н. Кустов1, Л.Г. Елисеенко1, В.В. Темченко1, О.Г. Алехина1, Б.А. Семерков1, Н.Э. Кравченко2, И.Н. Банных2, Д.Ю. Данько2

(Российская таможенная академия, Владивостокский филиал, г. Владивосток, Российская Федерация)

РЕЗЮМЕ

Работа выполнена в рамках контракта № 12601 «Совершенствование методик применения технических средств таможенного контроля за незаконным перемещением делящихся и радиоактивных материалов (ДРМ) через таможенную границу», заключенного между Международным агентством по атомной энергии (МАГАТЭ) и Владивостокским филиалом Российской таможенной академии (ВФ РТА) 22.09.2003 г.

Научный руководитель контракта – Борисенко А.В., начальник Учебно-методического центра таможенного контроля делящих и радиоактивных материалов (УМЦ ТКДРМ) ВФ РТА.

Организация-исполнитель контракта – Владивостокский филиал Российской таможенной академии.

Субподрядчик – временный творческий коллектив, состоящий из сотрудников УМЦ ТКДРМ ВФ РТА и Службы ТКДРМ Федеральной таможенной службы (ФТС России).

Исследования по программе проводились в два этапа. В ходе выполнения первого этапа были разработаны методические рекомендации по

применению различных модификаций стационарной таможенной системы обнаружения (СТСО) ДРМ «Янтарь», а также поискового измерителя РМ1703GN для осуществления таможенного контроля ДРМ.

Целью второго этапа являлась разработка методических рекомендаций по применению поисковых измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для обнаружения, локализации ДРМ в объектах таможенного контроля и измерения их радиационных характеристик.

Итогом работы явилось: издание учебного [4] и учебно-методического пособий [5], подготовка к изданию отчетов о выполнении первого [6] и второго [7] этапов контракта. По результатам работы изданы учебник [1] и учебно-методические пособия [2, 3]. Частично результаты работы изложены в статьях сборников [8, 9].

1 Владивостокский филиал Российской таможенной академии, г. Владивосток. 2 Федеральная таможенная служба, г. Москва.

СОДЕРЖАНИЕ Введение 1. Общие положения 2. Стационарная таможенная система обнаружения (СТСО) ДРМ «ЯНТАРЬ»

2.1. Эксплуатация, техническое обслуживание, текущий контроль и проверка работоспособности СТСО ДРМ «Янтарь»

2.2. Процедуры по применению СТСО ДРМ «Янтарь» при таможенном контроле пассажиров, товаров и транспортных средств 2.2.1. Методика проведения контроля на пешеходных переходах с помощью

систем «Янтарь-1П, -1П1», «Янтарь-2П» 2.2.1.1. Общие замечания 2.2.1.2. Порядок проведения радиационного контроля пассажиров и их багажа

2.2.2. Методика проведения контроля на автомобильных пунктах пропуска с помощью СТСО ДРМ «Янтарь-1А, -2А» 2.2.2.1. Общие положения 2.2.2.2. Контроль автобусов с пассажирами 2.2.2.3. Контроль грузового автотранспорта

2.2.3. Методика проведения контроля на железнодорожном транспорте с помощью СТСО ДРМ «Янтарь-1Ж, -2Ж» 2.2.3.1. Общие замечания 2.2.3.2. Контроль грузовых железнодорожных составов

3. Поисковые измерители РМ1703М, РМ1703GN, PM1401GN и универсальные радиометры-спектрометры МКС-А03 и IdentiFINDER-NGH 3.1. Общие положения 3.2. Тестовые испытания поисковых измерителей РМ1703М, РМ1401GN, РМ1401GN и

радиометра-спектрометра МКС-А03 3.3. Методические рекомендации по использованию поисковых измерителей РМ1703М,

РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для таможенного контроля пассажиров, багажа и транспортных средств. 3.3.1. Методические рекомендации по подготовке и применению поисковых

измерителей РМ1703М, РМ1703GN, PM1401GN 3.3.1.1. Назначение и область применения поисковых измерителей 3.3.1.2. Процедуры по подготовке к работе 3.3.1.3. Процедуры по применению поисковых измерителей при таможенном контроле

пассажиров, товаров и транспортных средств 3.3.1.4. Особенности применения поисковых измерителей с нейтронным каналом

3.3.2. Методические рекомендации по подготовке и применению универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH 3.3.2.1. Подготовка радиометров-спектрометров к работе, их функциональные

возможности 3.3.2.2. Методические рекомендации по применению универсальных

радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH

Заключение

Список литературы

ВВЕДЕНИЕ

Согласно правилам, определенным международными договорами, конвенциями и соглашениями, перемещение ядерных и других радиоактивных материалов в пределах государства и между государствами в целях обеспечения безопасности должно происходить с соблюдением строгих регулирующих, административных и инженерно-технических мер контроля. В случае делящихся материалов определены дополнительные требования относительно их физической защиты, учета и контроля. В каждом государстве указанная система контроля строится на основании существующих национальных законодательств, норм и правил, с учетом рекомендаций Международного агентства по атомной энергии.

Важным элементом этой системы является контроль делящихся и радиоактивных материалов (ДРМ) на границах, решающий две главные задачи: пресечение незаконного перемещения ДРМ через границу и регулирование внешнеэкономических операций при осуществлении легальных коммерческих сделок с ДРМ. В различных странах, в зависимости от национального законодательства, эти задачи решаются по-разному. В одних – пресечение незаконного перемещения ДРМ – задача пограничных служб, а регулирования – таможенных, в других, например, в Российской Федерации, обе эти задачи выполняют таможенные органы.

Несмотря на различие в подходах, эффективность решения обеих задач в значительной степени зависит от комплекса организационно-технических мероприятий, связанных с использованием технических средств радиационного контроля. Этот комплекс включает следующие элементы:

– совокупность форм и этапов контроля; – номенклатура технических средств, применяемых на каждом этапе; – методическая обеспеченность проведения форм, этапов контроля и особенностей

применения технических средств применительно к ним и к объектам контроля. На эффективность контроля особенно влияет последний элемент, поскольку ни

отличные технические характеристики приборов, ни их количество не позволят решить должным образом задачу, если не будут разработаны и строго использованы процедуры подготовки и применения технических средств применительно к этапам и объектам контроля. Это объясняется, прежде всего, особенностями объектов контроля (радиационная опасность, незначительные габаритные размеры, неотличимость по внешним признакам от нерадиоактивных объектов, использование маскировки в случае их незаконного перемещения) и технических средств контроля (внедрение значительного количества новых приборов, специально предназначенных для решения задач контроля ДРМ, сложность их эксплуатации, зависимость показаний приборов от внешних факторов, способов применения и т.п.).

Таким образом, разработка процедур применения технических средств радиационного контроля в целях повышения эффективности контроля ДРМ на границах особенно актуальна и важна. Однако до настоящего времени практически отсутствовали должным образом разработанные методические рекомендации по применению технических средств радиационного контроля, поступающих на вооружение в таможенные и другие правоохранительные органы, в том числе рекомендованные МАГАТЭ для применения при организации контроля ДРМ на границах в целях:

– обнаружения ДРМ, – измерения их радиационных характеристик и испускаемого им ионизирующего

излучения, – идентификации объектов контроля. В связи с этим начатые Владивостокским филиалом Российской таможенной академии

при поддержке МАГАТЭ, в рамках контракта №12601, исследования по разработке эффективных методов использования технических средств радиационного контроля, отличаются новизной и актуальностью.

В число этих технических средств ТКДРМ, прежде всего, входят: – различные модификации стационарной таможенной системы обнаружения

(СТСО) ДРМ «Янтарь»; – поисковые измерители РМ1401, PM1703 и их модификации; – радиометры-спектрометры МКС-А02, -03 и их модификации; – радиометр-спектрометр IdentiFINDER и его модификации. Приборы данного типа активно поступают на вооружение таможенных и других

правоохранительных органов большинства государств СНГ и хорошо зарекомендовали себя в реальных условиях таможенного контроля ДРМ. Кроме того, некоторые из указанных приборов (PM1401, PM1703) успешно прошли испытания по программе испытания приборов радиационного контроля ITRAP и рекомендованы МАГАТЭ к применению в различных государствах. Поэтому разработанные в рамках данного контракта методические рекомендации должны представлять интерес для потенциальных пользователей большинства государств СНГ, а также других стран, создающих национальные системы таможенного контроля перемещения ДРМ. В этом заключается практическая значимость проводимых исследований.

В процессе исследований, которые проводились в два этапа, решены следующие задачи: – разработаны методические рекомендации1 по подготовке к работе и использованию

СТСО ДРМ «Янтарь» (Янтарь-1А, -2А, -1С, -1П, -2П, -1Ж, -2Ж) (изложены в [4]); – разработаны методические рекомендации по обнаружению ДРМ в объектах

таможенного контроля с помощью различных модификаций СТСО «Янтарь» (изложены в [6]; – разработаны предшествующие таможенному контролю ДРМ процедуры по подготовке к

работе поисковых измерителей РМ1703М, РМ1703GN, PM1401GN, универсальных радиометров-спектрометров МКС-А03, IdentiFINDER-NGH и методические рекомендации по их использованию для таможенного контроля пассажиров, багажа и транспортных средств (изложены в [5]);

– издано учебное справочное пособие «Стационарная таможенная система обнаружения ДРМ «Янтарь-1П» [4];

– подготовлены к изданию учебное справочное пособие «Технические средства ТКДРМ: поисковые измерители РМ1703М, РМ1703GN, PM1401GN, универсальные радиометры-спектрометры МКС-А03, IdentiFINDER-NGH: методические рекомендации по применению» [5] и отчеты о выполнении первого и второго этапов работы [6, 7].

Приведённые ниже методические рекомендации с учётом ограничения объёма статьи изложены с определёнными сокращениями. Это касается как СТСО ДРМ «Янтарь» так и поисковых измерителей и переносных идентификаторов, прежде всего, если речь идет о вопросах реагирования на срабатывания технических средств ТКДРМ, связанные со спецификой ТКДРМ в Российской Федерации. Не приведены методические рекомендации по применению поискового измерителя РМ1401 и его модификаций.

Полное изложение результатов научных исследований в [4–7].

1 В понятие «методические рекомендации» входит подробное описание правил и порядка применения технического средства того или иного типа при осуществлении таможенного контроля ядерных и других радиоактивных материалов, перемещаемых через таможенную границу в различных видах багажа, товаров и транспортных средств.

1. ОБЩИЕ ПОЛОЖЕНИЯ В настоящее время все разнообразие приборов радиационного контроля для обнаружения

радиоактивных материалов на границах, в соответствии с рекомендациями МАГАТЭ [11, 13], может быть разделено на следующие категории: стационарные автоматические системы (fixed, radiation portal monitors), персональные поисковые измерители (Personal Radiation Detectors (PRDs) (далее – поисковые измерители), универсальные переносные идентификаторы для поиска и идентификации радиоизотопов (Multi-purpose Handheld Radioisotope Identifier Devices (RID).

Если движение товаров, транспортных средств или людей может быть сосредоточено в узких границах, известных как центральные пункты пропуска, для радиационного контроля лучше всего использовать стационарные системы.

Поисковые измерители и переносные идентификаторы особенно полезны в местах, где контроль производится на обширных площадях типа аэропортов или морских портов. Например, поисковый измеритель может иметь при себе каждый таможенник при исполнении должностных обязанностей.

Переносные идентификаторы более чувствительны к обнаружению ДРМ и выполняют гораздо больше задач, по сравнению с поисковыми измерителями, но они обычно более дорогие.

Переносные идентификаторы главным образом применяются в случаях: - когда имеется подозрение в незаконной торговле ДРМ; - для локализации источника ионизирующего излучения; - для измерения мощности дозы; - для идентификации радионуклида.

Существует несколько направлений использования технических средств таможенного контроля ДРМ, и каждое из них определяет выбор соответствующего прибора.

В общем виде основные задачи использования технических средств ТКДРМ с целью пресечения незаконного перемещения ДРМ через границу могут быть сформулированы следующим образом [11]:

- обнаружение – прибор необходим, чтобы выдать сигнал тревоги, если превышен некоторый пороговый уровень излучения;

- проверка – как только возникла тревога, необходимо проверить, действительно ли она подлинная. Для чего необходимо произвести повторное обследование объекта, убедиться, что сигнал срабатывания не был ложным;

- поиск, локализация и оценка радиационной безопасности – при реальной тревоге требуется найти и локализовать источник происхождения радиации, провести оценку радиационной безопасности, чтобы гарантировать безопасность таможенных служащих и окружающих, а также для определения соответствующего уровня реагирования;

- идентификация – после локализации источника излучения необходимо идентифицировать радионуклид.

Эффективность решения указанных задач и, соответственно, эффективность мер по предотвращению незаконного оборота ядерных и радиоактивных материалов на границах государств зависит не только от технических характеристик приборов радиационного контроля, но и в значительной степени от комплекса организационных и технических мероприятий, связанных с использованием технических средств радиационного контроля, определяемого особенностями национального законодательства.

В Российской Федерации задачи пресечения незаконного перемещения ДРМ через границу и регулирование внешнеэкономических операций при осуществлении легальных коммерческих сделок с ДРМ возложены на Федеральную таможенную службу, которая выстраивает свою систему таможенного контроля ДРМ руководствуясь, прежде всего, Таможенным кодексом РФ и Федеральным законом «Об использовании атомной энергии» –

основными документами, регламентирующими общие вопросы таможенного дела и атомного законодательства.

В соответствии с Таможенным кодексом РФ применение технических средств при проведении таможенного контроля ДРМ (радиационного контроля) происходит в ходе осуществления следующих форм таможенного контроля и этапов радиационного контроля:

- таможенное наблюдение (этап первичного радиационного контроля); - таможенный осмотр (этап дополнительного радиационного контроля товаров и

транспортных средств без их вскрытия); - таможенный досмотр (этап дополнительного радиационного контроля и углубленного

радиационного обследования). Задача первичного радиационного контроля (таможенное наблюдение) – выявление

объектов с повышенным относительно естественного радиационного фона уровнем ионизирующего излучения. Его осуществляют должностные лица таможенных органов, участвующие в таможенном контроле товаров и транспортных средств (требования по допуску к работам с источниками ионизирующего излучения (ИИИ) к указанным должностным лицам не предъявляются). На этом этапе таможенного контроля используются стационарные системы обнаружения ДРМ или одно-, двухканальные поисковые измерители.

Дополнительный радиационный контроль осуществляется в рамках таможенного осмотра и досмотра товаров и транспортных средств. Его задача – поиск и локализация ИИИ в составе объекта, со вскрытием либо без вскрытия объекта, измерение радиационных характеристик и оценка степени радиационной опасности. Его осуществляют должностные лица таможенных органов, имеющие допуск к работам с ИИИ. Используемые технические средства – поисковые измерители, персональные дозиметры, универсальные радиометры-спектрометры. Основания для его проведения – результаты таможенного наблюдения и дополнительного радиационного контроля в рамках таможен-ного осмотра.

Углубленное радиационное обследование (УРО) проводится в рамках таможенного досмотра товаров и транспортных средств со вскрытием объекта. Его задача – максимально возможная локализация, первичная идентификация ИИИ и предварительное отнесение ИИИ к одной из следующих групп: ядерные материалы или изделия на их основе, радиоактивные вещества или изделия на их основе, радиоактивные отходы; иные товары и транспортные средства с повышенным содержанием радионуклидов. Его осуществляют должностные лица таможенных органов, имеющие допуск к работам с ИИИ. Используемые технические средства – универсальные радиометры-спектрометры, гамма-спектрометры. Основанием для проведения УРО являются результаты дополнительного радиационного контроля.

Уровень реагирования на каждом этапе радиационного контроля зависит от оценки радиологической ситуации. Большинство ситуаций, с которыми сталкиваются сотрудники таможенных органов, связаны с небольшой опасностью и в значительной степени могут быть урегулированы лицами, не являющимися специалистами в области радиационной безопасности и таможенного контроля ДРМ – так называемый, оперативный уровень реагирования.

В случае же если, реализуется любая из нижеследующих ситуаций: - уровень радиации больше, чем 1 мкЗв/ч на расстоянии 0,1 м от поверхности объекта; - подтвержденное обнаружение нейтронного излучения; - идентификация ядерного материала; - поверхностное загрязнение (в пролитом или рассыпном виде) радиоактивных

материалов, то реагирование должно быть проведено специально обученным в области радиационной безопасности и таможенного контроля ДРМ специалистами.

ПРИМЕЧАНИЕ: Уровень в < 1 мкЗв/ч на расстоянии 0,1 м выбран на основании того, что в соответствии нормативными документами, действующими в Российской Федерации, при работе с такими источниками излучения не требуется никаких разрешений контролирующих органов. В то же время в других странах могут действовать другие нормы для того или иного реагирования, поэтому необходимо руководствоваться национальными нормативными актами.

2. СТАЦИОНАРНАЯ ТАМОЖЕННАЯ СИСТЕМА ОБНАРУЖЕНИЯ (СТСО) ДРМ «ЯНТАРЬ»

В учебном пособие [4] и методических рекомендациях [6] подробно описаны характеристики, принцип работы, порядок эксплуатации, технического обслуживания. Ниже, в качестве примера, приведены в сокращенном изложении процедуры по проведению текущего контроля и проверки работоспособности СТСО ДРМ «Янтарь-1П, -1П1».

2.1. Эксплуатация, техническое обслуживание, текущий контроль и проверка

работоспособности СТСО ДРМ «Янтарь» СТСО ДРМ «Янтарь-1П» устанавливается на пешеходных пунктах пропуска и имеет в

своём составе одну стойку. Ширина пешеходного прохода не более 1 м, скорость движения пешехода не более 5 км/ч. Отличие СТСО ДРМ «Янтарь-1П» от СТСО ДРМ «Янтарь-1П1» состоит в месте прохода пешехода – справа от стойки (-1П) или слева от стойки (-1П1).

Проверка работоспособности СТСО ДРМ «Янтарь» включает в себя следующее: - общая проверка работы системы; - проверка исправности функционирования нейтронного и гамма-каналов регистрации

ионизирующего излучения (ИИ); - проверка параметров настройки; - проверка количества ложных срабатываний; - проверка состояния аккумуляторных батарей.

ПРИМЕЧАНИЕ: Порядок проверки за исключением проверки состояния аккумуляторных батарей несколько различается у разных модификаций СТСО ДРМ «Янтарь».

Общая проверка работы системы заключается в проверке сигнализации и выполнении следующих действий:

- произвести внешний осмотр стоек и пульта управления. Стойки должны быть закрыты и опломбированы;

- проверить включение питания. На лицевых панелях модулей мониторов и на пульте - управления должны светиться индикаторы СЕТЬ; - если питание системы отключено или возникли неполадки (отказы), следует

доложить об этом начальнику смены, а также лицу, ответственному за эксплуатацию системы, и сделать соответствующую запись в специальном журнале с указанием признаков неисправности (в том числе, кому и когда доложено о неисправности);

- после включения питания в течение 3 мин осуществляется самодиагностика системы и её адаптация к фону. Не рекомендуется допускать какие-либо перемещения в контролируемом пространстве стоек до окончания времени адаптации. Световая сигнализация «тревога» на стойках должна гореть ровным светом;

- если программы самодиагностики обнаружили ошибки в работе системы (счёт от детекторов не соответствует предельным значениям фона; после включения системы или смены параметров идёт подготовка к работе; сработал датчик вскрытия), то световая сигнализация мигает;

- необходимо проверить точность установки часов пульта, т.к. по ним происходит синхронизация часов мониторов и, соответственно, привязка ко времени архивной информации, хранящейся в мониторах.

В течение рабочего дня необходимо следить за показаниями средней скорости счета импульсов фона. Резкое увеличение или уменьшение показаний фона при отсутствии объектов в зоне контроля чаще всего свидетельствует о возможной неисправности системы.

Необходимо постоянно контролировать частоту ложных срабатываний: если она увеличилась сверх установленной для данной системы или, наоборот, срабатывания длительное время не наблюдаются, необходимо немедленно проверить исправность работы системы в соответствии с инструкцией по ее эксплуатации.

Проверка исправности функционирования нейтронного и гамма-каналов, в соответствие с рекомендациями МАГАТЭ, должна производиться ежедневно при помощи стандартных источников ИИ (со значениями активности, не подпадающими под действие Норм радиационной безопасности).

Проверка СТСО ДРМ «Янтарь» по гамма-каналу производится с помощью источника Сs137 активностью 8÷12 кБк из комплекта образцовых спектрометрических источников ОСГИ-3-2-1Р. Источник необходимо поместить напротив стойки в месте расположения детектора гамма-излучения (отмеченным соответствующей меткой). При этом за время не более 3 с должна включиться световая и звуковая сигнализация ТРЕВОГА.

Проверка СТСО ДРМ «Янтарь» по нейтронному каналу производится с помощью источника Сf252 c выходом нейтронов на дату проверки более 1000 н/с следующим образом:

− источник необходимо поместить напротив стойки в месте расположения детектора нейтронного излучения (отмеченном меткой); после установки за время Т0 не более 3 с должна включиться световая и звуковая сигнализации ТРЕВОГА. При этом первый пороговый коэффициент L1 должен быть равен 4;

− если из-за относительно небольшого периода полураспада изотопа Сf252 выход нейтронов на дату проверки менее 1000 нейтрон/с, необходимо выбрать параметры системы в зависимости от срока эксплуатации источника. Значения этих параметров в зависимости от времени эксплуатации источника приведены в таблице 1:

− время одного измерения Т = t⋅m (t – экспозиция, m – количество интервалов выбираются в меню ПАРАМЕТРЫ МОНИТОРА на пульте ПВС-01); первый пороговый коэффициент L1 (выбирается в меню ПАРАМЕТРЫ МОНИТОРА на пульте ПВС-01); время адаптации к фону Тф (выбирается в меню ПАРАМЕТРЫ МОНИТОРА на пульте ПВС-01); время ожидания – Т0;

− после проверки работоспособности нейтронного канала необходимо восстановить с пульта ПВС-01 значения измененных параметров t, m и L1.

Таблица 1 Данные для контроля срабатывания нейтронного канала

Срок эксплуатации источника Сf252 от даты аттестации, лет* 0–2 3–4 5–6 7-8 9–10 11–12

Т, с 1–2 3 8 22 15 30 Тф с, не менее 30 90 240 660 450 900 Т0, с 3 5 12 30 25 50 L1 4 4 4 4 2 2 *)Выход нейтронов источника Сf252 на дату аттестации должен быть не менее 1000 нейтрон/с

ПРИМЕЧАНИЕ: Практика эксплуатации СТСО ДРМ «Янтарь» показала, что при проведении контрольных проверок по нейтронному каналу в зимнее время при отрицательных температурах заметно снижается интервал срабатывания. В методических указаниях детально описан порядок проверки параметров пульта, правильности настройки дискриминаторов, частоты ложных срабатываний и состояния аккумуляторных батарей всех типов СТСО ДРМ «Янтарь».

2.2. Процедуры по применению СТСО ДРМ «Янтарь» при таможенном контроле

пассажиров, товаров и транспортных средств

Таможенный контроль ДРМ с помощью различных типов СТСО ДРМ «Янтарь» существенно различается, поэтому в методических рекомендациях [6] рассматривается порядок действий таможенного контроля ДРМ отдельно для каждой системы. В тоже время, из-за ограничения на объем, из настоящей статьи исключены методика проведения радиационного контроля на СВХ с помощью СТСО ДРМ «Янтарь-1С» и процедуры контроля пассажирского железнодорожного транспорта с помощью СТСО ДРМ «Янтарь-1Ж, -2Ж», а остальные изложены в сокращении. Полностью все процедуры приведены в [6].

2.2.1. Методика проведения контроля на пешеходных переходах с помощью систем «Янтарь-1П, -1П1», «Янтарь-2П»

2.2.1.1. Общие замечания

На пешеходном переходе могут быть установлены одиночные мониторы (СТСО ДРМ «Янтарь-1П, -1П1») или двойные мониторы (СТСО ДРМ «Янтарь-2П»). В первом случае пешеход должен проходить в пределах 1,0 м от стойки системы. Там же, где ширина прохода пешеходов больше, чем 1,5 м, должны быть установлены две стойки по обеим сторонам прохода.

Необходимо проследить, чтобы стойка была размещена как можно дальше от массивных дверей, которые могут создавать ложные тревоги в результате экранировки детектора стойки дверями и увеличения колебания фона.

Датчик присутствия должен быть размещен так, чтобы он срабатывал только от пешехода, находящегося в зоне контроля. Поэтому необходимо строго следить за тем, чтобы вблизи зоны срабатывания датчиков присутствия по обеим сторонам стоек в момент прохождения пешехода через зону контроля не находились другие пешеходы.

При проведении радиационного контроля объекта он перемещается через контролируемую зону (между или вблизи стоек системы). При этом не допускается:

− неравномерное движение объекта в контролируемой зоне или движение со скоростью более 5 км/ч;

− одновременное нахождение в контролируемой зоне двух и более объектов; − остановка объектов контроля в ожидании прохода через контролируемую зону ближе 3

м от стоек. Остановка пешехода в зоне контроля даже на очень короткое время (порядка 20 с) может

привести к понижению порога срабатывания системы, в результате чего после ухода из зоны контроля увеличивается вероятность ложного срабатывания системы.

В случае срабатывания СТСО ДРМ «Янтарь-1П, -2П» не надо позволять пешеходу находиться вблизи стойки даже на очень короткое время, так как тем самым будет искусственно повышен порог сигнализации и при повторном проходе срабатывание может не произойти даже при наличии источника излучения.

В случае, если СТСО ДРМ «Янтарь-1П» подключен к серверу, необходимо обеспечить надежную связь между оператором видеосервера и сотрудником ОТОиТК на пешеходном переходе.

Объект считается прошедшим радиационный контроль, если световая и звуковая сигнализация стоек и пульта не сработали.

2.2.1.2. Порядок проведения радиационного контроля пассажиров и их багажа

За 20 мин до начала оформления пассажиров сотрудник ОТОиТК обязан: − включить пульт ПВЦ (если он не был включен или при отсутствии видеосервера); − убедиться в постоянном свечении индикатора на стойке СТСО ДРМ «Янтарь-1П, -2П»; − при отсутствии стационарного ограждения убедиться в установке леерного ограждения

в зале досмотра для движения пассажиров через зону контроля СТСО ДРМ «Янтарь-1П, -2П»; − при обнаружении неисправности системы сообщить в подразделение ТКДРМ. В случае срабатывания сигнализации сотруднику ОТОиТК необходимо: − выявить пассажира, на которого произошло срабатывание сигнализации; − перекрыть проход к стойкам и удалить всех пассажиров из контролируемой

(обозначенной) зоны на время, пока не закончится режим тревоги; − предложить пассажиру, вызвавшему срабатывание СТСО «Янтарь», повторно пройти

через стойки для исключения ложного срабатывания; − если срабатывание при повторном проходе не произойдет, срабатывание признать

ложным, доложить об этом дежурному по отделу и продолжить досмотр в обычном порядке;

− если сигнализация сработает при повторном проходе, необходимо сопроводить пассажира вместе с ручной кладью (багажом) в комнату досмотра, внимательно следя за тем, чтобы пассажир не избавился от опасного содержимого багажа;

− в случае, если неизвестно, от кого из пассажиров или от какого багажа сработала СТСО ДРМ «Янтарь», инспектор ОТОиТК должен срочно закрыть входы и выходы той зоны, где находится эта стойка, и доложить об этом оперативному дежурному таможни;

− доложить о срабатывании сигнализации дежурному сотруднику таможенных органов, находящемуся у видеосервера или пульта ПВЦ-01 и узнать у него, по какому каналу произошло срабатывание СТСО ДРМ «Янтарь» для выбора средств РК;

− провести с помощью переносных средств РК радиационный контроль пассажира или багажа без его вскрытия: выявить место нахождения источника излучения в вещах и измерить мощность дозы излучения на поверхности багажа;

− если источник радиации определен, например, расположен в одном из видов багажа, рекомендуется просветить его на рентгеновском аппарате, чтобы определить, существуют или нет защитные экраны от гамма-излучения;

− поместить выявленный багаж в металлический сейф (рентгеновский аппарат) до прибытия сотрудников подразделения ТКДРМ, ограничить нахождение сотрудников таможни в помещение, где находится источник излучения.

Дежурный отдела при срабатывании сигнализации на пульте ПВЦ СТСО ДРМ «Янтарь-1П, -2П» обязан:

− получить информацию от сотрудников отдела, проводящих досмотр, о выявленном багаже и мерах по задержанию пассажира и сообщить им, по какому из каналов произошло срабатывание;

− немедленно доложить оперативному дежурному таможни и подразделению ТКДРМ о срабатывании сигнализации СТСО ДРМ «Янтарь-2П», о принятых мерах по выявлению и задержанию владельца багажа, содержащего источник ионизирующего излучения, о мощности дозы излучения;

− заполнить графы Журнала срабатываний сигнализации СТСО ДРМ «Янтарь-1П, -2П» как при ложном, так и при фактическом его срабатывании;

− если система подключена к видеосерверу, заполнить и сохранить электронный протокол реагирования на сигнал тревоги.

В случае, когда причиной срабатывания СТСО ДРМ «Янтарь» является пассажир, сотруднику ОТОиТК, осуществляющему контроль, необходимо:

– устно опросить пассажира о возможности прохождении им курса лечения радиофармпрепаратами или операции на сердце, связанной с установкой кардиостимулятора;

– проверить у пассажира наличие медицинских документов, подтверждающих проведение этих процедур;

– с помощью дозиметра измерить МЭД по всей поверхности тела пассажира и в области больного органа (щитовидная железа, сердце и т.п.).

По результатам инструментального контроля (независимо от наличия медицинских документов) принимается решение о пропуске пассажира через таможенную границу Российской Федерации.

ПРИМЕЧАНИЕ: Мощность дозы гамма-излучения не должна превышать при выходе из радиологического отделения 3 мкЗв/ч. на расстоянии 1 м от пациента, которому с терапевтической целью введены радиофармацевтические препараты. В случаях, когда результаты измерений не дают однозначного подтверждения

проводившегося курса лечения, с разрешения соответствующего должностного лица таможни проводится личный досмотр пассажира с участием сотрудника подразделения ТКДРМ в строгом соответствии с таможенным законодательством. В случае возникновения внештатных ситуаций (например при отключении промышленной электроэнергии или в случае неработоспособности СТСО ДРМ «Янтарь-1П, -2П»), сотрудник ОТОиТК должен сообщить об этом в подразделение ТКДРМ, а досмотр пассажиров организовать переносными приборами радиационного контроля.

В случае выхода из строя СТСО ДРМ «Янтарь-1П, -2П» все выявленные неисправности системы записываются в Журнал аварий и поломок технических средств радиационного контроля, о чем немедленно докладывается в подразделение ТКДРМ.

2.2.2. Методика проведения контроля на автомобильных пунктах пропуска

с помощью СТСО ДРМ «Янтарь-1А, -2А» 2.2.2.1. Общие положения

СТСО ДРМ «Янтарь-1А, -2А» предназначена для радиационного контроля на автомобильных таможенных переходах. На этих переходах технологическая схема таможенного контроля предусматривает, как правило, следующее разделение транспортного потока:

- автобусы; - легковые автомобили; - грузовой автотранспорт; - отдельный таможенный контроль физических лиц, следующих на пассажирском,

легковом и грузовом транспорте. Допуск автотранспорта на территорию таможенного контроля как с территории

Российской Федерации, так и с территории сопредельного государства производится пограничным нарядом. Въезжающий автотранспорт проезжает между стойками СТСО ДРМ «Янтарь-1А, -2А» и останавливается на линейке досмотра на соответствующей полосе.

Таможенный досмотр автотранспорта производится совместно пограничным нарядом и досмотровой группой таможни с привлечением при необходимости других служб, за исключением досмотра ручной клади, багажа и личного досмотра физических лиц, которые производятся только таможенной службой.

При движении автомобилей через контролируемую зону не допускается: − неравномерное движение объекта в контролируемой зоне или движение со скоростью,

отличающейся от установленной (5 км/ч); − одновременное нахождение двух и более объектов в контролируемой зоне, остановка в

ожидании проезда через контролируемую зону ближе 5 м от стоек. Проезд каждого автомобиля должен осуществляться только после выезда из

контролируемой зоны предыдущего автомобиля. Въезд автомобилей в контролируемую зону должен осуществляться в месте, контролируемом видеосистемой. 2.2.2.2. Контроль автобусов с пассажирами

Если срабатывание СТСО ДРМ «Янтарь-1А, -2А» произошло после прохождения автобуса с пассажирами, сотрудник ОТОиТК докладывает о срабатывании старшему смены, входит в автобус, проверяет при необходимости списки пассажиров (если группа следует по спискам) и направляет пассажиров с ручной кладью и багажом в досмотровой зал для прохождения таможенного и паспортного контроля.

Если СТСО ДРМ «Янтарь-1А, -2А» подключена к серверу (снабжена видеомониторами), сотрудник ОТОиТК связывается с дежурным и уточняет канал срабатывания системы.

С разрешения старшего смены автобус направляется на повторный проезд мимо стоек системы.

Если произошло повторное срабатывание, сотрудник ОТОиТК докладывает об этом старшему смены и ставит автобус на площадку задержанных машин. Старший смены вызывает сотрудника подразделения ТКДРМ, который производит радиационный контроль автобуса с помощью приборов ручного или карманного типа с целью обнаружения, локализации и измерения характеристик источника ИИ.

В случае невозможности обеспечить повторный проезд, автобус направляется на площадку задержанных машин, где сотрудник ОТОиТК производит радиационный контроль автобуса с помощью приборов ручного или карманного типа в целях подтверждения наличия источника ИИ.

Водитель и пассажиры автобуса с багажом проходят радиационный контроль в зале досмотра с помощью СТСО ДРМ «Янтарь-1П, -2П». 2.2.2.3. Контроль грузового автотранспорта

СТСО ДРМ «Янтарь-1А, -2А» обычно сопряжена с системой видеонаблюдения, пульт управления которой установлен в комнате дежурного по автомобильному переходу или в определенном помещении, где в рабочее время находится специально назначенное должностное лицо ОТОиТК. При срабатывании СТСО ДРМ «Янтарь-1А, -2А» видеоизображение объекта, вызвавшего срабатывание, фиксируется на мониторе.

Информация о срабатывании СТСО ДРМ «Янтарь-1А, -2А» с указанием канала срабатывания через дежурного или специально назначенных должностных лиц ОТОиТК доводится до соответствующей досмотровой группы.

Досмотровая группа повторно пропускает грузовой автомобиль через стойки системы. В случае невозможности обеспечить повторный проезд, что весьма неудобно с большегрузными автопоездами, грузовой автомобиль направляется на площадку задержанных машин, где сотрудник ОТОиТК производит радиационный контроль автомобиля с помощью приборов ручного или карманного типа с целью удостовериться в наличии источника ИИ.

Если произошло повторное срабатывание, или с помощью приборов ручного или карманного типа обнаружено наличие источника ИИ, сотрудник ОТОиТК докладывает старшему смены и ставит грузовой автомобиль на площадку задержанных машин. Старший смены вызывает сотрудника службы ТКДРМ, который производит радиационный контроль автомобиля с помощью приборов ручного или карманного типа с целью обнаружения, локализации и измерения характеристик источника ИИ.

Большегрузные автомобили, содержащие сельскохозяйственные удобрения, изделия из табака, некоторые руды, фарфор и древесину, вызывающие сигнал тревоги, при наличии заключения Госсанэпиднадзора и соответствия уровней мощности дозы, указанной в заключении, пропускаются через границу в установленном порядке.

2.2.3. Методика проведения контроля на железнодорожном транспорте

с помощью СТСО ДРМ «Янтарь-1Ж, -2Ж» 2.2.3.1. Общие замечания

СТСО ДРМ «Янтарь-1Ж, -2Ж» устанавливается на железнодорожных пунктах пропуска. Особенностью контроля с помощью этой системы является большая протяженность железнодорожных составов. Движение поездов происходит по графику, поэтому на работу досмотровой группы по проведению таможенного контроля часто отводится мало времени.

При проведении радиационного контроля железнодорожных пассажирских перевозок имеются определенные трудности, связанные с большим количеством людей и необходимостью 100 %-го контроля.

При проведении радиационного контроля подвижного железнодорожного состава он перемещается через контролируемую зону (между стойками системы). При этом не допускается:

− движение объекта в контролируемой зоне со скоростью более 25 км/ч; − одновременное нахождение в контролируемой зоне двух и более объектов; − остановка подвижного состава в ожидании проезда через контролируемую зону ближе

10 м до стоек системы. Остановка подвижного железнодорожного состава в зоне контроля даже на очень

короткое время (порядка 10 с) может привести к дополнительной экранировке детекторов стоек, что приведет к понижению порога срабатывания системы, в результате чего после ухода подвижного состав из зоны контроля увеличивается вероятность ложного срабатывания системы.

В случае если СТСО ДРМ «Янтарь-1Ж, -2Ж» подключен к серверу, необходимо обеспечить надежную связь между оператором видеосервера и сотрудником ОТОиТК на железнодорожном пункте пропуска.

Объект считается прошедшим радиационный контроль, если световая и звуковая сигнализация стоек СТСО ДРМ «Янтарь 1-Ж, 2Ж» не сработали. 2.2.3.2. Контроль грузовых железнодорожных составов

В случае срабатывания СТСО ДРМ «Янтарь-1Ж, -2Ж» при помещении через зону контроля товаров и транспортных средств необходимо:

− удалить все объекты из контролируемой зоны на расстояние не менее 10 м от стойки на время, пока не закончится режим тревоги, определить на пульте ПВЦ, либо с помощью видеоизображения с сервера, на какой именно колесной паре (от какого вагона) произошло срабатывание, и получить его видеоизображение;

− наружным досмотром при помощи переносного поискового измерителя типа РМ1401GN с использованием выдвижной штанги проверить срабатывание СТСО ДРМ «Янтарь-1Ж, -2Ж» на этой колесной паре и в случае если показания поискового измерителя не превышают уровень естественного фона в 1,5 раза, срабатывание можно считать ложным;

− если срабатывание системы не является ложным, то инспектор ОТОиТК должен определить зону, в которой уровень гамма-излучения превышает 1 мкЗв/ч, установить ограждение по периметру этой зоны, выставить охрану и сообщить об этом в подразделение ТКДРМ и оперативному дежурному таможни.

Железнодорожные вагоны, содержащие сельскохозяйственные удобрения, изделия табака, некоторые руды, фарфор и древесину, вызывающие сигнал тревоги, при наличии заключения Госсанэпиднадзора, соответствия уровней мощности дозы, указанной в заключении, подтверждения того, что повышенная активность товара носит распределенный характер (при наличии возможности – проведение экспресс-идентификации груза, имеющего повышенный уровень ИИ радиометром-спектрометром типа МКС-А03), пропускаются через границу в установленном порядке.

При отключении промышленной электроэнергии (или в случае неработоспособности систем «Янтарь-1Ж, 2Ж») необходимо сообщить об этом в подразделение ТКДРМ таможни, а досмотр пассажиров организовать переносными приборами радиационного контроля (типа РМ1401GN).

3. ПОИСКОВЫЕ ИЗМЕРИТЕЛИ РМ1703М, РМ1703GN, PM1401GN И

УНИВЕРСАЛЬНЫЕ РАДИОМЕТРЫ-СПЕКТРОМЕТРЫ МКС-А03 И IdentiFINDER-NGH

3.1. Общие положения

Разработка методических рекомендаций по применению поисковых измерителей

РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для обнаружения, локализации ДРМ в объектах таможенного контроля и измерения их радиационных характеристик явилась целью второго этапа работ по совершенствованию методик применения технических средств таможенного контроля за незаконным перемещением ДРМ через таможенную границу.

Для реализации этой цели требовалось решить следующие задачи: - разработать предшествующие таможенному контролю ДРМ процедуры по

подготовке к работе поисковых измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH;

- разработать методические рекомендации по использованию поисковых измерителей РМ1703М, РМ1703GN, PM1401GN для таможенного контроля пассажиров, багажа и транспортных средств;

- разработать методики применения универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для таможенного контроля пассажиров, багажа и транспортных средств.

При решении этих задач было проведено: − изучение технико-эксплуатационной документации к указанным приборам; − изучение существующих международных и национальных, прежде всего российских,

законодательных, нормативно-правовых и иных документов, в т.ч. норм и правил, регламентирующих проведение организационно-технических мероприятий, связанных с использованием технических средств радиационного контроля;

− знакомство с результатами проводившихся ранее научно-исследовательских работ по разработке методик применения приборов радиационного контроля;

− проведение тестовых лабораторных и полевых испытаний приборов; − анализ поступающей из таможенных органов Российской Федерации информации об

опыте эксплуатации приборов радиационного контроля и проблемах, возникающих в связи с этим;

− разработка методических рекомендаций и процедур по использованию поисковых измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для таможенного контроля пассажиров, багажа и транспортных средств.

Ограничения на объем статьи не позволяют полностью представить результаты работы. Ниже, в качестве примера, приведены в сокращенном изложении результаты тестовых испытаний поисковых измерителей РМ1703М, РМ1401GN и РМ1401GN, процедуры по их подготовке и применению и методические рекомендации по применению универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH.

Полностью результаты выполнения второго этапа работы по контракту №12601 приведены в учебно-методическом пособии [5] и отчете [7].

3.2. Тестовые испытания поисковых измерителей РМ1703М, РМ1401GN, РМ1401GN и

радиометра-спектрометра МКС-А03

Одним из направлений проведения работ по разработке методических рекомендаций явились тестовые испытания приборов.

Испытания проводились в лабораторных условиях, а также в условиях пассажирского и автомобильного перехода. Цель испытаний – проверка соответствия прибора требованиям проведения ТКДРМ, изложенным в руководящих документах ФТС (ГТК) России и рекомендациях МАГАТЭ к техническим характеристикам поисковых измерителей, и на основе этих испытаний – выработка методических рекомендаций по применению прибора. При этом решались следующие задачи:

– проверка частоты ложных срабатываний при различных значениях числа среднеквадратичных отклонений «n» текущего радиационного фона;

– определение порога срабатывания N поисковых измерителей при различных значениях числа среднеквадратичных отклонений «n»;

– определение зависимости расстояния, при котором приборы выдают сигнал тревоги, от значения числа среднеквадратичных отклонений «n» текущего радиационного фона для различных гамма-источников;

– определение зависимости расстояния, при котором приборы выдают сигнал тревоги, от значения числа среднеквадратичных отклонений «n» текущего радиационного фона для нейтронного источника;

– особенности проведения поиска ИИИ в условиях повышенного радиационного фона. Для радиометра-спектрометра МКС-А03 были проведены следующие тестовые

испытания:

– сравнение возможностей применения поисковых измерителей и универсальных радиометров-спектрометров для таможенного наблюдения;

– изучение влияния сопутствующего гамма-излучения на показания радиометра-спектрометра МКС-А03 при измерении плотностей потоков альфа- и бета-частиц.

Испытания проводились с использованием локальных источников гамма-излучения Co-60, Cs-137, Ba-133, Am-241 с различными значениями активности. При испытаниях локальные источники размещались в обычных пассажирских видах багажа (чемоданы, сумки и др.) без применения специальных видов защиты; все измерения проводились при температурах +20 ºС и -15 ºС, с соблюдением требований руководящих документов ФТС (ГТК) России и рекомендаций МАГАТЭ, касающихся правил применения приборов при проведении ТКДРМ. Результаты всех измерений, полученные при этих температурах, были практически одинаковы. Измерения производились при фоновом излучении Нф = 0,08÷0,09 мкЗв/ч, измеренном со статистической погрешностью 3–5 %.

Ниже приведены в изложении некоторые результаты испытаний. I. Проверка частоты ложных срабатываний при различных значениях числа

среднеквадратичных отклонений «n» текущего радиационного фона. По рекомендациям МАГАТЭ для поисковых приборов частота ложных срабатываний не

должна превышать одного в течение 12 часов (при фоновой мощности дозы 0,2 мкЗв/ч), что означает выбор числа n ≥ 6. Тестовые испытания поисковых измерителей PM1703M и PM1401GN показали (табл. 2), что при n = 4 частота ложных срабатываний (~ 1 в час) хотя и превышает рекомендуемую МАГАТЭ, тем не менее, оптимально с точки зрения обнаружения минимального количества ДРМ с высокой вероятностью.

Таблица 2

Среднее время (период) между двумя последовательными ложными срабатываниями и частота ложных срабатываний при различных значениях «n»

Значение n Период ложных срабатываний (с) Частота ложных срабатываний в минуту

1 2÷3 20÷30 2 8÷12 5÷7 3 120÷150 0,4÷0,5 4 3000÷6000 0,01÷0,02 5 более 6 ч –

II. Определение порога срабатывания (Н) поисковых измерителей при различных значениях числа среднеквадратичных отклонений «n».

Для обнаружения и локализации ДРМ существенное значение имеет выбор и установка оптимального порога срабатывания сигнализации «n» в зависимости от конкретных условий применения прибора. Рекомендации МАГАТЭ и некоторых нормативных актов ФТС (ГТК) России в этом отношении несколько различаются. В целях уменьшения частоты ложных сигналов МАГАТЭ рекомендует устанавливать порог сигнализации, в три раза превышающий уровень естественного фонового излучения. С другой стороны, согласно руководящим документам ФТС России, критерием реагирования, в случае отсутствия стационарных систем радиационного контроля или для принятия решения о проведения углубленного радиационного обследования, следует принимать превышение величины мощности дозы гамма-излучения (Низм), над значением естественного фона (Нф), измеренного в зоне таможенного контроля, на величину, равную 0,5 Нф, т.е. Низм ≥ 1,5·Нф. Из тестовых испытаний следует, что, по рекомендациям руководящих документов ФТС (ГТК) России, необходимо выбирать значение

n = 3–4 (Низм= 1,5·Нф), несмотря на то, что по рекомендациям МАГАТЭ – значение n = 6–7 (Низм = 3·Нф ).

III. Определение зависимости расстояния, при котором приборы выдают сигнал тревоги, от значения числа среднеквадратичных отклонений «n» текущего радиационного фона для различных гамма-источников.

По результатам проведенных исследований (рис.1) можно сделать следующие выводы: – с увеличение значения «n» незначительно уменьшается расстояние, на котором

обнаруживается источник гамма-излучения; – с увеличением значения «n» снижается абсолютная погрешность определения

расстояния, что может способствовать более точной локализации источника. – изменение значения «n» через 0,1, предусмотренное производителем, не имеет

практического значения ввиду слабой зависимости расстояния, на котором обнаруживается источник, от значения «n».

Рис. 1. Зависимость расстояния, при котором прибор выдает сигнал тревоги, от значения числа среднеквадратичных отклонений «n», для точечных источников гамма-излучения: 1 – Ba-133 (А=132,5 кБк); 2 – Со-60 (A=95,8 кБк) IV. Определение зависимости расстояния, при котором приборы выдают сигнал

тревоги, от значения числа среднеквадратичных отклонений «n» текущего радиационного фона для нейтронного источника. Для определения зависимости расстояния, при котором прибор (РМ1401GN) выдает сигнал тревоги, от значения числа среднеквадратичных отклонений «n» использовался точечный нейтронный источник на основе Cf-252 с выходом нейтронов 3·103 нейтр/с. В комплект поискового измерителя РМ1401GN входит камера-замедлитель нейтронов, поэтому измерения проводились, как с камерой-замедлителем, так и без нее.

Результаты показывают (рис.2), что применение камеры-замедлителя нейтронов существенно увеличивает вероятность обнаружения нейтронных излучателей, что ведет к увеличению расстояния обнаружения нейтронного источника почти на порядок. Зависимость расстояния обнаружения от значения «n» не так заметна, как при обнаружении гамма-источников, особенно для прибора без камеры-замедлителя.

2

1

Рис. 2. Зависимость расстояния, при котором прибор выдает сигнал тревоги, от значения числа среднеквадратичных отклонений «n», для нейтронного источника Сf-252 (ρ=3⋅103 нейтр/с): 1 – с камерой-замедлителем; 2 – без камеры-замедлителя.

V/. Особенности проведения поиска ИИИ в условиях повышенного радиационного фона При проведении ТКДРМ поисковыми приборами в условиях повышенного

радиационного фона рекомендуется проводить калибровку по фону непосредственно на объекте, где проводится обследование.

Выводы

Таким образом, принимая во внимание экспериментальные результаты, при проведении ТКДРМ поисковыми приборами рекомендуется установить следующие значения коэффициентов «n»:

– для гамма-канала − 4; – для нейтронного канала − 3. В этом случае достигается оптимальное соотношение между приемлемым числом ложных

срабатываний и максимально возможным расстоянием обнаружения ИИИ. Результаты тестовых испытаний радиометра-спектрометра МКС-А03 показали, что:

• при таможенном наблюдении чувствительность МКС-А03 по гамма- и нейтронному каналам превосходит поисковые приборы;

• измерение плотности потока альфа-, бета-частиц с помощью МКС-А03 позволяет корректно выявлять минимально допустимые уровни радиоактивного загрязнения товаров и транспортных средств в процессе проведения таможенного контроля;

• при измерении плотности потока альфа-, бета-частиц дополнительное сопутствующее гамма-излучение может оказывать несущественное влияние на показания радиометра-спектрометра МКС-А03.

3.3. Методические рекомендации по использованию поисковых измерителей РМ1703М,

РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH для таможенного контроля пассажиров, багажа и транспортных

средств Методические рекомендации включают в себя: – процедуры подготовки к работе и проверки работоспособности технических средств; – порядок эксплуатации технических средств; – порядок и последовательность действий при проведении радиационного контроля

пассажиров, багажа, товаров и транспортных средств на пешеходном и автомобильном пунктах пропуска через таможенную границу.

3.3.1. Методические рекомендации по подготовке и применению поисковых измерителей РМ1703М, РМ1703GN, PM1401GN

3.3.1.1. Назначение и область применения поисковых измерителей

Поисковые измерители, к которым относятся РМ1703М, РМ1703GN и РМ1401GN, предназначены для:

− обнаружения, поиска и локализации радиоактивных и ядерных материалов по скорости счета гамма- и нейтронного излучения (РМ1703М – только гамма-излучения);

− оценки МЭД фотонного излучения по линии изотопа Cs-137 в коллимированном излучении.

Они используются для проведения следующих форм таможенного контроля: − таможенное наблюдение (первичный радиационный контроль); − таможенный осмотр (дополнительный радиационный контроль); − таможенный досмотр (дополнительный радиационный контроль).

3.3.1.2. Процедуры по подготовке к работе:

• провести внешний осмотр прибора на отсутствие механических повреждений; • проверить комплектность прибора; • включить прибор – он имеет функцию самоконтроля, позволяющую проверить

правильность его функционирования (включая состояние батарей) перед каждым использованием. По завершении тестирования прибор переходит в режим калибровки по уровню фона. Процессор рассчитывает среднюю скорость счета импульсов за время калибровки (Nф) и величину порога (П). Для проверки работоспособности прибора необходимо поместить его, если возможно,

вблизи контрольного источника и наблюдать за тем, как он регистрирует излучение. 3.3.1.3. Процедуры по применению поисковых измерителей при таможенном контроле

пассажиров, товаров и транспортных средств Методические рекомендации включают в себя: – методические рекомендации по применению поисковых измерителей в процессе

таможенного контроля физических лиц и багажа; – методические рекомендации по применению поисковых измерителей в процессе

таможенного контроля автотранспорта; – особенности применения поисковых измерителей с нейтронным каналом. Основные положения, касающиеся контроля физических лиц – пассажиров,

заключаются в следующем: − пассажир в контролируемой зоне должен перемещаться со скоростью не более 5 км/ч; − в зоне контроля должен находиться только один пассажир со своим багажом; − должны соблюдаться правила подготовки и общий порядок проведения радиационного

контроля пассажиров; – рекомендуется считать уровень радиации больше, чем 1 мкЗв/ч на расстоянии 0,1 м от

поверхности объекта, уровнем реагирования.

ПРИМЕЧАНИЕ. Согласно IAEA-TECDOC-1312 (МАГАТЭ, 2003) и нормативным документам ФТС (ГТК) России рекомендуемое значение скорости сканирования – не более 20 см/с на расстоянии 5÷10 см от объекта. Такая скорость сканирования обеспечивает измерение отдельных участков объекта протяженностью не более 5 см (при времени измерения прибора 0,25 с). Рекомендуемое расстояние является оптимальным с точки зрения наименьшей близости к объекту и необходимости предохранять прибор от непосредственного контакта с объектом исследования.

Методические рекомендации по применению поисковых измерителей в процессе таможенного контроля автотранспорта приведены отдельно для легковых автомобилей и автобусов, а также для грузовых автомобилей.

Поиск и локализация ИИИ в легковом автомобиле и автобусе, где находятся пассажиры, должны производиться отдельно для пассажиров с их багажом и транспортного средства. В Методических рекомендациях подробно расписан порядок проведения радиационного контроля автобуса, легкового и грузового автомобиля, последовательность контроля отдельных мест и устройств транспортного средства, подчеркивается необходимость всестороннего обследования и повторного поиска даже после обнаружения и изъятия источника ИИИ из автотранспорта. 3.3.1.4. Особенности применения поисковых измерителей с нейтронным каналом

Методические указания и алгоритмы по обнаружению, поиску и локализации источника гамма-излучения в различных условиях аналогичны вышеприведенным методическим рекомендациям.

Главное отличие РМ1703GN от РМ1401GN – в наличии детектора для регистрации нейтронного излучения. ПРИМЕЧАНИЕ: Устойчивое срабатывание прибора по нейтронному каналу с большой степенью вероятности

говорит об обнаружении ядерного материала или техногенного нейтронного излучателя.

Для существенного повышения (более чем в 20 раз) чувствительности прибора РМ1401GN к нейтронам необходимо использовать камеру-замедлитель.

Если превышение обнаружено по нейтронному каналу, то слышна звуковая сигнализация, отличающаяся от сигналов превышения по гамма-каналу. Это позволяет различать на слух, по какому из каналов произошло превышение порогового значения.

При локализации нейтронного источника звуковая и вибросигнализация будет подавать сигналы, характерные для превышения порога скорости счета нейтронного канала без реакции на приближение или удаление источника. Поэтому локализацию рекомендуется проводить, наблюдая визуально изменение скорости счета в нижней строчке ЖКИ (нейтронный канал).

3.3.2. Методические рекомендации по подготовке и применению универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH

Многоцелевые радиометры-спектрометры МКС-А03-1Н и IdentiFINDER-NGH могут использоваться для проведения следующих форм таможенного контроля ДРМ, а также перемещающихся через таможенную границу физических лиц, товаров и транспортных средств:

– таможенное наблюдение; – таможенный осмотр; – таможенный досмотр.

3.3.2.1. Подготовка радиометров-спектрометров к работе, их функциональные

возможности Для того чтобы произвести проверку работоспособности и калибровку прибора МКС-А03,

его необходимо установить на устройство зарядки и калибровки. Прибор самостоятельно производит подстройку калибровки по энергии и контроль работоспособности основных узлов. При включении прибора IdentiFINDER-NGH он также проводит контроль работоспособности основных узлов и подстройку энергетической калибровки за счет использования встроенного радиоактивного калибровочного источника и проблескового светодиода для контроля усиления фотоэлектронного умножителя.

Отметим, что приборы существенно различаются по своим функциональным возможностям и режимам работы.

Радиометр-спектрометр МКС-А03 имеет следующие режимы работы: – режим «ПОИСК /М. ДОЗЫ»; – режим «НЕЙТРОН М.ДОЗЫ»; – режим «АЛЬФА/БЕТА»; – режим «АНАЛИЗ»; – режим «МОНИТОР».

Радиометр-спектрометр IdentiFINDER-NGH предусматривает работу в одном из двух режимов: упрощенном («EASY USER MODE») и экспертном («EXPERT USER MODES»).

Основные положения, касающиеся таможенного наблюдения с помощью многоцелевых радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH:

– использовать радиометры-спектрометры для целей таможенного наблюдения целесообразно в исключительных случаях: только в отношении малогабаритных объектов (например, международных почтовых отправлений, пассажирского багажа) и только при отсутствии или неисправности стационарных систем обнаружения ДРМ;

– для проведения таможенного наблюдения необходимо перевести прибор в режимы: − для МКС-А03 – «МОНИТОР»; − для IdentiFINDER-NGH – «FINDER MODE», если наблюдение предполагается вести

только по гамма-каналу, или в режим «DOSE RATE MODE», если наблюдение ведется одновременно по гамма- и нейтронному каналам;

• осуществить калибровку по фону; • поместить прибор как можно ближе к месту перемещения контролируемых объектов,

при этом место размещения должно позволять слышать срабатывание звуковой сигнализации и визуально фиксировать объект, по которому оно произошло;

• при срабатывании сигнализации следует определить канал, по которому произошло срабатывание, задержать объект, вызвавший срабатывание, и провести его дополнительное радиационное обследование. 3.3.2.2. Методические рекомендации по применению универсальных радиометров-

спектрометров МКС-А03 и IdentiFINDER-NGH Дополнительный радиационный контроль в рамках таможенного осмотра (досмотра)

товаров и транспортных средств включает в себя решение следующих задач: − поиск и локализация ДРМ, вызвавших срабатывание тревожной сигнализации на этапе

таможенного наблюдения (первичный радиационный контроль); − измерение радиационных характеристик и оценка степени радиационной опасности

выявленных объектов, в составе которых могут находиться ДРМ. Измерение радиационных характеристик и определение степени радиационной

опасности объекта проводится в режиме «ПОИСК/М.ДОЗЫ» – для МКС-А03 и DOSE RATE MODE – для IdentiFINDER-NGH.

При проведении дополнительного радиационного контроля с помощью спектрометра-радиометра МКС-А03 необходимо провести измерение уровня поверхностного загрязнения контролируемого объекта αβ-излучающими радионуклидами. В случае обнаружения поверхностного загрязнения, необходимо проверить его вид (снимаемое или не снимаемое).

Углубленное радиационное обследование в рамках таможенного досмотра имеет своей

задачей максимально возможную локализацию и первичную идентификацию ДРМ с отнесением их к одной из следующих групп:

− ядерные материалы или изделия на их основе; − радиоактивные вещества или изделия на их основе; − радиоактивные отходы; − иные товары и транспортные средства с повышенным содержанием радионуклидов.

Для проведения идентификации выявленных в процессе таможенного досмотра ДРМ с помощью радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH необходимо выполнить следующие основные действия:

1.) расположить прибор так, чтобы измерительная часть гамма-детектора находилась рядом с обнаруженным источником излучения;

2.) запустить программу набора спектра; 3.) в зависимости от сообщения на экране пододвинуть прибор поближе или, наоборот,

отодвинуть подальше от источника и удерживать его в таком положении в течение всего времени набора спектра;

4.) при достижении заданной экспозиции измерение будет остановлено и прибор произведет попытку идентифицировать изотопный состав источника излучения;

5.) если радионуклид не обнаружен, то это может свидетельствовать о том, что: − интенсивность источника мала, − калибровка по энергии недостаточно точна, − данный радионуклид отсутствует в библиотеке. В этом случае следует:

− увеличить экспозицию и повторить измерение, − тщательно прокалибровать прибор по энергии, − пополнить библиотеку прибора.

ЗАКЛЮЧЕНИЕ

Результатом работы по контракту МАГАТЭ № 12601 явилась разработка методических

рекомендаций по обнаружению ДРМ, представляющие собой формализованный свод правил, приемов и способов применения указанных выше приборов в процессе таможенного контроля в определенных условиях, с учетом технических характеристик и возможностей каждого из приборов.

Методические рекомендации не являются официальным нормативным документом. Приведенные алгоритмы действий являются типовыми и не отражают всех возможных особенностей конкретных пунктов пропуска и действий должностных лиц таможенных органов, поэтому требуется их адаптация к технологиям таможенного контроля конкретных таможен.

Методические рекомендации предназначены для использования сотрудниками таможенных и других правоохранительных органов, осуществляющих контроль за перемещением ДРМ на границах. Сотрудники таможенных и других правоохранительных органов впервые получают детально и последовательно изложенные правила осуществления радиационного контроля на пешеходных, автомобильных и железнодорожных переходах, а также на складах временного хранения.

Методические рекомендации по обнаружению ДРМ будут способствовать качественному улучшению как таможенного контроля ДРМ, так и подготовки сотрудников таможенных и других правоохранительных органов, осуществляющих такой контроль.

СПИСОК ЛИТЕРАТУРЫ

[1] Организация таможенного контроля за делящимися и радиоактивными материалами: Учебное пособие/И.Н Банных,. А.В. Борисенко, Н.Э. Кравченко и др. и др. – Москва.: Святигорпресс, 2003. – 411 с. – ISBN 5-93-189-004-1.

[2] Методические рекомендации по применению технических средств ТКДРМ: учебно-метод. пособие. Т. 1. СТСО ДРМ «Янтарь», поисковые дозиметры РМ1401, РМ1401 М, РМ1401К-01, индивидуальные дозиметры РМ1203, РМ1621/И.Н. Банных, А.В. Борисенко, Д.Ю. Данько и др.; под общ. ред. Н.Э. Кравченко; ВФ РТА. – М.: ФТС России; Владивосток; ВФ РТА, 2005. –140 с.

[3] Методические рекомендации по применению технических средств ТКДРМ: учебно-метод. пособие. Т. II. Радиометры-спектрометры РСУ-01 «Сигнал-М», МКС-А02, МКС-А03, МКС-РМ1401К / И.Н. Банных, А.В. Борисенко, В.А. Гайфутдинов и др.; под общ. ред. Н.Э. Кравченко; ВФ РТА. – М.: ФТС России; Владивосток; ВФ РТА, 2005. – 152 с.

[4] Стационарная таможенная система обнаружения делящихся и радиоактивных материалов «Янтарь – 1П»: учебное пособие/авт.-сост.: А.В. Борисенко, Л.Г. Елисеенко, В.В. Темченко и др.; под ред. А.В. Борисенко. – М.: РИО РТА; Владивосток: ВФ РТА, 2004. – 160 с. – ISBN –5-9590-0006-7. Серия «Технические средства таможенного контроля делящихся и радиоактивных материалов».

[5] Технические средства ТКДРМ: поисковые измерители РМ1703М, РМ1703GN, PM1401GN; радиометры-спектрометры МКС-А03, IdentiFINDER-NGH: Методические рекомендации по применению: учебно-метод. пособие/авт.-сост.: А.В. Борисенко, В.Н. Кустов, В.В. Темченко и др; под ред. А.В. Борисенко. – Владивосток: ВФ РТА, 2006. – 148 с. – Серия «Технические средства таможенного контроля делящихся и радиоактивных материалов».

[6] Совершенствование методик применения технических средств таможенного контроля за незаконным перемещением делящихся и радиоактивных материалов (ДРМ) через таможенную границу. Часть 1. Разработка методических рекомендаций по обнаружению ДРМ в объектах таможенного контроля с использованием стационарной системы таможенного контроля ДРМ «Янтарь»: отчет по контракту МАГАТЭ, № 12601\Фонд ядерной безопасности. – Владивосток, 2004. – 84 с.

[7] Совершенствование методик применения технических средств таможенного контроля за незаконным перемещением делящихся и радиоактивных материалов (ДРМ) через таможенную границу. Часть 2. Разработка методических рекомендаций по обнаружению ДРМ в объектах таможенного контроля с использованием поисковых измерителей РМ1703М, РМ1703GN, PM1401GN и универсальных радиометров-спектрометров МКС-А03 и IdentiFINDER-NGH: отчет по контракту МАГАТЭ, № 12601\Фонд ядерной безопасности. – Владивосток, 2006. – 80 с.

[8] Темченко В.В., Борисенко А.В., Елисеенко Л.Г. Возможности и ограничения современных технических средств таможенного контроля по идентификации делящихся и радиоактивных материалов // Современное состояние, проблемы и перспективы таможенного дела на Дальнем Востоке России: сб. научных трудов. – Владивосток: ВФ РТА, 2004. – С. 41-44. – ISBN 5-9590-0002-4.

[9] Борисенко А.В., Елисеенко Л.Г., Темченко В.В. Проблемы обнаружения и идентификации делящихся и радиоактивных материалов // Современное состояние, проблемы и перспективы таможенного дела на Дальнем Востоке России: сб. научных трудов. – Владивосток: ВФ РТА, 2004. – С. 56-61. – ISBN 5-9590-0002-4.

[10] Предотвращение непреднамеренного перемещения и незаконного оборота радиоактивных материалов: IAEA-TECDOC-1311.– Вена: МАГАТЭ, 2003.

[11] Обнаружение радиоактивных материалов на границе: IAEATECDOC-1312. – Вена: МАГАТЭ, 2003.

[12] Реагирования на события, связанные с непреднамеренным перемещением и незаконным оборотом радиоактивных материалов: IAEA-TECDOC-1313. Вена: – МАГАТЭ, 2003.

[13] Technical and Functional Specifications for Border Monitoring Equipment: Reference Manual. Vienna: – IAEA, 2006. – Technical guidance. Nuclear security series No.1.

Research and Development of Hand-Held Neutron Monitor, and Feasibility Studies of a 6LiI(Eu) Crystal Based RID and SPRD

M. Majorov Federal State Unitary Enterprise,

Scientific Engineering Center Nuclear Physics Research

Abstract

According to the common approach to detection of nuclear and other radioactive materials at border

crossings and checkpoints, the primary automated radiation detectors – portal radiation monitors –

and secondary means – hand held instruments (RIDs or PRDs) – are widely used at any points

where there is potential threat of illicit trafficking. In order to be an efficient mean to provide

verification of alarms, the performance of the second set instruments should fit the sensitivity of fix-

installed systems, i.e. to provide detection of minimum mass or activity detectable with fix-installed

monitors. The main goal of the project was to bridge a gap between neutron sensitivity of the

secondary set of instruments and typical fix-installed systems. In this respect our activity included:

1) research and development of new tool – a highly sensitive hand-held neutron monitor

(He3-based) and 2) improvement of neutron detection capabilities for standard hand-held monitors

which could enhance sensitivity of these instruments and made the secondary inspection of vehicles

and pedestrians more reliable and efficient. The first year of Project focused on the development of

a new category of neutron search instrument – the “Neutron Search Detector” (NSD), which would

have matching sensitivity to that of a portal monitors. Upon completion of this task (including

instrument development and tests by the IAEA), project extensions for the second and third years

focused on the development of a new approach to the design of hand-held portable instruments: i.e.

to use one radiation detector involving simultaneous and separate neutron and gamma ray detection.

This approach would allow a gain in the net size and volume of the detector element of the hand-

held instruments, and thus (hopefully) allow an improvement of overall neutron sensitivity. For this

new approach, two new instruments were developed, and and delivered to the IAEA for evaluation.

Contents

Introduction 31. Scope of the Project 42. Research and development of neutron search detector 4 2.1. Fundamentals of neutron detection 4 2.2. Basic instruments for neutron detection 5 2.2.1. Gas-filled proportional counters for thermal neutron

detection 5

2.2.2. Proportional counters for fast neutron detection 6 2.2.3. Fission Chambers 6

2.2.4. Scintillation detectors 63. Feasibility study of neutron detector for NSD 10 3.1. Results for 3He counters based neutron detector 11 3.2. Results for 6LiI(Eu) crystal based neutron detector 154. Development of Neutron Search Detector KSAR1U.06 13 4.1. Detailed description of NSD KSAR1U.06 13

4.2. Algorithm of operation 17 4.3. Results of trial operation and laboratory tests 25 4.3.1. Informal type tests, test of a new Neutron Search Device

KSAR1U.06 at Seibersdorf performed in 2003-12-18 by Dr. M. Swoboda, SGTS/TNS

25

4.3.2. Laboratory Acceptance Test of the Neutron Search Device “Hand Held Neutron Monitor KSAR1U. 06” performed by Dr. Yu. Kulikov 25.02.2004

29

4.3.3. Comparative test of the NSD (Neutron Search Device) No 1 and No 2 performed by Dr. Yu. Kulikov on 23 March 2004

34

4.3.4. The Neutron Search Device (NSD) Functional/Technical Specifications and “Hand Held Neutron Monitor” KSAR1U.06 Acceptance Criteria and Check List

35

4.3.5. Improvement of NSD according to IAEA experts recommendations and suggestions

39

5. Feasibility study of LiI(Eu) based detector RID and SPRD 41 5.1. Scientific background of the research work 41 5.2. Technical approach and methodology 426. Experimental study of LiI(Eu)detector 437. Development of LiI(Eu) detector based RID 47 7.1. Result of the acceptance tests 508. Development of LiI(Eu) detector based SPRD 53 8.1. Preliminary results of the acceptance tests 55 Sammary of the Project 56 References 57

3

Introduction

According to the common approach to detection of nuclear and other radioactive materials at border crossings and checkpoints the primary automated radiation detectors – portal radiation monitors and secondary means – hand held instruments (RIDs or PRDs) are widely used at any points where there is potential threat of illicit trafficking. The first response to the gamma ray/neutron alarm, generated by PRM, is verification of the event with the second set of radiation detection equipment. For instance, if a portal alarm is activated, the first responder should utilize radiation pagers (PRD), hand-held survey meters or some other radiation detection equipment (RID) to verify the presence of radiation. If the initial alarm cannot be verified by the second instrument within the reasonable time, it may be assumed that the first indication was a false alarm, or that the information received was false (due to malfunction of the fix-installed system). If the presence of radiation is confirmed by the verification process, further investigation is performed (localization, radiological hazard, isotope identification of radioactive source).

In order to be an efficient mean to provide verification of alarms, the performance of the second set instruments should fit the sensitivity of fix-installed systems, id est to provide detection of minimum mass or activity detectable with fix-installed monitors. Meanwhile, while gamma ray sensitivity of hand-held instruments available in the market is sufficiently enough not only for verification, but also for localization the source detected by portal radiation monitors, neutron sensitivity did not match sensitivity of portals and therefore hardly provide verification procedure not speaking about localization of the source.

The main goal of the project was to bridge a gap between neutron sensitivity of the secondary set of instruments and fix-installed systems. In this respect our activity included: 1) research and development of new tool - highly sensitive hand-held neutron monitor and 2) improvement of neutron detector for standard hand-held monitors – radioisotope identifying device (RID) and personal radiation detector (PRD), which could enhance sensitivity to mass form factor of these instruments and made the secondary inspection of vehicles and pedestrians with RIDs and PRDs more reliable and efficient.

4

1. Scope of the project

There are two major items defining the technical performance of neutron search instruments. They are algorithm of raw data analysis and performance neutron detector. While algorithm implemented for search instruments is quite developed item, there is an essential variation in type of neutron detectors which could be used for hand-held instruments.

The first year of Project focused on the development of NEW CATEGORY of neutron search instrument – Neutron Search Detector (NSD), matching sensitivity of portal monitors. The purpose was to carry out feasibility analysis of various types of neutron detectors for NSD, to work out the functional and technical specification for NSD and to develop production prototype of such an instrument.

Project extensions for the second and third years focused on the development of NEW APPROACH to design of hand-held portable instruments. That was the idea to use one radiation detector involving simultaneous and separate neutron and gamma ray detection. This approach could allowed to gain a net size and volume of the hand held instruments and thus to use this gain for improvement of overall neutron sensitivity.

2. Research and development of neutron search detector.

Today there are many types of neutron detectors, commercially available in the market. However, only a few types of neutron detectors are applicable for being used in hand-held instruments indented for detection of neutron sources, because of the following reasons: - Detector should be as much as sensitive to neutrons with energies corresponding to fission spectrum under field conditions; - It should be as much as insensitive to gamma rays with energies below 3 MeV; - Detector should be rugged enough to be used at hand-held instrument: intrinsic strength to shock and vibration should be considered important.

Below we discuss contemporary methods of neutron detection and make primary selection of the neutron detector type being applicable for NSD.

2.1. Fundamentals of neutron detection

Neutron is elementary particle interacting in all known processes [1]: Weak: n→p+e-+νe - β-decay of neutrons Nuclear:

n+6Li→4He+3H+4.78 MeV Gravitational:

mnc2=939.5 MeV mn=1.67×10-27 kg Electromagnetic:

│qn│<10-21│qe│ no charge μn=-1.91 μB, μB=eh/4πmpc magnetic momentum

5

However, in practice only nuclear interaction makes possible neutron detection due to the following reasons:

− neutron-nuclei collision is usually following by energetic particle emission (high energy access)

− there is no significant delay between the time of neutron interaction and the appearance of reaction products (charged particles or gamma ray).

The major channels of neutron-nuclei interaction providing a mechanism of neutron detection are:

Table of Neutron interaction processes [2]

2.2. Basic instruments for neutron detection.

Two major groups of neutron detectors [3] are in use today. They are gas-filled counters and scintillation ones.

2.2.1. Gas-filled proportional counters for thermal neutron detection.

Gas-filled proportional counters are the regular proportional counters filled either with boron fluoride or 3He at high pressure about 5 atm. The thermal neutrons are captured by absorber (10B or 3He) and thereby induce nuclear reactions:

n+10B→7Li*+α+2.31 MeV (93.7%) 7Li*→7Li+γ+0.48 MeV

n+10B→7Li+α+2.79 MeV (6.3%) or

n+3He→p+3H+0.765 MeV The reactions yield charged particles, detected in proportional counter. Cross section for the reactions above for thermal neutrons are 3840 barns and 5330 barns respectively and decreases essentially with increasing of neutron energy providing only 1 barn at 1 MeV.

Proportional counters are widely used in different applications. They also may be used for fast neutron detection. In that case proportional counter is surrounded with moderator (usually polyethylene).

The main advantages of the 3He and 10B proportional counters are: − commercial availability in different forms, sizes, pressure (length

from 10 cm to 1.5 m, pressure up to 10 atm, diameter 10÷50 mm)

6

− high stability − high efficiency for thermal neutrons (Ø 25mm, 4 atm. 3He provides

~70%) − weak sensitivity to gamma rays allows operation into intensive gamma

ray fields, producing dose rate up to 10 Sv/h The disadvantages of the proportional counters are:

− high pressure may involve a problem of transportation − microphonic effect (low resistance to shock and vibration)

2.2.2. Proportional counters for fast neutron detection.

In elastic scattering projectile neutron share its initial kinetic energy with the target nuclei. Being charged, the recoil nuclei lose its energy ionizing the working gas in proportional counter. This mechanism is used in hydrogen, helium and methane CH4 –filled detectors, providing the detection of neutrons with energies above hundred keV. The main disadvantage of such detectors is the low efficiency of detection.

2.2.3. Fission Chambers.

Another type of gas-filled neutron detector is an absorber-deposed chamber such as fission chambers.

The mechanism of neutron detection by such detectors is the detection of fission fragment from thermal neutron induced fission of 235U or 239Pu – isotopes having large fission cross section. Energy release of the fission is about 160 MeV. This makes possible to use gas-filled chambers in their ionization mode, without secondary gas amplification.

To avoid full stopping of the fission fragments in the layer of fissioning materials, the typical layer thickness is 1÷2 mg/cm2. Therefore the efficiency of the fission chamber is relatively low, about 0.5÷1.0%. However due to high kinetic energy of fission fragments, the fission chambers are practically insensitive to gamma ray and may operate up to the dose rate of 104 Sv/h.

2.2.4. Scintillation detectors

Today there are many different types of scintillators, suitable for both slow and fast neutron detection. There are three main types of scintillators: inorganic, organic and mixed type.

Inorganic materials: 6LiI(Eu) crystals 6Li-glass Ce3+ doped solid state fibers 6Li+ZnS(Ag) detectors Organic materials:

Stilbene crystals Plastic scintillators – NE102, NE104, BC454 (5% boron loaded) Liquid – BC501, NE311(5% boron loaded), NE323(1% gadolinium loaded) Mixed type: Hornjak detector – composition of organic glass with ZnS(Ag)

7

Table 1. Neutron detectors: efficiency, sensitivity to gamma ray and applicability for hand-held instrument

Detector type

Size

Absorber/

Converter

Energy of

incident

neutron

Efficiency,

%

Sensitivity

to gamma

ray• , Sv/h

Applicability within NSD

Plastic

scintillator

NE102

5 cm

thick

1 Н

1 MeV

78

0,0001

Low (no pulse shape discrimination feature

– gamma ray are indistinguishable from

neutrons)

Liquid

scintillator

BC501

5 cm

thick

1 Н

1 MeV

78

0,001

Complicated

(require application of neutron/gamma ray

discrimination technique, relatively high

detection threshold

∼0.5 MeV)

CH4

proportional

counter at

7 atm

∅ 5 cm

1 Н

1 MeV

1 0.01

Low (efficiency)

4 Не at 18 atm

∅ 5 cm

4 Не

1 MeV

1 0.01

Low (efficiency)

Thermal neutron detectors – moderator is required

6LiI(Eu)

1 mm

thick

6 Li

0.025 eV

75

0.05

High (simultaneous and separate detection

of gamma ray and neutrons)

6Li-glass

1mm

thick

6 Li

0.025 eV

50

0.01

Medium (relatively high sensitivity to

gamma rays, possible to use in form of

solid state fibers)

3 Не, 4 atm

and 5% СО2

∅ 2.5 cm

3 Не

0.025 eV

77

10

High (the only way to built high sensitive

neutron detectors)

• T

he u

pper

lim

it of

gam

ma

dose

rate

whe

n de

tect

or is

still

abl

e to

ope

rate

8

BF3 at 0.66

аtm

∅ 5 cm

10B

0.025 eV

29

10

Low (efficiency)

BF3 at1.18

аtm

∅ 5 см

10B

0.025 eV

46

10

Low (efficiency)

Fission

chamber

2.0 mg/cm2

235 U

0.025 eV

0,5

106 –10

7 Inapplicable

9

Recent improvement of crystal growing technology has resulted in development of LiI(Eu) crystal with excellent properties for simultaneous detection both gamma ray and neutron.

The absorption of a neutron in the nuclear reaction n + 6Li 4He + 3H + 4.78 MeV

is accompanied by the kinetic energy release of reaction products. Being charged, they generates into the crystal 6LiI(Eu) the same quantity of light as that accompanying the complete absorption of electron with energy 3.2 MeV. This makes it possible to determine directly the number of measured gamma quanta with energies less than 3.2 MeV and the number of neutrons as the number of events in different ranges of one energy scale, i.e., to perform the n/γ separation using the amplitude selection method [4].

0 500 1000 1500 2000 2500 3000 35001

10

100

1000

10000

1460

keV

Neut

rons

2614

keV

Coun

ts

Energy, keV

Fig. 1. Amplitude distribution of background gamma ray and thermal neutron peak,

corresponding to complete absorption of the neutrons detected.

Owing to the fact that the energy distribution of background gamma quanta is confined mainly at energies lower than 2.6 MeV (208Tl), the contribution of background gamma-ray into the energy “window” of neutrons is negligibly small. The use of such gamma ray and neutron detector is mainly important for safeguards purposes, because its low weight and high efficiency for gamma ray and neutrons. Energy resolution of the crystal is also suitable for primary selection of the type of radioactive source found. Typically at 0.662 MeV the energy resolution is 7.5÷8.0% [5].

Solid-state 6Li-glass Ce 3+ doped fibers is a variation of regular scintillation detector. Thermalized neutrons are captured by 6Li and thereby induce a nuclear reaction, yielding an alpha particle and a triton. The charged

10

particles excite Ce3+, which is contained as a dopant in the glass fibers. Each fiber is coated with a low refractive index silicon polymer, which maximize the amount of light captured in the glass fiber. Generated light is guided to photomultiplier tube at the end of the fibers, thus providing the neutron detection.

Gamma ray induced events in the optical fibers can be distinguished from neutron signals by adjusting an appropriate threshold level for the pulse height.

Compared with other conventional neutron detectors, the main advantage of such a detector is very high neutron detection efficiency resulting from the large effective area up to 500 cm2. In addition the solid-state fibers are flexible and more robust than gas-filled proportional counters.

One of the interesting applications of recently developed semiconductive polymers is the use in thermal neutron detectors. A solid state 235U, 239Pu or 6Li thin converter are interposed between two layers of semiconductive polymer. Charged particles or fission fragment resulted from neutron capture in the converter are detected by polymers similar to conventional diode detector. The thickness of the individual detector is about 0.1÷0.3 mm, but the assembly of 3 detectors absorbs 85% of incident thermal neutrons.

3. Selection of neutron detector for NSD

Analysis of the detector characteristics listed in Table 1 displays only three possible candidates for being used in hand-held neutron instrument: stilbene (BC-501A or equivalent) based detectors, 6LiI(Eu) crystals or 3He proportional counters both in moderator.

Development of pulse shape discrimination technique resulted in number of organic scintillators, such as stilbene or liquid scintillators – NE 213 or BC501A, having become one of the most powerful instrument for studying fast neutrons. Many of high precision neutron spectrometers [7, 8] utilize the organic scintillator as an excellent detector for fast neutron registration even in the background of the accompanying gamma rays.

It’s well known [7], [8] and [9], that there is possibility of separate neutron detection in field of gamma ray exceeding neutron flux over 103. However, the pulse shape discrimination technique does not provide the reliable separation in low amplitude region (below 1 MeV of neutron energy) and relatively high gamma ray flux (<103 1/s⋅cm2)). The energy distribution of fission neutrons, emitted by nuclear materials may be well (accuracy about 5÷10%) represented by Watt distribution or by the more simple Maxwell shape with corresponding parameters of nuclear temperature T (Table 2).

TE

eET

En−

⋅⋅⋅

=3

2)(π

[6]

11

Table 2. Parameter of nuclear temperature [11]

Nuclei Fission process T, MeV

233U Thermal neutron induced fission 1.35

235U Thermal neutron induced fission 1.35

239Pu Thermal neutron induced fission 1.40

240Pu Spontaneous fission 1.24

241Pu Spontaneous fission 1.39

252Cf Spontaneous fission 1.42

Simple estimation shows, however, that contribution of neutrons below the detection threshold 1 MeV (where gamma/n separation becomes reliable) exceeds 34% for 240Pu and 30% for 252Cf. In field, fast neutrons emitted by bare source are moderated by source vicinity, construction materials, ground etc, and thus the contribution of neutrons below the threshold becomes more, dramatically degrading efficiency of detection.

3.1. Results for 3He counters and LiI(Eu) based neutron detector

As 3He and LiI detectors are low sensitive to fast neutrons, they should be surrounded by appropriate amount of moderator. The shape of the moderator involving the highest detection efficiency was always the subject of intensive study [10]. In the present report the criteria for optimization was the minimum mass of the detection unit at assigned sensitivity. In first phase, numerical modeling of the moderator shape was done with Monte-Carlo MCNP code.

Initial data for calculation was a rough estimation of the overall mass of the assembly, number and type of neutron counters, diameter and thickness of LiI(Eu) crystal.

Optimum shape of the moderator and 3He counters is shown in Fig. 2 with geometrical characteristics given in table 3.

Fig. 2. 3He neutron detection unit.

12

Table 3. Optimum geometry of 3He-based NSD Detector Three 3He proportional counters @3atm.

pressure, ∅32×180 mm Length, mm 200 Width, mm 150

Thickness, mm 90 Forward layer, mm 14 Backward layer, mm 4

Gap between counters, mm 36

Experimental study of the detector built according to the calculated data gave the following results: it was found that background count rate of three 3He tubes was 0.15÷0.20 CPS, sensitivity to fast neutrons – 21.4 cm2, increasing its value with moderating of neutrons in vicinity of the radioactive sample and intrinsic efficiency to neutrons – 7.1 %. The net weight of the detector was found to de 2.24 kg.

LiI(Eu) based detector was simulated as a polyethylene thick-wall cylinder Fig. 3. LiI(Eu) crystal with 6Li enriched up to 90% was mounted inside the cylinder.

Fig. 3. 6LiI(Eu) scintillation crystal Ø70×20 mm in polyethylene thick-walled

cylinder. Optimum shape is presented in Table 4. Table 4.

Detector 6LiI(Eu) 90% enrichment on 6-Li isotopeCylinder length, mm 130 Outer diameter, mm 220 Inner diameter, mm 80 Forward layer, mm 30

Intrinsic background was found to be 0.34 cps, gross weight of the LiI(Eu) + moderator detector was found 6.4 kg, sensitivity – 20.1 cm2.

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4. Development of Neutron Search Detector KSAR1U.06

Results of calculation presented above have shown the pronounced advantage of using 3He-counter based neutron search detector because the requested sensitivity of the detector may be obtained with the less mass of the detection unit. However, the use of LiI(Eu) based detector is defensible in the systems, if 1) high neutron sensitivity is desirable, but not a critical parameter 2) gamma ray spectrometry is considered important. For the development of Neutron search detector (NSD) 3He-counter based neutron detector was selected.

Following to IAEA recommendations the following functional and technical recommendations to NSD were considered important:

- High efficient moderated neutron detector with reasonable insensitivity to gamma radiation and resistively to vibration and shock

− Visual and acoustic (tone height being proportional to alarm threshold exceeding value) indicators of alarm

− Enhanced statistical algorithm moving-average criteria − Friendly interface with both digital and graphical indication

simplifying the neutron source localization process − Single-handed operation − Selectable radiological safety alarm threshold shown on the display − Rugged design for outdoor use in a wide range of temperature and

humidity − Illuminated large area display − Rechargeable batteries providing at least 8 hours of continues

operation − 2 operational buttons and 1 “power/ illumination on” button − 4096 alarm records available in power-independent memory − USB interface

4.1. Detailed description of NSD KSAR1U.06

Detailed description of NSD hardware is given in this chapter. Fig. 4 shows the top view of NSD. Both neutron detector and electronics are

integrated in one instrumental case. Rechargeable batteries and analog electronic socket are mounted on the either ends of the moderator block. LCD, LED "Alarm", operating buttons F1 and F2, connector for connection of the charger and PC via USB port (charger/USB), piezoelectric siren, and protected button "Switch On" with the built-in LED are mounted on the top cover of NSD case. Vibromotor generating an additional alarm signal is mounted inside the case. NSD has two lightweight carrying handles. Three 3He proportional counters SNM-88, diameter 32 mm, 200 mm long, are assembled to form the neutron detector. 3He pressure in the counters is 3 atm. The counters are placed into polyethylene moderator to improve

14

sensitivity to fission neutrons. Fission spectrum neutrons ambient dose rate is evaluated from the detected neutrons flux value. GM counter SBМ-20 is intended for detecting and evaluating of the gamma ray ambient dose rate over the energy range 0,06-3.0 MeV.

Fig. 4. General view onto neutron search detector. Upper panel – top view, bottom panel – detector with cover open.

Fig. 5 presents NSD functional diagram . Ni-MH rechargeable battery (RB)

3,8 A·h/7,2 V provides power for NSD. RB serviceability is analyzed by power-on control circuit which prevents turning NSD on at low voltage. Power-on control circuit applies supply voltage URB to HV power supply, electronic keys, LED "Switch On", divider, LF amplifier (LFA), and DC/DC low voltage former (DC/DC LVS), after the signal from the button "Switch On" is received in case the output voltage on RB is satisfactory. Supply voltage URB is converted by DC/DC LVS to voltage +5V and -5V for CSA, discriminator, pulse former, and controller supply so that NSD is in "On" mode. Voltage URB is applied to analog-to-digital converter (A/DC) input via divider. A/DC is built in controller. The divider transforms URB so as to get in the measurable by A/DC voltage range. In case URB steps down below the specified level controller produces a signal to

15

turn NSD off. Voltage +400 V is generated by HV power supply for GM counter that measures gamma ray ambient dose rate. Voltage +1600 V is generated for neutron detector (ND) detecting neutron radiation. LED indicates the NSD "On" mode.

Signals generated by ND follow to the discriminator via CSA. The discriminator produces logical signals, duration 40 μs, if ND signal amplitude exceeds ĀT/4. ĀT is an average pulse amplitude distribution value corresponding to the peak of thermal neutron full absorption in 3He proportional counters. Logical signals come to the controller’s counter where their number per unity of time is digitized. GM counter signals also come to the controller that conducts statistically data processing, via pulse former. The pulse shaper forms logical signals 40 μs length. NSD operation modes are changed with buttons F1, F2, and "Switch On". The "Switch On" button becomes functional when NSD is turned on.

Information from the controller comes to LCD so that NSD modes of operation and the results of data processing are indicated on the screen. Controller-PC data exchange interface is provided by USB-controller. LED "Alarm", vibrator, and piezodynamic are triggered by controller through electronic keys K2 and K3, and through LF amplifier. LED illumination is controlled via electronic key K1.

In order to save in RAM circular buffer data obtained as the result of measurements and to supply the internal timer NSD is provided with the subsidiary Ni-MH battery with capacity 0,07 A·h and voltage 3,6V (SB RAM).

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Fig. 5. Functional diagram of NSD KSAR1U.06.

4.2 Algorithm of operation

NSD modes of operation are TEST, WARM UP, BACKGROUND, SEARCH, INTEGRAL, and SETUP.

When it turns on NSD goes on to the TEST mode and checks the proper functioning of intrinsic systems. Acoustic indicator, display illumination, LED and vibrosignal are all switched on for approx. 1 second. Software version, time, and date are displayed on the screen (Fig. 6). Duration of TEST mode is 5 s.

Fig. 6. TEST mode screen view. WARM UP mode is active on completing the TEST mode. Fig. 7 shows the WARM UP screen view. Duration of this process is 60 s.

Fig. 7. WARM UP mode screen view.

Indication of current operational status and elapsed time in WARM UP mode is displayed in the center of the screen.

The current time indication from the internal RTC microcontroller timer is displayed in the right lower corner. Since the timer is operated from the subsidiary Ni-MH battery, the current time is not lost when NSD is turned off. Rectangle in the lower right corner of the screen indicates the RB status. Filled rectangle indicates the maximum RB charge, blank frame warns of forthcoming discharge of the NSD and that RB is to be charged. If the latter is the case acoustic signal (0,5 s, 0,5 kHz) is generated at 10 s intervals. If it is unnecessary to warm NSD up (e.g., a short-term break), WARM UP mode can be canceled by pressing the button F1. On completion up warming, NSD turns into BACKGROUND mode. Fig. 8 shows BACKGROUND mode screen view.

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Fig. 8. BACKGROUND mode screen view. Neutron background flux value is needed for searching neutron source measures while operating in SEARCH/INTEGRAL mode. BACKGROUND mode cannot be cancelled in contrast to WARM UP mode. When pressing the button F1, NSD turns into SETUP mode. If during the background measurement no neutron counts were detected, the system warns about fatal malfunction (Fig. 9). Attempt to press F1 button leads to repeated Background measurement.

Fig. 9. No neutron counts screen view. BACKGROUND mode also analyses count rate in gamma-ray detection channel (GM counter). Contrary to above mentioned neutron malfunction, gamma-ray malfunctioning is not a fatal error and after displaying the corresponding message (Fig. 10) the NSD allows operator to continue operation by pressing a button (F1 or F2).

Fig. 10. No gamma counts screen view. On completion operation in BACKGROUND mode, NSD turns into the SEARCH mode that is used for searching and localization neutron sources as well as verification of alarms. Fig. 11 shows the SEARCH mode screen view.

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Fig. 11. SEARCH mode screen view The right upper corner of the screen indicates specified sampling time (dT), prompt count rate (N), average background count rate (B) and alarm threshold in standard statistical deviation units (Th). Neutron (Dn) and gamma ray (Dg) ambient dose rate values and specified radiological safety alarm threshold (SA) in μSv·h-1 are indicated lower. Dose rate updating interval is 10 s and is not user-selectable in contrast to the sampling time and detection threshold.

To estimate a tendency of their change, obtained count rate values over the last 70 s are shown by diagram. Scaling is controlled automatically. For quick Y-axes rescaling, shift NSD into the SETUP or INTEGRAL mode followed by going to the SEARCH mode. Active alarm indicators (acoustic, LED and vibro) are displayed under the diagram with corresponding symbol. All three indicators are switchable and can be activated / deactivated independently in the SETUP mode. Beneath there is a formalism of generating alarm signal. The system triggers an alarm if

2+⋅+≥ BThBN , where N - is number of neutron counts within sampling time τ; B - is background count rate within sampling time; Th - sigma multiplier. On exceeding of the prompt count rate value over specified alarm threshold, the corresponding indication is displayed (Fig. 12), LED flashes with 0,5 Hz frequency, acoustic alarm turns on, tone height being proportional to alarm threshold exceeding value.

Fig. 12. SEARCH mode screen view on exceeding alarm threshold.

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A circular buffer for alarm related information recording is provided for the device. Alarm time, date, sampling time, N , sigma multiplier and B are recorded in one event. The buffer capacity is 4096 events. Data from memory can be read out by the PC. NSD can be shifted in SETUP or INTEGRAL mode by pressing F1 or F2-button correspondingly. INTEGRAL mode is intended to detect and verify weak neutron sources that cannot be detected in the SEARCH mode because of the insufficient measurement time. Fig. 13 shows the INTEGRAL mode screen view.

Fig. 13. INTEGRAL mode screen view. NSD displays the following items in this mode: TIME – Elapsed time (since START button (F1) was pressed) N – Number of neutron counts accumulated since START button (F1) was pressed B – Number of background neutrons simply estimated as product of time left and average background count rate earlier in BACKGROUND mode

CPS – Average count rate over the elapsed time, TIMENCPS =

ERR – statistical uncertainty of average count rate with confidence level 0.9 The data are updated on the screen once a second. On pressing STOP (F1) button NSD compares N and B and generates the message

(Fig. 14) BN > if the expression BBN ⋅+≥ 3 is correct. Otherwise the “N-B” underline is not displayed.

Fig. 14. INTEGRAL mode screen view on exceeding background. Measurement cycle repeats on pressing F1-button. NSD returns to the SEARCH mode if F2-button is pressed.

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SETUP mode is intended to optimize NSD performance to the user requirements and operation conditions. To shift in the SETUP mode, the password is required (Fig. 15).

Fig. 15. Entering password. A digit above the marker shifts from 0 to 9 on pressing F1. To shift to the next option press the F2-button. When the last digit of the password is entered, in case the password is legal, the main menu of the SETUP mode is displayed on the screen, otherwise message “ILLEGAL PASSWORD” displays on the screen. (Fig. 16). Pressing any button returns the device to the previous mode before shifting in the SETUP mode. The password is given in the NSD certificate.

Fig. 16. Illegal password entered. Main menu of the SETUP mode is shown in Fig. 17.

Fig. 17. Main menu of the SETUP mode. Menu option is realized by pressing the F1-button, and by F2-button to confirm the choice. When the F2-buttton is pressed in “Return” option, the device returns to the mode in which it operated before shifting in the SETUP

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mode. The device shifts in BACKGROUND mode in case “Review BACKGROUND” option is chosen. All NSD options are set up in the manufacturer specified condition when “All Default” option is chosen (Table 5). In this case, the confirmation menu is displayed on the screen (Fig. 18).

Fig. 18. Confirmation menu of the “All Default” option.

Any other option means turns into the corresponding submenu of the SETUP mode. Setting options in the BACKGROUND and SETUP mode is realized in the submenu “MEASURE OPTIONS” (Fig. 19).

Fig. 19. Submenu “MEASURE OPTIONS”.

To choose an option, press the F1-button, and to confirm the choice press the F2-button. Sampling time in SEARCH mode is specified by the option “DWELL TIME” and varies from 1 to 10 s (step 1 s).

Duration of the background count rate measurement in the BACKGROUND mode is specified by the option “BKGD TIME” and takes on values 100, 200, 300,

500, and 1000 s. Note, that at normal background conditions ( 2015.0 cmsn

⋅)

the background count rate is to be 0.1÷0.2 CPS. Therefore, to get a sufficient statistical accuracy of the data the minimum duration 300 sec should be applied.

“BKGD CORRECT” option is YES/NO. It means whether correction of average background value in the BACKGROUND mode takes place in the SEARCH mode depending on the current count rate fixed. In order to avoid variation of average background count rate on approaching the neutron source, this option is not corrected in case the current count rate exceeds the alarm threshold.

“ALARM TRSH” (sigma multiplier) parameter directly reflects to probability of detection and false alarm rate, because it sets alarm threshold. This

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value has the dimensions of the standard deviation unit and varies from 1 to 6. “ALARM LED”, “ALARM BEEP”, and “ALARM VIBR.” options activate/deactivate the corresponding alarm indicators. The options are YES/NO.

Radiological safety alarm threshold of total neutron and gamma ray dose rate value is set in “SAFETY ALARM” submenu.

Fig. 20. Alarm threshold total dose rate specifying menu.

Radiation dose rate are analyzed once in 10 s. Alarm indication is turned on in case alarm threshold is exceeded: LED flash frequency is 0,25 Hz, permanent acoustic alarm appears and corresponding message is shown on the screen. Dose rate is monitored in all operational modes, and alarm threshold indication is of priority. For example, in case the search threshold and the radiological safety dose rate threshold are exceeded simultaneously, indication corresponds to the dose rate threshold.

NSD operates in two languages - Russian or English. The language is selected by means of submenu “LANGUAGES”.

Fig. 21. Submenu “LANGUAGES”.

Switch on/off illumination of the screen is controlled by “LIGHTING” submenu.

Fig. 22. Illumination control panel.

Display illumination remains switched off in case “ALWAYS OFF” option is chosen except only illumination switch on for 40 s at every press on the “Switch On” button provided for changing illumination conditions in the dark. Display is permanently illuminated in “ALWAYS ON” mode.

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“40 SEC” mode illuminates the display for 40 s at every press on any button, at changing operation modes (including automatic modes), at exceeding of the detection threshold in “SEARCH” mode, and at exceeding the radiological safety alarm threshold. Note, that display illumination is one of the most power consuming NSD part. Avoiding frequently use of “ALWAYS ON” option provides longer operational life time of NSD without recharging batteries.

Use “TIME/DATE” menu to change current time and date. Time is indicated on the bottom right corner of the screen, and date indication is in square brackets. Press F1 button to select the option and F2 button to vary it.

Fig. 23. “TIME/DATE” mode screen view.

To turn NSD off, press “Switch on” button and hold it for 5 s. Corresponding screen message is shown in Fig. 24.

Fig. 24. Shutdown menu.

Shutdown is possible in any operation mode. “YES” option turns NSD off, “NO” option returns NSD into the current operation mode all data being retained. Use F1 and/or F2 buttons to select and confirm your choice.

NSD is linked with PC via USB-port, software-assisted being delivered. Executive instructions are provided to inquire about the current status of NSD and current neutron and gamma ray dose characteristics, to read the data on specified detection alarm threshold exceeding events from the buffer, to clear the event buffer, to inquire about the current measurement date in SEARCH or INTEGRAL mode, and to control NSD with the instructions that emulate pressing F1 and F2 buttons. To link the device with PC:

- connect NSD to output connector of PC USB-port with the cable form the delivery set;

- run setup.exe from CD disk supplied; - follow the instructions from the “Inquiry” menu of the NSD interface

program.

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Table 5. Specified options № Option Purpose Value range Default 1 DWELL TIME Sampling time in SEARCH

mode, s 1…10, step 1 4

2 BACKGROUND Duration of background count rate measurement, s

100, 200, 300, 500, 1000

100

3 BKGD CORRECT Automatic adjustment of the average background value depending on the current count rate

YES/NO YES

4 ALARM TRSH Sigma multiplier, σ 1, 2, 3, 4, 5, 6

2

5 ALARM LED Activation / Deactivation of LED indicator under alarm condition

YES/NO YES

6 ALARM BEEP Activation / Deactivation of acoustic signal indicator under alarm condition

YES/NO YES

7 ALARM VIBR. Activation / Deactivation of vibro-indicator under alarm condition

YES/NO YES

8 SAFETY ALARM Radiological safety alarm threshold dose rate, μSv·h-1

20, 50, 100, 200, 500

200

9 LANGUAGE Working language Russian, English

English

10 LIGHTING Display illumination YES, NO, 40 s 40 s

4.3. Results of trial operation and laboratory tests.

The acceptance test and trial operation were performed at Federal state Unitary Enterprises Scientific Engineering Center Nuclear Physics Research, Saint Petersburg, Russia, at Headquarters IAEA, Vienna, Austria and at Austrian Research Centre, Seibersdorf.

After the first test phase the number of recommendations and suggestions were made to improve functional and technical performance of NSD. Then the necessary changes were applied to its design. The final, production prototype was tested at Scientific Engineering Center Nuclear Physics Research.

Beneath, there are reports on the results of NSD test, performed by group of experts from IAEA – Dr. Rolf Arlt, Dr. Yury Kulikov and Dr. Martha Svoboda.

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4.3.1. Informal type tests, test of a new Neutron Search Device KSAR1U.06

at Seibersdorf performed in 2003-12-18 by Dr. M. Swoboda, SGTS/TNS

The neutron sensitivity of RID’s is sufficient to trigger an alarm caused by a neutron source corresponding to the ITRAP specifications for fixed installed systems (20000n/s Pu) But the neutron sensitivity now is much higher than requested in the ITRAP specifications. A neutron source hidden in a car detected by the fixed installed system cannot be detected by RID’s. We already tested the MKC-A02 in summer 2003 (Report: Testing the Hand-held MKC-A02 in Comparison with the Exploranium Border Monitor). The results in summer 2003 have shown that a new hand-held neutron detector is needed to match the sensitivity of RID’s. For the tests now we used a new neutron search device (KSAR1U.06, Prototype developed by the “Scientific Engineering Centre, Nuclear Physics Research”), one MKC-A02 and one NRM-477. Different neutron sources were hidden into a car and a search process was performed. The new neutron search device has a sufficient sensitivity for searching and locating a neutron source detected with a border monitor system having a sensitivity of 9350n/s for 8km/h.

a. Car detection test

b. Search test

Used instruments

Equipment: Exploranium border monitoring system (AT-920) MKC-A02 NRM-477 NSD (New Search Device) NSD Setup: Dwell time 2s BG measurement time 100s BG correction Yes Alarm Threshold 2σ Safety Alarm 1500cps

Fig. 25. NSD, MKC-A02, NRM-477.

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A. Car test

The first used source corresponds to the Technical/Functional Specifications for Border Radiation Monitoring Equipment Draft Rev 18.5, 19 December 2003, especially to the dynamic test with moving sources. All four used Pu sources together correspond to 150g WGPu or a neutron rate of 9350n/s. A car containing four Pu sources in a lead container passed the detection pillars of the fixed installed system AT-920 (Exploranium). In the first run the car speed was too high therefore only a g-alarm was triggered. The car speed was reduced for the second run and the result caused by the same sources on the same place was a neutron alarm Source: All four PuO2 CBNM Standards in the transport box (10mm lead)

B. Neutron vehicle search

Different selections of a set of Pu sources were hidden (each after another) on different places into the car and a search process was performed. The first step was always to locate the source from outside of the car. For this purpose all three instruments were used. General procedure: Step1: Using the instrument with the most sensitive side towards the car (Fig. 28) go once around the car with a search speed of <50cm/s at a height of about 60cm about ground The test was counted successful if the user was able to get a neutron signal determining from outside the quadrant of the car where the source was located. Step2: Access the quadrant (open trunk, door, motor cover, etc.) and find the source. If there was no signal during searching in step 1, the source was searched using the step 2 procedures, what is much more time consuming. If the source was found within 10 minutes, the test was still counted successful.

Fig. 26. Search process with the most sensitive side towards the car.

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Table 6. Plutonium sources used for evaluation Source Equivalent mass of weapon grade Pu [g] Neutrons/second Pu93 O2 10.25 680 Pu84 O2 21 1390 Pu70 O2 46 3060 Pu61 O2 64 4220

The total neutron production rate is therefore at least 9350n/s. The total mass is approximately equivalent to about 150g of weapon grade metal plutonium. B.1 Used sources Pu93 O2, Pu84 O2, Pu70 O2 and Pu61 O2 in the lead transport container Searching Results MKC-A02 was detected, but very slow, scanning movement was required NRM-477 was not able to detect the source from outside of the car NSD has clearly indicated the location of the source from outside of the car B.2 Used sources Pu93 O2, Pu84 O2, Pu70 O2 and Pu61 O2 in the lead transport container on a different location as in B.1.

Fig. 27 Lead transport container in the trunk of the car.

Searching Results MKC-A02 the detection process was marginal NRM-477 it was not possible to detect the sources from outside of the car NSD The location of the sources was clearly indicated from outside of the car. B.3 Used sources Pu 61 O2 source (reactor grade Plutonium) in the lead transport container. Before the car test was performed a simple detection check was done. Result: MKC-A02 Neutron alarm triggered in a distance of 10 cm from the surface of the transport box

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NRM-477 There was no possibility to detect the source with this instrument The Pu 61 O2 source in the lead container was hidden in the trunk of the car Searching Result MKC-A02 There was no possibility to find the source from outside of the car NRM-477 --- not used NSD It was possible to locate at first from outside the part of the car, which was containing the source. A search process inside the car also exactly located the source. B.4 Used sources Pu 84 O2 source (nearly weapon grade Plutonium) in the lead transport container. The Pu 84 O2 source in the lead container was hidden in the cabin of the car MKC-A02 There was no possibility to locate the source from outside. NRM-477 There was no possibility to detect the source at all (even if not hidden) NSD It was only with a slow speed scanning method possible to find the source from outside. It was marginal and time consuming. Summary on the results of the test

The NRM-477 does not fulfill the requirements of neutron search device at all. It was also the only of those three instruments, which was not able to directly detect the Pu 61 O2 source (reactor grade plutonium). The NSD is at least 5.5 times more sensitive than the NRM 477. All three tests were also performed with the MKC-A02 (old model), which is about two times more sensitive than the NRM-477. This MKC is quite on the detection limit and the search process was slow. The new neutron search device, NSD passed all searching tests. This device has the highest neutron sensitivity of all tested devices. It has also a bar graph indication for the search process as usually used in the gamma search process simultaneously with an audio signal. This instrument is still a prototype and some usability features need improvement.

4.3.2. Laboratory Acceptance Test of the Neutron Search Device “Hand Held

Neutron Monitor KSAR1U. 06” performed by Dr. Yu. Kulikov 25.02.2004

Technical Performance Specifications

A. The Acceptance Test has been performed according to the Chapter 6.4 of the “Technical/Functional Specifications for Border Radiation Monitoring Equipment” Draft 18.5 (December 2004), which contains the following items:

A.1 Absolute detection efficiency: shall be not less than 20 cps per n/s cm² for fission spectrum neutrons;

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A.2 SNM detection limit: shall be not more than 100 g of WGPu at 1 m distance and 10 s measurement time (80% detection probability at 95% confidence level) in standard neutron background;

A.3 False Alarm Rate (FAR): shall be less than 1/10 min at a 2 s dwell time at 95% confidence level in standard neutron background;

A.4 Safety Alarm: shall be triggered, when the dose rate on the detector surface exceeds 100 μSv/h. The safety alarm level shall be user settable.

A.5 Gamma insensitivity: shall be no neutron alarm at a gamma dose rate (Co-60) not less than 100 μSv/h at the detector surface;

A.6 Search mode: acoustic and visual indication of the neutron source localization process with changing audio signal repetition rate and (optional) MCS graph on the display depending on the source – detector distance. The alarm algorithm should have the option to adapt to a continuously or slowly changing background;

A.7 Environmental: Temperature range – 20 + 50 degrees Centigrade; non-condensing humidity 90% at 35 degrees Centigrade;

A.8 Electromagnetic compatibility: not considered in this report. Requirements are listed in paragraphs 5.4.18 – 5.4.22 of the Draft 18.5;

A.9 Battery life: 8 hours without alarm and 3 hours at alarm conditions;

A.10 Physical dimensions: 300x200x150 mm and 4 kg;

A.11 Ruggedness: waterproof under rain conditions; resistant to drop test in its shipping case (1 m height).

B. Test of the main Functional Requirements 1.1 through 1.6.

Test source.

The strongest 252Cf source available at the headquarters # K867 was used for sensitivity assessment in both static (1.1 and 1.2) and dynamic “search modes.” (1.6)

The source supplier data: 11.1E+5 Bq (3E-5 Ci) on 1989-01-15, which corresponds to No = 12.9 E+4 n/s intensity [ 1Bq = (2.646/85.5)*3.757 = 0.116, reference: “Passive non-destructive assay of nuclear materials”, LA-UR-90-732 ]. Presumably, this value is subject to 10% uncertainty. Taking into account 2.646 y half-decay time, the K867 source activity at 2004-02-10 shall be:

N = No * exp [ - 0.693 * 15.07 / 2.646 ] = 2500 +/-250 n/s

The actual neutron intensity was also estimated by a standard IAEA HLNC counter of 0.175+/-0.010 absolute detection efficiency for fission neutrons

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on 2003-07-02, and resulted in 437+/-7 cps, which corresponds (on 2004-02-10) to

N = (437 / 0.175) * exp [ - 0.693 * 0.58/2.646 ] = 2150 +/- 130 n/s.

Finally the K867 source intensity used for the test results assessment was arbitrarily set as 2300 n/s

B.1 Absolute detection efficiency

Absolute detection efficiency was measured in INTEGRAL mode at different (from 20 up to 120 cm) distances with less than 5% statistical uncertainty each. Background counts (0.15 cps) were always subtracted. Results are presented in the following Table 8:

Table 7.

L, cm 150 120 100 80 60 40 20 10 N, cps 0.41 0.55 0.72 0.89 1.52 2.42 6.31 14.3 F, n/scm² 0.008 0.013 0.018 0.029 0.050 0.114 0.458 1.83 N/F 51 42 40 31 30 21 14 8

Neutron flux, F, was calculated for “point” geometry and no scattering effects, and both assumptions are not valid. Instead of 1/R² law the N counts follow just 4/3 power. Detector has about 100x250 mm sensitive area. Nevertheless, the absolute detection efficiency requirement can be considered as met: 40 cps per 1 unit of flux at 1m distance, and 30 cps per 1 unit of flux at 80 cm distance. Declared value is 24 cps at 1 m. Specs minimum requirement is 20 cps.

B.2. SNM detection limit

Specs 18.5 test method requires at least 45 alarms at 50 tests when Cf252 source of 6500 n/s intensity is measured at 1 m distance with 10 s dwell time. It is equivalent to 2300 n/s at 60 cm distance. Actual measurement results are as follows:

Table 8.

L, cm 60 70 80 90 100 Alarms 50/50 46/50 37/50 32/50 21/50

SNM detection limit requirement is met.

B.3 False Alarm Rate (FAR)

Calculations show that the required FAR 1/300 at 95% confidence level can be achieved at the average FAR probability equal to 0.0012 (binomial distribution). Assuming Poisson statistics at 0.11 cps background (FAR was measured at these background conditions), it corresponds to the alarm

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threshold of 1.50 cps at 2 s dwell time, or about 2.3 cps at 1 s dwell time. According to the manufacturer, the alarm threshold algorithm is:

A = B’ + SIG sqrt (B’) + 2.

To satisfy the above FAR requirement actual alarm threshold A shall be higher than 1.50 at 2 s dwell time, and higher than 2.3 at 1 s dwell time (at 0.11 cps background)

The above relation allows to estimate that at dwell time 2 s and SIG=2, the alarm A = 1.58 cps. At dwell time 1 s and SIG=2, the alarm A = 2.77 cps. Hence, 2 SIG should be enough.

Results of actual measurements were as follows:

At 1s dwell time and 2 sigma: 11 alarms at 14 hours. At 2 s dwell time and 2 sigma: 1 alarm at 2 hours; at 2s dwell time and 3 sigma: 0 alarms at 4 hours. All results satisfy the FAR acceptance criteria at 2 sigma and 0.11 cps background.

However, calculations show that the FAR requirements at background higher than 0.2 cps can be reliably (better than 95% confidence level) met at 3 sigma only.

Final recommendation: At < 0.2 cps background 2 sigma meet the FAR

requirement; at > 0.2 cps background 3 sigma shall be used.

B.4 Safety Alarm.

The test was performed at SIL with 4 Cf252 sources (## PP770, 771, 772, 773) of total activity 1.27 MBq = 150 000 n/s. Assuming 1/R² law such source creates 100 µSv/h at approximately 12 cm distance. Direct measurement with calibrated MKS-A02 detector registered 100 µSv/h at 17 cm, which can be accepted keeping in mind scattering effects. The tested NSD has a selectable safety alarm threshold from 50 cps till 2000 cps. Direct measurements with above Cf 252 sources resulted in the following “Safety Alarm Threshold – Neutron dose rate” numbers:

Table 9.

Alarm, cps 50 100 200 500 1000 1500 2000 µSv/h 6.2 12.5 25 62 125 190 250 Dist., cm 120 73 47 24 13 8 4

For example, the Safety Alarm Threshold of 1000 cps corresponds to 125 µSv/h. In real practice this value seems to be too high. It can be recommended to set the Safety Alarm Threshold at 200 cps, or about 25 µSv/h.

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The acceptance criterion is met.

B.5 Gamma insensitivity. To be done…

B.6 Search mode.

18.5 Specs require movement of 20000 n/s Cf252 source from about 2 m till about 10 cm distance at about 0.1 speed and 2 s dwell time. The audio alarm signal should trigger at about 1.2 m and increase its frequency until the maximum value when the minimal distance is reached. The MCS graph should adequately reflect the process of signal intensity change.

Source of 20000 n/s intensity at 1.2 m distance creates about 0.11 n/s cm² flux. The used K867 source of 2300 n/s creates such flux at about 40 cm distance, or slightly more because of scattering effects, and the net neutron counts rate of about 2.4 cps (see paragraph B.2).

The search mode simulation was realized with the rotating machine, as illustrated in the following figure.

Fig. 28 Search mode scheme. The source is rotated on the end of the pole of R = 40 cm length. The maximum “source-detector” distance is 2R+L=90 cm, and minimum is L=10 cm. The rotator speed is variable from 0.75 till 20 circles

per minute.

The test was performed at the source linear speed of about 3 cm/s, 9 cm/s, and 18 cm/s. This is equivalent to the time intervals of about 23 s, 8 s and 3 s, when the “source-detector” distance is within 40 cm.

Background value was 0.19 cps. Alarm level at 2 s dwell time and 3 sigma criterion was

0.38 + 3*0.62 + 2 = 4.2/2 = 2.1 cps.

All tests have given positive results, and the MCS graphs clearly indicated rise and fall of the signal level. The maximum counts rate at the shortest distance was about 10 – 15 cps which agrees with the static mode tests (2.2).

The test has been repeated with artificial background value of 0.8 cps, which is equivalent to the largest “source-detector” distance of 90cm. At these

34

conditions the detector alarm level increases up to 3.7 cps at 2 s dwell time and up to 5.5 cps at 1 s dwell time. Again all tests were positive, but the duration of the alarm signal was shorter. At the rotator high speed there was delaying of the alarm sound of about 1 - 2 s. The search mode test results are positive. Criterion is met.

B.7 Environmental

The test was only partly performed. At – 20 degrees the device was still functional, but the battery life was considerably shorter: 4–5 hours instead of 8 hours at normal conditions.

B.8 Electromagnetic

Not considered here.

B.9 Battery life.

It was checked as 8 hours at least without alarm conditions. Operational during recharging time, which is about 2–3 hours.

B.10 Physical dimensions

300x160x120 and 4.6 kg – slightly more than recommended (4 kg), but acceptable.

B.11 Ruggedness

Not tested – we have one prototype only!

4.3.3. Comparative test of the NSD (Neutron Search Device) No 1 and No 2

performed by Dr. Yu. Kulikov on 23 March 2004

No 1 KSAR1U.06

No 2 Berthold LB 6414

A. Test at the Prater Atom institute

A.1 Insensitivity to Gamma radiation.

Cs137 source producing 1.2 mGy/h 1t 1 m distance was used. At the same room there was a strong PuBe neutron source producing 5E+6 n/s. When it was moved to the longest possible distance from the measurement station the neutron background was still about 2 cps (KSAR1U) and 0.8 cps (LB 6414). Both NSD were exposed to the Cs137 source gamma radiation of about 0.075 mGy/h or approximately 100 µSv/h.

The neutron counts rate results were:

No 1: Standard gamma background: 2.1+/-5% cps; 100 mkSv/h: 2.15+/-5% cps.

No 2: Standard gamma background: 0.8+/-7% cps; 100 mkSv/h: 0.8 +/-7% cps.

35

Conclusion: Both NSD are not sensitive to gamma radiation of 100 mkSv/h

A.2 Neutron detection efficiency

Measurements, performed with Cf252 source of 2300 n/s intensity resulted in:

No 1: 40 cm^2 at 100 cm, 31 cm^2 at 80 cm, 30 cm^2 at 60 cm, 21cm^2 at 40 cm

No 2: 20 cm^2 at 100 cm, 16 cm^2 at 80 cm, 13 cm^2 at 60 cm, 10 cm^2 at 40 cm

PRST: 40 -50 cm^2 at 40 cm.

Results

1. No1 NSD has at least 2 times higher neutron efficiency comparative to No2.

2. PRST has shown 2–2.5 higher efficiency than No 1. Keeping in mind, that the sensitive area of the PRST is 500 cm^2 (about 250 cm^2 for No 1), the intrinsic efficiency ratio PRST/No 1 will be 1.0 till 1.25, i.e. about the same.

4.4.5 The Neutron Search Device (NSD) Functional/Technical Specifications

and “Hand Held Neutron Monitor” KSAR1U.06 Acceptance Criteria and

Check List

Acceptance criteria of NSD were established in Chapter 6 of the “Technical/Functional Specifications for Border Radiation Monitoring Equipment”, Version 18.5 (December 2004). The following is a detailed point-by-point comparison of the NSD functional requirements formulated in 18.5 (pages 48–52) versus parameters of the prototype device KSAR1U.06, resulting from the tests and trial operation described in 4.4.1-4.4.3.

Acceptance Test Items and Check List are presented in Table 10.

Acceptance Test Item Yes/No Comments

6.3.1 Generic Requirements

a) Single-handed operation Y

b) Selectable Safety alarm threshold shown on display

Y Restricted access ??

c) Rugged design for outdoor use Y To be checked

d) Illuminated display Y

e) Battery life for use at least 8 hours Y

f) Critical parameters re-settable to factory defaults


g) Minimum buttons to operate Y (3)

h) Easy decontamination Y Not checked

36

i) Suitable documentation N Not available

j) Power on diagnostics, date and time Y - N Date and Time only

6.3.2 Detector

a) High efficiency e4quivalent to a moderated He3

Y 3 tubes in polyethylene

b) Reasonable insensitivity to gamma radiation

Y To be tested

c) Integrated in one instrument case Y

d) Optional – simple gamma detector for

safety alarm

N To be incorporated

6.3.3 Display and visual indication

a) Adequate display size, illuminated user settable, contrast adjustment, manual adjustment, if needed

N Size is OK, no adjustments

b) Capability for multi lingual support Y (?) English only

c) Indications in timer counter mode: time,cps,error…

Y Integer mode

d) Set up for search mode: dwell time, number of sigma

Y

e) Indication in search mode: cps, dwell time, sigma

Y - N Cps only

f) Permanent indications: battery status, safety alarm th., sigma, dwell time, alarm sound status

Y - N Battery status and alarm threshold

g) POST message, SW and HV, date/time, memory

Y - N Date and time only

h) Mode indication (INT, SEARCH, SET) Y

i) Graphic signal intensity indication (MCS – optional)

Y MCS

j) Alert/alarm indication and sigma multiplier

Y Sigma in set up only

k) neutron dose rate indication N Cps only

l) indication of remaining battery life Y

m) date and time Y

37

6.3.4 Acoustic signals

a) Alarm sounds, tone height, LED, vibration Y Tone height and LED

b) Increasing rate N

c) Visual and acoustic dose rate/cps alarm for safety

Y Visual and acoustic,at preset cps

d) Visual and acoustic low battery alarm Y-N Visual indication only

e) Silent and audible alarm capability: screen massage

??

6.3.5 Internal memory and PC link

a, d) Storage for >1000 alarms; standard file info,

all setup and diagnostic data and measurement results

Y-N Circular buffer, 2048

Date/time, cps, dwell time

b) PC with standard interface Y USB

c) ASCI data format Y

6.3.6 Power supply

a) Rechargeable and non-rechargeable battery

option

N Rechargeable only

b) Setup and data not lost, if main battery

is dead

Y (?) To be tested

c) Auto voltage setting, world wide AC adapter, car

Y - N

d) Switch to trickle charge, if left on mains power

?

e,f,g) Charging and full charge indication. Time is less than discharge time. Charge is possible during operation

Y

6.4 Technical and functional performance

6.4.1 Neutron sensitivity

6.4.1.1/2 Absolute efficiency: not less 20 cps/n/s cm²

Y See test report

6.4.1.3/4 SNM detection: not more than 100 g WGPu

Y See test report

6.4.1.5/6 FAR less than 1/300 at standard Y See test report

38

background

6.4.1.7/8 Safety Alarm: at 100 µSv/h; user settable

Y See test report

6.4.2/3 Gamma insensitivity: not less than 100 µSv/h

Y 100 μSv/h Cs-137

6.4.4 Alarm setting: visual / acoustic; displayed on the screen; adjustable within full range (restricted access)


6.4.5/6 Search mode: source localization process indication; MCS (optional); Y rescaling; bkg update

Y See test report

6.4.7 Environmental: operational at –20 - + 50 degrees

? Tested at –2O only

6.4.8 Electromagnetic compatibility ? Not considered here

6.4.9 Battery life: 8 h without alarm, 3 h with alarm; operational during recharge; recharge time < 8 hours

Y ? 8 h without alarm; alarm not tested; recharge is 2 hours

6.4.10 Physical dimensions: max 300x200x100; 4 kg

Y 300x160x120; 4.6 kg

6.4.11 Ruggedness: waterproof under rain; drop test

?? Not tested

Preliminary assessment.

A. Most of the technical and functional requirements, including high neutron sensitivity, are met.

B. A few parameters still require testing:

- environmental test at + 50 degrees (6.4.7) Presumably OK

- alarm condition battery life (6.4.9) Presumably OK

- data, set up parameters and time stamp shall not be lost, when battery is dead (6.3.6 b); Yes

- electromagnetic compatibility (6.4.8)

- ruggedness (6.4.11)

- easy decontamination (6.3.1.h) Presumably OK

C. Some of the required by Specs 18.5 parameters are not realized at present prototype device version:

a) Suitable documentation. Absolutely necessary.

39

b) Simple gamma detector for safety alarm. Optional, but very useful feature.

c) Safety alarm at INTEGRAL mode. It is necessary

d) Some relevant alarm event parameters are not stored in the memory file (Alarm threshold, number of sigma). It is necessary

e) Display contrast adjustment.

f) Bilingual support (English only). Presumably can be provided with any

requested language (?)

g) Neutron dose rate indication (cps only). Very desirable for fission

neutron spectra

h) No non-rechargeable battery option. Is it possible to realize?

i) Set up option is not password protected for restricted access to parameters adjustment. Desirable

j) acoustic low battery alarm . Desirable.

Additional requests:

1. The device can be used in unattended mode, but in such a case a red light flush shall be provided (optional: remote signalization on the custom officer desk?)

2. It is desirable to provide measures against accidental turn on/off of the instrument (during transportation, or accidentally by the user).

4.4.6. Improvement of NSD according to IAEA experts recommendations and

suggestions.

As it stated in Table 11 the number of improvement need to be made to enhance NSD usability. On the second phase of the Project all suggestions and recommendations were compiled, analyzed and the items to be improved were listed. Perfomance of the second generation of NSD was improved according to suggestions of acceptance tests and trial operation. The list of improvement is given in Table 11 Status in NSD KSAR1U.06 ver. 1. Status in NSD KSAR1U.06 ver. 2.1 Suitable documentation was not provided Operating manual including

circuit diagrams and connection diagrams

2 No gamma detector GM detector 0.06-3.0 MeV to indicate ambient gamma ray dose rate

3 No safety alarm in INTEGRAL mode Done 4 Relevant alarm event were not stored in

the NSD memory Done

40

5 Display contrast adjustment No 6 Multilingual support Russian, English 7 No neutron dose rate indication Done 8 No non-rechargeable batteries No 9 Free access to setup parameter

adjustment Password protected

10 No low battery alarm signal Yes 11 No LED indicator for using instrument in

unattended mode LED super bright LED on the face panel

12 No measures against accidental turn on/off of the instrument (during transportation, or accidentally by the user).

Additional protective washer to prevent from accidental switching

13 USB/Charge connector is not splash water protected

New IP 55 Connector

14 Some suggestions to user interface Corrected In addition, the weight of the NSD was decreased to 4.2 kg, additional handle and vibro signal were added.

The resulting NSD technical specification is given in Table 12. # Parameter Value 1 Absolute detection sensitivity of NSD to fission neutrons at

1 m, 1/n cm2 23

2 Alarm threshold, corresponding to the probability of detection 0.8, velocity of scanning 0.5 m/s, and distance of closest approach to the source 1 m, n/s

1.2×104

3 False alarm rate 1:600 s 4 Intrinsic background of NSD at standard background, CPS 0.2 5 The range of Ambient gamma ray dose rate, μSv·h-1 0.14÷1400 6 Duration of continues operation under no alarm condition

(LED lightning always off), h 8

7 No alarm generated when exposed to Co-60 gamma ray source, producing at the sensitive surface on NSD ambient dose rate, μSv·h-1

100

8 Dimensions, length × width × height, mm 300×160×1309 Range of operating temperatures, 0C -20÷+50 10 Sealed to IP 55 11 Net weight, kg 4.2

41

5. Feasibility study of LiI(Eu) detector based RID and SPRD.

The goal of the second part of the Project was a development of innovative approach for radiation detection and implementation such a technique for radioisotope identification device and spectral personal radiation detector.

The identification of the isotope causing a radiation alarm at a border monitor or within the country is essential for a quick resolution of the case. Commonly, hand-held radioisotope identification devices (RIDs) are used to solve this task. Although a number of such devices are commercially available, there is a gap between the requirements of the final users and what is technically feasible.

Several components are essential for the performance of such devices: − The quality of the raw data – the gamma spectra; − Isotope identification software matching the specifics of a

certain gamma detector; − Intrinsic sensitivity to gamma rays and neutrons.

The goal of the second phase of the project was the feasibility study of 6LiI(Eu) detector based RID as one of the promising approach to the problem of radioisotope detection, localization and identification in field.

5.1. Scientific background of the research work

The performance of hand-held radioisotope identification devices still lags the needs and expectations of their users. Gamma peak locations drift by more than the desired 1-2 % if the devices are exposed to high dose rates or significant temperature changes. The non-linearity of the light output requires corrections, which are often less than perfect. This causes spurious isotope indications, problems detecting shielded sources, and samples containing multiple or weak sources. The serious problem of the available RID is insufficient sensitivity to neutrons.

RIDs, on the other hand, play a decisive role when used by front line officers in the response-to-detection chain [12]. Large scale radiation monitoring at borders and inside a country can only be successful if the front line officers can quickly decide on the spot whether a radiation alarm was caused by naturally occurring radioactive material (NORM), by a medical isotope or whether there is a case of illicit trafficking of radioactive or nuclear material.

Although there is a new class of automated spectral radiation portal monitors emerging [13], [14] and [15] a RID remains as the regular instrument for the handling of this task. RIDs are also expected to be needed in the future to support a spectral border monitor to verify the reason for an alarm in the event that was not caused by NORM or a medical isotope.

As RID is a multipurpose instrument, which measures gamma ray energy distribution and dose rate as well as detects neutrons. Spectral detectors, used in a RID, have the following functions, which determine the requirements:

42

− Gamma spectrometry sufficient to identify the isotope; − Sufficient gamma ray and neutron sensitivity to allow the user to

quickly find a weak source which caused a radiation alarm of a portal monitor;

− Relatively short response time (either decay time of the light flash or the charge collection time in case of a solid state detector) to withstand a high count rate;

− Ruggedness to effective application in field; − Reasonable price and availability in large numbers.

Recently new gamma ray and neutron detectors were implemented into the

technology of radiation detection. These are LaCl3, 6LiI(Eu), CdWO4 scintillating detectors and CZT hemispheric and planar detectors with LaCl3 and 6LiI(Eu) - the most promising detectors among them [16].

Under this part of the Project experimental study of 6LiI(Eu) detector was to be done and R&D of hand-held 6LiI(Eu) based RID and SPRD were to be performed.

5.2. Technical approach and methodology

The most of commercially available RIDs use NaI(Tl) based detectors – IdentiFINDER-NGH (Target) [18], MKC-A03 (ASPECT)[19], GR-135 (Exploranium)[20] and Inspector 1000 (Canberra)[21]. Advantages of NaI(Tl) detectors include good availability, low price and high light output. There are some disadvantages however, which are usually taken into account by RID vendors. They are: non-linearity of the energy scale in low energy (<150 keV) area and strong dependence of light output on temperature.

Some instruments implements CsI(Tl)+pin-diode assemply - PM1801, PM1802 (Polymaster)[22]. This combination allows designing very compact, low power detection assemblies. Advantages of this approach are: 1) No high voltage is needed; 2) Low power consumption, small and compact design, 3) Low price, 4) Better resolution for energies above 300 keV. Meanwhile the detectors are sensitive to temperature and thus are hardly applicable for detection of gamma and X-rays with energies below 100 keV at temperatures above 40 C0.

Semiconductor detectors provide the ultimate solution to isotope identification. Room temperature semiconductor detectors – such as CdTe, CdZnTe, HgI2 – are used in some RIDs (fieldSPEC-CZT, GR-135 and ICS 4000 (XRF Corporation). Simultaneously with better energy resolution, linearity and lower temperature drift, there are some problems, which limit application of these detectors – price and small size.

General problem of all RID available in the market now – low neutron sensitivity, what recently resulted in the appearance of new class of hand-held instrument – neutron search devices. Even the most sensitive RID – MKC-A03 – is generally used only for verification of neutron alarm but neither for search nor for localization of the week neutron sources.

43

The big interest to highly enriched 6LiI(Eu) crystal (>90% of 6Li) as neutron detector due to a high detection efficiency to thermal neutrons. 6Li nuclei have a large cross section of 940 barns for (n,α) reaction with kinetic energy release of 3.78 MeV:

HLin 36 +→+ α This principal feature of the crystal as an efficient neutron detector

makes the scintillator very competitive in application to border monitoring [23]. Moreover, scintillators with good performance for gamma ray and neutron detection are of great importance for RID, because in all contemporary instruments without exception the combination of two detectors (gamma and neutron) is used.

6. Experimental study of 6LiI(Eu)

As it was noted earlier gamma ray spectrometry of the detector within RID must be sufficient to identify the isotope. Energy resolution, light yield and its non-proportionality to the incident energy, scintillation pulse shape and temperature drift are the most critical parameters for gamma ray spectrometry in RID.

Same sample spectra obtained with LiI(Eu) detector is shown in Fig. 1.

0 500 1000 1500 2000 2500 3000 3500 40000

2000

4000

6000

8000

10000

Cf-2

52 (n

eutro

ns)

Co-

60Cs-

137

Amplitude, keV

Fig. 29. Amplitude spectra of gamma from Cs-137,Co-60 and neutrons

from Cf-252.

44

Energy resolution of the crystal was measured with 6LiI(Eu) detector ∅45×20 mm coupled to 2” bialkali PMT Hamamatsu R6231-01. Anode signal was processed with Spectrometric 2048-channel Analyzer. Some results of the measurements are presented in Figs 30-31.

0 500 1000 1500 20000

1x103

2x103

3x103

4x103

5x103

6x103

6LiI(Eu) 45x20 mm

7.5 %

Cou

nts

Energy, keV

100 1000

5

10

15

20

25

30

Th-2

32

Eu-1

52C

o-60

Ce-

139

Cs-

137

Am-2

41

Eu-1

52

Ener

gy re

solu

tion,

%

Energy, keV

Fig. 30. Energy distribution of gamma rays emitted in Cs-137 decay.

Fig. 31. Energy resolution vs gamma rayenergy.

The data obtained (7.5% for 0.662 MeV) are in a good agreement with that measured in [17] with 6LiI(Eu) ∅50×5 + PMT XP5200.

The photoelectron number per 1 MeV gamma ray, defined as a ratio between position of the gamma ray peak and the position of single photoelectron peak was measured in [17] and found to be 15000±1500 phe/MeV (∼40% of NaI(Tl)).

Fig. 32 presents the non-proportionality characteristic of 6LiI(Eu) in comparison to that of NaI(Tl) and CsI(Tl) crystals [17]. Non-proportionality is defined here as the ratio of photoelectron yield measured for photopeak at specific gamma-ray energy relative to the yield at 662 keV gamma peak [16].

10 100 10000,90

0,95

1,00

1,05

1,10

1,15

1,20

1,25

1,30

1,35 LiI NaI(Tl) CsI(Tl)

Ligh

t yie

ld [%

of 6

62 k

eV]

Energy [keV]

Fig. 32. Non-proportionality of the LiI(Eu) detector.

45

The non-proportionality of 6LiI(Eu) follows the general behavior of the halide crystals with an excess of light at low energies. However, the value of non-proportionality relative to that of NaI(Tl) crystal is much less, what makes the use of crystal more attractive for detection of low energy gamma ray.

Sufficient gamma ray and neutron sensitivity of RID allows the user to quickly find a weak source which caused a radiation alarm of a portal monitor. Therefore, intrinsic detection efficiency is also very important characteristics of the detector.

The main properties of LiI(Eu) crystal are compiled in Table 13 [14].

Table 13. Characteristics of scintillators. Parameters CdZnTe CdWO4 LaCl3(Ce) 6LiI(Eu) NaI(Tl) CsI(Tl) Density [g/cm3] 6 7.9 3.86 4.08 3.67 4.51 Decay time constant [μs]a

0.1-0.2 14 0.025 1.2 0.23 1

Light output [ph/MeV] -- 27000 36000 15000 40000 53000 Temperature drift [%/˚C]

0.01 ±0.01 ±0.1 -0.3 ±0.1 ±0.13

Effective atomic number

48.5 65 59.5 53 50 54

Peak emission [nm] -- 475 335 475 420 550 Non Prop. of light yield[%]b

-- -20 ±3 +13 +20 +15

Energy resolution Cs-137 [%]

2.5 6.6 4.2 7.5 6.7 6.1

Energy resolution Am-241 [%]

-- 21 13.6 21 10.4 13.7

Sensitivity for therm. neutrons

No no no very high

no no

a. For the CdZnTe detector, the charge collection time; b. Non proportionality of the light yield at 20 keV.

Density and effective atomic number of 6LiI(Eu) crystals is slightly higher

then that of NaI(Tl), what makes the gamma ray detection efficiency of the 6LiI(Eu) greater as well.

As it follows from the table 13, the decay time of LiI detectors is longer than that of NaI one. We tested a withstanding of LiI based detector to high countrate. Ra-226 0.5 mCi source was used for this test. Results are presented in Fig. 33.

46

0 250 500 750 100010

100

1000

10000Peak shift 1.7%

Ra-226

ADC channel

2 kCPS 5 kCPS 12 kCPS 30 kCPS 50 kCPS

Fig. 33. Test of withstanding to high count rate.

The peak shift observed was found to be 1.7% with shaping time 1 μsec. Detection of fast neutrons with 6LiI(Eu) detector is more efficient when

the neutrons are moderated up to thermal energies. Therefore, 6LiI(Eu) detector for RID should be surrounded by a H-enriched material, such as polyethylene. Optimization of the moderator size/shape was performed during RID design on the second stage of the Project.

The obtained results have shown an attractiveness of the 6LiI(Eu) detector for the RID because of the following reasons:

1. The crystal performance is good enough to provide the reliable gamma ray spectrometry. 6LiI(Eu) has about the same energy resolution as NaI(Tl) does, non-proportionality of light output is less.

2. Gamma ray detection efficiency of 6LiI(Eu) detector is slightly greater than that of NaI(Tl).

3. 6LiI(Eu) allows simultaneous detection of gamma ray and neutrons with about the same efficiency as He-3 proportional counters in moderator of the same mass. An advantage of scintillation detector is it’s ruggedness to vibration and shock (no microphonic effect).

47

7. Development of 6LiI(Eu) based RID

Within the framework of the Project 12602 6LiI(Eu) based RID “SIGMA-n” (Spectrometer Identifier GamMA ray and Neutron”) KSAR1U.05-03 was designed and built. The design process covered three main directions:

− Hardware development (6LiI(Eu) detector, Multichannel Analyzer, instrument case and associated accessories);

− User interface development; − Software development – algorithm of identification. Data on gamma ray spectrometry has shown that, regardless to conventional

point of view, 6LiI(Eu) detector performance is sufficient for isotope identification. Another important parameter of RID detector is the neutron sensitivity, allowing the user to fulfill not only verification, but also search and localization of neutron sources 2÷3×104 n/s in cargo, cars, vehicles or pedestrians. The required performance can only be achieved provided neutron sensitivity is greater then that of today the most sensitive RID – MKC-A03, with mass upper limit of 3 kg.

A series of numerical (Monte-Carlo) calculations aimed the optimization of the shape and size of the moderator was performed within the Project. Initial parameters of calculation were:

− Mass of the detector should not exceed 1.4 kg; − Thickness of 6LiI(Eu) crystal should not be less then 20 mm (due

to required sensitivity to Co-60); A criterion for optimization was maximum specific sensitivity of the assembly ξ to fission neutrons:

cmx mmm ++=

εξ

where ε - sensitivity to neutrons, mm , cm - masses of the moderator and

crystal, xm - constant, denoting the expected mass of the RID electronics,

case, batteries etc, xm =1.1 kg). Results of optimization are shown in Table 14 and Fig. 34.

48

Table 14. Results of Monte-Carlo calculation Parameter

Value, mm

Crystal

Diameter 45 Thickness 20

Moderator

Bigger diameter of cone 138 Smaller diameter of cone 89 Height 87

Forward wall thickness 22

Fig. 34. RID detector. 1-6LiI(Eu) crystal, 2 - polyethylene moderator, 3 - PMT, 4 - detector case.

Multichannel Analyzer includes spectroscopic amplifier, baseline restorer,

overlap rejecter and 2048 channel flash ADC. The Analyzer is controlled by MSP430-8 MHz microprocessor with 10 kB of RAM. An additional feature involves statistical leveling, providing differential nonlinearity less then 1%.

The main microcontroller is built on the platform of STM 269-40 MHz processor with 512 kB ROM and 256 kB RAM.

To provide the dose rate measurement up to 10 mSv RID contains Geiger counter SBM-20.

RID has an USB ver.2 internal interface for transferring recorded spectra onto PC. Software is running under Windows 98/ME or 2000/XP operating systems.

Alarm indication of the RID includes three items: silent indicator (vibro-), LEDs and piezosiren. Each indicator can be activated/deactivated from the SETUP of RID.

RID case was developed with KOMPAS-3D V7 Plus computer-aided design kit by rapid prototyping method. Original layout of the detector in RID across the symmetry axis is caused by agronomic requirement – to decrease the load onto the user’s wrist.

User interface of Sigma-n was developed in accordance with “Usability Guide for Manufacturer of Radiation Monitoring Devices”, ver.1.0 by Michael Alexander. New approach to the design process was implemented – first, the software emulator of RID was developed and sent to IAEA for preliminary examination. Second, all suggestions and recommendations were accumulated and implemented into RID software.

49

RID user interface involves 3 button operation (Fig. 35) with the following modes (Fig. 36-39):

Fig. 35. User panel of RID. 1 – Neutron alarm LED indicator 2 – Enter key 3 – Cursor up/left 4 – Cursor down/right 5 – Power/Lightning on/off 6 – Instrument On 7 - Gamma alarm LED indicator

8 – LCD 82×62 mm 320240 pix mono

ALARM

SI GMA-n

1

2 3 4 5

6

7

8

Fig. 36. RID in Search mode.

Fig. 37. RID in Identification mode.

Fig. 38. RID in Integral mode.

Fig. 39. RID in Setup mode. Instrument has the battery independent memory to store up to 64 gamma ray spectra. User interface has Multilanguage support.

Isotope identification software of RID is a combination of two traditional approaches to the radioisotope identification issue – templates matching and peak analysis method. Both methods have their intrinsic advantages and

50

disadvantages. Preliminary results of RID trial operation have shown the higher reliability of the coincidence scheme.

7.1. Test and trial operation.

“SIGMA-n” was tested in accordance to IAEA Specification at Scientific Engineering Center “Nuclear Physics Research”. Results of the test are compiled in Table 15.

51

Table 15. Compilation of “SIGMA-n” test results. S

pecification

Results

Parameter

IAEA

ANSI N42.34

6LiI(Eu) RID

SIGMA-n

1 Energy range of

the detected gamma

ray, MeV

0,03-3,0

- 0,03-3,0

2 Neutron detection

+ +

+

3 Sensitivity to

gamma ray

0,05 μSv/h

in 1 s

(Am-241, Cs-137, Co-60)

0,05 μSv/h

in 1 s

(Сs-137)

Am-241 - 0.008 μSv/h

in 1 s

Ba-133 - 0.017 μSv/h

in 1 s

Cs-137 - 0.061 μSv/h

in 1 s

Co-60 - 0.142

μSv/h

in 1 s

4 Sensitivity to

neutrons

2.0×10

4 n/s @ 0.2 m in 5 s

2.3×10

4 n/s @ 0.25 m in 2 s

1.5×10

4 n/s @ 0.2 м in 5 s

5 FAR

1:60 (gamma), 1:3600

(neutron)

- 1:60 (gamma), 1:3600 (neutron)

6 Dose rate, μSv/h

0,01

÷100

0,01

÷100

0,01

÷100

6 Number of ADC

channels

1024

- 2048

7 Isotopes to be

identified

U-233,235,238, Np-237, Pu,

Ga-67, Cr51, Se-75, Tc-99m,

Pb-103, In-111,

I-123,125,131, Tl-201,

Xe-133, K-40, Ra-226,

Th-232+ daughters, Co-57,

U-233,235,238, Np-237, Pu,

Ga-67, Cr-51, Se-75,

Tc-99m, Pb-103,

In-111, I-123,125,131, Tl-201,

Xe-133, K-40, Ra-226,

Th-232 + doughters, Co-57, Co-

U-233,235,238, Np-237, Pu,

Ga67, Tc-99m,

I-123, K-40, Ra-226, Th-232+

daughters, Co-57, Co-60, Ba-133,

Cs-137, Ir-192,

Eu-152, Am-241, Cf-252, Pu-Be,

52

Co-60, Ba-133, Cs-137, Ir-

192, Tl-204, Am-241, Eu-152

60, Ba-133, Cs-137, Ir-192, Tl-

204, Am-241

Ce-139, Sn-113, Y-88

Simultaneous identification up

to four isotopes

8 Layout

Single handled, no external

cables


cables


cables

9 Overall

dimensions, mm

300×200×150

- 304×185×150

10

Weight, kg

3 -

2,6

11

Duration of the

continues

operation without

charging, h

8 2

6

12

Operating

temperatures and

humidity,

0 С, % at

+35

0 С

-20÷+50,

90

-20÷+50,

90

-20÷+50,

90

53

8. Development of LiI(Eu) detector based SPRD

Improvement of the radiation detection technologies and associated electronics has recently resulted in the development of new category of radiation detection instruments – Spectral Personal Radiation Detectors. Being the most advanced instruments for radiation monitoring equipment, these devices utilized the advantages of RID (internal MCA, reliable identification of radioactive isotope caused an alarm and low weight and size of pocket type instruments.

It is obvious, that the idea to use LiI detector simultaneously as spectroscopic gamma and neutron detector is even more attractive for a SPRD then for an RID. Therefore, the third part of the Project was dedicated to development and test of LiI(Eu) based SPRD.

General view of the spectral personal radiation detector SIGMA-n personal is shown in Fig. 40.

54

SPRD has the same functional units as RID does. They are: LiI(Eu) based radiation detector with incorporated into the crystal LED pulser and Hamamatsu R1924 photomultiplier tube (1), main PC Board (2) with multichannel analyzer, microconroleer and high voltage power supply, buttons (3), graphical LCD (4), and Li-ion rechargable batteries. Fig. 8 shows the cross section of SPRD.

Fig. 41. Inner view of Sigma-n personal SPRD

1 – Detecting unit 5 – On/Off button 2 – Main board 6 - Li-ion batteries 3 – Operating button PCB 7 – Clip 4 – LCD

Neutron and gamma rays are detected with ∅18×18 mm 90% enreached with Li-6 LiI(Eu) crystal. Multichannel Analyzer includes spectroscopic amplifier, baseline restorer, overlap rejecter and 2048 channel flash ADC. The Analyzer is controlled by MSP430-8 MHz microprocessor with 10 kB of RAM. An additional feature involves statistical leveling, providing differential nonlinearity less then 1%.

The main microcontroller is built on the platform of STM 269-40 MHz processor with 512 kB ROM and 256 kB RAM. SPRD has an USB ver.2 internal interface for transferring recorded spectra onto PC. Software is running under Windows 98/ME or 2000/XP operating systems. Similar to RID’s user interface the SPRD has four modes of operation: setup, search, integral and identification. Physical dimensions and weigth of LiI(Eu) based SPRD are 171×96×54 mm, 0.46 kg.

55

8.1. Preliminary results of laboratory test.

Laboratory tests were performed at Scientific Engineering Center “Nuclear Physics Research”. The tests are still underway but some preliminary results are compiled in Table 16. Table 16.

Results

Parameter

IAEA 6LiI(Eu) SPRD

SIGMA-n personal

1 Energy range of the detected gamma ray, MeV

0,06-1,330 0,03-3,0

2 Neutron detection + +

3 Sensitivity to gamma ray

0,5 μSv/h when passed by with 0.5 m/s

Am-241 - 0.06 μSv/h

Cs-137 - 0.20 μSv/h

Co-60 - 0.60 μSv/h

4 Sensitivity to neutrons

2.0×104 n/s when passed by with 0.5 m/s @ 0.1 m

1.5×104 n/s

5 FAR 1 in 3600 sec 1 in 3600 sec

6 Dose rate, μSv/h 0,01÷100 0,01÷100

6 Number of ADC channels

- 2048

7 Isotopes to be identified

U-233,235,238, Np-237, Pu, Ga67, Tc-99m,

I-123, K-40, Ra-226, Th-232+ daughters, Co-57, Co-60, Ba-133, Cs-137,

Ir-192, Eu-152, Am-241, Cf-252, Pu-Be, Ce-139, Sn-113, Y-88; Simultaneous

identification up to four isotopes

8 Overall dimensions, mm

200×100×50 171×96×54

9 Weight, g 400 0.460 kg

56

10 Duration of the continues

operation without charging, h

400 8

11 Operating temperatures and humidity, 0С, % at +35 0С

-20÷+50,

90

-20÷+50,

90

Preliminary results of the test have shown that LiI(Eu) based SPRD meets IAEA requirement demanded for PRD. Moreover, the use of LiI(Eu) detectors enhance neutron sensitivity of the instrument and provide simultaneous detection of gamma ray and neutrons with one detector.

Summary of the Project

Under the IAEA Project#12602 feasibility study of neutron detector for hand-held search instrument was performed. It was shown, that 3-He proportional counters in polyethylene block as fast neutron moderator are the most useable detector for hand-held neutron search instrument, matching sensitivity of portal radiation monitors. Within the Project production prototype was developed and tested by IAEA experts.

Technical and functional performance achieved enable the use of NSD prototype - KSAR1U.06 as a necessary supplement for vehicle and pedestrian monitors to do an independent verification of neutron alarm of a large border monitor, allowing first-line officers to verify the alarm and localize the source.

Feasibility study of 6LiI(Eu) detector for radioisotope identification device was performed during the second year of the Project. It was shown, that application of the detector in portable devices is attractive for the following reasons:

− relatively good spectrometry; − very good sensitivity to neutrons.

RID KSAR1U.05-03 “SIGMA-n” and SPRD KSAR1U.08 “SIGMA-n personal” both on the base of 6LiI(Eu) detector were built and tested. Results have confirmed the conclusion made about good applicability of 6LiI(Eu) in isotope identification device.

The RID “SIGMA-n” and SPRD “SIGMA-n personal” are expected to be sent for independent testing to the IAEA.

57

References

[1] N.A. Vlasov., Neutrons. Moscow, Science, 1971. [2] Yu. N., Pepyolyshev, Neutron Registration. Present State and Prospects of

Development, Preprint JINR P13-86-719,1986. [3] Passive Nondestructive Assay of Nuclear Materials, Edited by: D. Reilly,

N. Ensslin and H. Smith, Jr, NUREG/CR-5550 LA-UA-90-732, 1991. [4] Majorov, M., Lebedev, A., Materials of 4-th International Conference

“Radiation Safety: ecology-nuclear power” 24-28 September 2001, St.Petersburg, Russia, p.284.

[5] A. Syntfeld, M. Moszyński, R., Arlt, M. Balcerzyk, M. Kapusta, M. Majorov, R. Marcinkowski, P. Schotanus, M. Swoboda, and D. Wolsky, 6LiI(Eu) in Neutron and γ–ray Spectrometry – a High Sensitive Thermal Neutron Detector, Materials of the 2-nd Research Coordination Meeting of IAEA on “Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials”, 4-8 October, Sochi, Russia 2004.

[6] Technical/Functional Specifications for Border Radiation Monitoring Equipment, Draft 18.5, IAEA, December 2004.

[7] Batenkov O., et. al., Nuclear Instruments and Methods in Phys. Research A329:235, 1997.

[8] Ju.I. Kolevatov, V.P. Semenov, L.A. Trykov, Spectrometry of neutrons and gamma ray in radiation physics. Moscow, Energoatomizdat, 1990.

[9] Majorov, M., Lebedev, A., Materials of 4-th International Conference “Radiation Safety: ecology-nuclear power” 24-28 September 2001, St.Petersburg, Russia, p.281.

[10] Ronald, I., Ewing and Keith W. Marlow, Nuclear Instruments and Methods in Phys. Research A299 (1990), 559-561.

[11] V. Frolov, Nuclear-physics methods of fissionable materials control, Moscow, Atomizdat, 1976.

[12] IAEA TECDOC-1312, Detection of radioactive materials at borders, September 2002.

[13] M. Schrenk, R. Arlt, P. Beck, H. Boeck, F. Koenig, T. Leitha, “A Real Time, Isotope Identifying Gamma Spectrometer for Monitoring of Pedestrians”, presented at the IEEE 2004.

[14] Draft IAEA-TECDOC, Technical/Functional Specifications for Border Radiation Monitoring Equipment, October 2004.

[15] N42.38 D, American National Standard Performance Criteria for Spectroscopy Based Portal Monitors used for Homeland Security.

[16] Balcerzyk, M. Moszynski, M. Kapusta, “Comparison of LaCl3: Ce and NaI(Tl) scintillators in γ-ray spectrometry”, 2001 Elsevier Science.

58

[17] Syntfeld, M. Moszynski, M. Balcerzyk, m. Kapusta, M. Majorov, P. Schotanus, R. Marcinkowski, M. Swoboda and D. Wolski, “6LiI(Eu) in Neutron and γ-ray Spectrometry-a High Sensitive Thermal Neutron Detector,” (presented at the IEEE 2004).

[18] IdentiFINDER-NGH, Target, http://www.target-systems-gmbh.de/. [19] MKC-A02, MKC-A03, Aspect, http://aspect.dubna.ru/. [20] GR-135, Exploranium, http://exploranium.com. [21] Inspector 1000, Canberra, http://www.canberra.com. [22] PM1801, PM1802, Polimaster, http://polismart.net. [23] M. Majorov, A. Lebedev,”LiI(Eu) crystals application in radiation control

hardware”, Proc. Of the 4-th Int. Conf.: “Radiation Safety: Ecology-Nuclear Power”, 2001, St. Petersburg, Russia.

[24] WinMCA, Forschungszentrum Rossendorf, http://fz-rossendorf.de. [25] M. Swoboda, R. Arlt, V. Gostilo, A. Lupilov, M. Majorov, M. Moszynski,

A. Syntfeld “Spectral Gamma Detectors for Hand-Held Radioisotope Identification Devices (RIDs) for Nuclear Security Applications” (presented at the IEEE 2004).

[26] ICS 400, XRF Corporation, http://www.xrfcorp.com.

Radiation Monitoring with Norm Detection of Vehicles at Borders at Stand-Still

V.D. Petrenko, Yu.N. Karimov, N.N. Shipilov, A.I. Podkovirin, M.I. Fazilov, B.S. Yuldashev

Uzbekistan Academy of Sciences, Institute of Nuclear Physics, Uzbekistan

Abstract

In this report the results of works performed under the Research Contract 12607/R0 are

presented. All accomplished works can be divided into three parts:

In the first part, the method is described used for radiation control of vehicles crossing

customs and border check-points. At detection of the radioactivity in a vehicle, it is detained and

the NORM materials and other radioactive materials are identified. The radiation monitor model

was developed with realisation of the described vehicle control concept in it. This model was

tested on-site on real vehicles crossing the customs check-point of the Republic of Uzbekistan.

In the second part, based on the activities conducted in the first part, it was demonstrated

that this method can be realised in practice for the purpose of its application at customs and

border check-points for detection of nuclear and fissile materials with identification of

radioactive isotopes in NORM-materials and other radioactive materials, as well as for decision

making upon further possible transportation of such cargo. These results, showed in this report,

were analysed and demonstrated that the 100x50 mm NaI(Tl)-crystal based scintillation detector

is applicable, so that the identification capabilities become better when such detector is placed at

50 cm distance and closer to the vehicle. Other single crystal can be also used; however, one

should take into account its geometry depending on the sizes of the crystal.

In the third part, new sophisticated methods for radioactive materials detection were

described, and the possibility for creation of radiation monitoring instrumentation based on the

described above concept was demonstrated.

LIST OF ABBREVIATION, SYMBOLS, UNITS AND TERMS

ADC Analog to Digital Converter

SCC State Customs Committee

MCA Multi Channel Analyzer

IAEA International Atomic Energy Agency

INP Institute of Nuclear Physics

NORM Naturally Occurring Radioactive Material

R&D DEPARTMENT Research and development department

TIR Transportation International Regulations

PM Photo-multiplier

RID Radioisotope Identifier Device

2

INTRODUCTION

Prevention of nuclear weapon and its components spread is important task in the modern world. For this aim, the control over the transportation means should be carried out at the customs border check-points to discover radioactive and fissile materials transported in them. In case when the transportation mean cannot be opened (for example, TIR system vehicle) within the control procedure it is important upon the alarm signal to determine the source location and to estimate its radioactive threat. It is also important to identify the radioactive source in term of its possible use as nuclear weapon component despite the “clean” or “dirty” bomb this could be.

The given report presents the results of efforts for the № 12607/RO Research Contract performed in years 2005/2006 at the Institute of Nuclear Physics of the Academy of Sciences of the Republic of Uzbekistan in cooperation with the IAEA. This project anticipated the assembly and testing of the 100х50 mm NaI(Tl) crystal based detector and the appropriate software for the automatic identification of nuclides in the transportation means at stand-still by real-time identification of spectrum from the NaI(Tl) detector.

Such approach is especially important for the case of the NORM materials (naturally occurring radioactive material) transportation, which can be determined as industrial materials, like tile, fertilizers, organic fertilizers, porcelain toiletry and others, containing 40K, 232Th, 226Ra, natural uranium, and causing so-called “innocent alarms”.

Identification of 40K is especially important as it is contained in many everyday life materials.

1. CONCEPT AND STRUCTURE OF THE RADIATION MONITOR FOR TRANSPORTATION AT STAND-STILL

In our opinion, the concept of the stationary radiation monitor at vehicle stand-still when excessive radiation is detected and the alarm is set off can be presented as follows:

• In case of alarm, at detected radioactivity in the object, it is important to obtain the radioactive materials distribution along the length. • The object with the radioactive material must stop in the control area for spectrum measurement and identification of the radioactive material. • After this the decision upon the further destiny of the vehicle is made.

3

Picture 1 shows the reciprocal location of the major parts of the installation — scintillation plastic detector, NaI(Tl) crystal based scintillation detector, traffic lights and object sensor realizing the proposed concept.

Figure 1: A representation of the measurement process for a cargo vehicle. The graph represents the measurement of a gamma signal as a function of time (i.e. as a function

of position of the truck as it passes by the detector). In this example, the truck generates a gamma alarm.

Our researches demonstrated that combination of the plastic and NaI(Tl) scintillators in one casing makes the determination of the isotopes characteristics more complex when the isotope is distributed unevenly along the length or there are several isotopes at different points of the object. Therefore, we selected the following structure of our installation:

The first monitor contains plastic detector and is in charge for the following:

- detection of radioactive material in an object;

- measurement of radioactive material’s distribution along the whole object in percents;

- alarm signal forming and switching on the “STOP” signal (red signal on the traffic-lights) when radioactive material is detected.

4

The second monitor has NaI(Tl) detector and performs the following operations:

- it measures spectrum for 5–10 minutes when the alarm received from the first detector monitor and the object appears in the second monitor’s control area;

- it forms the signal on isotope identification when the measurements are finished;

- based on the information on radioactive material distribution and the type of the radiation source, the logical unit makes decision whether to give permission to leave or to detain for detailed radiation control. If the decision made for the vehicle to leave the traffic-lights sets off green light, permitting further transportation. On the contrary, the decision upon detailed radiation control is drawn and the object is redirected to the examination area.

The layout of the installation is shown on Fig. 2.

Figure 2: A diagram of the electrical interface between components and the recording computer.

5

Fig. 3 demonstrates the logical circuit of the detector.

Figure 3: The alarm logic of the detection system.

Isotope identification Идентификаци

NORM material

NORM

Prohibited material Запрещенный

YES ДА

Presence of object Наличие объекта

Traffic lights СветофорGreen light ON

Вкл. Зеленый свет

Red light ON Вкл. красный

свет

Permission to leave

Разрешить проезд

Passage not allowed

Запретить проезд

Detailed check Углубленный

досмотр

NO НЕТ

YES

ДА

Plastic detector Yes

ДА

Background excess

Превышение фона


Isotope identification

Идентификация изотопа

NORM material NORM материал

Prohibited material Запрещенный материал

YES ДА


NaI(Tl) detector Детектор

NaI(Tl)

6

Based on the given algorithm the software was developed. When there is no the object the first detector measures background, which is indicated on Fig. 4.

Figure 4: An example where no vehicle has entered into the detection region.

One can see from the Fig. 4 that the background level is lower than the alarm level indicated with bold line. In the second case, when the object appears and does not contain radioactivity, the background stays close to the earlier detected level shown on Fig. 5.

Figure 5: An object enters the detection region, but does not contain any radioactive material.

7

Figure 6: A vehicle enters the detector, carrying radioactive material. An alarm is generated.

Let us now consider the case when the object moves by the monitor. Fig. 6 shows that there is a material with excessive radioactivity (fertilizers occupying 21% of the object’s volume) in the central part of the object.

Figure 7: A modelled example of a vehicle containing multiple sources.

8

We have modelled the situation when the radioactivity sources were placed at different points of the vehicle, and as a result we have obtained two clear peaks indicating two sources caused this alarm. This is shown on Fig. 7.

To conduct all these measurements we have prepared this radiation monitor based on 600х200х95 mm3 plastic detector. Fig. 8 shows this detector (1) with the PM FEU-118 (2) matched with the signal analyzer (3). Further the information is sent onto the industrial computer (4). This installation is placed in metal casing (5), and can be mounted at the determined points for radioactivity detection and for analysis of obtained spectrum.

Figure 8: A radiation monitor that used

plastic scintillator. Figure 9: A radiation monitor that used

NaI(Tl).

Fig. 9 shows the radiation monitor composed of the 40х40 mm2 NaI(Tl) crystal based scintillation detector (1). All the data is analyzed by the industrial computer (2). The equipment is placed into the casing (3), and can be carried along the vehicle.

Unfortunately, we did not have the large size scintillation detector, and therefore, the main aim of the research was a detailed study of identification of the isotopes referred as the NORM-materials.

3

2

1

4

1

2

3

5

9

2. ASSEMBLY AND TESTING OF THE 100x50 mm NaI(Tl) CRYSTAL BASED DETECTOR

Later on in our further researches for radioactive NORM-materials identification, we used equipment (Fig. 10) kindly provided by the IAEA’s Department of Safeguards Nuclear Security Equipment Laboratory (NSEL). This equipment included: a NaI(Tl) crystal based detector (1), a miniМСА-166 multichannel analyzer (2), a portable computer TOSHIBA with the Specmon software (3).

The Specmon software is designed to identify passing sources in a spectroscopic based portal monitor system. To increase the count statistics, the new recorded spectrum gets added to the prerecorded spectra of the same alarm. After background elimination the spectrum will get identified into the memory of the computer (online). In case of a radioactive medical isotope the activity alarm is flagged out by a green light. A red light is given in the case of another isotope or if the identification was not possible. A logbook shows the results at the display. This program combines the recording software winSPECa and the identification software Identify with the subprogram Quick Identify. All four programs were designed by the company GBS.

Figure 10: Loaned equipment from the IAEA’s Nuclear Security Equipment Laboratory (NSEL).

1

2

3

10

Figures 11 and 12 show this equipment in the casing of the radiation monitor.

Figure 11

Figure 12

11

3. TESTING OF THE SYSTEM AT THE CUSTOMS POINTS

Since the transportation of the radioactive NORM-materials through the customs check-points in Uzbekistan does not happen everyday the SCC is concurred and agreed on the possible immediate response of the INP experts by visiting the Yallama, Navoi, Gisht-Kuprik, Bekabad and Alat customs check-points for examinations and identification of NORM materials. All the abovementioned customs check-points are equipped with the Yantar-type portal radiation monitors produced by R&D Centre Aspect (Dubna, Moscow region).

3.1. Customs check-point Yallama.

Fig. 13 shows the Yantar-type portal radiation monitors produced by the R&D Centre Aspect (Dubna, Moscow region) installed at the Yallama Customs check-point. The vehicle radioactive monitors are installed at both sides of the road. The passage for pedestrians is equipped with dual portal monitor for control of the radioactivity of passing through pedestrians and passengers. Every monitor is equipped with the surveillance camera.

Figure 13: The Customs Check Point Yallama.

When this portal radiation monitor detects some radioactive cargo the server computer displays the vehicle or the face of the passenger who caused the alarm signal (Fig. 14), as well as the distribution at gamma or neutron channel, at which this alarm was set off (Fig. 15). One can see from the Fig. 15 that the radioactivity source is located at the rear part of the vehicle as the marker line is drawn at the activity peak, and the corresponding video-image shows the rear part of the trailer.

12

Figure 14 Figure 15

The personnel of the customs point detains this vehicle and calls for the INP experts. Upon arrival on-site the INP experts analyze the radioactive cargo distribution along the vehicle by taking into account the diagram from the Fig. 15 with the following measurement of the spectral composition of the detected ionizing radiation source.

Figure 16 Figure 17

It is occurred that in the sack located in the rear side the last on the left-hand side contained molybdenum oxide contaminated with radioactive isotopes. Figures 16 and 17 show this vehicle with discovered radiation source inside (sacks with molybdenum oxide).

13

After this, the concurred with customs personnel standard procedure of radioactivity source identification and decision making takes place.

Figure 18: A NaI(Tl) measurement at the front of the trailer of the vehicle shown in Figs 14–17. The distance to the trailer was 1 meter horizontally, and the detector

was 2 meters above the road surface.

Fig. 18 presents the spectrum – the result of identification with the 100х50 NaI(Tl) crystal based detector on 2 meter height from the road level and at the distance of 1 meter from the side of the vehicle at the beginning of the trailer. At the upper side of the Figure the measured spectrum is shown. In the right side of the spectrum one can see the 1440 keV energy peak of the К-40 radionuclide. In the lower part of the Figure the identification results are presented showing the date and the time of measurements together with the type of the identified radionuclide and the preliminary evaluation of the radionuclides activity. In the given case one can see from the Figure that the identified radionuclide occurred to be К-40 isotope. On the Figure 15 one can also see that these sacks are placed along the axis of symmetry of the vehicle and, therefore, the distance between the detector and the sacks is more than two meters.

Further, the identification results were obtained at different measurement conditions. For example, Fig. 19 shows the spectrum measured at 2 meter height from the road level and at 0.5 meter distance from the lateral side of the vehicle at its beginning. As one can see from this Figure, the number of detected gamma-radiation increased, which is especially clear from the scattered radiation peak, showing again the К- 40 radionuclide.

14

Figure 19: A NaI(Tl) measurement at the front of the trailer of the vehicle shown in Figs 14–17. The distance to the trailer was 0.5 m horizontally, and the detector was

2 m above the road surface.

Fig. 20 presents the spectrum taken at 2 meter height from the road level and 1 meter distance from the lateral side of the vehicle at its middle part. As one can see from the Figure, the increase in number of the detected gamma-radiation is still observed. Unclear peaks of U-235, U-238, Pu-240 have appeared, and the К-40 radionuclide cannot be identified against their background.

Figure 20: Measurement at 1m horizontally from the middle of the trailer.

15

Fig. 21 shows the spectrum measured at the distance of 0.5 m from the lateral side of the vehicle at the height of 2 meters from the road level at its middle part. From the Figure, it is clear that the number of unclear peaks decreased to single Eu-152, and К-40 radionuclide is once again identified.

Figure 21: Measurement at 0.5m horizontally from the middle of the trailer.

Fig. 22 presents the spectrum measured at 1 meter distance from the lateral side of the vehicle and at 2 meter height from the road level at the rear part of the trailer. From the Figure, one can see that approaching detector does identify the К-40 radionuclide at the rear part of the vehicle, but it does identify the peaks corresponding to uranium and its chain.

Fig. 22: Measurement at 1m horizontally from the Rear of the trailer.

Thus, Fig. 23 presents the spectrum measured at 0.5 meter distance from the lateral side of the vehicle and at 2 meter height from the road level at the rear part of the

16

trailer. One can see that the U-238 radionuclide was identified. Based on the measurement results one can conclude that single point measurement identification does not give full information on all radionuclides in the examined object, especially if the measurements are conducted not on the activity peak.

Figure 23: Measurement at 0.5 m horizontally from the rear of the trailer.

As the spectrometer MCA-166 with 100x50 mm NaI(Tl) crystal was borrowed we could not use it in our further studies. It was important to compare the obtained results with those obtained by the available hand-held spectrometers. Thus, Fig. 24 shows the energy spectra of the cargo detained at the Customs Point Yallama on March 7, 2006, and measured by a RID (model Target identiFINDER, Version: E3/9V S/N: 04f3/2465).

Figure 24: Spectrum from RID: Field 4F2465_2006-03-07_11-07-55-012.SPC; Date: 07.03.2006 11:07:12.

17

The spectrum shown on Fig. 24 was measured directly on the surface of the package after opening of the vehicle at additional customs examination. By using the identiFINDER the whole set of cargo was examined and only one package (the last sack in the trailer) occurred to be with the radioactive contamination (U-235). The decision was made to send the samples of the cargo for further analysis. The analysis was conducted in the INP AS RU. The measurements were performed by means of a 80 cm3 High Purity Germanium detector based Canberra DSA 1000 semiconductor spectrometer. The measurement revealed the presence of uranium-235 and a number of its fission products.

Fig. 25 shows the spectrum measured by this detector and the identification results.

Figure 25: Results of measurement by the HP Ge system, located at the Institute of Nuclear Physics AS RU.

Fig. 26 shows the just the γ-spectrum of the sample, where traces of U-235 and Pa-234m can be clearly observed. When natural radioactivity measured the U-238

18

contents can be determined either by U-235 (185.7 keV line) with the following recalculation on U-238 or by 766.6 and 1001 keV lines which are emitted by Pa-234m – the third daughter radionuclide in the U-238 radioactive chain:

>⋅

>>>⋅

−−

yearsU

daysPa

daysTh

yearsU m

5234234234

9238

105.217.11.241047.4αββα

Pa-234m is almost in radioactive equilibrium with U-238 or it is accumulated relatively fast (in 2-3 months) up to the radioactive equilibrium, upon gaining which the activities of Pa-234m and U-238 become equal.

Figure 26: The measured gamma spectrum for the sample.

In addition to the gamma measurements, samples of the material were analysed by means of neutron-activation analysis. The samples were irradiated in the heat channel of the INP AS RU WWR-SM nuclear reactor for 16 hours, and, as a result, gave the average value of 0.9±0.06 wt. % of uranium. Obviously, such cargo could not be used for its direct purpose. It was found to be contraband and the investigation was started.

3.2. Customs check-point Navoi.

Fig. 27 shows the Yantar-type portal radiation monitors produced by the R&D Centre Aspect (Dubna, Moscow region) and installed at the Navoi customs point. Navoi differs from the Yallama post with its single portal pedestrian radiation monitor.

19

Figure 27: Customs Check Point Navoi.

In case of alarm from the portal radiation monitors the vehicle is identified and is detained by the customs officers based on the video-image displayed on the server computer (Fig. 28).

Figure 28 Figure 29

The personnel of the customs point called up the experts from the INP. Upon arrival the INP experts analysed the radioactivity distribution in the sample along the vehicle. According to the diagram on the Fig. 29, the activity is distributed evenly from the middle-part to the rear part of the trailer. The spectral composition of the detected source was conducted by the 100x50 mm NaI(Tl) crystal based detector as shown on Fig. 30.

20

Figure 30.

Fig. 31 shows the spectrum measured at 1 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the beginning of the trailer. One can see from the figure that the radionuclides are not identified, whereas, however, it is possible to say that it is the К-40 isotope by its peak.

Figure 31: Measurement at 1m horizontally from the front of the trailer shown in Figures 28–30.

21

Fig. 32 shows the spectrum measured at 0.5 meter distance from the lateral side of the vehicle and at 2 meter height from the road level at the beginning of the trailer. As one can see from the Figure the number of detected gamma increased, and the radionuclide К-40 was identified.

Figure 32: Measurement at 0.5 m horizontally from the front of the trailer.

Fig. 33 presents the spectrum measured at 1 meter distance from the lateral part of the vehicle and at 2 meter height from the road level at the middle of the trailer. As one can see from the Figure the К-40 radionuclide was identified.

Figure 33: Measurement at 1m horizontally from the middle of the trailer.

22

Fig. 34 shows the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter height from the road level at the middle of the trailer. As one can see from the Figure, here the К-40 radionuclide was also identified.

Figure 34: Measurement at 0.5 m horizontally from the front of the trailer.

Fig. 35 presents the spectrum measured at 1 meter distance from the lateral part of the vehicle and at 2 meter height from the road level at the rear part of the vehicle. Figure shows that the К-40 radionuclide can still be clearly identified.

Figure 35: Measurement at 1m horizontally from the rear of the trailer.

23

Fig. 36 shows the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and 2 meter distance from the road level at the end of the trailer. As one can see from the Figure the К-40 radionuclide can be still clearly identified.

Figure 36: Measurement at 0.5m horizontally from the front of the trailer.

At the additional customs examination the cargo was studied with the hand-held RID MKS-A02. The results it displayed were: • Nuclides were not found • Dose power on the lateral part surface was - 0,12-0,15 μSv/h, at the background - 0,10-0,12 μSv/h

The sensitivity of the MKS-A02 was lower than that of the Target produced identiFINDER and the identification was conducted without opening the vehicle, which clearly influenced by the identification result — “absence of the radionuclides”.

The declared cargo type was cutlery of colourless and dyed glass and crystal (country of origin: Russia). This is consistent with the measurements made.

3.3. Customs check-point Gisht-Kuprik

Fig. 37 shows the Yantar-type monitors produced by the R&D Centre Aspect (Dubna, Moscow region) installed at the Gisht-Kuprik customs check-point located on the border with Kazakhstan Republic.

24

Figure 37: Customs Check Point Gisht-Kuprik.

When the portal radiation monitors discover the radioactive cargo the vehicle caused the alarm is displayed on the server computer (Fig. 38) together with the distribution on the gamma channel on which the alarm was set off (Fig. 39).

Figure 38 Figure 39

The personnel of the custom point detained this vehicle. Based on the concurred procedure the personnel of the customs called up the INP experts for additional measurements. When the information from the Yantar radiation monitors was analysed the radioactive cargo distribution character was determined along the vehicle. It is obvious from the Fig. 39 that the radioactivity is localised in the middle part of the trailer and in the end of the trailer.

Further, the spectral composition of the detected ionising radiation source was measured. Since it was not possible to keep the vehicle detained for a long time, all measurements were conducted at minimum of required equipment, which is shown

25

on Fig. 40. The vehicle was measured according to the standard procedure described above.

Figure 40

Thus, Fig. 41 shows the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the end of the trailer. One can see that the К-40 radionuclide was identified.

Figure 41: Measurement at 0.5 m horizontal distance from the front of the trailer shown in Figures 38–40.

Fig. 42 shows the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the middle of the trailer. One can see from the figure that once again the К-40 radionuclide was identified.

26

Figure 42: Measurement at 0.5 m horizontal distance from the middle of the trailer.

Figure 43: Manifest of vehicle, consistent with measurements.

Fig. 43 presents the invoice containing the information about the transported cargo.

The declared contents of this cargo were: tiles (weight: 12200 kg), country of origin: China. Results of the analysis conducted by the hand-held MKS-A02

27

spectrometer: Dose on the surface of lateral part - 0,13-0,15 μSv/h, at the background - 0,9-0,11 μSv/h.

3.4. Customs check-point Bekabad Rail-road

Fig. 44 presents the Yantar-type portal radiation monitors produced by the R&D Centre Aspect (Dubna, Moscow region) installed at the Bekabad Railroad customs check-point. The monitors are installed beyond the station area at the point where rails come together to one line and are fenced with metal construction against acts of vandalism.

Figure 44: Customs check-point Bekabad.

The images displayed on the server computer when the portal radiation monitors detect radioactive cargo are presented on Figures 45 and 46.

Figure 45 Figure 46.

28

These figures are similar to those imaged at other posts, but the object caused this alarm in this case is train, very long, and the peaks on the gamma channel activity distribution diagram correspond to one particular carriage. As one can see from the Fig. 46, the indicator is located at one of the peaks, and the video-image on the Fig. 45 shows some part of the carriage, which therefore means that the next carriage also carries radioactivity presented by the next peak on gamma channel activity distribution. In this case these are the freight carriage with fertilisers, which is proved by the conducted measurements. Due to its special purpose the carriage is designed so that the loose material empty the carriage itself via the hatch at the bottom, therefore the carriage has cone form, and only one measurement at the centre of the carriage was possible limited with the requirements on approaching to carriages.

Fig. 47 shows the spectrum measured at 2.1 meter distance from the lateral side of the carriage and at 2 meter height from the rails level. As one can see from the figure the К-40 radionuclide was identified.

Figure 47: Measurement at 2.1 m horizontal distance from the carriage of the train shown in Fig. 45 and discussed in the text.

The radiation portal monitors were installed at the Bekabad rail-road recently, and there was no hand-held equipment at this customs point. Thus, no additional RID measurements were made.

29

3.5 Customs check-point Alat

Figures 48 and 49 present the Yantar-type radiation portal monitor produced by the R&D Centre Aspect (Dubna, Moscow region) installed at the Alat customs check-point.

Figure 48 Figure 49

Similarly to the Gisht-Kuprik the Alat has separately installed truck and automobile vehicles monitors. The pedestrian monitor is installed inside the checking building shown on Fig. 49.

When the portal radiation monitors detect the radioactive cargo the server computer displays the vehicle caused this alarm (Fig. 50) and the distribution on the channel on which the alarm as set off (Fig. 51).

Figure 50 Figure 51

30

The personnel of the customs point provided the standard vehicle detaining procedure and called up the INP experts. Upon arrival, the INP experts analysed the distribution of the radioactive cargo along the vehicle in accordance with the activity distribution diagram on the gamma channel (Fig. 51), with the following measurement of the spectral composition of the ionising radiation source by means of 100х50 NaI(Tl) crystal based detector.

Figures 52 and 53 show the truck caused this alarm and its cargo occurred to be the zinc powder filled sacks.

Figure 52 Figure 53

Fig. 54 shows the spectrum measured at 1 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the beginning of the trailer. As one can see from this Figure, the К-40 radionuclide was identified, however, peaks corresponding to other radionuclides were also observed.

Figure 54: Measurement at 1 m horizontal distance from the front of the trailer shown in Figures 50–53.

31

Fig. 55 presents the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the beginning of the trailer. One can see from this Figure that the К-40 radionuclide is still identified. But only Cs-137 left of all elements under suspicion, and the number of detected gamma increased.

Figure 55: Measurement at 0.5 m horizontal distance from the front of the trailer.

Fig. 56 shows the spectrum measured at 1 meter distance from the lateral part of the vehicle and at 2 meter height from the road level at the middle of the vehicle. As one can see from the Figure two radionuclides К-40 and Cs-137 were identified.

Figure 56: Measurement at 1 m horizontal distance from the middle of the trailer.

32

Fig. 57 shows the spectra measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the middle part of the trailer. As one can see from the Figure two radionuclides К-40 and Cs-137 were identified.

Figure 57: Measurement at 0.5 m horizontal distance from the front of the trailer.

Fig. 58 shows the spectra measured at 1 meter distance from the lateral part of the vehicle and at 2 meter distance from the road level at the end of the trailer. One can see from the Figure that the identified radionuclide occurred to be the К-40 radionuclide. This is supported by the intensity distribution on the gamma channel obtained from the Yantar monitor (Fig. 42), which demonstrates that the intensity becomes lower by the end of the trailer.

Figure 58: Measurement at 1 m horizontal distance from the rear of the trailer.

33

Fig. 59 shows the spectrum measured at 0.5 meter distance from the lateral part of the vehicle and at 2 meter height from the road level at the end of the trailer. As one can see from the figure the Cs-137 radionuclide was identified. This supports the fact that there is a contamination of the cargo transported at the end of the vehicle.

Figure 59: Measurement at 0.5 m horizontal distance from the rear of the trailer.

Quite high radiation intensity and the discovered Cs-137 radionuclide caused a detailed examination of the container. The cargo was additionally studied by the Target identiFINDER hand-held dosimeter, which is demonstrated on Fig. 60.

Figure 60

Fig. 61 shows the energy spectrum of the samples from the cargo detained at the customs point Alat on 17.02.06. It was measured by a RID (model Target identiFINDER, Version: E3/9V S/N: 04f3/2465).

34

Figure 61: Target identiFINDER spectrum for the contents in the vehicle referenced in Figures 50-60.

Later on, the radioactivity of every sack was measured, and the degree of Cs-137 contamination of their contents was evaluated. The cargo was detained, and an investigation is ongoing.

35

Summary

1. In the given report the results of activities on the research contract 12607/RO

are presented. 2. All performed works can be divided into three parts:

a) creation of the concept and apparatus for detection and determination capabilities of NORM-materials and radioactive isotopes transported in vehicle at its stand-still, and laboratory tests;

b) tests of the developed apparatus and isotope identification software on real object at the customs check-points of Uzbekistan;

c) determination of possibilities for development of new methods and apparatus for detection of nuclear fissile and radioactive materials, as well as the results of nuclear smuggling countering in Uzbekistan described in attached articles.

In the first part, the radiation monitoring of vehicles at the border at stand-still for detection of NORM-materials was developed, the prototype of the radiation monitor was designed with the concept realized, and the preliminary tests were performed on the real customs check-points of Uzbekistan. Advantages of the proposed control method comparing to the traditional approaches are: • Automatic determination of innocent alarms; • Automatic identification of nuclides in vehicle; • Display of the ionizing radiation source distribution diagram along the vehicle. Disadvantages of the proposed method: • Necessary organizational measures for vehicle control with the given method,

namely — in case of emergency the vehicle is forced to stop by the radiation monitor for 5 minutes to have measurement of the spectral composition of the ionizing radiation. On our opinion, it is almost impossible to analyse the cargo without forced stopping of vehicle;

• Use of additional indicating equipment (traffic-lights). In the second part, based on the conducted activities the practical possibility for identification of NORM-materials and decision making on the further destiny of the vehicles transporting this cargo are presented. The analysis of the obtained results presented in this report demonstrated that it is possible to use the 100х50 mm NaI(Tl) crystal based scintillation detector, with better identification at the distance between the detector and the vehicle is of 50 cm and less. The third part shows the possible ways and importance of the development of new advanced methods and apparatus for detection of nuclear fissile and other NORM-materials and their identification by using the methods described in these articles.

36

The authors wish to express their deepest gratitude and thanks to the staff of the IAEA, namely, Teresa Benson, Anita Nilssen, Rolf Arlt, Ken Baird, Pavel Dvornyak, Inna Cherkasskaya and others who paid continuous attention and supported this project determining its successful completion. The authors believe that this research on development of new methods for detection and identification of nuclear fissile and radioactive materials will continue.

Publications produced during this CRP

[1] V.D. Petrenko, Yu.N. Karimov, A.I. Podkovirin, N.N. Shipilov, B.S. Yuldashev, M.I. Fazylov Efforts of Uzbekistan to prevent terrorism and smuggling of radioactive and nuclear materials //Applied Radiation and Isotopes, Applied Radiation and Isotopes 63 (2005) 737–740.

[2] Yu.N. Karimov, N.N. Shipilov, M.I. Fazylov, V.D. Petrenko, B.S. Yuldashev. On the possibility of detecting fissile nuclear materials using coincident gamma rays //Applied Radiation and Isotopes 63 (2005) 655–658.

[3] N.N. Shipilov, M.I. Fazylov, A.I. Podkovirin, Yu.N. Karimov, V.D. Petrenko, B.S. Yuldashev Identification of radioactive materials in moving objects // Applied Radiation and Isotopes 63 (2005) 783–787.

[4] V.D. Petrenko, The experience of the State Customs Committee in use of the apparatus for the radiation monitoring of the vehicles and passengers in Uzbekistan // “International Workshop on Radiological Sciences and Applications”, Albuquerque, USA, 16–18 June 2003.

[5] Yu.N. Karimov, B.S. Yuldashev, V.D. Petrenko, N.N. Shipilov, M.I. Fazylov, Vehicle control methods witch automatic nuclides identification and innocent alarms determination // The third Eurasian conference “Nuclear science and its application”, 5–8 October 2004, Tashkent, Uzbekistan, p.144.

[6] N.N. Shipilov, B.S. Yuldashev, V.D. Petrenko, Yu.N. Karimov, M.I. Fazylov, A.I. Podkovirin Video surveillance as an indispensable part of radiation control // The third Eurasian conference “Nuclear science and its application”, 5–8 October 2004, Tashkent, Uzbekistan, p.144.

[7] V.D. Petrenko, B.S. Yuldashev, Yu.N. Karimov, N.N. Shipilov, M.I. Fazylov Radiation monitoring at customs border posts of Uzbekistan //The third Eurasian conference “Nuclear science and its application”, 5–8 October 2004, Tashkent, Uzbekistan, p.145.

[8] IAEA DRAFT TECDOC “Technical/Functional Specifications for Border Radiation Monitoring Equipment”, Rev. 20B (2004).

Ruggedized Detection Probe for Field Use with Large Volume Coplanar CdZnTe Detector

L. Grigorjeva, V. Gostilo, A. Loupilov, I. Lisjutin, R. Rjabchikov University of Latvia, Institute of Solid State Physics, Riga, Latvia

Abstract

The aim of the present Project was the development of ruggedized detection probe with large

volume coplanar CdZnTe detector on the basis of CdZnTe detector with maximum possible high

efficiency for field use.

Coplanar grid contacts applied for the detectors based on 15x15x10 mm3 crystals provided the

total absorption efficiency of registration about 8%, energy resolution less than 2% on the energy

of 662 keV and the peak/compton ratio better than 11. The best energy resolution of 10.5 keV

(1.59%) was achieved at HV=2 kV and the optimal voltage of 60 V between the grids. The energy

resolution of the detectors on energies 59.6 and 1332 keV was 7.9 and 14.5 keV, respectively.

A gamma-radiation Detection Probe (DP) based on a CdZnTe spectrometric coplanar detector was

developed for joint operation with standard MCA-166. DP has dimensions 165x60x50 mm and it

comprises: a CdZnTe coplanar detector, installed on Peltier cooler; a 2-channel differential charge

sensitive preamplifier with resistive feed back; thermostabilisation device of detector; HV filter;

HV power supply with voltage divider.

The test of the DP, supplied by power source of МСА-166, shows that all characteristics in the

requirements specifications of the technical assign are realised except the temperature stability.

To rise insufficient stability of DU caused by insufficient power of the supply source +12V/50

mА in МСА-166 analyser the additional power supply +14.5 V/100 mA with Li-Ion Rechargeable

Battery was applied what has provided the required stability.

Introduction The aim of the present Project is the development of ruggedized detection probe for field use with large volume coplanar CdZnTe detector on the basis of CdZnTe detector with maximum possible high efficiency.

In the first year of the development the basic effort was paid to the development of production technology of CdZnTe detectors with coplanar electrode structure. The development of the detectors of such electrodes structure reuquires significant improvement of photolithography technologic processes. That task was solved successfully what is allowed to fabricate detectors with large volume.

Investigations of crystal performance of different producers have also demanded for considerable efforts. Those performances were inspected carefully in manufacture process to adapt manufacture technology to the crystals properties.

Two coplanar grid detectors were fabricated with application of the technology developed for the fabrication of pixel and strip detectors. Interpixel resistivity on different grids was within 1–6 GOhm. Energy resolutions achieved are 8.2 keV on 59.9 keV (13.8%); 8.04 keV on 122 keV (6.6%); 11.4 keV on 662 keV (1.72%); 16.0 keV on 1332 keV (1.2%) correspondingly.

Evaluated registration efficiency of developed detector CZT2-4-2 was 18.75 mm2 or 8.3% for irradiation from the grids and 16.9 mm2 or 11.2% from the end face.

All results of that researches and performances of the developed detectors are presented in details in the chapter I Development of CdZnTe Coplanar Grid Strip Detectors.

Chapter II Design Development of Detection Probe shows the results of the development of the outer view and design of the device itself. First prototype was fabricated and delivered to IAEA for testing. In this probe we proposed the solution with thermo stabilization of detector which should provide high temperature stability of Detection Probe and expand the temperature working range. This solution was not proposed by somebody till now.

Chapter III Correction of Detection Probe Design shows those changes which were done in Detection Probe in accordance with results of testing of the first prototype: correction of electronics circuitry; integration of additional circuitry for generator and peak stabilizer and preamplifier gain tuning; optimization of Peltier cooler performance on results of testing; correction of technical documentation. After finishing of fabrication of the second prototype in October 2006 it will be delivered to IAEA for the testing in November 2006.

I. DEVELOPMENT OF CdZnTe COPLANAR GRID STRIP DETECTORS

The aim of this part of contract was the develop of CZT coplanar detectors with the largest possible efficiency. 1. Input Inspection of CZT crystals 1.1. Inspection of sizes We received three crystals grown by Yinnel Tech Inc. Their sizes are as follows (Tab. 1):

Table 1

THE DESCRIPTION OF YINNEL TECH.INC. CRYSTALS

№ Number of the crystal Shape Sizes, mm

1 CZT2-4-1 rectangular parallelepiped* 17. Х17. Х11. 2 CZT2-4-2 rectangular parallelepiped * 17. Х17. Х11.

3 CZT2-4-3 (side piece)

irregular cross-section: 11.1Х9.7

* there was a small deviation from flatness of sides (about 0.3 mm). 1.2. Visual inspection The crystals have polished surface (the polishing was presumably made with the powder M7-M14). The edges of the crystals were small chips (not more than 0.3 mm) All crystals have no visible structure violations on the surface (borders of blocks, twins, inclusions). 1.3. The preparation of the samples for the measurements All crystals were subjected to the standard procession: grinding, polishing, chemical polishing in bromine ethil etchant. On two opposite edges the gold electrodes were deposited with the deposition of gold from 5% solution of hydrochloroarurate acid. On the sample No.CZT2-4-3 the electrodes were deposited only on the central part of edges. The area of each electrodes was about 1 cm3. 1.4. The measurements of electrodes mobility - μe and the quantity (μτ)e The values (μτ)e and μe were measured by modified time of flight method with application of alpha-particles method. The average values of the measured quantities (μτ)e and μe are shown in the Table 2.

Table 2

AVERAGE VALUES OF THE MEASURED QUANTITIES (μτ)E and μE

№ Number of crystal Thickness of detector, mm μe, cm2/Vs (μτ)ex10-3, cm2/V 1 CZT2-4-1 11.6 740 6.8 2 CZT2-4-2 11 750 6.9 3 CZT2-4-3 (side piece) 11.1 710 5.5

There were no considerable changes in values of the measured quantities (μτ)e and μe at the change of the side of radiation. The registered alpha spectra had no false, double or rather distorted in shape peaks. That shows the homogeneity of the samples by area. To measure the value of the quantities (μτ) and μ for the holes was not possible due to their low value. 1.5. I-V characteristics measurement I-V characteristics were measured at the ambient temperature about +27 0С (Table 3). The time of exposition in the dark before the measurements was about 20 hours. The value of the current was fixed in 30 seconds after application of a new value of the voltage. After that application of the voltage the value mof the current is not a stable one. In the range of the applied voltages (up to 1000 V) the value of the current is decreased over the time. The stabilisation of the value of the current could appear in 15–60 minutes after the voltage is applied.

Table 3

I-V CHARACTERISTICS, MEASURED AT THE AMBIENT TEMPERATURE OF +270С

Number of crystal U, V I, nA

CZT2-4-1

100 200 300 400 500 600 800 1000

5 9 14 18 23 28 39 49

CZT2-4-2

100 200 300 400 500 600 800 1000

7 15 23 33 41 50 68 86

CZT2-4-3 (side piece)

100 200 300 400 500 600 800 1000

3 7 11 14 17 20 27 33

2 The change of the polarity of the detectors switch on does not result in the considerable change of I-V characteristics. All measured I-V characteristics in the applied range of the voltages have the nature close to the linear one. The specific resistance of the samples (3–6)*1010 Ohm*cm. 1.6. Measurement of spectra The energy resolution on line 59.6 keV of isotope Ам-241 for the various operating voltages are shown in the Table 4:

Table 4

THE ENERGY RESOLUTION ON LINE 59.6 KEV OF ISOTOPE АМ-241 FOR THE VARIOUS OPERATING VOLTAGES

Number of

crystals Operating voltage, V

Energy resolution, keV

CZT2-4-1

400 600 800 1000

8.6 5.5 4.5 4.2

CZT2-4-2

400 600 800 1000

8.8 7.2 6.9 6.7

The spectrum of the isotope Ам-241 registered by the sample № CZT2-4-1 at the operating voltage 1000 V is presented in Fig. 1. Fig. 2 shows the spectrum of isotope Cs-137 registered by the same sample at the same voltage. 1.7. After input testing next conclusions were made: - The examined samples have rather high homogeneity. - The examined samples have rather high value of the parameter (μτ)e = 6.8x10-3 cm2/V and low value of that parameter for the holes. - The samples have specific resistivity (3-6)*1010 Ohm*cm. Final conclusion is the samples are suitable for the fabrication of the detectors with coplanar lattice. 1. Layout and photo masks Layout of coplanar electrodes for the detectors was developed by Michigan University. We didn’t fined any technological limitations for fabrication of detectors with such layout. Two mirror photo masks with the field 15mmx15mm were fabricated by us for coplanar electrodes on two opposite sides of the detector. 2. Fabrication of coplanar electrodes After input testing solid electrodes were deleted from the crystals. The surface of crystals was processed again and fabrication of coplanar electrodes were done by the technology developed for pixel and strip detectors. But special technological processes were used on photolithography for passivasion of surface and to provide as higher as possible interstrip resistivity.

3Electrodes were done by deposition of gold (Fig. 4). To our regret the contacts of both crystals were made over several times due to the insufficient feed through I-V characteristics of the fabricated detectors. The search for the reasons has revealed that inner cavities of the crystal have appeared on the surface of that crystals in the result of all treatments (Fig. 5). A lot of crystal material has to be taken off to delete those cavities. And the second crystal has subjected to the treatments as well. As the result the final thickness of the crystal CZT2-4-1 was ~9 mm, and the crystal CZT2-4-2 ~ 9,5 mm. After detectors crystals fabrication the measurements of the interstrip resistance was made. After the thermal treatment of the crystals on air at the temperature Т = 120 °С within t = 120 minutes the measurements were repeated again. The results of the measurements of interstrip resistance of the detector crystals before and after annealing are shown in Table 5 and in Figs 6 and 7.

Table 5

INTERSTRIP RESISTANCE (in GOhms) BEFORE AND AFTER ANNEALING

Crystal CZT2-4-1 Crystal CZT2-4-2 before thermal

treatment after thermal

treatment before thermal

treatment after thermal

treatment U, В

R1 R2 R1T R2T R1 R2 R1T R2T

10 1,159 0,198 1,653 0,571 0,119 0,943 0,755 3,521 20 1,543 0,208 2,283 0,722 0,142 1,124 0,881 4,598 30 1,871 0,229 2,752 0,823 0,166 1,316 0,953 5,19 40 2,162 0,256 3,12 0,909 0,192 1,493 1,011 5,674 50 2,415 0,282 3,39 0,978 0,215 1,667 1,048 6,002 60 2,609 0,308 3,576 1,03 0,236 1,807 1,07 6,166 70 2,726 0,329 3,676 1,065 0,254 1,902 1,048 6,228

As the table shows after annealing the interstrip resistance on the different grids has increased from 1.5 to 5 times in comparison with initial value. The final interstrip resistance at the voltage 60 V on the various surfaces of the fabricated detectors was from 1 to 6 GOhm. The difference of the interstrip resistances on the different surfaces of the detector could be explained most obvious by the different stehiometry of CdZnTe on those surfaces. Those difference could be different due to the growth reasons as well as due to the distinctions appeared in procession of those surfaces at chemical polishing. It was noticed at the time of the fabrication process that not only interstrip resistances value on the various surfaces have been changed but the ratio between resistance on the surfaces to the opposite one. However the detailed study of that processes was not the aim of this project. 3. Design and Assembling After testing the crystals were assembled in the case. With the account of the research aspect of the present project we have developed the design (Fig. 8), that allows the multiple assembly and disassembly of the crystal and its duplicate application in the further projects. The crystals were placed on the dielectric substrate and wire bonded by gold wire to the contact areas at the substrate (Fig. 9). Dielectric substrate with the crystal was placed in commercial aluminum case with thickness of the walls of 1 mm . The sealing of the detectors case was made also . The wire bonding of the connector is shown at the Fig. 8. 4. The measurement of the detector CZT2-4-1 characteristics After the assembly the interstrip resistances were remeasured again. After that the I-V characteristics were measured. All measurements were made at the ambient temperature +230C.

4For the testing of the spectrometry characteristics the simplest preamplifier was made in the accordance to the recommendation of [P. N. Luke, “Unipolar Charge Sensing with Coplanar Electrodes - Application to Semiconductor Detectors”, IEEE Trans. Nucl. Sci., vol. 42, no. 4, pp. 207–213, Aug. 1995.] We have not executed any special actions to provide the least noise level of preamplifier. The noise level and energy resolution were defined mostly by the adjustment (balancing) of the preamplifier. The amplification and shaping of the signals of preamplifier were made with the standard spectrometry device MULTISPECTRUM manufactured by Baltic Scientific Instruments (BSI), www.bsi.lv . 5.1 The measurement of the detector CZT2-4-1 I-V parameters 5.1.1. The resistance between grids at the voltage 60 V is :

- the side А (blue) - 1.7 GOhm - the side B (red) - 1.03 GOhm.

5.1.2. I-V characteristics between grids on the sides А and В are shown in Table 6 and Fig. 11.

Table 6

I-V CHARACTERISTICS BETWEEN GRIDS ON THE SIDES А AND В U, V 10 20 30 40 50 60 I,nA (grids A)

8 16 23 27 31 36

I,nA (grids B)

10 21 32 41 48 58

5.1.3. I-V characteristics of the detector (table 7) are shown for the connection of the outputs listed below:

- grids B and guard ring B are connected together (cathode) - guard ring A is connected to GND - grid A (anode)

Table 7

I-V CHARACTERISTICS OF THE DETECTOR

U, V -100 -200 -400 -600 -800 -1000 -1200 -1400 -1600 -1800 -2000I,nA 0.4 0.7 1.2 1.7 2.2 2.6 3.1 3.6 4.1 4.5 5.0

U – the voltage supplied to the cathode. I-V characteristics is not changed at the connection of grids A-1 or grids A-2 or connecting them together.

5.1.4. I-V characteristics of the detector (Table 8) are shown for the connection of the outputs as follows: - grids B and guard ring B are connected together (cathode) - guard ring A is not connected to GND - grid A (anode)

Table 8


U, V -100 -200 -400 -600 -800 -1000 -1200 -1400 -1600 -1800 -2000 I,nA 0.9 1.6 2.7 3.7 4.9 5.9 7.0 8.1 9.2 10.4 11.7

5 U – the voltage supplied to the cathode. The change of the polarity of the detector switch on does not provide the essential change in I-V characteristics. All measured I-V characteristics have the view close to linear. The switch of the guard ring at the side of the signal grids results in the increase of the detector current in 2 times approximately. 5.2. The measurement of the spectrometry characteristics of the detector CZT2-4-1. 5.2.1. The measurement of the spectra were made to connect the outlets in accordance to p.5.1.3, additionally – grids A are switched on differentially [P. N. Luke, “Unipolar Charge Sensing with Coplanar Electrodes - Application to Semiconductor Detectors”, IEEE Trans. Nucl. Sci., vol. 42, no. 4, pp. 207–213, Aug. 1995]. The measurements were made at the voltages on the cathode within the range of 500–1500 V and on the anode of 10–60 V. The optimal parameters of the measurements were as follows: the optimal voltage -1000 V was supplied to cathode, +40 V to anode, shaping time constant was 3 μs. 5.2.2. The measurement of spectrometry characteristics requires a sharp selection of amplification in the second measurement channel at the summation of the signals, the adjustment is made by energy resolution of gamma radiation source. The energy resolution depends mainly on the correct adjustment of the summary signal. 5.2.3. The proper energy resolution of the preamplifier is 3.0 keV at the nominal of the load resistors of 100 MOhm, and 2.1 keV at 500 MOhm. 5.2.4. When detector was exposed with light over a certain time and then connected to preamplifier, when HV negative polarity was supplied to cathode and positive polarity to anode, there appeared large excess current noise which could be explained by the residual charge after exposition of the detector. That effect disappears when all grids together with HV filters are short-circuit on the case of the preamplifier. 5.2.5. The typical spectra of the various gamma-radiation sources for the detector CZT2-4-1 are shown in Figs 13–17. 6. The measurement of the detector CZT2-4-2 parameters 6.1. The measurement of the detector CZT2-4-2 I-V parameters

6.1.1. The resistance between grids at the voltage 60 V is:

- the side А (blue) – 5.0 GOhm - the side B (red) – 1.28 GOhm

6.1.2. I-V characteristics between grids on sides А and В are shown in Table 9 and Fig. 18.

Table 9

I-V CHARACTERISTICS BETWEEN GRIDS ON SIDES А AND В

U, V 10 20 30 40 50 60 I,nA (grids A)

2.1 4.5 6.8 8.4 10.1 12.0

I,nA (grids B)

7.0 16.8 24.9 32.4 39.1 46.8

6 6.1.3. I-V characteristics of the detector (Table 10) are shown for the outlets connection as follows:

- grids B and guard ring B are connected together (cathode) - guard ring A is connected to GND - grid A (anode)

Table 10


U, V -100 -200 -400 -600 -800 -1000 -1200 -1400 -1600 -1800 -2000I,nA 0.4 0.7 1.2 1.7 2.2 2.7 3.1 3.6 4.1 4.5 5.0

U – the voltage supplied to the cathode. At the connection of grids A-1 or grids A-2, or connecting them together, the I-V characteristics is not changed. 6.1.4. I-V characteristics of the detector (Table 11) are shown for the following outlets connection: - grids B and guard ring B are connected together (cathode) - guard ring A is not connected to GND - grid A (anode)

Table 11


U, V -100 -200 -400 -600 -800 -1000 -1200 -1400 -1600 -1800 -2000I,nA 1.0 1.7 2.9 4.1 5.3 6.6 7.8 9.0 10.2 11.4 12.7

U – the voltage supplied to the cathode. The change of the polarity of the detector switch on does not provide the essential change in I-V characteristic. All measured I-V characteristics have the view close to linear. The switch of the guard ring at the side of the signal grids results in the increase of the detector current 2–2.5 times. 6.2. The measurement of the spectrometry characteristics of the detector CZT2-4-2. 6.2.1. The measurements of the spectra were made to connect the outlets in accordance to p.2.2.2 , additionally – grids A are switched on differentially [P. N. Luke, “Unipolar Charge Sensing with Coplanar Electrodes - Application to Semiconductor Detectors”, IEEE Trans. Nucl. Sci., vol. 42, no. 4, pp. 207–213, Aug. 1995.]. The measurements were made at the voltages on the cathode within the range of 500–1500 V and on the anode of 10–60 V. The optimal parameters of the measurements were as follows: Optimal parameters of the measurements were as follows: the optimal voltage -1000 V was supplied to cathode, +40 V to anode, shaping time constant was 3 µs. 6.2.2. The typical spectra of the various gamma-radiation sources are shown in Figs 18–24. 6.2.3. We have found that energy resolution of detector is better if to irradiate it from the end face of crystal but not from the grids face. For example Figs 22 and 24 show such spectra for Cs-37 correspondingly. Instead of energy resolution 12,0 keV detector demonstrates resolution 11,4 keV. Probably it’s due to more uniform conditions for charge collection in this case, but investigation of this effect was not the aim of this project. 6.2.4. We have evaluated also registration efficiency of the detector CZT2-4-2 for Cs-137. It was 18.75 mm2 or 8.3% for irradiation from the grids face and 16.9 mm2 or 11.2% from the end face.

77. Chapter I Conclusions.

1. Two coplanar grid detectors were fabricated with application of the technology developed by BSI for the fabrication of pixel and strip detectors.

2. Interpixel resistivity on different grids was within 1- 6 GOhm. 3. Energy resolutions achieved are 8.2 keV on 59.9 keV (13.8%); 8.04 keV on 122 keV

(6.6%); 11,4 keV on 662 keV (1.72%); 16.0 keV on 1332 keV (1.2%) correspondingly. 4. Evaluated registration efficiency of developed detector CZT2-4-2 was 18.75 mm2 or 8.3%

for irradiation from the grids and 16.9 mm2 or 11.2% from the end face.

8

0 100 200 300 400 5000

500

1000

1500

2000

2500

3000

coun

ts

channels

Am-241

Fig. 1. Spectrum of Am-241, shaping time = 1μs, U = 1000V, Toper. = -27°C.

0 100 200 300 400 5000

1000

2000

3000

4000

5000

6000

7000

8000

coun

ts

channels

Cs-137

Fig. 2. Spectrum of Cs-137, shaping time = 1 μs, U = 1000V, Toper. = -27°C.

9

Fig.3 Layout of coplanar grid structure.

Fig.4. Coplanar grid strip detector CZT2-4-2 (15x15x9.5 mm3) based on Yinnel Tech Inc. crystal.

10

a)

b)

11

c)

d)

Fig. 5. a), b),c), d) defects of coplanar strip grid detector.

12

e)

Fig. 5e. 1 mm scale for evaluation of defects sizes on CdZnTe crystal.

13

Fig. 6. The results of the measurement of interstrip resistance of detector CZT2-4-1.

14

Fig. 7. The measurement results of interstrip resistance of the detector CZT2-4-2.

15

Fig. 8. Construction of the detector assembly.

16

Fig. 9. Assembling of the detector crystal inside the case.

17

Fig. 10. Coplanar grid strip detector in case.

18

0 10 20 30 40 50 600

5

10

15

20

25

30

35

40

45

50

55

60

65 grids A grids BI, nA

U, V

Fig. 11. I-V characteristics between grids on the sides A and В (CZT2-4-1).

0 500 1000 1500 20000

2

4

6

8

10

12 guard ring GND- off guard ring GND-onI, nA

U, V

Fig. 12. I-V characteristics of the detector CZT2-4-1.

19

Fig. 13. (CZT2-4-1). The spectrum of Am-241( Res 59.6 keV = 8.4 keV).

Fig. 14. (CZT2-4-1) The spectrum of Со-57 (Res 122keV = 8.04keV).

20

Fig. 15. (CZT2-4-1). The spectrum of Cs-137 (Res 662keV = 13.2 keV).

Fig. 16. (CZT2-4-1) The spectrum of Cs-137 and Test (Res. 662keV = 13.8 keV, Res test = 8.2keV).

21

Fig. 17. (CZT2-4-1) The spectrum Co-60 (Res. 1332keV = 19.9 keV).

22

0 10 20 30 40 50 60

0

10

20

30

40

50

grids A grids B

I, nA

U,V

Fig. 18. I-V characteristics between grids on sides A and В (CZT2-4-2).

0 500 1000 1500 20000123456789

1011121314

guard ring GND-on guard ring GND-off

I,nA

U, V

Fig. 19. I-V characteristics of the detector (CZT2-4-2).

23

Fig. 20. (CZT2-4-2) The spectrum of Am-241 (Res. 59.6 keV = 8.2 keV).

Fig. 21. (CZT2-4-2) The spectrum of Co-57 (Res122keV – 7.9 keV).

24

Fig. 22. (CZT2-4-2) Spectrum of Cs-137 (Res. 662 keV=12,0 keV( 1.8 %), p/c =11.5, p/valley = 103, sens. = 18.75 mm2 (8.3%)

at exposition from detector’s B side.

Fig.23. (CZT2-4-2) Spectrum of Co-60 (Res. 1332 keV = 16.0 keV).

25

Fig. 24. (CZT2-4-2) Spectrum of Cs-137 (Res 662keV = 11,4 keV (1.7%), p/c = 10.9, p/valley = 78, sens. = 16.9 mm2 (11.2%)).

26

II. Design Development of Detection Probe Gamma-radiation Detection Probe (further DP) based on a CdZnTe spectrometry coplanar detector was developed for joint work with standard MCA-166 MultiChannel Analyser. A photo of the first prototype of the developed DP with closed and opened case is presented on the Figs 25 and 26. The DP is comprised of : - a detector based on CdZnTe coplanar detector, which is installed on a thermoelectric cooler (TEC) to provide thermostabilization, - a 2-channel differential charge sensitive preamplifier (PA) with resistive feed back; - a HV filter - HV power supply with voltage divider

1. Structure chart A structure chart of the Detection Probe is shown in Fig. 27. 2. Detector unit 2.1 The detector unit is a sealed chamber containing the TEC with heat exchanger, two 6-pins connectors, the detector and a temperature sensor PT-100 on a substrate, both glued to the upper TEC stage (possible with input PA stages) . Approximate sizes of TEC are 30х30х8 mm and the detector is- 15х15х10 mm. There are minimal distances (≤ 2.5 mm) between inner surface of the cover and the installed elements. 2.2 The approximate sizes of the cylindrical detector unit are diameter of 45 mm and height 35 mm. 2.3 The fastening of the PA flange is a rectangular under standard Al profile. 3. PA unit 3.1. The PA unit is comprised of: - two channel charge sensitive preamplifiers with summator and HV filter with HV divider. - detector temperature stabilisation device. Overall dimensions of printed board – 75 х 55 х 20 mm. 3.2. Detector temperature stabilization device provides maintain of set up detector temperature by TEC with maximum current 50 mA and voltage ± 12 V. Temperature measurement is made with temperature sensor РТ-100. 3.3. Ambient temperature could change within 0 ÷ + 400С, and average value of detector temperature was +20 (±0.5)0С. 3.4. TEC operates in cooling and heating mode. 3.5. Possible temperature setting on the detector in the range of +10 to +200C. Developed DP is satisfied to specification presented in Table 12.

27

Table 12 TECHNICAL CHARACTERISTICS of DETECTION PROBE

N Parameters Value Measurement results Remarks 1 2 3 4 5 1. Measuring gamma-radiation range,

keV 50 - 1500 50 - 1500 ок

2. Maximum loading on energy 662 keV (Cs-137), cps on shaping time constant 1μs

≥ 50000

50000

ок

3. Operating mode time setting, hour ≤ 0,5 0,5 ок 4. Continuous operation time, hour ≥ 12 12 ок 5. Energy resolution on energy

662 keV (Cs-137), keV (at optimal shaping time 3 μs and temperature + 22 0C)

≤ 16.5 (2.5%)

16.5 (2.5%)

ок

6. Geometric sizes of the detector, mm 15x15x10 12.45x12.45x7.5 ок? 7. Peak/Compton ratio on energy

662 keV (Cs-137) ≥7 6.5 ок?

8. Detector supply voltage of negative polarity , V

≤ 2000 1800 ок

9. Output signal polarity negative negative ok 10. Registration sensitivity on energy

662 keV (Cs-137), mm2

≥15

9.5

ок?

11. Integral nonlinearity should be, % ≤0,1 0.1 ок 12. Overall dimensions, mm ≤200x100x80 165x60x50 ок 13. In operating temperatures range

0 ÷ +40 0С| - conversion coefficient temperature instability , %/0С - energy resolution temperature instability, %/0С

≤ 0,01

≤ 0. 5

0.02

1.4

? ?

14. Preamplifier supply voltage, V/mA

+12/60 -12/60 +24/30 -24/30

+12/50 -12/50 +24/33 -24/19

ок

15. Shaping time constants MCA-166, μs

1, 2 1,2 ок

16. MCA-166 should have additional supply buses ± 5V on supply connector Preampl. Power Supply

+5 V/100мА-5 V/100мА

284. Chapter II Conclusions: The analysis of the results presented in Table 12 shows that all characteristics in the requirements specifications of the technical assign are realised except Technical Specification p.13 for temperature stability. Results mismatch in specification p.7 and p.10 is provided by the geometry dimensions of the detector, (Spec. p.6). If the detector with the dimensions specified in specification is installed then Numbers 7 and 10 are executed automatically. Poor temperature stability of DU (Spec. p.13) is caused by insufficient power of the supply source + 12V/50 mА in analyser МСА-166. To provide more high stability for DU the additional power supply + 14.5 V/100 mA is required. Additional power supply with Li-Ion Rechargeable Battery could be used for that. That is why the debugging of the detection probe in the development of the additional supply unit is required.

Fig. 25. Detection probe with closed case.

29 Fig. 26. Detection probe with opened case.

30

Fig. 27. Structure chart of Detection Probe.

31 III. Correction of Detection Probe Design

As it was shown in the conclusion of Chapter II, the development of the additional power supply is required to provide the necessary temperature stability of the detection probe. The development should comprise of :

• The development of the circuitry of the power supply • The correction of the detection probe design • The correction of the documents • The manufacture of the sample • The testing

The development of Li-Ion Rechargeable Battery with commutation and control diagram and Universal “smart Charge”r was made and Structure chart of Detection Probe is shown in Fig. 28. The housing design of Detection Probe was debugged and additional housing of the developed power supply was added to it. A new design is presented in Fig. 29. Conclusion of Chapter 3 As the result of the second year of the project the correction of the technical documents was made for the ruggedised probe for design part as well as for the electrical one. In accordance to the corrected documents the components were purchased and the manufacture of the second prototype of the detector was started. The manufacture will be over at the end of the October and after the testing in November 2006 the prototype will be supplied for the studies in IAEA.

32

Fig. 28. Structure chart of Detection Probe.

33

Alexander Loupilov Chief Spectrometrist Riga September 4, 2006

Fig. 29. Overall view of Detection Probe

Investigation of Different Scenarios that can be Used to Mask Nuclear Material with Other Gamma Emitters

M.I. Reinhard, D. Prokopovich, D. Alexiev, D. Hill Australian Nuclear Science and Technology Organisation (ANSTO)

N. Dytlewski International Atomic Energy Agency (IAEA)

Abstract Efforts to interdict trafficking in illicit nuclear and other radioactive materials take place at border control

points. Here, following the alarming of radiation portal monitors, border control officers equipped with hand

held isotopic analyzers examine suspect materials or packages. In order to not unnecessarily impede or

disrupt the flow of traffic through the border control point the outcome of inspections is required promptly

using a few minutes of measurement time. This task can present a significant challenge to hand held isotopic

analyzers due to the limitations in the radiation detection technology employed, and the limited competency

of officers in gamma-ray spectrometry.

The Australian Nuclear Science and Technology Organisation completed three programs of work to

experimentally investigate different scenarios that could potentially be used to mask the presence of illicit

nuclear materials. The first two programmes were concerned with the masking of HEU and plutonium. Two

separate reports were filed previously with the IAEA CRP. The third programme, described here,

investigated the comparative performance of the isotope analyser algorithms incorporated in a portable

HPGe detector from ORTEC with those of the IDENTIFY software of GBS.

INTRODUCTION

Efforts to interdict trafficking in illicit nuclear and other radioactive materials take place at border control points. Here, following the alarming of radiation portal monitors, border control officers equipped with hand held isotopic analyzers examine suspect materials or packages. In order to not unnecessarily impede or disrupt the flow of traffic through the border control point the outcome of inspections is required promptly using a few minutes of measurement time. This task can present a significant challenge to hand held isotopic analyzers due to the limitations in the radiation detection technology employed, and the limited competency of officers in gamma-ray spectrometry.

Of concern is the potential use of legal shipments of radioactive materials, such as radiopharmaceuticals, to mask the presence of illicit nuclear materials. The philosophy behind the masking of illicit nuclear materials is to prevent, or confuse the hand-held isotope analysers from obtaining the positive signatures that they require to unambiguously identify the radioisotopes of concern. The algorithms that isotope analysers use are proprietary, so one has to forward guess their detect-and-identify strategies. Knowledge of the underlying deficiencies of radiation detection instruments and inherent difficulties of gamma ray spectrometry can be exploited to circumvent the detection of nuclear and other radioactive materials. This same information might also be used to improve the performance of detector systems or to formulate better strategies to counteract attempts to mask the presence of illicit materials. The Australian Nuclear Science and Technology Organisation completed three programs of work to experimentally investigate different scenarios that could potentially be used to mask the presence of illicit nuclear materials. The first two programmes were concerned with the masking of HEU and plutonium. Two separate reports were filed previously with the IAEA CRP [1, 2]. The third programme investigated the comparative performance of the isotope analyser algorithms incorporated in a portable HPGe detector from ORTEC with those of the IDENTIFY software of GBS is described here.

1. ORTEC “The Detective” versus GBS “IDENTIFY”

1.1. Introduction/Background The experimental programme carried out and described in our previous reports definitively demonstrated the advantage owed to radiation detectors with good energy resolution at uncovering the presence of masked nuclear materials. In this respect HPGe is far superior to NaI or CZT. The need to operate HPGe detectors at cryogenic temperatures (<100K) has precluded the wide spread adoption of this type of detector in handheld isotope analysers for border security applications. This is due to difficulties associated with accessing a reliable supply of liquid nitrogen in the field as well as technical issues associated with the incorporation of cryogenic components into hand held instrumentation. Quite recently however, ORTEC, suppliers of radiation detection and measurement systems, introduced to the market a portable HPGe based isotope analyser. The instrument utilises a Stirling cycle mechanical system to cool the front end detector components thereby completely eliminating the need for liquid nitrogen. The instrument weighs approximately 8 kg making it portable albeit with more physical effort than that required for existing isotope analysers based on NaI or CZT. The overall performance of an isotope analyser is not just governed by the front end detector technology bit also on the algorithms used to identify peaks in the measured gamma-ray spectrum and correctly attribute them to the appropriate isotope. An experimental programme was

undertaken to compare the performance of the software algorithms incorporated into ORTEC’s “The Detective” with those incorporated into GBS’s IDENTIFY software package at detecting nuclear material when shielded or combined with other gamma ray emitters.

1.2. Experimental Methodology A variety of test scenarios were formulated which involved the combination of nuclear material with other gamma ray emitters and a combination of different shielding material configurations. All scenarios consisted of a linear placement of nuclear material, shielding material, other isotope and “The Detective”. The HEU and plutonium materials described in previous reports were utilised. Shielding materials included lead plate (2mm to 1 cm thicknesses), steel plates (1 mm to 1 cm thicknesses), cadmium plate (2 mm thickness), combinations of the afore mentioned materials or no shielding material. The isotopes used included 99mTc, 60Co, 137Tc, 131I, 133Ba, 152Eu, 226Ra and combinations thereof. The quantities of isotopes and distance of separation between the source and “The Detective” were varied in order to test the detection performance of the algorithms over a broad range of degrees of difficulty from easy to highly challenging. In some scenarios the object of the exercise was to push the technology to the point of failure as evident by the production of a false negative result. A total of 135 different scenarios were tested; 74 involving HEU, 54 involving Pu and 7 without any nuclear materials present to test for false positives. All gamma ray spectra were measured using “The Detective” instrument. For each scenario two separate measurements were made using acquisition times of 1 minute and 10 minutes. The short measurement times were necessary to test algorithm performance with low statistical data likely to be encountered in actual border security applications. During acquisition “The Detective” provides a running analysis of the identified isotopes which are displayed as a list on the onboard LCD display. The isotopes identified by “The Detective” during acquisition were recorded manually by the testing officer. Following the spectral acquisition using the “The Detective” the acquired data was downloaded from the instrument and uploaded into the GBS “IDENTIFY” software programme. Spectral analysis was repeated using “IDENTIFY” and the identified isotopes recorded. The results from both algorithms were then compared in terms of the rates of false positives and false negatives.

1.3. Results

1.3.1. HEU The numbers of false negatives and false positives from the 74 HEU scenarios is given in Table 5.

Table 5: False positives and false negatives in masking scenarios involving HEU.

Technology False Negatives False Positives “The Detective” 31 0

IDENTIFY 11 0 The number of false negatives produced by “The Detective” was found to be approximately three times higher than that produced by “IDENTIFY”.

No false positives were observed by either technology.

1.3.2. Plutonium The numbers of false negatives and false positives from the 54 plutonium scenarios is given in Table 6.

Table 6: False positives and false negatives in masking scenarios involving plutonium.

Technology False Negatives False Positives “The Detective” 9 0

IDENTIFY 6 0 The number of false negatives produced by “The Detective” was found to be approximately three times higher than that produced by “IDENTIFY”. No false positives were observed by either technology during the course of this study.

1.4. Discussion It is important to note as a prelude to further discussion that the observed high rates of false negatives were somewhat expected as the technology was intentionally pushed to failure using quite challenging masking scenarios. The experience of the team with isotope analysers containing NaI or CZT type detectors demonstrated that the challenging scenarios used to this section all exceeded the performance limitations of such moderate energy resolution technology. For this reason the observed results were not in any way indicative of questionable performance of a HPGe front end but rather a direct result of the experimental methodology employed. The higher rate of false negatives produced by “The Detective” in comparison to that of “IDENTIFY” in scenarios involving both HEU and plutonium was indicative of superior performance of the algorithms contained in “IDENTIFY” at detecting masked nuclear material. With all spectra acquired using “The Detective” a small advantage may have been afforded to the algorithms of this instrument over the algorithms of “IDENTIFY”. This is due opportunity afforded to the manufactures of “The Detective” to optimise the parameters of the peak detection algorithms to the known characteristics of the front end HPGe detector on board the instrument. In the case of “IDENTIFY” performance parameters of the HPGe detector on board “The Detective”, most notably the energy resolution as a function of gamma ray energy, had been experimentally measured and entered into “IDENTIFY” as part of the prior calibration. Poor setting of detector performance parameters into the peak detection algorithms was observed by the testing team to degrade the peak detection performance and therefore isotope identification. All efforts were made to ensure that the detector performance parameters of “The Detective” were accurately measured and recorded in the “IDENTIFY’ software settings. Knowledge of the settings used within “The Detective” could not be obtained. As such a definitive advantage of “The Detective” over “IDENTIFY” cannot be claimed. The greater rate of false negatives produced by both algorithms in the case of the scenarios involving HEU as opposed to plutonium is suggestive of the greater difficulty required to unmask HEU using a HPGe front end. The lower energy of the HEU emissions combined with the low emission intensities makes detection of HEU particularly problematic. In the case of plutonium the indicative emissions are of higher energy thereby less likely to be attenuated through any shielding material placed laterally to the nuclear material. This result would not be expected for lower energy

resolution front end detectors due to the inability of all but HPGe to resolve the closely spaced gamma ray lines of 239Pu. The lack of any false positives from either algorithm was not unexpected. This was due to the simple fact that nuclear material was present in approximately 95 % of the scenarios thereby reducing the opportunity for a false positive to arise. It could be postulated that due to the extremely good energy resolution of HPGe and the lack of any gamma ray emissions from the masking isotopes with energies within a FWHM of those of either HEU or plutonium that false positives are not likely to be a significant problem for “The Detective”. In all other experimentation and experience with this instrument during the trial period no false positives were recorded. It is difficult to speculate further on the reasons behind the superior performance of “IDENTIFY” over that of “The Detective” with respect to false negatives. As part of the analysis process the “IDENTIFY” software displays on the measured spectrum each of the identified peaks and to which isotope they are attributed. The positions in the spectrum of missing peaks associated with isotopes detected on the basis of other peaks are shown. This avails the user with the opportunity to form expert judgement as to presence of low statistical peaks not identified by the peak detection algorithms. Such information is useful in assisting an understanding of how well the algorithms work. In the case of “The Detective” the results of the analysis are not presented to the user apart from the final isotope list. This is not meant to infer a criticism of “The Detective” but rather an indication of why a more sophisticated analysis of the comparative performance results cannot be undertaken. The limited results displayed by “The Detective” is probably instructive in making this instrument simpler for customs officers not trained in gamma ray spectroscopy easier to use.

1.5. Conclusion In summary “The Detective”, with front end detector technology based on HPGe, affords the user with a superior capability to unmask the presence of illicit nuclear material than that available in other isotope analysers which make use of NaI or CZT. At eight kilograms the extent to which the instrument is portable is the subjective opinion of the user. What is indisputable is that the instrument is mobile and not limited to localities serviced with a regular supply of liquid nitrogen. In terms of the rate of false negatives the algorithms on board “The Detective” were inferior to those of “IDENTIFY” at uncovering masked nuclear material. Improvements to the algorithms used in “The Detective” may be an important upgrade opportunity for ORTEC.

REFERENCES [1] Dytlewski, N., Reinhard, M.I. and Alexiev, D., “Report on IAEA Co-ordinated Research

Project: Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and other Radioactive Materials, Task 6: Investigate different scenarios that can be used to mask nuclear material with other gamma emitters, PART I: HEU,” ANSTO, February 2003 [Safeguards-in-Confidence].

[2] Reinhard, M.I., Alexiev, D., Dytlewski, N. and Thomson, S., “Report on IAEA Co-ordinated

Research Project: Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and other Radioactive Materials, Task 6: Investigate different scenarios that can be used to mask nuclear material with other gamma emitters, PART II: Plutonium,” ANSTO, March 2004 [Safeguards-in-Confidence].

[3] Sampson, T.E., “Plutonium Isotopic Composition by Gamma-Ray Spectroscopy: A Review,”

LA-10750-MS, UC-10, September 1986. [4] Dytlewski, N. and Ensslin, N., “An Isotopic Analysis of ANSTO’s Plutonium Inventory by

Gamma Ray Spectroscopy,” AP/TN. 218, October 1988.

ACKNOWLEDGEMENTS

The ANSTO authors would like to thank ORTEC and Nucletron for the generous loan of “The Detective” for use as part of this programme.

Reduction of the Frequency of Innocent Alarms in Border Monitoring

H. Böck, S. Hengster, M. Schwarz,

M. Schrenk, M. Swoboda Vienna Institute of Technology Atominstitute

Abstract

Under the Research Agreement no 12585 four major tasks were carried out at the Vienna Institute of Technology-Atominstitute by four students. In particular the tasks were

A. M. Swoboda: “Dose rate and Isotope Identification Algorithm for CZT detectors for a possible use in SPRDs

B. M. Schrenk: “A Real Time, Isotope Identifying Gamma Spectrometer for Monitoring of Pedestrians by Michael Schrenk

C. S. Hengster: “Comparison of Simulation and Measurement of Radioactive Sources by the Usage of SuperSynth

D. V. Schwarz: “RID Hardware and ID-Software tests to compare different RIDs and ID methods by Vinzenz Schwarz”

The first three projects could be finished within the contract period while the last project is still ongoing. These individual reports are described below.

A: Dose rate and Identification Algorithm for CZT detectors for a possible use in SPRDs by Martha Swoboda A.1. Introduction A new instrument group is going to be established called SPRDs (Spectral Radiation Detectors). These instruments are hybrids between PRDs (Personal Radiation Detectors) and RIDs (Hand-held Radioisotope Identification Devices). SPRDs will be developed in the first line to allow the user to distinguish quickly in case of a radiation alarm between medical isotopes and other isotopes, because most of the radiation alarms for pedestrians are triggered by medical isotopes and a RID is not quickly available. Nevertheless the algorithm is a full isotope identification algorithm for a certain number of unshielded isotopes. This is needed because a nuclear or an industrial isotope shall never be indicated as a medical isotope. Also a mixture of a medical isotope with another isotope should be correctly recognized. In order to avoid false positive results a full algorithm is needed. Whether detailed isotope identification results and/or an isotope categorization only are indicated by the instrument is a question of the end-user requirements. This type of instruments will be very small (like PRDs or even a wrist watch) and can therefore easy be used for uncovered or covered operations. The energy compensated dose rate indication is then also the basis for safety functions similar to the PRDs. This new instrument type will not replace the RIDs for various reasons (not for NORM in vehicles, not for heavily shielded isotopes, less sensitive because using CdZnTe detectors). The work done under this research agreement was related to two functions of this new instrument type:

1. Energy compensated dose rate indication based on spectral information 2. Isotope identification, Isotope categorization.

Used Detector and detector setting CZT500, #9928/069 (Ritec, Latvia) MMCA-166, #9980/599 (GBS Elektronik, Germany) Settings 1μs shaping time selected Number of channel: 2048 Usable number of channels 2048-64=1984 Energy range: 3MeV

1.1. Used detector system.

A.2. Dose rate algorithm The dose rate algorithm is based on the spectroscopic information. The detector efficiency of CdZnTe detectors is strongly dependent on the energy of the incoming gamma radiation. For high energy gamma radiation the efficiency is very low in comparison to low energy gamma radiation at the detector. A number of isotopes, mainly emitting monoenergetic gamma lines, were selected in order to cover with all those gamma lines the whole energy range up to 3MeV. ROI I-125 Am-241 Tl-201 Co-57 U-235 Cr-51 F-18 Cs-137 Mn-54 Co-60 [K-40]* Ra-Th1 27keV 2 60keV 3 70keV 4 135keV 122keV 4 136keV 5 167keV 186keV 6 320keV 7 511keV 8 662keV 9 834keV 10 1173keV 10 1333keV 11 1460keV 12 ***

2.1. Selected isotopes for the dose rate algorithm ***…There is only one calibration factor for energies >1500keV ROIs were selected in such a way that the whole energy range is covered by ROIs and the peaks of the isotopes mentioned above are located in one of the ROIs. Twelve ROIs were set around the peaks listed above (see 2.1). A set of spectra of the isotopes listed above was taken and the dose rate at the location of the detector measured with a calibrated reference dose meter. The calibration factor for each photopeak was determined. Furthermore the peak to Compton ratio for all ROIs below the photopeak was determined.

A software program based on Visual Basic 6.0 was created. The algorithm now starts looking from the highest energy for an occupied ROI and goes down to low energies. The calculation starts with the first ROI, which is occupied (occupied means that a certain threshold is exceeded). The peak to Compton ratios in the ROIs below according to the occupied ROI are then calculated and subtracted. The information in the first occupied ROI is then multiplied by the calibration factor. The algorithm is now looking for one more occupied ROI below. If there is still one, the procedure will be repeated. Results are given in table 2.3.

2.2. Schematic picture of the dose rate calculation method for an isotope mixture or a

multipeak spectrum.

2.3. Results of the dose rate algorithm (Blue coloured row contains the isotopes used to

figure out the calibration factor).

The dose rate accuracy for this type of instruments should be within ±50 percent. Therefore the dose rate results based on this algorithm are within the required limit. Pu 61 02 is the source name of reactor grade Plutonium and U446 is the source name for a 4.46% enriched Uranium source. A.3. Isotope Identification algorithm for a CZT 500 detector The goal was to use a fast algorithm, applicable also to spectra with bad statistics. Furthermore the power consumption must be low. Therefore using the normal procedure of a peak search algorithm combined with decision logic does not meet the requirements. A method based on ROIs was used for this algorithm. Only a certain number of isotopes were selected for this method as library. Isotope library Industrial: Na-22, Co-57, Co-60, Ba-133, Cs-137, Eu-152, Ho-166m, Ir-192, Bi-207, Am-241 Medical: F-18, Ga-67, Tc-99m, I-131, Tl-201 Nuclear (U-233), U-235, (Np-237), U-238, Pu WG, Pu RG, NORM (Naturally Occurring Radioactive Material) K-40, Ra-226, Th-232 Method A number of peaks were selected for each isotope of the library. ROIs were set around selected peaks. The result is a number of 39 partially overlapping ROIs. The algorithm now is using the information given in the ROIs, a set of rules (decision logic) and the library in order to identify the isotope or isotope mixture. Identification criteria Identification criteria for single isotopes: Measurement time ≤10 min. 0.5μSv/h at the location of the detector above Background Identification criteria for Mixtures: Identification for at least the same dose rate

No shielded sources A spectra catalogue containing spectra of all the isotopes in the library was created. The spectra were taken with different dose rate at the detector and different measurement time. WinSPEC-T was used for data evaluation. A Visual Basic 6.0 code is currently being developed to test the identification algorithm. A.4. Reference: The algorithm development was done in close cooperation with the FZR (Forschungszentrum Rossendorf, Germany) in view of their development of the SpecWatch (see Research Agreement 12517). A.5. Summary: The Forschungszentrum Rossendorf is developing a miniaturized personal device for automatic detection and identification of gamma radiation sources packed in a wrist watch (called SpecWatch). The identification and dose rate algorithm will be tested by FZR in view of the possible use in this device.

B: A Real Time, Isotope Identifying Gamma Spectrometer for Monitoring of Pedestrians by Michael Schrenk B.1. Abstract: The demand for installation and use of radiation monitors at border crossing points [1, 2] and other locations in a country has significantly increased due to the fact that terrorist threats may involve the use of radiation dispersal devices (RDDs, dirty bombs). One of the problems that customs officers and security forces experience is a high frequency of “innocent” radiation alarms caused by airport passengers who have undergone a medical treatment. Since the half-life of the isotopes that are used for medical treatment ranges from hours to several days, the dose-rate for days and even weeks after the treatment is high enough to trigger a radiation alarm of a border monitor. In this report we describe the development and testing of a real time gamma spectrometer based on a commercially available large volume NaI detector and a computer coupled multi channel analyser (MCA) with fast data collection, stabilisation of the energy scale and isotope identification software. The system is capable of measuring a burst of gamma spectra in second intervals, to identify the isotopes and to produce a “green” alarm in real time when a medical isotope is present and a “red” alarm in other cases. The system has successfully been tested under laboratory conditions, as well as at an international airport and on patients in the radiation ward of hospitals. This work has been performed under an IAEA Research Agreements with the Atom Institute of the Austrian Universities and the International Atomic Energy Agency. B.2. Introduction: The Vienna International Airport (VIA) is a medium sized airport with more than 12 million passengers per year. Passengers that are arriving into the European Union (EU), and having nothing to declare, leave the arrival area with their luggage through the green channel. This Exit is radiation-monitored by a Yantar Pedestrian Monitor [3]. The ASPECT Yantar 1U is a single pillar detection unit containing a plastic szintillator for gamma radiation detection and two He3-tubes which are embedded in a polyethylene moderator for neutron detection. The dimensions of the pillar are 175 x 56 x 25 cm and the detection height is from the ground level up to 200 cm. An external controller (PVC-01) is evaluating the data, making the decision for an alarm and storing all values of gamma and neutron alarms in cps units. The system was installed under the ITRAP (Illicit Trafficking Radiation Assessment Program) [4] [5]. During a test phase, about 6–8 alarms per day occurred when the detector operated at the recommended sensitivity. In the case when an alarm occurs, the custom officers have to detain the persons and send the person back through the radiation monitor. If the pedestrian radiation monitor confirms the alarm, additional manual measurements with a radioisotope identification device (RID) are made to obtain information about the location, the dose rate and the isotope of the source. A large rate of medical isotopes in people was found. These alarms are called innocent alarms and are caused by passengers who recently had undergone therapy or diagnosis with radioactive medical isotopes. Therefore, it was necessary to find a technical solution to reduce the time-consuming process of manually detecting and identifying medical isotopes using a RID. A Real Time Gamma Spectrometer, based on a commercially available large volume NaI(Tl)-detector, a MCA, and a standard laptop with an online identification program, should be able to flag most of the innocent alarms, allowing for a reduction in the workload for the officers.

B.3.Experimental details: B 3.1. A Real Time Gamma Spectrometer (offline mode) This system was used for data collection at the airport and elsewhere to develop and optimise the method. The offline data evaluation should give information about the frequency of the occurrence of commonly used medical isotopes at the VIA. It also should give information about what percentage of the alarms at the VIA could be flagged in real-time as innocent alarms. Figure 1 gives the schematics of the system. A large volume NaI(Tl)-detector (5”x1”) is used for the spectral measurements with high efficiency. A MiniMCA [6] is used for data transmission to the computer in one-second real-time (RT) intervals. The computer program WinSPECa [6] distinguishes between spectra that are under a certain trigger level, and spectra, that have a higher count rate than the preset level. Only in the second case are the spectra recorded on the disk for later evaluation. The identification software versions used for this are IdentPro [7] and Identify [6]. Figure 2 gives an overview of the installation of the systems at the airport. The NaI detector is placed face to face with the Yantar Monitor. In between these detectors is the green channel.

Fig. 1. Schematic of the Offline Real

Time Gamma Spectrometer.

Fig. 2. Ground plan of the green channel, the detector positions and the commonly used pathways (2 and 3). In case of an alarm the

passengers between the monitor and the exit gate are send back to confirm the alarm again

(pathway 1). Measurements of medical isotopes which could cause innocent alarms at border monitors The first data from the airport show a huge difference between the shape of the spectra from sources that were shown in commonly used spectral catalogues and sources carried in passengers due to scattering of the gamma rays. Some measurements at the hospital had to be made on patients to obtain additional information on gamma spectra of radioactive medical isotopes in patients. Based on a list of the commonly used isotopes [8], we were able to measure 12 different medical isotopes in the department of nuclear medicine and the institute of radio oncology of the “Vienna Danube Hospital (SMZ-Ost)”. All these medical isotopes are injected or implanted in the human body. Several days and even weeks after the treatment the dose-rate of these medical isotopes can still be high enough to trigger a radiation alarm. Three different

detectors for the measurements were used: a 3”x3” NaI-, a CZT-1500- and sometimes a portable HPGe-detector. To get more information about the medical isotopes and their spectra (to optimise isotope identification), the following measurements were made: i) Measurements to create a catalogue of in vitro (in the bottle) and in vivo (in the person) spectra. ii) Short-time measurements with low dose rates to get initial information on the detection limits of the Real Time Gamma Spectrometric system. iii) Measurements of Pd-103 and I-125 seeds for prostate gland therapy with their low energetic peaks at 20 keV. These generated questions about the detectability of the isotopes. iv) Measurements on a Perspex phantom to answer a concern of whether we are able to distinguish between a source that is in a person and a source that is carried (smuggled) next to a person’s body, e.g., in the luggage.

The Online Real Time Gamma Spectrometer The evaluation of all the data from the airport and from the hospital measurements showed the direction in which the software developer had to optimise the Online Real Time Gamma Spectrometer. Several field tests of the online version, which is based on the offline set up with the additional new online identification software Specmon [6], were then used to show that the system can fulfil the expectation.

Testing the System for use as a Spectral Portal Monitor The International Atomic Energy Agency (IAEA) will publish in the near future specifications and minimum requirements for spectral portal monitors [9]. These minimum requirements recommend the use of a NaI-detector with a size of at least 6”x2” for spectral pedestrian monitoring. The detector of the system used in this paper is a little smaller but should show, in principle, similar performance. Among others, two special points of interests are investigated in this paper: i) The vertical and horizontal sensitivity distributions for the spectral portal monitor should be comparable to the detection range of a standard pedestrian portal monitor. A decrease of observed count rate of less than 50% at the edges of the range is required. ii) A high count rate can cause a distortion of the gamma ray spectrum. The system should still identify the isotope and be able to quickly recover normal operating conditions after the event. B.3.2. Results 3.2.1. Results of the airport measurements

i) Run through sequences:

These sequences are produced by plotting Integral Counts of each spectrum of an alarm as a function of time. There are two different uses of the time axis. In the first method, each spectrum that is recorded over a period of 1 or 2 seconds in real-time is used to label the axis (Figure 3). In the second method, the real-time axis uses the recorded time stamp in seconds from the ASCII code of the spectral files (Figure 4). This time correction shows gaps that were produced by count rates below the trigger level (i.e., no recorded spectra). However,

there are also gaps that were caused by insufficient writing speed (12 ms, Win 95) to the hard disk.

Fig. 3. Run through sequence of the alarm 102 with the recorded files as the time axis.

Fig. 4. Run through sequence of the alarm 102 from file VIA055481 to file VIA055498 with the time axis uses the

recorded time stamp. This leads to an expansion of the region of figure 3 and shows more details. The gaps were produced by not recorded spectra with count rates below the trigger level and by the insufficient writing speed of

the hard disk (clearly seen at second 7 and 34). Several of the high count rates alarms delivered peaks in the run through sequences with increasing flanks over a time period of 15 seconds and single spectra with Integral counts of more than 20.000 cps. Low count rate alarms (about 1000 cps above background) show quit shorter increasing flanks with about five seconds. The first increase in the count rate shown in the figures corresponds to the peak that triggered the alarm for both, the spectrometer and the Yantar portal monitor. All of the following peaks were caused by the interaction of the customs officers with the passengers. Figure 2 shows the geometry of the two detectors and the possible pathways of the sources. The path commonly taken by passengers is the direct line starting from the corner of the Yantar that goes straight to the corner of the walls (pathway 2). If an alarm occurs, customs officers detain the persons in the area between the monitor and the exit gate. These passengers are sent back through the detection area into the customs control section (pathway 1 parallel to the wall). The officers make their first re-check with the help of the Yantar monitor. Now, the persons were sent separately through the detection area to confirm the alarm. Therefore the first peak is usually the smallest, which is caused by the greater distance of the source to the NaI-detector. Figure 5 shows one of the most interesting events recorded. The spectrometer triggered 8 minutes before the person with the source caused an alarm by passing the pedestrian monitor. The first interaction with the customs officers in front of the detector produced, over a period of one minute, spectra with more than 20.000 counts per seconds. In a room next to the NaI-detector, the person was measured with the RID and the data were noted in the logbook. This procedure produced additional 4 minutes of spectra where the spectrometer system was triggered. The person left the transit area through the red channel (Customs declaration exit).

Fig. 5. The run through sequence of the alarm 112.

ii) Frequency of the alarm giving isotopes:

For this evaluation, a comparison with the logbook from the customs office was necessary. Table I gives a short overview of some differences in the detecting systems.

System Yantar Spectrometer detector Plastic 4.6 litre 5“x1“ NaI(Tl) Crystal background 345 cps (variable) 240 cps (variable) trigger level 8 σ (var. with bkgrd.) 295 cps, constant detecting range small corridor small corridor + part of the luggage hall

Table 1 Comparison of the monitoring systems In 241 days of data collection the spectrometer recorded 163 events. It missed four events that were noted in the logbook. Eleven events were recorded from the spectrometer showing no notification in the customs logbook, which means that the Yantar had not alarmed (only the spectrometer triggered an alarm). The rest of the Yantar alarms and the spectrometric events were at the same time.

The 163 events were: • 154 medical sources • 3 NORM sources (stones and minerals) • 5 not identifiable events • 1 Yantar source test event

The 154 medical sources (100%) were: • 97 Tl-201 (62.99%) • 27 I-131 (17.53%) • 22 Tc-99m (14.28%) • 4 Ga-67 (2.60%) • 3 I-123 (1.95%) • 1 In-111 (0.65%)

About 95% of all medical alarms were caused by only three isotopes (Tl-201, I-131, Tc-99m). The not identifiable isotopes delivered spectra with low statistics, having count rates in the range of the trigger level. Additionally, the alarm sequence was too short. Even summing of all spectra of this alarm event did not give a clear identification result. It was only possible to exclude Tl-201 and Tc-99m. A possible explanation for the eleven extra events could be that they were caused by the detection range of the NaI-detector reaching out into the luggage hall and the blue channel (Exit for EU internal flights), where a source could also exit this luggage area. The spectrometric data gave additional information about the Yantar false alarms, which were noted in the customs logbook. Of these, 24 out of 25 false alarms could be identified as medical.

iii) Time of identification

The other important point of this evaluation was to determine the time when the first of a sequence of spectra is identified correctly, and that all following spectra gave the same result, until the maximum of the run through peak is reached. It requires one second to record and transmit a spectrum to the computer, and less than one second to identify it. Therefore, the spectrum taken at least 2 seconds before the maximum is reached had to be identified correctly in order to generate a timely alarm. This evaluation was done with the peak search mode of the program Identify [6]. The Identification software Identify is based on a peak search algorithm. The peaks found by the algorithm are compared with the peak energies from the isotope library. A decision logic is responsible for the identification results. The most significant peaks in the run through sequence are the first peak and the highest peak. The evaluation of the first peak gives information about how many of the alarms could be flagged in a timely manner as innocent for the case of the geometry of the green channel of the VIA (pathway 2 for the source. Passing the detector in a straight way, shown in figures 2 by the second part of the pathway 1, at a minimum distance of one meter is the typical example of the geometry of a portal monitoring system. Therefore, the data evaluation of the best peaks should provide information for the operation of these kinds of monitors. The evaluation was made with two different methods. The normal method is to evaluate each spectrum of the peak. The summing method is to sum a spectrum to the previous spectra of the same run to get better statistics. In the case of the best peak, only the spectra of the short run through sequence of this peak were taken for the evaluation (not all previous spectra from this alarm).

Evaluation of the first peak Evaluation of the best peak medical isotope normal adding normal adding

I-131 30.8% 50.0% 38.5% 61.5% Tc-99m 59.1% 68.2% 63.6% 86.4% Tl-201 31.3% 56.3% 37.5% 68.6%

Table II the evaluation of the alarm sequences Table II shows the percentage of innocent alarms, which were flagged in a timely way. It can be seen that Tc-99m is the easiest isotope to be identified and that I-131 shows the worst identification properties of the three candidates. The other result shows that the summing method should be a part of the online software version since it gives a higher hit rate. With a specifically tuned software, which is able to evaluate correctly the X-ray peaks of the Tl-201 spectrum, the hit rate of this isotope could be increased enormously. Spectra that are distorted due to high count rates, however, must be excluded. Although not all alarms can be identified

(flagged) in time, this result shows a significant reduction of the manual follow-up which would be required in the case of the use of only a gross gamma detector. B 3.2.2. Measurements with radioactive medical isotopes which could cause innocent alarms at border monitors i) The ICRP 80 Catalogue (from the year 1998)[8] lists 73 different isotopes in use for nuclear medicine. The following isotopes were available at the hospital and gave a good cross-section of the standard isotopes used in Europe: F-18, Co-57, Ga-67, Y-90, Tc-99m, Pd-103, In-111, I-123, I-125, I-131, Sm-153, Tl-201.

All these isotopes were measured in vitro and several were also measured in vivo. Based on this, a comparison of in vitro and in vivo spectra was made. Figure 6 shows clearly the differences. On the one hand, there is a huge region of scattered gamma rays, which is caused by the human body. On the other hand, there is a shielding effect of the body which leads to a decrease in the lower energetic peaks of some isotopes, e.g., I-123.

Fig. 6. Comparison of in vitro and in vivo spectra (recorded with a 3x3” NaI-detector).

ii) The short time measurements at low dose rates show that single peak in vitro spectra with a low energetic peak can be identified correctly even at 50 nSv/h above background and a measurement time of 1 s.

iii) The measurements with Pd-103- and I-125 seeds produced the following results: The system is able to measure the 20 keV Pd-103 and the 27 keV I-125 X-ray peaks even in the case of short measurement times. The situation changes if the seeds are placed into the prostate gland or into a Perspex phantom. The X-ray line of the Pd-103 isotope is shielded so well by the body that in long time measurements only a bad Pd-103 spectrum was obtained. No useable spectra of this isotope in short time measurements were obtained. The 27 keV line of I-125 seeds delivers a usable spectrum in both cases. iv) Several tests with I-131 and a Perspex phantom lead to the following results: Due to the huge scatter area of Tc-99m, I-123 and I-131 a subroutine should be able to determine if the source is in the body or next to the body. Isotopes like I-131 and Ga-67 with peaks in the low energy and in the middle- or high-energy range can give information about shielded or unshielded isotopes. This can also be computed by a corresponding subroutine.

v) Isotopes for the library of the identification programs: In summary, all these measurements show that only a few of the medical isotopes need to be included in the identification library.

Several of the medical isotopes cannot be taken into account because of: a) Too short a half life: F-18 b) Only Bremsstrahlung spectra: Y-90 c) Not measurable in a Real Time Gamma Spectrometric System: Pd-103 d) Seldom used: Co-57 f) Impurities allowed: Sm-153 (dominant Eu-152 spectrum after a few days).

The medical isotopes for the library should include:

Basic: Ga-67, Tc-99m, In-111, I-123, I-131, Tl-201 Additional: I-125, Xe-133 These basic isotopes were found at the airport measurements.

B.4. The Online Real Time Gamma Spectrometer

The requirements for online systems are: • Notebook 600 MHz, 128 MB, Win 98 or higher • background updating and a trigger level of 5 sigma • summing spectra mode + background subtraction • real-time spectral identification • spectrum library should be optimised for the application with emphasis on medical

isotopes • colour indication to flag the kind of isotope: green for medical isotopes and red for

other isotopes, mixture of medical with other isotopes and unknown isotopes • never give an indication of a green alarm if other than medical isotopes are present • capability to detect nuclear and industrial isotopes in the presence of medical isotopes • spectra recording for later evaluation by an expert • electronic logbook of the alarms and identified isotopes

A prototype system was field tested in the hospital. Because of the high dose rates, a 3”x3” NaI-detector was sufficient. People with the following sources were tested: F-18, Ga-67, Tc-99m, In-111, I-123, I-131 and Tl-201. All these isotopes gave a green light indication several seconds before the person was passing the detector. The program used in the Specmon device [6] is the link between the recording- and identification-software. A small window (Figure 7) on the PC-display appears if the program is operating. The window is split into three areas. On the top there is the alarm light indication. A yellow light for the indication that the system is working properly. A red light is the sign that the isotope is not medical (or not identified). The middle part of the display shows the nominal and current background and the trigger level in cps. In addition, the identification results in real time of each new generated spectrum can be seen. The logbook of the last three alarms can be seen in the lowest section of this window.

In the future, the system should work with an occupancy sensor and an external light indication.

Fig. 7. The portal monitor window at the PC-display.

Fig. 8. The horizontal and vertical performance of the detector.

B.5. Testing the System for Use as a Spectral Portal Monitor i) Vertical and horizontal performance To simulate the geometry of a portal monitor the following tests were made. To test the horizontal performance of the detector, it was placed in a horizontal position one meter above the floor. A 73 µSv/h Cs-137 source was placed at a one-meter distance in front of the detector. In 10 cm steps it was moved from the floor (-100 cm) to a height of +100cm. The Integral Counts were compared with the 0 cm position. The decrease in each source position should be less than 50%. The same procedure was used with the detector in a vertical orientation. The horizontal distance from the centre of the detection sensible volume to the Cs-137 source was one meter. The vertical distances were variable in 10 cm steps from minus 100 to plus 100 cm. The Integral Counts were compared with the 0 cm position. A second graph was generated by a comparison of the Integral Counts at the position where the larges amount of counts appeared. ii) System performance at high count rates If a strong source is passing the detector, a distortion of the spectra could occur. The recovery time of the system should be minimal after the collapse and the system should guarantee that after this event the performance is not distorted. A peak shift to higher energies is also the result of higher count rates. If the system is working in a summing spectra mode, the peak shift decreases the resolution of the sum spectrum. Figures 9 shows the spectral distortion for four different isotopes. Each line gives the position of the main peak of that isotope in comparison to the count rate. Each peak has a shift to higher energies. The amount depends on the peak energy. A maximum occurs at 200 to 300 kcps. Then the shift changes direction toward lower energies until a rapid shift shows the start of the collapse.

Fig. 9. The peak centroids of the main peaks of four different isotopes as function of the count rate.

Fig. 10. The dynamic test with an I-131 source. The figure shows the peak shift of the 365 keV peak and the collapse when a source with high activity is passing the

detector. Figure 10 is based on a dynamic test with an I-131 capsule that is moved toward the detector and back away from the detector. It can be clearly seen that after the collapse (files 152–158) the system returns to normal operating conditions. B.6. Summary The work described here had the goal of flagging innocent alarms caused by medical isotopes for cases when a conventional pedestrian monitor would give a gamma alarm. Based on a large volume NaI detector (minimum size of 6”x2”), the system shows a gamma sensitivity that compares well with that of a gross gamma counter. However, it has the advantage of providing the spectrometric information, allowing the identification of isotopes. Therefore, a new generation of spectral portal monitors for monitoring pedestrians has been developed and demonstrated. The decision to be made in the situation of a green alarm indication is still the responsibility of the user (manual follow up or not, depending on the threat level). However, this monitor cannot fully replace a RID, which is still needed for verification of “red” alarms. The most important function of such a device is the identification of medical isotopes. All other isotopes (or unclear cases) must be flagged “red” and be followed up with a RID. The system should never flag any other isotope green. Therefore, the system should have an acceptable sensitivity to detect nuclear material and industrial isotopes in presence of a masking medical isotope. This work will be continued to investigate in more detail performance limitations and improvements of the method, based on the results of ongoing field tests. B.7. Acknowledgement

The author wish to thank the GBS Company for the provision of the online version of the data taking and evaluation program for some of the described tests.

B.8. References [1] IAEA-TECDOC-1312, “Detection of radioactive materials at borders“, Sept. 2002. [2] IAEA-TECDOC-1313, “Response to events involving the inadvertent movement or illicit

trafficking of radioactive materials”, Sept. 2002. [3] ASPECT, http://www.aspect.dubna.ru/english/.

[4] P. Beck, “Illicit Trafficking Radiation Detection Assessment Program”, Austrian Research Center Seibersdorf Report, OEFZ—GS—0002, February (2000), http://www.arcs.ac.at/G/system/itrap.

[5] P. Beck, F. Reichart, W. Klösch, G. Stehno, “ITRAP-Airport”, ARC Seibersdorf research Report, ARC Sr--G--0001, April (2004).

[6] User manuals for Identify, WinSPECa, Specmon and MCA-166, GBS-Elektronik GmbH. Dresten, Germany. Available: http://www.gbs-elektronik.de/.

[7] R. Gunnink, R. Arlt, “Ident Pro: Isotope identification software for analyzing illicit trafficking spectra”, presented at the ESARDA-Symposium 2003, Session 10: Combating Illicit Trafficking, Stockholm, 2003.

[8] Radiation Dose to Patients from Radiopharmaceuticals: ICRP 80 (International Commission of Radioactive Protection, Publication 80) Volume 28, No. 3 1998.

[9] Draft IAEA-TECDOC, Technical/Functional Specifications for Border Radiation Monitoring Equipment, October 2004.

C. Comparison of Simulation and Measurement of Radioactive Sources by the Usage of SuperSynth by Sabine Hengster

C.1 Abstract The main issue of this work was the investigation of the possibility to obtain simulated spectra similar to measured spectra with SuperSYNTH [1], a program to define a set of input parameters for MCNP [2]. C.2 Work performed C.2.1 Comparison of simulation to measurement for several NaI detector configurations

C.2.1.1 Simple source (Cs137)

SuperSYNTH was tested with a Scionix Na(I) 25B25/1-E2-X 1*1” detector [3], which had a simple end cap geometry. In order to get the fitting data for the device, various source configurations and end cap materials were tested. As this did not produce a similar spectrum to the measured, a neutron radiography and a X-ray picture were taken to get an appropriate set of input parameters for the detector (see table C.1).

material Distance distance from detector face

end cap Al 0,4 window Al2O3 0,3

end cap spacing vacuum (predefined) 0,1 absorber 1 silicone rubber 0,1 0,4 absorber 2 Al 0,1 0,5

Table C.1 Input set for a Scionix Na(I) 1*1” detector

A spectrum of this set of variables for a Cs137 source can be seen in figure C.1.

Figure C.1. Simulated and measured spectrum of a Cs137 source for a Scionix [3] detector.

The next type of detector to test were target fieldspec 1*1” and 1*2” Na(I) detectors [4]. The build up of this detector was known as presented by Target Systems, but in this type of detector the detecting crystal was surrounded at a distance of 1.7 cm by the electronics of the detecting device. The input of SuperSynth allowed only the definition of one end cap and entrance window material, therefore the a averaging and weighting of the different materials had to be performed. As the best set of input for the target fieldspec 1*2” Na(I) detectors the following data were obtained: For the detector input:

• RCC NaI with a diameter of 3.54 cm and a length of 5.1 cm, • User defined endcap with a thickness of 1.071 cm, consisting of an interpolated

mixture of 2 H, 2 C, 1 B, 1 Si, 1 S, 1 Cr, 1 Ni, 89 Al atoms and an interpolated density of 3g/cm3. End cap spacing 0.1 cm,

• User defined Teflon end cap window with an thickness of 0.2cm, consisting of 2 F and 1 C Atom and a density of 2.17 g/cm3.

For the absorber input: The missing layers in front of the detector crystal were defined as absorbers in front of the detector face:

• From 0.3 to 0.35 cm an Al disk with a radius of 1.91 cm and a density of 2.707 g/cm3, • From 0.351 to 0.668 cm an Epoxy disk with a radius of 1.27 cm and a density of

1.4 g/cm3, consisting of 10 H, 6 C and 1 O atom, • Form 0.84 to 1.08 cm a second Al disk with a radius of 2.31 cm.

For a resulting spectrum of a Cs137 source see figure C.2.

Figure C.2. Simulated and measured spectrum of a Cs137 source for a target fieldspec

detector. This set of input parameters was used for further simulations concerning Uranium and Plutonium in fertilizers. C.2.1.2 “Masked” source (fertilizer)

Strong sources of Uranium and Plutonium were tried to hide behind several PVC boxes of fertilizer with a dimension of 43.5*29*13 cm. As the composition of the fertilizer was unknown, it was determined by X-ray fluorescence analysis, for details see table 2.

element concentation [μg/g] deviation [μg/g] concentration [%] Si 1221.668 31.378 21.27 P 46.892 5.435 0.82 S 136.639 12.215 2.38 Cl 2342.708 32.438 40.78 K 1923.608 21.688 33.49 Ca 19.372 1.455 0.34 Zn 12.607 0.179 0.22 Br 34.203 0.13 0.6 Rb 6.884 0.104 0.12

Table C.2 Composition of fertilizer

A minimum of 30 centimetres of fertilizer were necessary to shield a bundle of Uranium fuel plates (active length: 584-610mm, active width: 60 mm with a tolerance of 5mm) with an enrichment of 80% and 160g of U 235 at a Live Time of 600 s, see figure C.3.

Figure C.3. Fuel plates and measuring conditions. For the measurement and simulation of Uranium two CBNM [5] sources were used: CBNM-295-028 with an enrichment weight of 2.95% (5g U235, 165g U238) and CBNM-446-028 with an enrichment weight of 4.46% (7.6g U235, 162.4g U238). The nuclear material was measured behind one box of fertilizer, for the measuring conditions see figure C.4.

Figure C.4. CBNM source behind one fertilizer box. For the resulting spectra see figure C.5 and C.6.

Figure C.5. Spectrum of 2.95% enriched Uranium, measured and simulated

at a Live Time of 600s (energy in keV).

Figure C.6. Spectrum of 4.46% enriched Uranium, measured and simulated

by a reduced Live Time of 150s (energy in keV).

For the shielding of a Pu239 source (with a total mass of 6.65g) and an enrichment of 93% about 30 centimetres of fertilizer were necessary, and an 61% enriched source could be shielded by 40 centimetres at a Live Time of 600s. The number of particles of the CBNM [5] sources was too high to simulated, therefore the simulation for the sources had to be simplified by reducing the distance in order to run a simulation with a weaker source. Each source was in form of a sealed capsule containing plutonium oxide in pellet form and a deepening as entrance window, see figure C.7. The sources had a diameter of 1.7 cm, a height of 1.5 cm and an overall mass of 0.45 g.

Figure C.7. Shape of the 239Pu-samples.

The structure of the simulated spectrum was similar to the measured one (see figure C.8).

Figure C.8. Pu239-94% behind 6cm of fertilizer at a Live Time of 180s.

C.2.1.3 “Shielded” source (concrete)

The absorbing concrete was approximated with CaO to SiO2+Al2O3+Fe2O3 mixed with H2O, and a the spectrum of a Cs137 was measured and simulated with a 5*1” Na(I) detector. The Live Time for the simulation had to be reduced, but the structure of the simulated spectrum was similar to the measured (see figure C.9).

Figure C.9. Measured spectrum (black, Live Time 1399s) and simulated spectrum

(red, Live Time 10s and multiplied with 7).

C.2.1.4 Further aspects

• Determination of the energy resolution and shift of the FWHM of the GR-135 Exploranium [6], for details see figure C.10.

Figure C.10. Resolution of the GR-135 Exploranium at 662 keV.

• The measured spectra of Cs137 with a low Z-material as absorber and of a pure beta-

Emitter. A 500MBq Source was measured at a distance of 500cm while the GR-135 Exploranium detector was shielded 100cm of water, and a Sr90 source was measured in order to get two similar spectra shapes – for the spectra see figure C.11.

Figure C.11. Spectra of Cs137 and Sr90.

C.3. Summary This work had the aim to see, how radioactive sources could be masked and shielded by common materials, and to investigate in which extent the spectra could be simulated by Monte Carlo simulation. The programming interface SuperSYNTH [1] proved to be easy to use and produced some spectra with a similar shape as the measured one. As the program could not handle a NPS (number of particle) value > 231-1, and the Live Time had to be reduced, the simulations had to be approximated. The simulation of simple detectors proved to be more accurate than detectors with an unknown or complicate surrounding of the detecting crystal, but for not to sophisticated detectors and geometries SuperSynth proved to be a powerful tool for simulations.

C.4. References [1] W. Hensley, SuperSYNTH, Version 0.40 Beta_1, Pacific Northwest National

Laboratory, Richland, WA, 2004. [2] Monte Carlo N-Particle code, Version 5, Los Alamos National Laboratory, 2003. [3] Scionix, http://www.scionix.nl/. [4] identiFINDER, Target Systems, http://www.target-systems-gmbh.de/frames-e.htm [5] CBNM NRM 271, Central Bureau of Nuclear Measurements Certified Nuclear

Reference Material. [6] Exploranium, http://www.saic.com/products/security/gr-135/.

D. RID Hardware and ID-Software tests to compare different RIDs and ID methods by Vinzenz Schwarz

D.1. Abstract This diploma-thesis deals with the detectability of various isotopes tested with so called hand held radionuclide identification devices (RIDs) under various, partly hardened conditions. Such devices are not only used to measure the count rate that the radiation of an isotope produces, such as a simple Geiger-Müller Counter would do. They are capable of recording spectra of the emitted gamma-radiation and can even identify the radiating isotope. The examined devices are the “IdentiFINDER” by Target systemelectronic GmbH and the MKC-A03 from the Russian company Aspect.

Target IdentiFINDER Aspect MKC-A03 The results of this work should give an in-depth overview of the time required to detect a specific isotope with a certain device or software. This is important because those devices are mostly used by non-professional staff (in terms of the technical background) for border-monitoring, contamination monitoring, surveillance of crowds of people (such as Olympic Games), waste monitoring and much more. For this range of application it is important to know, how the devices respond to different conditions. The first part of the work consists of a presentation of the devices and the software in use and it gives a brief overview over the functionality and the range of application. One can get an idea of how and by whom those devices are commonly used and furthermore take a look at the workflows at borders around the world. The IAEA gives precise specifications on how the device should work, and how easy they should be to handle. This has been published in the document “IAEA Nuclear Security Series No.1 – Technical Guidance” (NSS#1). The handling is particularly important for the staff that has to work with the devices like customs authorities, fire brigades police, and other non-professional users. Those specifications will find place in this part of the work as well as suggested changes to it.

The next section enters into the technical background of the devices. The setup and the physical background are explained and the miscellaneous types of detectors inside of RIDs are accurately described. It is important to highlight the differences between the devices in use and especially between the varying identification algorithms, such as “Template Matching”, “Peak Search” and the “ROI-Method” (Region Of Interest).

Display of Target IdentiFinder, viewing the recorded spectra in advanced mode

Not only the hardware is examined, but also the software which is used to analyse spectra, recorded with the RIDs. Some of the programs are “Identify” by J. Brutscher and “IdentPro” by R. Gunnink. While R.Gunnink uses an enhanced ROI-Method, Identify works efficiently with a peak-search algorithm. The goal is to point out the advantages and disadvantages of each of the algorithms and to give an impression of how the could be combined to get even better results.

Identify by J. Brutscher is using peak-search algorithm

In the next chapter the measurements are the central topic. Setup, conditions and measurements are described and results are being presented. It is dealt with identification limits of both, devices and software and how those limits can be predicted and tuned. It is very interesting to observe the (sometimes completely different) results the devices and the software are yielding from exactly the same setup. The identification quality and the time required to even get a result sometimes varies in a very wide range. In most cases the PC-software is much faster – probably because of the higher computing power. This has to be analysed.

D.2. The main measuring method to determine the minimum time required to identify an isotope with “IdentPro” and Target “identiFinder”: Setup: a single-line-isotope behind a box of fertilizer (K40) First step is to take a long-term spectrum of an isotope behind a box of fertilizer. The fertilizer is used to create a higher count-rate. This has to be done because the IdentiFinder needs a certain amount of counts to identify nuclides properly. Naturally, the resulting spectrum can be easily identified by the identiFinder and IdentPro. Now the intention is to find out, at what real-time of a spectrum IdentPro recognises the isotope with 99% certainty. Therefore the “Simulation Mode” of IdentPro is used to simulate the given spectrum at a shorter runtime. IdentPro is able to not only downgrade the spectrum but to really simulate the spectrum as well as the statistics for a specific runtime. The chosen runtime has to be shorter than the real-time of the given spectrum. The simulation is started at zero runtime (t=0). Then the runtime (t) is increased bit by bit as soon as IdentPro recognises the desired isotope within the simulated spectrum. The next step is to repeat the simulation for 500 times for the determined runtime (t) and to find out whether the isotope can be detected in more then 99% of the 500 simulations. If this is not the case, the runtime is increased and the spectrum is simulated for 500 times again. This procedure is repeated as long as the 99% certainty is reached. The result is a value for the real-time of a spectrum where IdentPro recognises the specific isotope in 99 out of 100 spectra. This value has to be compared with the identiFinder. Therefore 20 (later more…) measurements with the given setup and the evaluated real-time (t) are done. The identiFinder should as well identify the isotope at least in 99% of the measurements. The target-software “wintmca” is used to batch the measurements with the identiFinder.

original spectra simulations with IdentPro batched measurements IdentiFinderisotope realtime [s] simulated time [s] detectability not detected [%] detectabilityCs137+K40 300 50 100,00 0 79%Udep+K40 600 100 13,16 >50% <10%Udep+K40 600 500 20,00 10% ~10%Ba133+K40 600 120 22,25 >20% <10%Ba133+K40 600 360 17,74 0% 16%Eu152+K40 600 170 10,00 0% 56,50%Ra226+K40 600 300 10,28 0% 28% Results from measurements series with IdentPro and Target IdentiFinder. Detectability means the certainty of detection. For example: When examining depleted Uranium, identPro was able to identify the simulated spectra within a simulated time of 100 seconds, while the IdentiFinder needs a measuring time of 500 seconds to deliver suitable results.

Illicit Trafficking: Monte Carlo Modelling of Shielded Uranium Source Gamma Spectra from the NaI Detector

P. Ragan, S. Abousahl

Institute for Transuranium Elements, Joint Research Centre, European Commission, Karlsruhe, Germany

Abstract

The detection of uranium is of greatest interest because it is the type of nuclear material

so far mostly encountered in illicit trafficking. However, its detection and identification is

generally difficult and sometimes even impossible, especially if the uranium is highly

enriched and concealed under shielding.

In this report we will present a complete study of the uranium gamma spectra seen by a

typical hand-held monitor with a NaI detector through various shielding materials.

Resulting gamma spectra were both obtained experimentally and by Monte Carlo

simulation studies. From a review of the spectral data we propose a new approach for the

identification of uranium which makes also use of the shape of the measured gamma

continuum observed from shielded uranium. This complementary information can be

helpful in cases of poor statistical data for the full photon energy peaks of uranium

gamma rays. The applicability and limitations of this analysis approach will be discussed.

Contents

1.1. Avant Propos

1.2. Introduction

2. Methods

2.1. Simulation software

2.2. Detector

2.3. Geometry of the problem

2.4. Decay and daughter products in uranium

2.5. Photon energies

2.6. Bremsstrahlung

2.7. Primary and scattered radiation

3. Calculation of doses

4. Final shape of the spectra

5. Comparison to measured spectra

6. Investigation of an alternative approach

7. Conclusion

1.1. Avant Propos

Within the framework of the Coordinated Research Project (CRP) “Improvement of

Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and other

Radioactive Materials”, IAEA has set a list of R&D topics that are needed. The

participating laboratories to the CRP were encouraged to lead one of these topics. The

ITU-JRC as member of this CRP and also member of the consulting group on

“Functional Specifications of Border Monitoring Equipment” showed interest in the

investigation of algorithms for the detection of shielded radioactive sources and

especially the case of Uranium materials.

The detection of uranium is of greatest interest because it is the type of nuclear material

so far mostly encountered in illicit trafficking. However, its detection and identification is

generally difficult and sometimes even impossible, especially if the uranium is highly

enriched and concealed under shielding.

In this report we will present a complete study of the uranium gamma spectra seen by a




identification of uranium which makes also use of the shape of the measured gamma




This report is made for the IAEA under the research agreement 12592 between IAEA and

ITU/JRC within the framework of the Coordinated Research Project (CRP) “Improvement

of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and other

Radioactive Materials”.

1.2. Introduction

On 17-11-2003 at 22h08 we received the following e-mail from a colleague in xxxx, “I need your help. Border police stopped truck from xxxx, in the post you visited previously. I went with hand held device (detector NaI) to identify source. The Hand held device could not identify and wrote "Not in library". Trailer was very big. I measured around it. Dose rate in one place was 2–3.5 microSv/h, but in all the rest was background dose rate level. I send you the spectrum. I think it can be bremsstrahlung from beta source, is it? Please have a look.” Best regards”. The material was in fact shielded uranium, which could not be detected with the mentioned instrument.

Nowadays fighting against illicit trafficking of nuclear materials is as important as never

before. At the customs only hand-held devices can be used. NaI and CZT detectors are

the most commonly applied hand-held spectrometers, as they are compact, easy-to-use

and need minimal support requirements. On the other hand their resolution capability is

limited. The detection of uranium is of greatest interest because it is the type of nuclear

material so far mostly encountered in illicit trafficking.

Gamma ray spectra from uranium samples depend on the enrichment. The 185.7 keV

gamma ray is the most prominent single gamma ray from a uranium sample enriched

above natural U-235 level. In the case of natural and low enriched uranium the 766.4 keV

and 1001 keV gamma lines from Pa, a daughter of U-238, are also useful for

identification. The identification and the determination of enrichment can be carried out

easily when no other isotopes and shieldings are present.

The identification of shielded radioactive or nuclear sources by gamma spectrometry is

still an unsolved problem. Hand-held devices used by custom officers at borders suffer

from this limitation. The shape of the spectra of a shielded source is specific to the nature

of the radioactive isotopes of the source and the nature and thickness of the shielding

materials. Highly enriched uranium is the most difficult nuclear material to be identified

when it is shielded. 235U is emitting gamma lines at low energies (up to approximately

200 keV). Daughter products from the 238U are emitting gamma lines at high energies

(766.4 and 1001 keV). It is then more easy to detect depleted than enriched shielded

uranium. However, an important contribution to the U spectra comes from beta particles

from the decay of 234mPa emitted with maximum energy 2.29 MeV. The beta particles are

producing bremsstrahlung in the uranium source itself. The Bremsstrahlung continuum is

representing a significant contribution to the shape of the measured spectra.

In this paper we will present a complete study of the uranium gamma spectra seen by a




identification of uranium that makes also use of the shape of the measured gamma




2. Methods

2.1. Simulation software

A Monte Carlo code MCNP ver. 4 has been used for simulations. This code can

simulate a transport of neutrons, photons and electrons in a 3D space in a geometry

consisting of various cells with different materials. Following interactions are taken into

account for the transport of photons: photoelectric absorption with possibility of emission

of x-ray fluorescence photons, coherent (Thompson) modified by scattering functions,

incoherent (Compton) scattering modified by form factors and pair production. The cross

sections of these interactions are contained in libraries for elements from hydrogen (Z=1)

to plutonium (Z=94). An electron library is available for the electron transport. Several

output tallies – current, flux across a surface, track length estimate of the flux in a cell are

available. One of the features is a pulse height tally, so it can simulate a spectral energy

distribution of particles detected by a detector.

2.2. Detector

We assumed as detector a NaI(Tl) detector with a 1x1” cylindrical crystal

(2.54 cm diameter and height as well). The material compositions and densities used to

build up the configuration of the simulation problem are listed in Table 1. The detector is

covered by a MgO reflector and enclosed by aluminium and on the one side there is a

glass coupling to a photomultiplier. The layer of MgO is 1 mm thick, and the Al housing

of the detector is 0.8 mm thick. Detector drawing is in Fig. 1.

Fig. 1. NaI(Tl) detector.

2.3. Geometry of the problem

The following three spherical geometries were chosen for the source material -

• 2 cm diameter, density 10 g . cm-3, mass 41.88 g of UO2, (pellet)

• 5 cm diameter, density 5 g . cm-3, mass 327.25 g of UO2, (pellets)

• 9 cm diameter, density 2.5 g . cm-3, mass 1000 g of UO2, powder

The source is positioned in a centre and is surrounded by air. The assumed lead shielding

is built as a layer at 5 cm from the centre and later on as a box with walls at 5 cm from

the centre. The thickness considered was 1 and 5 mm. First simulations were done with

one NaI(Tl) detector with the front side at 10 cm distance from the centre. The

simulations were improved by introduction of six identical detectors positioned on main

axes of the geometry (x, y, z). A schematic drawing of the configuration is shown in

Fig. 2.

NaI(Tl) Glass coupling to PMT

Al

MgO reflector

Fig. 2. The x-y cut of the geometry with six detectors. Spherical UO2 source is in the middle. Two remaining detectors are on the z axis. The thickness and material of the

shielding is determined by the thickness of the shielding box and is a subject of change.

Table 1. Composition of materials used in simulations

No Material Elemental composition

Fraction by weight

Fraction by elements

Density [g . cm-3]

Na 0.153184 I 0.845582

1. NaI(Tl)

Tl 0.001234

3.67

Mg 0.5 2. MgO O 0.5

2.5

3. Al Al 1 2.7 Si 1 4. Glass – quartz O 2

2.64

C 0.000124 N 0.755268 O 0.231781

5. Dry air

Ar 0.012827

0.001205

U 1 6. UO2 O 2

10, 5, 2.5

7. Pb Pb 1 11.35 Cr 0.18 Fe 0.72

8. Stainless steel ( used as alternative to lead)

Ni 0.1

7.9

C 1 9. Polyethylene (as detector cap) H 2

0.92

2.4. Decay and daughter products in uranium

Uranium is present in the Earth’s crust at an average concentration of 2 ppm.

Acidic rocks with high silicate, such as granite, have higher than average concentrations

of uranium, while sedimentary and basic rocks have lower than average concentrations.

Uranite or pitchblende (U3O8) are the most common uranium-containing ores. The

concentration of U3O8 in ores can vary from 0.5% in Australian ores to 20% in Canadian

ores. Uranium ore after mining is milled and the uranium content is then leached from it.

The final product of this separation procedure is called yellow cake (U3O8).

For 235U enrichment the yellow cake has to be transformed into the gaseous phase

of UF6. The enriched UF6 represents the starting material for the uranium fuel fabrication,

where it is first converted into uranium dioxide. In the fuel cycle the uranium dioxide

normally occurs in the form of powder (density ~2.5 g·cm-3), pressed green pellets

(density 6.5 g·cm-3) or sintered pellets (density ~10 g·cm-3).

Immediately after conversion from UF6 to UO2 the uranium materials are free

from daughter products. The in-growth of them is determined by the half-life of 231Th

(25.52 h) in the case of 235U, and by the half-life of 234Th (24.1 d) in the case of 238U.

Near-equilibrium is therefore reached after about 3 days for 235U, and after ~3 months for 238U. Table 2 lists the typical isotopic composition of different enrichment grades.

Table 3 gives characteristic decay data of uranium and relevant daughter products.

Table 2. Isotopic composition of uranium (atom %) Uranium U-235 U-234 U-238 Depleted 0.2 0.0012 99.7988 Natural 0.72 0.0055 99.2745 Enriched 2 – 5 0.019 – 0.048 Rest 3.5 0.03 96.47 High enriched > 20 > 0.019 Rest 90 0.88 9.12 Table 3. Properties of uranium isotopes and their daughters

Parent nuclide

Half-life| Decay mode

Decay energy[MeV]

Emission probability

[%]

Specific activity

[Bq . g-1]

Daughter nuclide

U-238 4.468E+9 y Alfa 4.27 100 12437 Th-234 SF 0.00005 Th-234 24.1 d Beta 0.273 100 Pa-234m Pa-234m 1.17 min Beta 2.271 98.84 U-234 IT 0.074 0.16 Pa-234 Pa-234 6.7 h Beta 2.197 100*0.16 U-234 U-235 7.038E+8 y Alfa 4.679 100 79960 Th-231 SF 7E-9 Ne 8E-10 Th-231 25.52 h Beta 0.39 100 Pa-231 Alfa 4.213 1E-8 Ra-227 Pa-231 32760 y Alfa 5.149 100 Ac-227 U-234 245500 y Alpha 4.859 100 2.3E+8 Th-230 SF 1.7E-9 Th-230 75380 Alpha 4.77 99.5 Ra-226

2.5. Photon energies

Uranium as a material of interest is produced in various grades of enrichment in 235U. Wide use of uranium as a fuel for NPP made it a “common” problem for illicit

trafficking. Natural and depleted uranium are also used as a relatively cheap and highly

efficient shielding material for therapeutic and industrial radioactive sources.

Photons produced by the radioactive decay of uranium are largely shielded by the

uranium material itself, with a high degree of absorption mainly at low energies below

300 keV. The number of photons flying out of a real sample is then different from the

total number of photons produced in a source. The analysis of the importance of photons

of different energies for detection therefore has to take into account the self-absorption in

the source. It is obvious that the photons of higher energies are of higher relevance.

The self-absorption was calculated using the f2 tally – flux averaged over a

surface for energies in an interval 30–2000 keV. The self-absorption in the spherical

source with diameter 2 cm and the density 10 g . cm-3 versus the photon energy is shown

in Fig. 3.

0.00.10.20.30.40.50.60.70.80.91.0

0 500 1000 1500 2000

Energy [keV]

Selfa

bsor

ptio

n

Fluence Photons of Primary Energy

Fig. 3. Self absorption of photons in the source of 2 cm diameter with a UO2 density of

10 g·cm-3. The points were calculated using MCNP and represent the number of photons flying out of the sphere per one starting photon of corresponding energy. Triangles are representing the primary and scattered photons, and squares are the photons of primary

energy only.

The uranium isotopes and their daughter products emit a variety of gammas and

x-rays. The importance of the different radiations depends on their emission probability

and on the self-absorption in the source as well. The relevance of the different gamma

lines for detecting is changing when we take into account the enrichment and self

absorption. The number of escaping photons per gram then determines to:

Y = Ag · EP · SA,

where Ag is specific activity [Bq · g-1] of uranium nuclides (for daughter products

equilibrium with parent U isotopes is assumed, EP is the emission probability and SA is

the self absorption. Table 4 lists the number of escaping photons from a UO2 sphere

(φ 2 cm, density 10 g·cm-3), ranked by intensity, for 4 different enrichment grades.

Table 4. Number of primary photons (Y) flying out of a UO2 sphere with diameter of

2 cm and density 10 g · cm-3 for four different enrichments in 235U.

HEU 90% 3.50% Natural 0.0072% Depleted 0.002% No. E[keV] Nuclide Y E [keV] Nuclide Y E [keV] Nuclide Y E [keV] Nuclide Y

1 185.7 U-235 2389 185.7 U-235 93 1001.0 234mPa 63 1001.0 234mPa 662 205.3 U-235 263 1001.0 234mPa 61 766.4 234mPa 19 766.4 234mPa 203 143.8 U-235 240 766.4 234mPa 19 185.7 U-235 19 92.8 Th-234 144 93.4 U-235 173 92.8 Th-234 14 92.8 Th-234 14 92.38 Th-234 145 84.2 Th-231 156 92.4 Th-234 14 92.4 Th-234 14 63.29 Th-234 76 163.3 U-235 156 205.3 U-235 10 63.3 Th-234 6 185.7 U-235 67 105.0 U-235 99 143.8 U-235 9 742.8 234mPa 5 742.8 234mPa 58 90.0 U-235 98 93.4 U-235 7 786.3 234mPa 3 786.3 234mPa 49 109.2 U-235 61 84.2 Th-231 6 205.3 U-235 2 112.8 Th-234 2

10 202.1 U-235 55 163.3 U-235 6 112.8 Th-234 2 1737.7 234mPa 2Brems. 37 390 401 403

Gammas in the low energy region are the most important ones for high enriched

uranium (HEU). With increasing amount of 238U in the materials, high energy gammas

from 234mPa at 1001 and 766.4 keV are becoming more and more important. For the

assessment of the number of the detected photons, however, also the dependence of the

detection efficiency of the gamma detector on the gamma energy has to be taken into

account. Fig. 4 shows the calculated detection efficiency for 2 different NaI(Tl) detectors.

1.00E-04

1.00E-03

1.00E-02

0.01 0.1 1 10

Energy [MeV]

Effic

ienc

y

1.000

1.500

2.000

2.500

Volume ratio = 2.16

Ratio

1x1"

Fieldspec - 1.2x1.5"

Fig. 4. The NaI(Tl) detector efficiency for a point isotropic source at 10 cm distance from

the detector front calculated by MCNP. The efficiencies were calculated for 1x1” and 1.2x1.5” detectors commonly used in hand-held instruments.

The final shape of the spectrum as detected by the detector is depending also on

the amount of scattered photons in the source itself, in materials surrounding the

measuring geometry and in materials present between the active detector volume and the

source. The most important are the materials covering the detector (aluminium and

magnesium oxide) and materials of an eventual shielding of the source.

2.6. Bremsstrahlung

Three isotopes are decaying by emission of beta- particle with a decay probability

of ~100% (see Table 3). Two of them (234Th and 231Th) are producing beta particles of

relatively low energy with a maximum at 0.2 and 0.3 MeV. The mean of the energy

distribution is approximately at 1/3 of the maximum energy. The third one (234mPa) has a

maximum of the energy distribution at 2.29 MeV. These beta particles are producing

bremsstrahlung inside the material of a source. The mean of the energy distribution of the

bremsstrahlung continuum is approximately at 1/3 of the mean of the beta particles

energy distribution. Low energy electrons are producing bremsstrahlung mainly at low

energies, with the mean of the distribution far below the 100 keV. In contrary the mean

energy of the betas coming from 234mPa is 0.82 MeV. The bremsstrahlung photons are

then produced with the mean of ~250–300 keV. These photons are partly absorbed by the

material of the source, so the final mean energy is shifted to higher energies

(approximately 400 keV). The actual mean energy is slightly depending on the shape and

the density of the source.

The production of the bremsstrahlung in a UO2 source was obtained by the

simulation of the electron transport in the source. The continuous beta distribution of beta

decay (Fig. 5) needed as input was modeled as a histogram (20 steps in energy), and

electrons were accounted to be emitted with isotropic distribution inside the volume of

the source. The bremsstrahlung photons flying out of the UO2 sphere were then tallied in

10 keV energy bins. (Fig. 6)

0

10

20

30

40

50

60

0 0.5 1 1.5 2 2.5

Energy [MeV]

Rel

ativ

e em

issi

on p

roba

bilit

y

Emax=2.29 MeV

Fig. 5. Spectrum of beta particles emitted by 234mPa.

0.E+00

1.E-06

2.E-06

3.E-06

4.E-06

5.E-06

6.E-06

7.E-06

0.0 0.5 1.0 1.5 2.0

Energy [MeV]

Phot

on fl

ux [c

m-2]

Fig. 6. Spectrum of photons from bremsstrahlung flying out of a 2 cm UO2 sphere with density 10 g · cm-3. Spectrum was tallied (f4 tally) in a cell, which was a shielding box 10x10x10 cm with thickness 1 mm filled with air in 10 keV intervals. Two significant

peaks are Kα and Kβ x-rays of uranium.

The bremsstrahlung spectra were calculated for the three source geometries as

defined above (spherical with diameter 2, 5 and 9 cm). These were then used as input

spectra for tallying the detector response for bare and shielded sources. The same

spectrum plotted in linear scale in Fig. 6 is also shown in logarithmical scale in Fig. 7.

The total number of photons flying out of the source per one beta particle/decay is

contained in table 130 of the MCNP output file (as “cell 1 exiting”). The numbers of the

bremsstrahlung photons which are escaping from the 2, 5 and 9 cm UO2 sources as

defined above are 0.0326, 0.0272 and 0.0290 per one beta particle starting in the source

respectively.

2.7. Primary and scattered radiation

The photon fluxes calculated by MCNP at 10 cm from the centre of the 2 cm UO2

sphere (with density of 10 g · cm-3) normalized per one starting photon of energy 93,

185.7, 205.3, 766.4 and 1001 keV are indicated in Fig. 7. The absorption in the low

energy region is important for primary and scattered photons as well. Results of photon

flux calculations at a distance of 10 cm from the centre of the source are given in Table 5

for photons of primary energy, for scattered photons and for the total photon flux. For

gammas at 185.7 keV the contribution of scattered photons to the total flux is 35% at

maximum. This contribution in the case of a shielded source decreases as the thickness of

the shielding is increasing, and it reaches a value of 23% for photons of the same energy

after 5 mm of lead.

The shielding is changing the spectral fluence of the scattered radiation. Below

about 400 keV the scattered radiation is higly absorbed by the shielding. Above this

energy the contribution of scattered radiation is increasing with increasing thickness of

the shielding. (Fig. 6)

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0Energy [MeV]

Flue

nce

at 1

0 cm

from

cen

tre

[cm

-2]

1001766.4

205.3

185.7

93

Bremsstrahlung

Fig. 7. The photon spectral fluxes at 10 cm distance from the centre of a 2 cm UO2 sphere

(density 10 g·cm-3) normalized for one photon generated in the source. The bremsstrahlung component is generated by beta decay of the 234mPa with betas with

maximal energy 2.29 MeV.

Table 5. Photon fluxes [cm-2] at 10 cm from the centre of a UO2 sphere with diameter 2 cm and density 10 g·cm-3 normalized to one photon of the corresponding primary energy generated in the source. Three combinations – without shielding, shielded by 1 and 5 mm of lead, total – photon flux, primary – flux of the photons of primary energy, scattered – contribution to the total flux by the scattered photons

No shielding 1mm Pb 5mm Pb Energy Total Primary Scattered Total Primary Scattered Total Primary Scattered

93 3.40E-05 3.26E-05 1.39E-06 7.43E-07 7.69E-09 7.35E-07 185.7 7.09E-05 4.58E-05 2.51E-05 1.32E-05 9.85E-06 3.35E-06 3.65E-08 2.80E-08 8.48E-09205.3 8.31E-05 5.80E-05 2.52E-05 2.17E-05 1.73E-05 4.40E-06 2.16E-07 1.68E-07 4.80E-08766.4 5.20E-04 4.26E-04 9.35E-05 4.77E-04 3.78E-04 9.90E-05 3.30E-04 2.33E-04 9.63E-051001 5.94E-04 4.86E-04 1.08E-04 5.62E-04 4.43E-04 1.19E-04 4.33E-04 3.05E-04 1.28E-04

The same as above as relative contributions 93 1 0.96 0.04 1 0.01 0.99

185.7 1 0.65 0.35 1 0.75 0.25 1 0.77 0.23205.3 1 0.70 0.30 1 0.80 0.20 1 0.78 0.22766.4 1 0.82 0.18 1 0.79 0.21 1 0.71 0.291001 1 0.82 0.18 1 0.79 0.21 1 0.70 0.30

UO2 sphere, 2 cm diameter, density 10 g.cm-3

0.00E+00

1.00E-07

2.00E-07

3.00E-07

4.00E-07

5.00E-07

6.00E-07

7.00E-07

8.00E-07

9.00E-07

1.00E-06

0.00E+00 2.00E-01 4.00E-01 6.00E-01 8.00E-01 1.00E+00 1.20E+00 1.40E+00 1.60E+00 1.80E+00 2.00E+00Energy [MeV]

Phot

on fl

uenc

e [c

m-2

]

1001 keV 0 mm Pb1001 keV 1 mm Pb1001 keV 5 mm PbBremsstrahlung

Fig. 8. Fluence spectra of the simulated 1001 keV gamma line from the shielded 2 cm

UO2 sphere compared to the bremsstrahlung continuum (simulated by MCNP).

3. Calculation of doses

When defining a photon field at a point (as flux across the surface for example), the dose

rate at this point can be simply calculated as follows:

⎥⎦⎤

⎢⎣⎡Φ∗∗⎟⎟

⎠

⎞⎜⎜⎝

⎛∗∗= −

24 )(10767.5

cmphoton

hGyEED

tissue

entissue

μρμ

tissue

en⎟⎟⎠

⎞⎜⎜⎝

⎛∗∗ −

ρμ410767.5 is the mass energy-absorption coefficient multiplied by a

constant, E is the energy in keV and )(EΦ the photon flux per cm2. Values for the

photon flux per cm2 are directly given by MCNP tally type 2. When the mass energy-

absorption coefficient multiplied by energy (Fig. 9) is inserted in an input file on tally

modification card, then the result will be directly the dose rate. The dose rates were

caculated for each component (93, 186, 205, 766, 1001, bremsstrahlung) separately and

the results are given in Table 6.

Fig 9. Dose rate conversion coefficients.

Table 6. Partial contributions to dose rate for natural U, enriched by 3.5% with 235U and

HEU with 90% of 235U in [nGy.h-1]. The uranium is in the form of sphere with 2 cm

diameter and mass of 42 g of UO2. (Note – the numbers are rounded, so the sum is

correct, although by manual summing the result will be slightly different.)

Energy [MeV] 0.093 0.186 0.205 0.766 1001 Bremsstrahlung Sum Nat. U 0 mm Pb 1 2 0 9 38 85 136 1 mm Pb 0 0 0 8 35 65 110 5 mm Pb 0 0 0 6 27 43 75 3.5% 0 mm Pb 1 11 2 9 36 83 142 1 mm Pb 0 2 0 8 34 64 109 5 mm Pb 0 0 0 6 26 42 73 90% 0 mm Pb 1 275 44 1 4 9 334 1 mm Pb 0 54 12 1 4 7 77 5 mm Pb 0 0 0 1 3 4 8

0.001

0.01

0.1

1

10

100

1000

10000

0.001 0.01 0.1 1 10

tissue

en⎟⎟⎠

⎞⎜⎜⎝

⎛ρμ

Etissue

en *⎟⎟⎠

⎞⎜⎜⎝

⎛ρμ

Etissue

en *10767.5 4⎟⎟⎠

⎞⎜⎜⎝

⎛∗∗ −

ρμ

4. Final shape of the spectra The investigation on the influence of the beta particles flying out of the unshielded

sources on the detector response shows that they are contributing directly as electrons

depositing their energy in the sensitive volume of the detector, and they are producing

bremsstrahlung in the materials covering the detector crystal as well. The contribution of

these parts is not very important, as can be seen from Fig. 10. It is usually absorbed by

the shielding. Final detector responses for the seven gamma lines, the bremsstrahlung in

the source and the beta particles from the source are displayed in Fig. 10 and in linear

scale in Fig. 11.

Enriched U 3.5%, 2 cm, NaI spectrum, no shielding

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

0.0 0.5 1.0 1.5

U-235 0.1857

Pa-234m 0.766

Pa-234m 1.001

Sum

Bremsstrahlung

Composite .093

U-235 0.205

U-235 0.1438

U-235 0.1633

Electrons

Electrons &brem

Fig. 10. Partial and final detector responses from 2 cm UO2 source without shielding.

5. Comparison to measured spectra The overall final shape of the calculated spectrum after taking into account the

contribution from the bremsstrahlung closely resembles real spectra measured with the

Fieldspec hand held instrument (Fig. 12).

Enriched U 3.5%, 2 cm, NaI spectrum, no shielding

0.000

0.005

0.010

0.015

0.020

0.025

0.030

0.035

0.040

0.0 0.5 1.0 1.5

U-235 0.1857

Pa-234m 0.766

Pa-234m 1.001

Sum

Bremsstrahlung

Composite .093

U-235 0.205

U-235 0.1438

U-235 0.1633

Fig. 11. The same as in Fig. 10 but in a linear vertical scale.

Fig. 12. Measured spectra with NaI(Tl) detector.

0.0001

0.001

0.01

0.1

1

10

100

1 1000

Channels

Phot

ons

BGR

U 4.3% + 4 mm Pb

U 4.3%

6. Investigation of an alternative approach The bremsstrahlung produced by betas from 234mPa represents an important part of

uranium spectra. It is changing the overall shape of the spectrum, and the uranium

gamma and X-ray lines used for the analysis and detection are superimposed on this

continuum. The mean of this continuum is approximately at 400 keV, so it is not so much

shielded as the 235U gammas at energies 186 and 205 keV (5 mm Pb will reduce photons

at this energy ~4 times, at 205 keV more than 200 times, and at 186 keV more than

850 times). An example how the appearance of the Uranium spectra changes in the low

energy with the lead shielding is shown in Figure 12.

The most difficult case for detection is highly enriched uranium (90% 235U) shielded by

lead (a lead thickness of 5 mm was assumed in the present studies), where the

characteristic gamma rays from 235U can no longer be detected. The dose rate in this case

is really low, and it is almost only due to bremsstrahlung and high energy gammas from 234mPa. Evaluation of the spectrum in terms of integral counts within suitable energy

intervals still will give a possibility to make a further analysis, if compared to the known

responses of similar sources. Table 3 shows that it is not possible to detect a smaller

amount of HEU (about 50 g) under shielding. However the ratio (signal + background/

background) is increasing with the uranium amount, and for 235U masses above about

300 g the detection of highly enriched uranium becomes possible using this new

approach.

Table 3. Sensitivity analysis for Highly Enriched Uranium for Fieldspec NaI(Tl) detector 1.2x1.5"

Shielded by1 mm of lead Shielded by 5 mm of lead

Energy interval

(keV) 400-600 400-1100 300-600 300-1100 400-600 400-1100 300-600 300-1100

Counting rate per sec

Background 5.58 10.90 12.74 18.06 5.58 10.90 12.74 18.06

42 g HEU 0.84 1.59 1.55 2.30 0.56 1.12 0.93 1.49

327 g HEU 5.78 11.08 10.52 15.83 3.86 7.85 6.35 10.34

1000 g HEU 18.92 36.16 34.61 51.85 13.13 26.78 21.46 35.12

(Signal + Background)/Background

42 g HEU 1.15 1.15 1.12 1.13 1.10 1.10 1.07 1.08

327 g HEU 2.04 2.02 1.83 1.88 1.69 1.72 1.50 1.57

1000 g HEU 4.39 4.32 3.72 3.87 3.35 3.46 2.68 2.94

7. Conclusion

The MCNP simulations were carried out (i) to model the emission characteristics of

relevant specific gamma lines and of the bremsstrahlung emitted from different

geometries of sources of and HEU, LEU, NU and DU materials under various gamma

shielding, and (ii) to model the characteristic gamma responses of typical scintillation

detectors (NaI) as used in hand-held monitors. The limits of the identification of uranium

under shielding by gamma spectrometry are then defined. The results of this study will be

used for testing the performance of any hand-held device in the identification of any

uranium materials. From a review of the spectral data we are proposing a new approach

for the identification of uranium which makes also use of the shape of the measured

gamma continuum observed from shielded uranium. This complementary information

can be helpful in cases of poor statistical data for the full photon energy peaks of uranium

gamma rays. Further investigations of this approach are still necessary, especially for

uranium mixed with other radioactive sources.

1

Gamma Spectra Measurements of Various Radiation Sources Using a LaBr3 Detector

P. Mortreau, A. Fernandes, R. Berndt JRC Ispra

Abstract

The participation of JCR Ispra in this Coordinated Research Project (CRP) involves

the measurement, compilation and report of the gamma spectra emitted by various

radioactive sources, using standard detectors under reference conditions.

Measurements have been performed in the Perla laboratory for a series of radiation

sources, including plutonium and uranium samples, using a (dia.1x1) in. NaI detector.

In the present work a LaBr3 scintillator detector with trade name BrillanCe-380 has

been used to perform a similar set of measurements. Besides becoming an extension

of the previous work and an additional contribution to a spectra database, this work

may provide some support for the assessment of potential benefits or disadvantages of

scintillators with a better energy resolution for illicit trafficking applications.

NOTE: A folder of spectra is provided. The text version of

these files is shown in ANNEX I-3.

1

INDEX

1. INTRODUCTION

2. EXPERIMENTAL SET-UP

2.1 DOSE RATE MEASUREMENTS

2.2 DETECTION CHAIN

2.3 BACKGROUND MEASUREMENTS

2.4 STABILISATION

2.5 RADIATION SOURCES

2.6 FILTERS

3. SPECTROMETER CALIBRATION

3.1 ENERGY CALIBRATION

3.2 ENERGY RESOLUTION

3.3 DETECTOR EFFICIENCY

4. DETERMINATION OF SOURCE-DETECTOR DISTANCE

5. MEASUREMENTS SERIES

5.1 SINGLE SOURCES

5.2 PLUTONIUM AND CS SOURCES

6. SUMMARY

REFERENCES

ANNEXES

Annex 1: The spectrum file format*.spe

Annex 2: List of measurements series

Annex 3: Energy calibration – list of spectra

2

1. INTRODUCTION

The participation of JCR Ispra in the Coordinated Research Project (CRP)

“Improvement of technical measures to detect and respond to illicit trafficking of

nuclear and radioactive materials” (Research agreement 12595) involves the

measurement, compilation and report of the gamma spectra emitted by various

radioactive sources, using standard detectors under reference conditions [1].


sources, including plutonium and uranium samples, using a (dia.1x1) in. NaI detector

[2]. In the present work a LaBr3 scintillator detector with trade name BrillanCe-380

[3] has been used to perform a similar set of measurements. Besides becoming an

extension of the previous work and an additional contribution to a spectra database,

this work may provide some support for the assessment of potential benefits or

disadvantages of scintillators with a better energy resolution for illicit trafficking

applications.

2. EXPERIMENTAL SET-UP

2.1. DOSE RATE MEASUREMENTS

Source spectra measurements were performed at a dose rate of 0.5 μSv h-1. A dose

rate meter (Scintillator Measuring Unit 6150 ADB) was used to locate the adequate

measurement point in the radiation filed of a source corresponding to this dose rate.

According to its technical specifications, the adequate photon energy range for the

detector operation is 23 keV – 7 MeV and the recommended dose range starts from

5 nSv h-1.

Dose rate profiles were measured in order to determine the measurement position (see

Section 4).

No information is provided about the energy dependence of this detector. Although

this may be a major origin of error in the dose rate estimation, it has been assumed

that the calibration factor is constant for the various energies considered in the present

work.

The source distance (r) is measured from the surface of the dose rate meter to the

surface of the source. The axis of the dose rate meter was aligned with the point

sources or with the axis of small volume sources and perpendicular to the axis of the

large cylindrical sources. In the case of the small volume source, the face turned to the

3

meter was that corresponding to the highest dose rate. The experiment was supported

on a table. The height (h) of the detector axis was 8 cm. A similar setup was used for

the spectra measurements, performed at a slightly larger height (9.5 cm).

An effort was made to minimize the amount of support material inducing radiation

scattering. For supporting the detector in a stable situation, lead bricks were used at

the detector end furthest away from the source. Figure 1 presents a schematic view of

the experimental set-up and

Figure 2 describes some of the source supports used.

Figure 1. Set-up of detectors and radiation sources.

(a) Point sources (e.g., Cs FV713) (b) Small volume sources

(e.g., Pu CBNM61)

Figure 2. Examples of source supports.

2.2. DETECTION CHAIN

Spectra measurements were performed with a LaBr3 detector type BrilLanCe-380

(nr. M-J900CSB) associated to a photomultiplier Model 276 and connected to a

miniaturized system (MCA-166 GBS Elektronik GmbH, Rossendorf, Germany) that

includes the electronics required to operate and control the detector, record and

analyze the measured spectra.

4

The multi channel analyzer was used with 4096 channels at approximately

1 keV/channel (peak of 137Cs at channel nr. 662). A level threshold of 2% was used to

reduce background counts. The detector was operated at a biasing voltage of +500 V

and time shaping constant of 1 μs. The coarse amplifier gain was 50. The pole-zero

correction and fine amplifier gain were visually adjusted to approximately +1376 mV

and 1.1, respectively. The fine adjustment on the amplifier gain was that required to

establish the 1keV/channel reference condition.

2.3. BACKGROUND MEASUREMENTS

Background measurements were performed prior and after each source measurement

series.

Figure 3 shows the background count rate measured during various days, the average

background count rates were 200 cps in PERLA laboratory.

198

200

202

204

206

208

210

212

214

216

0 100 200 300 400 500 600 700 800 900

time(h)

back

grou

nd c

ount

rate

(cps

)

Figure 3. Background count rate.

2.4. STABILISATION

Variations in LaBr3 detector responses as a function of time or temperature often

occur causing a modification in the keV/channel factor. For this reason, a stabilization

procedure was applied in order to maintain the 137Cs peak at channel nr. 662. The

stabilization consists of (i) accumulating counts in the region of interest (ROI) where

the 137Cs peak should occur; (ii) evaluating of the peak centroid in the ROI;

(iii) automatically adjusting the fine gain, in order to shift the peak centroid to channel

5

nr. 662. Each cycle (i)–(iii) is continuously repeated during the measurement and

defines a stabilization interval. The percentage gain correction is after each interval is

displayed.

It is worth referring that once the stabilization is switched off, the former gain

correction will no longer apply.

Before each measurement series, the stabilization procedure consisted of: (i) with the

stabilization switched off, measurement of a 137Cs source after adjusting the amplifier

fine gain in order to locate the peak centroid in channel nr. 662(1); (ii) with the

stabilization switched on, measurement of the 137Cs source until stabilization interval

nr. 5 is performed.

It should be noted that the automatic gain corrections are performed every time the

number of counts in the ROI achieves 50000, which may happen during the

measurement of sources other than 137Cs. This situation has occurred during

measurements of low energy-emitting sources shielded by Pb (e.g., 57Co and U). For

example, in the spectrum of 57Co shielded by 6mm Pb (Figure 4), the most intense

peaks at 122 and 136 keV (total intensity of 86%) have disappeared and those at 692

and 706 keV (total intensity of 0.154%) show up. As these peaks have an energy

similar to that of 137Cs, they will trigger an inconvenient modification in the fine gain

once the filtered source is measured during long time intervals.

Figure 4. Spectra of bare and shielded (6mm Pb) 57Co source.

In such situations, spectra were measured with the stabilization switched off. Prior to

such measurements, a 137Cs source was evaluated and the gain was re-adjusted by

hand whenever necessary. In the present work, stabilization corrections were

6

generally smaller than 2% (and always smaller than 3%) and therefore these situations

have never become problematic.

2.5. RADIATION SOURCES

Table 1 describes the various sources evaluated in the present work. These may be

grouped in (i) point wise standard sources; (ii) uranium sources (ii) plutonium sources

and Th (iiii).

Isotope Eor element (keV)

AEA Technology 511QSA GmbH, nr. FV 711,B1544, B1545, B1524, DC866

57Co MA 101 Point source 122.1AEA Technology 1173.2QSA GmbH, nr. DC 870 1332.5AEA Technology 79.6DD943,DD944, FV712, DC945,DC940, DC867

356

AEA TechnologyQSA GmbH, nr. FV 713AEA Technology 2614.5QSA GmbH, nr. HY 763 (from 208Tl)

CBNM 031 200g U3O8 powder, 0.316 wt% 235U 185CBNM 071 200g U3O8 powder, 0.711 wt%235U 185

U CBNM 446 200g U3O8 powder, 4.46 wt% 235U 185Perla U50 1751g UO2 powder, 19.88 wt% 235U 185

Perla U128 1751g UO2 powder, 92.42 wt% 235U 185

PERLA1-PL30 Al-UO2-Al sandwitch plate 625*70.75*1.27mm3, 61.05g U, 19.88

wt% 235U, active zone(500…600)*(60.4…65.1)*(…0.3

185

PERLA3-PL89 Al-UO2-Al sandwitch plate 625*70.75*1.27mm3, 8.37g U,

93.11wt% 235U, active zone(500…600)*(60.4…65.1)*(…0.3

185

6.625g PuO2 pellets. 238-242Pu wt% 6Jun02:

0.010, 93.533, 6.314, 0.104, 0.040Pu 6.625g PuO2 pellets. 238-242Pu wt%

6Jun02:1.097, 64.968, 26.366, 3.212, 4.358

9.4g Pu+1.5%Ga alloy. 238-242Pu wt% 6Jun02:

0.011, 93.922, 5.901, 0.134, 0.03218.8g Pu+5%Ni+3%Cu alloy. 238-242Pu

wt% 6Jun02: 0.004, 95.470, 4.458, 0.053, 0.015

PERLA06 Pu02 powder, 7.939g Pu, 2steel walls 1.6mm+1mm thick, 238-242Pu wt% 6-

June 2006: 1.618%, 61.252%, 26.084%, 5.092%, 5.954%, Am-241 6.681%

208(Pu241)

PERLA01 Pu02 powder, 7.939g Pu, 2steel walls 1.6mm+1mm thick, 238-242Pu wt% 6-

June 2006: 0.181%, 72.19%, 24.983%, 1.654%, 0.993%, Am-241 2.708%

208(Pu241)

Th PERLA 281T2 Natural powder thorium, 1500g 2614.5

PuGa nr. 32 208( Pu241)

PM2 208( Pu241)

CBNM 61 208( Pu241)

CBNM 93 208( Pu241)

137Cs Point source 661.6

228Th Point source

60Co Point source

133Ba Point source

Source Description

241Am Point source

Table 1. Evaluated radiation sources.

7

2.6. FILTERS

Three filters were used to simulate container walls: 3mm steel, 5mm steel and 6mm

lead. The filters were positioned closed to the detector.

3. SPECTROMETER CALIBRATION

The detection system was calibrated for energy, resolution and efficiency, using the

standard point sources using the following gamma radiation peaks.

________________________

Nuclide Energy(keV) ________________________

133Ba 79.6

57Co 122.1

137Cs 661.62

60Co 1173

1332

228Th(208Tl) 2614.5

Eu152 122

244

344

779

1086

1408

3.1. ENERGY CALIBRATION

The energy calibration was made with the 133Ba, 57Co, 137Cs, 60Co and 228Th sources:

Each point source was measured during 200 s, except the 137Cs source which was

registered for 2000 s.

At the beginning of the measurements series, the stabilization was switched on in the

measurements, so that the 661.6 keV peak of a 137Cs source would be positioned in

channel nr. 662.

In scintillators, the relation between the peak channel and energy normally differs

from the identity function (figure 5). Linear or quadratic functions are normally used

to describe this relation with a very good agreement. In this case, the maximum

8

deviation between measured and calculated peak channels has been reduced (Table 2)

from 130 (identity function) to 12 (linear function) and 4 (quadratic function).

0

500

1000

1500

2000

2500

3000

0 500 1000 1500 2000 2500 3000

channel

Ener

gy(k

eV)

measured points

identity function

quadratic fit (y=3E-05x2 + 0.9916x - 2.0063)

Figure 5. Energy calibration curve of the LaBr3 detector.

Isotope E 8 (keV) channel Identity function

Linear interpolation Quadratic interpolation

(y=x) (y=1.0582x -23.353) (as in Figure 5)

1173 1147.82 25.18 -18.27 1.361332 1297.98 34.02 -18.17 1.54

228Th 2614.5 2473 141.5 20.92 -0.4257Co 122.1 123.31 -1.21 14.97 1.86133Ba 79.6 82.18 -2.58 15.99 0.39

356 357.07 -1.07 1.50 0.81137Cs 661.7 661.82 -0.12 -15.28 -4.18

Channel deviation (calculated – measured)

60Co

Table 2. Peak position.

9

3.2. ENERGY RESOLUTION

The energy resolution is normally characterized by the Full Width at Half Maximum,

FWHM.

Table 3 and Figure 6 show the results obtained for isolated peaks. The linear relation

observed in a log-log scale (with a slope close to 0.5) is typical of scintillators, for

which the statistical contribution to the energy resolution is dominant [4].

Isotope E g (keV) FHWM(%)

1173 2.31332 2.2

228Th 2614.5 1.557Co 122.1 8.0133Ba 79.6 10.9

356 4.0137Cs 661.7 3.1

60Co

Table 3: Detector resolution.

133Ba

133Ba

137Cs

60Co60Co

228Th

1.0

10.0

100 1000 10000

Energy (keV)

Res

olut

ion(

%)

y=-0.5641x+4.8006

Figure 6. Resolution calibration curve of the La Br3 detector.

10

3.3. DETECTOR EFFICIENCY

The detector efficiency was determined with a 152Eu source (reference HP 809),

25 cm distant from the detector and behind a 0.1 cm thickness Cd filter.

The absolute full-energy peak efficiency was calculated as:

γ

γγε

peA

Adect

Treftot

meas

2/1

)2ln(−

= (1)

where reftotA : activity at a reference date ref

totA ,

measA γ : measured peak activity;

T1/2 : source half-life;

tdec : time span between reference and measurement dates.

γp : branching ratio of the gamma line,

Figure 7 shows the measured absolute-full energy peak efficiencies.of the LaBr3

detector and for comparison that of typical detectors [5].

0.00000001

0.0000001

0.000001

0.00001

0.0001

0.001

0.01100 1000 10000E(keV)

Abs

-Ful

l-Ene

rgy

Peak

Eff

icie

ncy

(cou

nts p

er e

mitt

ed p

hoto

n)

NaI diam. 7.6cm*7.6cmNaI diam. 7.6cm*1.9cmGe coaxial diam 5.2cm*5.4 cmGe planar diam. 3.6 cm*2 cmGe planar diam.1.6cm*1 cmCZT/1500 1.5*1.5*0.75 cm3CZT/500(56) 1.5*1.5*0.5 cm3CZT/500(13) 1.5*1.5*0.5 cm3SDP310 0.5*0.5*0.25cm3LaBR3

Figure 7. Absolute-full energy peak efficiency for different detector types [5].

11

4. DETERMINATION OF SOURCE-DETECTOR DISTANCE

The dose profile produced by a point source in vacuum is described by a 1/r2

dependence, where r is the distance to the source. In a practical situation, neither

sources or detectors are point wise; furthermore scattering occurs in the materials

surrounding the experiment (e.g., walls, supports, collimators, air) and in the detector

itself. These distortions are mainly relevant (i) very close to the detector, where the

source and detector dimensions and also the scattering in collimators and supports are

significant, and (ii) at large distances from the detector, where scattering from walls

and air may become important. Still, the 1/r2 dependence is expected to be valid in an

intermediate range of displacement between the source and the detector. In these

cases, the dose profile is described as

D(r)=A/(r+r0)2 (2)

where A is an intensity constant that depends mainly on the source activity and r0

locates the effective measurement point in the detector (see Figure ). The location of

the effective measurement point may depend on photon energy because the

interactions within the detector also depend on energy.

Figure 8. Schematic representation of the effective measurement point r0.

For the conditions of Eq. (2), there will be a linear relation between r and D-1//2:

r=-r0+a1x (3)

with x=D-1/2 and a1=A1/2

To determine the effective measurement point for the sources, dose measurements

were performed in several points, in which the source was placed at different

distances from the detector window. Figure 9 shows an example of the dose rate

profile as a function of the distance between the source and the detector. After fitting

12

of the dose distribution as a function of the distance to the detector, the parameter r is

determined.

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0 5 10 15 20 25Distance (cm)

Dos

e ra

te (m

icro

Sv/

h)

cbnm031

cbnm071

cbnm446

Figure 9. Dose rate as function of the distance source-detector.

The radioprotection instrument is equipped with a gas counter. For this reason, it was

assumed that the effective detector position was in the center of the gas volume

(ro=4 cm). This hypothesis was confirmed by some test measurements with Am and

Cs sources.

Table 4 summarizes the values. (r+ro) of the position corresponding to a dose rate of

0.5 μSv h-1.

13

Isotope

or element

No shield interposed 3 mm Fe 5 mm Fe 6 mm Pb

AEA TechnologyQSA GmbH, nr. FV 711,B1544, B1545,

B1524, DC86620 0.3(D=0.421mSv)* 0.5(D=0.118mSv)* 0.6(D=0.118mSv)*

57Co MA 101 60 45 45 10AEA TechnologyQSA GmbH, nr. AEA Technology

DD943,DD944, FV712,

DC945,DC940, DC867

AEA TechnologyQSA GmbH, nr. FV

AEA TechnologyQSA GmbH, nr.

HY 763

CBNM 031 16.5 14.5 13.8 9CBNM 071 16.5 15 14 10

U CBNM 446 18 16.5 16.5 10.5Perla U50 54 48 44 27.5

Perla U128 79 66 57 17PERLA1-PL30 24 16.5 13 0.6 (D=0.252mSv)*PERLA3-PL89 16 10 7 0.6 (D=0.078mSv)*

354 71 35Pu 86 30 9.5

194 31 1359 31 17

PERLA06 374 161 127 91PERLA01 200 80 65 40

Th PERLA 281T2 190 180 170 150

19.3

24.8 23.8 15.8

12

34.8 34.8 30.8

PM2 34

35 33

23.8

CBNM 93 37PuGa nr. 32 34

24.8

228Th 26.5

CBNM 61 110.5

133Ba 49

137Cs 26.5

r 0.5 (cm)

Source

241Am

60Co 39

Table 4.

*) the dose rate of 0.5 μSv/h could not be achieved. The distance between the source

and the detector corresponds to the thickness of the interposed shield.

14

5. MEASUREMENT SERIES

5.1. SINGLE SOURCES

For each source, a series of measurements was performed as described in Table5 [2].

Source Stabilisation Filter Measurement time (s)

Nr. of spectra

137Cs off - 100 1 137Cs on - 100 1

background on - 300 1 sample on - 3 10 sample on - 10 10 sample on - 180 1 sample on 3 mm

Fe 60 1

sample on 5 mm Fe

300 1

sample on* 6 mm Pb

1200 1

background on* - 300 1

Table 5. List of measurement series for each radiation source.

The spectra for the measurement series have the following name structure:

(i) Background spectra

Background spectra are identified as

bgdssxx.spe

where

bgd stands for background;

ss series name of the source (table 6)

xx is the spectrum number.

(ii) Monitor spectra

Monitor (137Cs) spectra are identifies as

monxxoo.spe

where

mon stands for monitor;

xx is the spectrum number;

15

oo indicates if the stabilization has been switched on or off, according to the

following scheme: oo=on stabilization switched on

oo=of stabilization switched off

(iii) Sample spectra

Sample spectra have been identified as

ssffttx.spe

where

ss identifies the source as defined in Table 6 ;

ff identifies the filter, according to the scheme ff=f0 no filter

ff=f3 filter 3 mm Fe

ff=f5 filter 5 mm Fe

ff=p6 filter 6 mm Pb

tt indicates the measurement time tt=t1 3 s

tt=t2 10s

tt=t3 180s

tt=t4 60s

tt=t5 300s

tt=t6 1200s

x is the spectrum number.

Isotope / Source Identification (ii) Isotope / Source Identification (ii)

241Am am Perla U50 u357Co c5 PERLA1-PL30 u460Co c6 Perla U128 u5228Th th PERLA3-PL 89 u6133Ba ba cbnm61 p0137Cs cs Perla06 p1

U CBNM031 u0 Perla01 p2U CBNM071 u1 cbnm93 p3U CBNM446 u2 PuGa n32 p4

PM2 p5

Table 6. Source identification in spectra files.

16

General information about each spectrum is given in Annex 2. The spectra are given

in electronic form and can be made available to the users

5.2. PLUTONIUM AND CESIUM SOURCES

To simulate an illegal transportation of plutonium transported in a legal transport of a

137Cs source, several measurements were performed using the sources and shielding

available in the laboratory:

1. a 18 mm thick lead cylinder containing a Cesium source of 370MBq was

placed a 25 mm thick Pb shielding ( figure 10 [2]) and the spectrum named

Cs.spe was measured,

2. a Plutonium sample (1000g Pu powder, isotopes:239…242:0.181%,

72.190%,24.983%, 1.654%, 0.993%, 241Am:2.708%, 2 stainless steel walls of

1.6 mm of 1.6mm and 1mm thickness) was placed behind the same shielding

of 25 mm lead and the spectrum Pu.spe was measured

3. The Pu and Cs sources were measured together and the spectrum PuCs.spe

was registered.

The three spectra are shown on figure 11.( see file PuandCs.xls)

Figure 10.

17

0

500

1000

1500

2000

2500

3000

3500

0 100 200 300 400 500 600 700 800 900 1000

channel

Nb

of c

ount

s

Shielding 137Cs:18+25 mm Pb

Shielding Pu: 25 mm Pb

Shielding 137Cs: 18+25 mm Pb, shielding Pu:25mm Pb

Figure 11.

6. SUMMARY


sources, including plutonium and uranium samples, using a LaBr3 detector with trade

name BrillanCe-380 [3] detector. The Source spectra measurements performed at a

dose rate of 0.5 μSv h-1are a contribution to the spectra data base for the trafficking

Coordinated Research Project (CRP) “Improvement of technical measures to detect

and respond to illicit trafficking of nuclear and radioactive materials”. These spectra

are made available to the users as ZIP. Files.

For additional questions, please contact: [email protected]

18

REFERENCES [1] Coordinated Research Project (CRP) “Improvement of technical measures to

detect and respond to illicit trafficking of nuclear and other radioactive materials”, International Atomic Energy Agency, Vienna, 25 April 2002.

[2] R. Berndt and G. Caravati, “Gamma spectra of different radiation sources,

measured with a NaI detector”, Technical Note Nr. 1.03.31, Institute for the Protection and Security of the Citizen, Joint Research Center - Ispra, February 2003.

[3] A. Iltis, M.R. Mayhugh, P. Menge, C.M. Rozsa, O. Selles, V. Solovyev,

NIM section A 563 (2006) 359-363 [4] G. F. Knoll, “Radiation detection and measurement”, 2nd Ed., John Wiley &

Sons, Inc., New York, 1988. [5] P. Mortreau, R. Berndt, “Handbook of gamma spectrometry methods for non-

destructive Assay of nuclear materials”, report EUR 19822 EN

20

ANNEX 1

Spectrum file format *.spe

21

File contents Comment

$MCA_166_ID: spectrometer type: MCA-166 SN# 283 serial no. of the instrument used for this spectrum HW# 9801 hardware version FW# 9804 firmware version AP# SPEC 1.23.12 (r) Gen.Date: 1998/06/16 software used: SPEC.EXE 1.23.12 (r) $SPEC_REM: user remarks: JRC NaI 8631 1"*1" remark 137Cs no. FV713 333kBq 06/06/02 ... filter 0mm, distance 430mm ... 120cps, dead time 0% ... A:\DATA\csnf0t33.spe path and spectrum name (as given after the measurement) $DATE_MEA: date and time of the measurement 06/18/2002 05:56:58 mm/dd/yyyy hh:mm:ss $MEAS_TIM: measurement time: 180 180 live time dead time $DATA: 0 4095 no. of first and last channel 0 counts in channel 0 0 counts in channel 1 0 0 0 6 12 6 8 14 15 14 21 18 ... 0 0 0 $ROI: 1 616 712 $ENER_FIT: -33.022671 0.000000 $ENER_DATA: 2 661.830017 661.700012 1300.869995 1332.500000 $ADC: 4096 0 3967 $PRESETS: Live Time (sec) 180 0 $PZC_VALUE: parameters for pole zero cancellation

22

2499 15 15 $FAST_DISCR: 400 $SLOW_DISCR: 400 $THR: 0 $GAIN_VALUE: 20 1.1317 $DTC: 1 $INPUT: Amplifier neg $PUR: on $STAB: gain stabilisation on gain stabilisation was ON 600 gain stabilisation evaluates the peak in the range from 724 channel 600 to 724 and brings the centroid into 662 channel 662 $POWER: +12= on -12= on +24= on -24= on $HV: detector high voltage +646V +646 V used in this measurement unused $MCS_CHANNELS: 4096 $MCS_INPUT: Input Rate $MCS_TIME: 10 $MCS_SWEEPS: 1 $MODE: MCA $TDF: 800 $POWER_STATE: I+12= 8mA I-12= 10mA I+24= 5mA I-24= 4mA IBAT= 159mA IHV = 18mA ICHR= 159mA UBAT=8049mV UHVs= 644mV $COUNTS: 17606 $PD_COUNTS: 18066

23

$RT: 180 $DT: 188 $BT: 188 $STAB_OFFSET: 47 $STAB_OFFSET_MIN: 47 $STAB_OFFSET_MAX: 47 $STAB_COUNTER: 5 $REC_COUNTER: 1073245 $REC_ERROR_COUNTER: 297 $SPEC_INTEGRAL: 18025 total number of counts in the spectrum $ROI_INFO: 1 616 712 664.31 13.14 1532 1229 127

24

ANNEX 2

Source spectra files

25

241Am series Source: AEA Technology no. FV711, B1544, B1545, B1524, DC 866 Total activity: 1873 KBq (08-May-2006) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 08-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon001of.spe 137Cs 661.6 no - 100 1848 2 672.93 mon002on.spe 137Cs 661.6 no - 100 1850 2 663.48 amf0t10.spe 241Am 59.6 no 20 3 2212 2.7 61.93 amf0t11.spe 241Am 59.6 no 20 3 2249 1.9 62.01 amf0t12.spe 241Am 59.6 no 20 3 2197 2.8 61.88 amf0t13.spe 241Am 59.6 no 20 3 2242 2.8 62.00 amf0t14.spe 241Am 59.6 no 20 3 2204 2.4 61.92 amf0t15.spe 241Am 59.6 no 20 3 2215 1.9 61.55 amf0t16.spe 241Am 59.6 no 20 3 2238 2.3 61.62 amf0t17.spe 241Am 59.6 no 20 3 2264 2.1 61.62 amf0t18.spe 241Am 59.6 no 20 3 2240 2.3 61.65 amf0t19.spe 241Am 59.6 no 20 3 2217 2.0 61.56 amf0t20.spe 241Am 59.6 no 20 10 2287 2.5 61.57 amf0t21.spe 241Am 59.6 no 20 10 2288 2.7 61.52 amf0t22.spe 241Am 59.6 no 20 10 2239 2.3 61.63 amf0t23.spe 241Am 59.6 no 20 10 2270 2.4 61.59 amf0t24.spe 241Am 59.6 no 20 10 2310 2.6 61.56 amf0t25.spe 241Am 59.6 no 20 10 2275 2.6 61.57 amf0t26.spe 241Am 59.6 no 20 10 2272 2.3 61.56 amf0t27.spe 241Am 59.6 no 20 10 2289 2.3 61.57 amf0t28.spe 241Am 59.6 no 20 10 2269 2 61.88 amf0t29.spe 241Am 59.6 no 20 10 2294 2.6 61.53 amf0t30.spe 241Am 59.6 no 20 180 2300 2.4 61.55 amf3t40.spe 241Am 59.6 3, Fe 0.3 60 4530 4.7 60.88 amf5t50.spe 241Am 59.6 5, Fe 0.5 300 762 0.8 61.05 amp6t60.spe 241Am 59.6 6, Pb 0.6 1200 186 0.3 - backam01.spe no - 300 216 0.5 -

26

133Ba series Source:AEA Technology no.FV712, DD943,DD944, DC945,DC940, DC867 Activity: 761 KBq (08-May-06) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 08-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon003of.spe 137Cs 661.6 no 10 100 1823 2 671.47 mon004on.spe 137Cs 661.6 no 10 100 1824 2 661.98 backba00.spe no - 300 207 0.3 baf0t10.spe 133Ba 356 no 49 3 1122 1.7 359.707 baf0t11.spe 133Ba 356 no 49 3 1055 1.6 360.16 baf0t12.spe 133Ba 356 no 49 3 1124 1.7 360.44 baf0t13.spe 133Ba 356 no 49 3 1132 1.1 360.24 baf0t14.spe 133Ba 356 no 49 3 1095 1.4 359.19 baf0t15.spe 133Ba 356 no 49 3 1100 1.6 360.57 baf0t16.spe 133Ba 356 no 49 3 1106 0.8 359.91 baf0t17.spe 133Ba 356 no 49 3 1085 1.6 361.02 baf0t18.spe 133Ba 356 no 49 3 1129 1.3 360.87 baf0t19.spe 133Ba 356 no 49 3 1111 1.1 360.10 baf0t20.spe 133Ba 356 no 49 10 1109 1.7 360.00 baf0t21.spe 133Ba 356 no 49 10 1114 1.6 359.59 baf0t22.spe 133Ba 356 no 49 10 1122 1.6 360.46 baf0t23.spe 133Ba 356 no 49 10 1123 0.8 359.30 baf0t24.spe 133Ba 356 no 49 10 1129 1.7 359.72 baf0t25.spe 133Ba 356 no 49 10 1128 1.7 360.26 baf0t26.spe 133Ba 356 no 49 10 1144 1.6 360.56 baf0t27.spe 133Ba 356 no 49 10 1107 1.5 361.02 baf0t28.spe 133Ba 356 no 49 10 1116 1.6 360.10 baf0t29.spe 133Ba 356 no 49 10 1128 1.6 359.77 baf0t30.spe 133Ba 356 no 49 180 1111 1.6 360.04 baf3t40.spe 133Ba 356 3, Fe 35 60 1331 1.4 360.33 baf5t50.spe 133Ba 356 5, Fe 33 300 1298 1.4 360.36 bap6t60.spe 133Ba 356 6, Pb 12 1200 450 0.5 - backba01.spe no - 300 210 0.3

27

57Co series Source:MA101 Activity:109 KBq (12-May-06) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 12-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon005of.spe 137Cs 661.6 no 10 100 1947 2.2 672.79 mon006on.spe 137Cs 661.6 no 10 100 1958 2.1 662.14 backc500.spe no - 300 299 0.5 c5f0t10.spe 57Co 122.1 no 60 3 3981 0.7 123.72 c5f0t11.spe 57Co 122.1 no 60 3 4098 0.7 123.94 c5f0t12.spe 57Co 122.1 no 60 3 4009 0.7 123.98 c5f0t13.spe 57Co 122.1 no 60 3 4056 0.7 124.05 c5f0t14.spe 57Co 122.1 no 60 3 4030 0.7 123.95 c5f0t15.spe 57Co 122.1 no 60 3 4030 0.7 123.65 c5f0t16.spe 57Co 122.1 no 60 3 3987 0.7 124.00 c5f0t17.spe 57Co 122.1 no 60 3 4070 0.7 123.61 c5f0t18.spe 57Co 122.1 no 60 3 3968 0.7 123.60 c5f0t19.spe 57Co 122.1 no 60 3 4006 0.7 124.32 c5f0t20.spe 57Co 122.1 no 60 10 4059 0.7 123.99 c5f0t21.spe 57Co 122.1 no 60 10 4051 0.7 123.66 c5f0t22.spe 57Co 122.1 no 60 10 4016 0.7 123.70 c5f0t23.spe 57Co 122.1 no 60 10 4047 0.7 123.94 c5f0t24.spe 57Co 122.1 no 60 10 3999 0.7 124.01 c5f0t25.spe 57Co 122.1 no 60 10 4071 0.7 123.70 c5f0t26.spe 57Co 122.1 no 60 10 4064 0.7 124.03 c5f0t27.spe 57Co 122.1 no 60 10 3997 0.7 123.75 c5f0t28.spe 57Co 122.1 no 60 10 4098 0.7 123.97 c5f0t29.spe 57Co 122.1 no 60 10 4055 0.7 123.98 c5f0t30.spe 57Co 122.1 no 60 180 3904 0.7 123.97 c5f3t40.spe 57Co 122.1 3, Fe 45 60 4632 0.9 125.87 c5f5t50.spe 57Co 122.1 5, Fe 45 300 3637 0.7 123.58 c5p6t60.spe 57Co 122.1 6, Pb 10 1200 1233 1.5 - backc501.spe no - 300 204 0.3

28

137Cs series Source:AEA Technology no.FV713 Activity:312 KBq (10-May-06) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 10-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon007of.spe 137Cs 661.6 no 10 100 1819 2 669.14 mon008on.spe 137Cs 661.6 no 10 100 1815 2 663.31 backcs00.spe no - 300 216 0.3 csf0t10.spe 661.6 no 26.5 3 499 0.8 662.32 csf0t11.spe 661.6 no 26.5 3 494 0.8 663.67 csf0t12.spe 137Cs 661.6 no 26.5 3 525 0.8 662.08 csf0t13.spe 137Cs 661.6 no 26.5 3 522 0.8 661.03 csf0t14.spe 137Cs 661.6 no 26.5 3 524 0.8 661.14 csf0t15.spe 137Cs 661.6 no 26.5 3 513 0.8 664.31 csf0t16.spe 137Cs 661.6 no 26.5 3 516 0.8 660.53 csf0t17.spe 137Cs 661.6 no 26.5 3 524 0.8 656 csf0t18.spe 137Cs 661.6 no 26.5 3 520 0.8 662.60 csf0t19.spe 137Cs 661.6 no 26.5 3 518 0.8 661.52 csf0t20.spe 137Cs 661.6 no 26.5 10 524 0.8 661.47 csf0t21.spe 137Cs 661.6 no 26.5 10 510 0.8 661.59 csf0t22.spe 137Cs 661.6 no 26.5 10 514 0.8 661.69 csf0t23.spe 137Cs 661.6 no 26.5 10 524 0.8 661.45 csf0t24.spe 137Cs 661.6 no 26.5 10 507 0.8 661.08 csf0t25.spe 137Cs 661.6 no 26.5 10 518 0.8 659.88 csf0t26.spe 137Cs 661.6 no 26.5 10 527 0.8 661.67 csf0t27.spe 137Cs 661.6 no 26.5 10 510 0.8 661.19 csf0t28.spe 137Cs 661.6 no 26.5 10 511 0.8 662.41 csf0t29.spe 137Cs 661.6 no 26.5 10 514 0.8 661.34 csf0t30.spe 137Cs 661.6 no 26.5 180 518 0.8 661.44 csf3t40.spe 137Cs 661.6 3, Fe 24.8 60 528 0.2 661.08 csf5t50.spe 137Cs 661.6 5, Fe 23.8 300 549 0.2 661.92 csp6t60.spe 137Cs 661.6 6, Pb 19.3 1200 477 0 661.93 backcs01.spe no - 300 220 0

29

60Co series Source:AEA Technology no.DC 870 Activity: 63 KBq (10-May-06) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 10-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon009of.spe 137Cs 661.6 no 10 100 1800 2.3 668.21 mon010on.spe 137Cs 661.6 no 10 100 1800 2 662.98 backc600.spe no - 300 240 0.3 c6f0t10.spe 60Co 1173.2 no 39 3 370 0.4 1154.84 c6f0t11.spe 60Co 1173.2 no 39 3 370 0.4 1163.00 c6f0t12.spe 60Co 1173.2 no 39 3 370 0.4 1141.26 c6f0t13.spe 60Co 1173.2 no 39 3 370 0.4 1144.00 c6f0t14.spe 60Co 1173.2 no 39 3 370 0.4 1152.0 c6f0t15.spe 60Co 1173.2 no 39 3 370 0.4 1118.00 c6f0t16.spe 60Co 1173.2 no 39 3 370 0.4 1319.0 c6f0t17.spe 60Co 1173.2 no 39 3 370 0.4 1152.88 c6f0t18.spe 60Co 1173.2 no 39 3 370 0.4 1145.22 c6f0t19.spe 60Co 1173.2 no 39 3 370 0.4 1153.41 c6f0t20.spe 60Co 1173.2 no 39 10 370 0.4 1157.47 c6f0t21.spe 60Co 1173.2 no 39 10 370 0.4 1153.43 c6f0t22.spe 60Co 1173.2 no 39 10 370 0.4 1161.00 c6f0t23.spe 60Co 1173.2 no 39 10 370 0.4 1158.1 c6f0t24.spe 60Co 1173.2 no 39 10 370 0.4 1163.32 c6f0t25.spe 60Co 1173.2 no 39 10 370 0.4 1157.88 c6f0t26.spe 60Co 1173.2 no 39 10 370 0.4 1169.83 c6f0t27.spe 60Co 1173.2 no 39 10 370 0.4 1165.57 c6f0t28.spe 60Co 1173.2 no 39 10 370 0.4 1160.03 c6f0t29.spe 60Co 1173.2 no 39 10 370 0.4 1161.69 c6f0t30.spe 60Co 1173.2 no 39 180 370 0.4 1160.29 c6f3t40.spe 60Co 1173.2 3, Fe 34.8 60 550 0.6 1157.57 c6f5t50.spe 60Co 1173.2 5, Fe 34.8 300 400 0.6 1155.77 c6p6t60.spe 60Co 1173.2 6, Pb 30.8 1200 380 0.5 1158.45 backc601.spe no - 300 210 0.3

30

228Th series Source:AEA Technology no.HY763 Activity: 193 KBq (12-May-06) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 10-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon011of.spe 137Cs 661.6 no 10 100 1700 1.9 663.19 mon012on.spe 137Cs 661.6 no 10 100 1700 1.9 662.05 backt800.spe no 300 230 0.4 t8f0t10.spe 228Th 2614.5 no 26.5 3 250 0.6 2428 t8f0t11.spe 228Th 2614.5 no 26.5 3 250 0.6 2407 t8f0t12.spe 228Th 2614.5 no 26.5 3 250 0.6 2508.33 t8f0t13.spe 228Th 2614.5 no 26.5 3 250 0.6 2431.00 t8f0t14.spe 228Th 2614.5 no 26.5 3 250 0.6 2451.00 t8f0t15.spe 228Th 2614.5 no 26.5 3 250 0.6 - t8f0t16.spe 228Th 2614.5 no 26.5 3 250 0.6 2520.00 t8f0t17.spe 228Th 2614.5 no 26.5 3 250 0.6 2476.00 t8f0t18.spe 228Th 2614.5 no 26.5 3 250 0.6 2499.00 t8f0t19.spe 228Th 2614.5 no 26.5 3 250 0.6 2406.00 t8f0t20.spe 228Th 2614.5 no 26.5 10 250 0.6 2506.29 t8f0t21.spe 228Th 2614.5 no 26.5 10 250 0.6 2512.00 t8f0t22.spe 228Th 2614.5 no 26.5 10 250 0.6 2522.64 t8f0t23.spe 228Th 2614.5 no 26.5 10 250 0.6 2503.20 t8f0t24.spe 228Th 2614.5 no 26.5 10 250 0.6 2512.20 t8f0t25.spe 228Th 2614.5 no 26.5 10 250 0.6 2505.75 t8f0t26.spe 228Th 2614.5 no 26.5 10 250 0.6 2511.20 t8f0t27.spe 228Th 2614.5 no 26.5 10 250 0.6 2505.00 t8f0t28.spe 228Th 2614.5 no 26.5 10 250 0.6 2483.67 t8f0t29.spe 228Th 2614.5 no 26.5 10 250 0.6 2479.00 t8f0t30.spe 228Th 2614.5 no 26.5 180 250 0.6 2494.70 t8f3t40.spe 228Th 2614.5 3, Fe 24.8 60 230 0.6 2503.73 t8f5t50.spe 228Th 2614.5 5, Fe 23.8 300 230 0.4 2491.41 t8p6t60.spe 228Th 2614.5 6, Pb 15.8 1200 170 0.5 2495.65 backt801.spe no - 300 60 0.3

31

U series u0 Source: CBNM031 U3O8 powder, 200g U3O8, 0.316wt% 235U, container wall: Al, 2 mm thick, plane Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 11-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon013of.spe 137Cs 661.6 no - 100 1700 10 672.07 mon014on.spe 137Cs 661.6 no - 100 1700 10 663.50 backu000.spe no - 300 200 0 u0f0t10.spe U 185.7 no 16.5 3 680 0 203.01 u0f0t11.spe U 185.7 no 16.5 3 680 0 192.75 u0f0t12.spe U 185.7 no 16.5 3 680 0 199.87 u0f0t13.spe U 185.7 no 16.5 3 680 0 188.81 u0f0t14.spe U 185.7 no 16.5 3 680 0 188.55 u0f0t15.spe U 185.7 no 16.5 3 680 0 193.17 u0f0t16.spe U 185.7 no 16.5 3 680 0 180.58 u0f0t17.spe U 185.7 no 16.5 3 680 0 - u0f0t18.spe U 185.7 no 16.5 3 680 0 187.61 u0f0t19.spe U 185.7 no 16.5 3 680 0 - u0f0t20.spe U 185.7 no 16.5 10 680 0 190.28 u0f0t21.spe U 185.7 no 16.5 10 680 0 188.74 u0f0t22.spe U 185.7 no 16.5 10 680 0 189.34 u0f0t23.spe U 185.7 no 16.5 10 680 0 188.90 u0f0t24.spe U 185.7 no 16.5 10 680 0 190.32 u0f0t25.spe U 185.7 no 16.5 10 680 0 185.83 u0f0t26.spe U 185.7 no 16.5 10 680 0 188.27 u0f0t27.spe U 185.7 no 16.5 10 680 0 184.89 u0f0t28.spe U 185.7 no 16.5 10 680 0 191.81 u0f0t29.spe U 185.7 no 16.5 10 680 0 184.11 u0f0t30.spe U 185.7 no 16.5 180 680 0 189.20 u0f3t40.spe U 185.7 3, Fe 14.5 60 720 0.5 189.89 u0f5t50.spe U 185.7 5, Fe 13.8 300 650 0.4 188.64 u0p6t60.spe U 185.7 6, Pb 9 1200 440 0.2 - backu001.spe no - 300 200 0

32

U series u1 Source: CBNM071 U3O8 powder, 200g U3O8, 0.711wt% 235U, container wall: Al, 2 mm thick, plane Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 08-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon015of.spe 137Cs 661.6 no 10 100 1900 10 671.99 mon016on.spe 137Cs 661.6 no 10 100 1900 10 663.83 backu100.spe no - 300 80 0 u1f0t10.spe U 185.7 no 16.5 3 160 0 179.71 u1f0t11.spe U 185.7 no 16.5 3 160 0 191.44 u1f0t12.spe U 185.7 no 16.5 3 160 0 192.16 u1f0t13.spe U 185.7 no 16.5 3 160 0 190.12 u1f0t14.spe U 185.7 no 16.5 3 160 0 186.83 u1f0t15.spe U 185.7 no 16.5 3 160 0 188.27 u1f0t16.spe U 185.7 no 16.5 3 160 0 191.75 u1f0t17.spe U 185.7 no 16.5 3 160 0 186.72 u1f0t18.spe U 185.7 no 16.5 3 160 0 190.91 u1f0t19.spe U 185.7 no 16.5 3 160 0 185.90 u1f0t20.spe U 185.7 no 16.5 10 160 0 188.64 u1f0t21.spe U 185.7 no 16.5 10 160 0 190.07 u1f0t22.spe U 185.7 no 16.5 10 160 0 191.03 u1f0t23.spe U 185.7 no 16.5 10 160 0 190.34 u1f0t24.spe U 185.7 no 16.5 10 160 0 189.29 u1f0t25.spe U 185.7 no 16.5 10 160 0 189.34 u1f0t26.spe U 185.7 no 16.5 10 160 0 187.19 u1f0t27.spe U 185.7 no 16.5 10 160 0 186.92 u1f0t28.spe U 185.7 no 16.5 10 160 0 189.80 u1f0t29.spe U 185.7 no 16.5 10 160 0 185.56 u1f0t30.spe U 185.7 no 16.5 180 160 0 188.86 u1f3t40.spe U 185.7 3, Fe 15 60 400 0.5 188.50 u1f5t50.spe U 185.7 5, Fe 14 300 380 0.5 189.18 u1p6t60.spe U 185.7 6, Pb 10 1200 250 0.5 - backu101.spe no - 300 70 0

33

U series u2 Source: CBNM446 U3O8 powder, 200g U3O8, 4.46wt% 235U, container wall: Al, 2 mm thick, plane Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 08-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon017of.spe 137Cs 661.6 no 10 100 1900 2 671.42 mon018on.spe 137Cs 661.6 no 10 100 1900 2 661.96 backu200.spe no - 300 220 0.3 u2f0t10.spe U 185.7 no 18 3 1050 1 190.34 u2f0t11.spe U 185.7 no 18 3 1050 1 186.38 u2f0t12.spe U 185.7 no 18 3 1050 1 189.71 u2f0t13.spe U 185.7 no 18 3 1050 1 190.36 u2f0t14.spe U 185.7 no 18 3 1050 1 188.93 u2f0t15.spe U 185.7 no 18 3 1050 1 188.95 u2f0t16.spe U 185.7 no 18 3 1050 1 189.32 u2f0t17.spe U 185.7 no 18 3 1050 1 189.35 u2f0t18.spe U 185.7 no 18 3 1050 1 189.54 u2f0t19.spe U 185.7 no 18 3 1050 1 189.67 u2f0t20.spe U 185.7 no 18 10 1050 1 188.53 u2f0t21.spe U 185.7 no 18 10 1050 1 188.45 u2f0t22.spe U 185.7 no 18 10 1050 1 189.12 u2f0t23.spe U 185.7 no 18 10 1050 1 189.13 u2f0t24.spe U 185.7 no 18 10 1050 1 189.40 u2f0t25.spe U 185.7 no 18 10 1050 1 189.46 u2f0t26.spe U 185.7 no 18 10 1050 1 188.88 u2f0t27.spe U 185.7 no 18 10 1050 1 188.67 u2f0t28.spe U 185.7 no 18 10 1050 1 188.54 u2f0t29.spe U 185.7 no 18 10 1050 1 188.67 u2f0t30.spe U 185.7 no 18 180 1050 1 188.90 u2f3t40.spe U 185.7 3, Fe 16.5 60 460 1 189.05 u2f5t50.spe U 185.7 5, Fe 15.5 300 850 1 188.92 u2p6t60.spe U 185.7 6, Pb 10.5 1200 370 0.5 - backu201.spe no - 300 200 0.2

34

U series u3 Source: PERLA U50 UO2 powder, 1751 g U, 19.88wt% 235U, container wall: Al, 2 mm thick, plane Rossendorf) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 31-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb mm s cps % channnel mon019of.spe 137Cs 661.6 no - 100 1800 2 651.08 mon020on.spe 137Cs 661.6 no - 100 1800 2 660.58 backu300.spe no - 300 200 0.3 u3f0t10.spe U 185.7 no 54 3 1400 0.5 188.45 u3f0t11.spe U 185.7 no 54 3 1400 0.5 187.76 u3f0t12.spe U 185.7 no 54 3 1400 0.5 188.96 u3f0t13.spe U 185.7 no 54 3 1400 0.5 188.15 u3f0t14.spe U 185.7 no 54 3 1400 0.5 188.31 u3f0t15.spe U 185.7 no 54 3 1400 0.5 187.87 u3f0t16.spe U 185.7 no 54 3 1400 0.5 188.35 u3f0t17.spe U 185.7 no 54 3 1400 0.5 188.53 u3f0t18.spe U 185.7 no 54 3 1400 0.5 187.93 u3f0t19.spe U 185.7 no 54 3 1400 0.5 187.77 u3f0t20.spe U 185.7 no 54 10 1400 0.5 188.52 u3f0t21.spe U 185.7 no 54 10 1400 0.5 188.93 u3f0t22.spe U 185.7 no 54 10 1400 0.5 188.20 u3f0t23.spe U 185.7 no 54 10 1400 0.5 188.76 u3f0t24.spe U 185.7 no 54 10 1400 0.5 188.89 u3f0t25.spe U 185.7 no 54 10 1400 0.5 188.56 u3f0t26.spe U 185.7 no 54 10 1400 0.5 188.44 u3nf0t27.spe U 185.7 no 54 10 1400 0.5 188.09 u3f0t28.spe U 185.7 no 54 10 1400 0.5 188.05 u3f0t29.spe U 185.7 no 54 10 1400 0.5 188.37 u3f0t30.spe U 185.7 no 54 180 1400 0.5 188.40 u3f3t40.spe U 185.7 3, Fe 48 60 1250 0.7 188.41 u3f5t50.spe U 185.7 5, Fe 44 300 1250 0.7 187.59 u3p6t60.spe U 185.7 6, Pb 27.5 1200 600 0 - backu301.spe no - 300 200 0

35

U series u4 Source: PERLA1, pl30; Al-UO2-Al sandwich plate 625 * 70.75 * 1.27mm3, 61.05g U, 19.88wt% 235U, active zone (500...600) * (60.4......65.1) * (...0.3...) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 10-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon021of.spe 137Cs 661.6 no 10 100 1800 2 672.91 mon022on.spe 137Cs 661.6 no 10 100 1800 2 663.31 backu400.spe no - 300 220 2.4 u4nf0t10.spe U 185.7 no 24 3 2400 2.40 188.96 u4nf0t11.spe U 185.7 no 24 3 2400 2.4 189.43 u4nf0t12.spe U 185.7 no 24 3 2400 2.4 188.51 u4nf0t13.spe U 185.7 no 24 3 2400 2.4 188.59 u4nf0t14.spe U 185.7 no 24 3 2400 2.4 189.65 u4nf0t15.spe U 185.7 no 24 3 2400 2.4 188.64 u4nf0t16.spe U 185.7 no 24 3 2400 2.4 189.25 u4nf0t17.spe U 185.7 no 24 3 2400 2.4 189.00 u4nf0t18.spe U 185.7 no 24 3 2400 2.4 188.76 u4nf0t19.spe U 185.7 no 24 3 2400 2.4 188.48 u4nf0t20.spe U 185.7 no 24 10 2400 2.4 188.60 u4nf0t21.spe U 185.7 no 24 10 2400 2.4 188.80 u4nf0t22.spe U 185.7 no 24 10 2400 2.4 188.63 u4nf0t23.spe U 185.7 no 24 10 2400 2.4 188.59 u4nf0t24.spe U 185.7 no 24 10 2400 2.4 188.95 u4nf0t25.spe U 185.7 no 24 10 2400 2.4 188.91 u4nf0t26.spe U 185.7 no 24 10 2400 2.4 188.95 u4nf0t27.spe U 185.7 no 24 10 2400 2.4 188.99 u4nf0t28.spe U 185.7 no 24 10 2400 2.4 188.92 u4nf0t29.spe U 185.7 no 24 10 2400 2.4 188.85 u4nf0t31.spe U 185.7 no 24 180 2400 2.4 188.90 u4nf3t41.spe U 185.7 3, Fe 16.5 60 2250 2.4 188.61 u4nf5t51.spe U 185.7 5, Fe 13 300 2300 2.4 188.87 u4np6t61.spe U 185.7 6, Pb 0.6 1200 4600 6.2 188.46 backu401.spe no - 300 200 0.3

36

U series u5 Source: PERLA U128 UO2 powder, 1667 g U, 92.42% 235U, container wall: Al, 2 mm thick, plane Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 21-April-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon023of.spe 137Cs 661.6 no 10 100 1900 2.2 656.50 mon024on.spe 137Cs 661.6 no 10 100 1900 2.2 661.57 backu500.spe no - 300 200 0 u5f0t10.spe U 185.7 no 79 3 1800 1.6 u5f0t11.spe U 185.7 no 79 3 1800 1.6 200.8 u5f0t12.spe U 185.7 no 79 3 1800 1.6 201.0 u5f0t13.spe U 185.7 no 79 3 1800 1.6 201.2 u5f0t14.spe U 185.7 no 79 3 1800 1.6 201.5 u5f0t15.spe U 185.7 no 79 3 1800 1.6 201.0 u5f0t16.spe U 185.7 no 79 3 1800 1.6 201.7 u5f0t17.spe U 185.7 no 79 3 1800 1.6 200.6 u5f0t18.spe U 185.7 no 79 3 1800 1.6 201.0 u5f0t19.spe U 185.7 no 79 3 1800 1.6 200.7 u5f0t20.spe U 185.7 no 79 10 1800 1.6 201.3 u5f0t21.spe U 185.7 no 79 10 1800 1.6 201.1 u5f0t22.spe U 185.7 no 79 10 1800 1.6 200.8 u5f0t23.spe U 185.7 no 79 10 1800 1.6 201.6 u5f0t24.spe U 185.7 no 79 10 1800 1.6 200.8 u5f0t25.spe U 185.7 no 79 10 1800 1.6 200.1 u5f0t26.spe U 185.7 no 79 10 1800 1.6 200.7 u5f0t27.spe U 185.7 no 79 10 1800 1.6 201.0 u5f0t28.spe U 185.7 no 79 10 1800 1.6 200.0 u5f0t29.spe U 185.7 no 79 10 1800 1.6 200.7 u5f0t30.spe U 185.7 no 79 180 1800 1.6 200.9 u5f3t40.spe U 185.7 3, Fe 66 60 1100 2 200.1 u5f5t50.spe U 185.7 5, Fe 57 300 1900 2 200.5 u5p6t60.spe U 185.7 6, Pb 17 1200 1200 2 203.8 backu501.spe no - 300 30 0

37

U series u6 Source: PERLA 3, pl83; Al-UO2-Al sandwich plate 625 * 70.75 * 1.27mm3, 8.37g U, 93.11wt% 235U, active zone (500...600) * (60.4......65.1) * (...0.3...) Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 21-April-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon025of.spe 137Cs 661.6 no 10 100 1900 2 670.52 mon026on.spe 137Cs 661.6 no 10 100 1900 2 662.96 backu600.spe no - 300 210 0.3 u6f0t10.spe U 185.7 no 16 3 2300 2 188.86 u6f0t11.spe U 185.7 no 16 3 2300 2 188.95 u6f0t12.spe U 185.7 no 16 3 2300 2 188.91 u6f0t13.spe U 185.7 no 16 3 2300 2 188.39 u6f0t14.spe U 185.7 no 16 3 2300 2 188.93 u6f0t15.spe U 185.7 no 16 3 2300 2 188.95 u6f0t16.spe U 185.7 no 16 3 2300 2 188.68 u6f0t17.spe U 185.7 no 16 3 2300 2 188.87 u6f0t18.spe U 185.7 no 16 3 2300 2 188.67 u6f0t19.spe U 185.7 no 16 3 2300 2 188.29 u6f0t20.spe U 185.7 no 16 10 2300 2 189.00 u6f0t21.spe U 185.7 no 16 10 2300 2 188.93 u6f0t22.spe U 185.7 no 16 10 2300 2 188.54 u6f0t23.spe U 185.7 no 16 10 2300 2 188.94 u6f0t24.spe U 185.7 no 16 10 2300 2 188.54 u6f0t25.spe U 185.7 no 16 10 2300 2 188.56 u6f0t26.spe U 185.7 no 16 10 2300 2 188.88 u6f0t27.spe U 185.7 no 16 10 2300 2 188.61 u6f0t28.spe U 185.7 no 16 10 2300 2 188.32 u6f0t29.spe U 185.7 no 16 10 2300 2 188.57 u6f0t30.spe U 185.7 no 16 180 2300 2 188.54 u6f3t40.spe U 185.7 3, Fe 10 60 2300 2 188.81 u6f5t50.spe U 185.7 5, Fe 7 300 2600 2 188.63 u6p6t60.spe U 185.7 6, Pb 0.6 1200 2000 2 188.60 backu601.spe no - 300 200 0.3

38

Pu series p0 Source: CBNM Pu61, PuO2 pellet, 6.626g Pu, thin metal window, 1mm PVC Pu 238-242 (6-Jun-02): 1.097%, 64.968%, 26.366%, 3.212%, 4.358%, Am/Pu:5.196% Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 26-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel NaI detector spectra: mon027of.spe 137Cs 661.6 no 10 100 1800 2.2 681.11 mon028on.spe 137Cs 661.6 no 10 100 1800 2.2 659.57 backp000.spe no - 300 200 0.3 p0f0t10.spe Pu 208 no 354 3 7500 3 - p0f0t11.spe Pu 208 no 354 3 7500 3 - p0f0t12.spe Pu 208 no 354 3 7500 3 - p0f0t13.spe Pu 208 no 354 3 7500 3 - p0f0t14.spe Pu 208 no 354 3 7500 3 - p0f0t15.spe Pu 208 no 354 3 7500 3 - p0f0t16.spe Pu 208 no 354 3 7500 3 - p0f0t17.spe Pu 208 no 354 3 7500 3 - p0f0t18.spe Pu 208 no 354 3 7500 3 - p0f0t19.spe Pu 208 no 354 3 7500 3 - p0f0t20.spe Pu 208 no 354 10 7500 3 210.1 p0f0t21.spe Pu 208 no 354 10 7500 3 209.5 p0f0t22.spe Pu 208 no 354 10 7500 3 212.79 p0f0t23.spe Pu 208 no 354 10 7500 3 209.74 p0f0t24.spe Pu 208 no 354 10 7500 3 208.6 p0f0t25.spe Pu 208 no 354 10 7500 3 203.79 p0f0t26.spe Pu 208 no 354 10 7500 3 205.66 p0f0t27.spe Pu 208 no 354 10 7500 3 214.00 p0f0t28.spe Pu 208 no 354 10 7500 3 210.02 p0f0t29.spe Pu 208 no 354 10 7500 3 206.41 p0f0t30.spe Pu 208 no 354 180 7500 3 211.36 p0f3t40.spe Pu 208 3, Fe 110.5 60 7800 3 210.35 p0f5t51.spe Pu 208 5, Fe 71 300 6000 3 210.43 p0p6t60.spe Pu 208 6, Pb 35 1200 2000 4 210.77

39

Pu series p1 Source: PERLA06, PuO2 powder, 7.939g Pu, 2 steel walls 1.6mm + 1mm thick, Pu238-242 (6-Jun-02): 1.618%, 61.252%, 26.084%, 5.092%, 5.954%, Am:6.681% Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 24-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon028of.spe 137Cs 661.6 no 10 100 1800 2.3 655.42 mon029on.spe 137Cs 661.6 no 10 100 1800 2.3 661.42 backp100.spe no - 300 300 0 p1f0t10.spe Pu 208 no 374 3 2600 9 211.88 p1f0t11.spe Pu 208 no 374 3 2600 9 206.00 p1f0t12.spe Pu 208 no 374 3 2600 9 212.69 p1f0t13.spe Pu 208 no 374 3 2600 9 208.15 p1f0t14.spe Pu 208 no 374 3 2600 9 212.94 p1f0t15.spe Pu 208 no 374 3 2600 9 205.89 p1f0t16.spe Pu 208 no 374 3 2600 9 207.56 p1f0t17.spe Pu 208 no 374 3 2600 9 208.31 p1f0t18.spe Pu 208 no 374 3 2600 9 211.50 p1f0t19.spe Pu 208 no 374 3 2600 9 208.55 p1f0t20.spe Pu 208 no 374 10 2600 9 209.50 p1f0t21.spe Pu 208 no 374 10 2600 9 210.90 p1f0t22.spe Pu 208 no 374 10 2600 9 211.73 p1f0t23.spe Pu 208 no 374 10 2600 9 216.22 p1f0t24.spe Pu 208 no 374 10 2600 9 208.16 p1f0t25.spe Pu 208 no 374 10 2600 9 211.13 p1f0t26.spe Pu 208 no 374 10 2600 9 215.75 p1f0t27.spe Pu 208 no 374 10 2600 9 211.36 p1f0t28.spe Pu 208 no 374 10 2600 9 207.16 p1f0t29.spe Pu 208 no 374 10 2600 9 219.15 p1f0t30.spe Pu 208 no 374 180 2600 9 210.97 p1f3t40.spe Pu 208 3, Fe 161 60 2200 7 209.48 p1f5t50.spe Pu 208 5, Fe 127 300 1800 6.3 209.64 p1p6t60.spe Pu 208 6, Pb 91 1200 1600 1.2 209.84 backp101.spe no - 300 200 0.3 -

40

Pu series p2 Source: PERLA01, PuO2 powder, Pu mass 8.009g, 2 steel walls 1.6mm + 1mm thick, Pu238-242 (6-Jun-02): 0.181%, 72.19%, 24.983%, 1.654%, 0.993%, Am/Pu: 2.708% Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 11-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channel mon030of.spe 137Cs 661.6 no 10 100 1800 2 668.56 mon031on.spe 137Cs 661.6 no 10 100 1800 2 662.97 backp200.spe no - 300 200 0.3 p2f0t10.spe Pu 208 no 200 3 2600 9 210.33 p2f0t11.spe Pu 208 no 200 3 2600 9 213.65 p2f0t12.spe Pu 208 no 200 3 2600 9 213.44 p2f0t13.spe Pu 208 no 200 3 2600 9 210.79 p2f0t14.spe Pu 208 no 200 3 2600 9 210.39 p2f0t15.spe Pu 208 no 200 3 2600 9 212.61 p2f0t16.spe Pu 208 no 200 3 2600 9 206.78 p2f0t17.spe Pu 208 no 200 3 2600 9 203.72 p2f0t18.spe Pu 208 no 200 3 2600 9 209.71 p2f0t19.spe Pu 208 no 200 3 2600 9 208.83 p2f0t20.spe Pu 208 no 200 10 2600 9 212.69 p2f0t21.spe Pu 208 no 200 10 2600 9 215.29 p2f0t22.spe Pu 208 no 200 10 2600 9 212.54 p2f0t23.spe Pu 208 no 200 10 2600 9 213.34 p2f0t24.spe Pu 208 no 200 10 2600 9 211.59 p2f0t25.spe Pu 208 no 200 10 2600 9 211.08 p2f0t26.spe Pu 208 no 200 10 2600 9 209.86 p2f0t27.spe Pu 208 no 200 10 2600 9 207.42 p2f0t28.spe Pu 208 no 200 10 2600 9 211.84 p2f0t29.spe Pu 208 no 200 10 2600 9 210.32 p2f0t30.spe Pu 208 no 200 180 2600 9 211.45 p2f3t40.spe Pu 208 3, Fe 80 60 900 6 212.17 p2f5t50.spe Pu 208 5, Fe 65 300 600 7.4 211.46 p2p6t60.spe Pu 208 6, Pb 40 1200 600 1.6 217.01 backp201.spe no - 300 200 0

41

Pu series p3 Source: CBNM Pu93, PuO2 pellet, 6.625g Pu, thin metal window, 1mm PVC Pu238-242 (6-Jun-02): 0.010%, 93.533%, 6.314%, 0.104%, 0.040%, Am/Pu: 0.223% Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 21-April-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon032of.spe 137Cs 661.6 no 10 100 2200 2.4 659.08 mon033on.spe 137Cs 661.6 no 10 100 2200 2.4 660.92 backp300.spe no - 300 30 0 p3f0t10.spe Pu 208 no 86 3 2300 1.8 - p3f0t11.spe Pu 208 no 86 3 2300 1.8 - p3f0t12.spe Pu 208 no 86 3 2300 1.8 - p3f0t13.spe Pu 208 no 86 3 2300 1.8 - p3f0t14.spe Pu 208 no 86 3 2300 1.8 - p3f0t15.spe Pu 208 no 86 3 2300 1.8 - p3f0t16.spe Pu 208 no 86 3 2300 1.8 - p3f0t17.spe Pu 208 no 86 3 2300 1.8 - p3f0t18.spe Pu 208 no 86 3 2300 1.8 - p3f0t19.spe Pu 208 no 86 3 2300 1.8 - p3f0t20.spe Pu 208 no 86 10 2300 1.8 210.92 p3f0t21.spe Pu 208 no 86 10 2300 1.8 208.24 p3f0t22.spe Pu 208 no 86 10 2300 1.8 217.18 p3f0t23.spe Pu 208 no 86 10 2300 1.8 - p3f0t24.spe Pu 208 no 86 10 2300 1.8 204.25 p3f0t25.spe Pu 208 no 86 10 2300 1.8 - p3f0t26.spe Pu 208 no 86 10 2300 1.8 222.91 p3f0t27.spe Pu 208 no 86 10 2300 1.8 - p3f0t28.spe Pu 208 no 86 10 2300 1.8 - p3f0t29.spe Pu 208 no 86 10 2300 1.8 211.70 p3f0t30.spe Pu 208 no 86 180 2300 1.8 209.43 p3f3t40.spe Pu 208 3, Fe 37 60 1000 1 206.75 p3f5t50.spe Pu 208 5, Fe 30 300 950 0.9 209.97 p3p6t60.spe Pu 208 6, Pb 9.5 1200 470 0.5 - backp301.spe no - 300 200 0.3

42

Pu series p4 Source:PuGa no.32, Pu+1.5%Ga alloy, 9.4g Pu, 0.6mm thick disc, 1 wall 0.5 mmFe Pu238...242 (6-Jun-02): 0.011%, 93.922%, 5.901%, 0.134%, 0.032%, Am/Pu: 0.046% Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 24-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb mm s cps % channnel mon034of.spe 137Cs 661.6 no 10 100 1800 2 659.64 mon035on.spe 137Cs 661.6 no 10 100 1800 2 661.86 backp400.spe no - 300 200 0 p4f0t10.spe Pu 208 no 194 3 2800 3 209.1 p4f0t11.spe Pu 208 no 194 3 2800 3 222.4 p4f0t12.spe Pu 208 no 194 3 2800 3 217.6 p4f0t13.spe Pu 208 no 194 3 2800 3 227.3 p4f0t14.spe Pu 208 no 194 3 2800 3 222.3 p4f0t15.spe Pu 208 no 194 3 2800 3 218.4 p4f0t16.spe Pu 208 no 194 3 2800 3 217.9 p4f0t17.spe Pu 208 no 194 3 2800 3 225.8 p4f0t18.spe Pu 208 no 194 3 2800 3 210.9 p4f0t19.spe Pu 208 no 194 3 2800 3 215.4 p4f0t20.spe Pu 208 no 194 10 2800 3 221.2 p4f0t21.spe Pu 208 no 194 10 2800 3 222.5 p4f0t22.spe Pu 208 no 194 10 2800 3 218.3 p4f0t23.spe Pu 208 no 194 10 2800 3 218.8 p4f0t24.spe Pu 208 no 194 10 2800 3 220.5 p4f0t25.spe Pu 208 no 194 10 2800 3 214.2 p4f0t26.spe Pu 208 no 194 10 2800 3 222.2 p4f0t27.spe Pu 208 no 194 10 2800 3 222.8 p4f0t28.spe Pu 208 no 194 10 2800 3 220.1 p4f0t29.spe Pu 208 no 194 10 2800 3 218.2 p4f0t30.spe Pu 208 no 194 180 2800 3 221.1 p4f3t40.spe Pu 208 3, Fe 34 60 12400 12 220.2 p4f5t50.spe Pu 208 5, Fe 31 300 5900 6 220.1 p4p6t60.spe Pu 208 6, Pb 13 1200 4900 4 220.4 backp401.spe no - 300 200 0.3

43

Pu series p5, source PM2 18.8g Pu+5%Ni+3% Cu alloy. 238-242Pu wt% 6 Jun02: 0.004, 95.470, 4.458, 0.053, 0.015, walls 3mm Al+1 mm Fe Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 31-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps (%) channnel mon038of.spe 137Cs 661.6 no 10 100 1800 2.2 657.40 mon039on.spe 137Cs 661.6 no 10 100 1800 2.2 661.04 backp500.spe no - 300 30 0.3 p5f0t10.spe U ore 208 no 59 3 2800 1 - p5f0t11.spe U ore 208 no 59 3 2800 1 212.10 p5f0t12.spe U ore 208 no 59 3 2800 1 204.00 p5f0t13.spe U ore 208 no 59 3 2800 1 195.65 p5f0t14.spe U ore 208 no 59 3 2800 1 - p5f0t15.spe U ore 208 no 59 3 2800 1 - p5f0t16.spe U ore 208 no 59 3 2800 1 216.25 p5f0t17.spe U ore 208 no 59 3 2800 1 210.85 p5f0t18.spe U ore 208 no 59 3 2800 1 - p5f0t19.spe U ore 208 no 59 3 2800 1 - p5f0t20.spe U ore 208 no 59 10 2800 1 213.71 p5f0t21.spe U ore 208 no 59 10 2800 1 - p5f0t22.spe U ore 208 no 59 10 2800 1 207.87 p5f0t23.spe U ore 208 no 59 10 2800 1 216.68 p5f0t24.spe U ore 208 no 59 10 2800 1 214.94 p5f0t25.spe U ore 208 no 59 10 2800 1 209.14 p5f0t26.spe U ore 208 no 59 10 2800 1 - p5f0t27.spe U ore 208 no 59 10 2800 1 - p5f0t28.spe U ore 208 no 59 10 2800 1 219.85 p5f0t29.spe U ore 208 no 59 10 2800 1 226.09 p5f0t30.spe U ore 208 no 59 180 2800 1 210.93 p5f3t40.spe U ore 208 3, Fe 34 60 1700 0.3 208.63 p5f5t50.spe U ore 208 5, Fe 31 300 1400 0.5 210.39 p5p6t60.spe U ore 208 6, Pb 17 1200 1000 0.1 - backp501.spe no - 300 200 0.3

44

Th powder Source:Perla 381T2 - Natural Thorium powder, 1500g Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 24-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position keV Fe/Pb cm s cps % channnel mon036of.spe 137Cs 661.6 no - 100 1800 2 653.54 mon037on.spe 137Cs 661.6 no - 100 1800 2 660.66 backth00.spe no - 300 200 0.3 thf0t10.spe Th 2614.5 no 194 3 550 0.7 - thf0t11.spe Th 2614.5 no 194 3 550 0.7 2495 thf0t12.spe Th 2614.5 no 194 3 550 0.7 2457 thf0t13.spe Th 2614.5 no 194 3 550 0.7 2408 thf0t14.spe Th 2614.5 no 194 3 550 0.7 2500 thf0t15.spe Th 2614.5 no 194 3 550 0.7 2448 thf0t16.spe Th 2614.5 no 194 3 550 0.7 2462 thf0t17.spe Th 2614.5 no 194 3 550 0.7 2490 thf0t18.spe Th 2614.5 no 194 3 550 0.7 2391 thf0t19.spe Th 2614.5 no 194 3 550 0.7 2395 thf0t20.spe Th 2614.5 no 194 10 550 0.7 2438 thf0t21.spe Th 2614.5 no 194 10 550 0.7 2527 thf0t22.spe Th 2614.5 no 194 10 550 0.7 2490 thf0t23.spe Th 2614.5 no 194 10 550 0.7 2506 thf0t24.spe Th 2614.5 no 194 10 550 0.7 2504 thf0t25.spe Th 2614.5 no 194 10 550 0.7 - thf0t26.spe Th 2614.5 no 194 10 550 0.7 2502 thf0t27.spe Th 2614.5 no 194 10 550 0.7 2479 thf0t28.spe Th 2614.5 no 194 10 550 0.7 2509 thf0t29.spe Th 2614.5 no 194 10 550 0.7 2465 thf0t30.spe Th 2614.5 no 194 180 550 0.7 2469 thf3t40.spe Th 2614.5 3, Fe 184 60 430 2 2497 thf5t50.spe Th 2614.5 5, Fe 174 300 470 0.7 2532.0 thp6t60.spe Th 2614.5 6, Pb 154 1200 770 1.1 2517.0 backth01 no - 300 200 0

45

137Cs plus Pu Sources: 1. PERLA111, PO2 powder, Pu mass 999.8g, 2 steel walls 2.6mm + 1.5mm thick, Pu238-242 (6-Jun-02): 0.181%, 72.19%, 24.983%, 1.654%, 0.993%, Am/Pu: 2.708% 2. 137Cs, 370MBq Detector: LaBr3 BrilLanCe-380 Crismatec no. J900CSB MCA: MCA-166 (GBS Rossendorf) 31-May-06 Dose rate at the detector position about 0.5 microSievert Spectrum Time Isotope E Filter Distan Meas Count Dead Peak name mm, ce time rate time position hh:mm keV Fe/Pb cm s cps % channnel Pu.spe 23:46 Pu ... 25mm Pb - 300 800 0.5 - PuCs.spe 23:54 137Cs 661.6 43mm Pb - 300 1050 1 677.04 Pu ... 25mm Pb - Cs.spe 17:56 137Cs 661.6 43mm Pb - 300 500 0.5 676.90

46

Annex III: Energy calibration - list of spectra Co57.spe Co60.spe Ba133.spe Cs137.spe Th.spe

Feasibility Study of a Wrist Watch Based Radiation Survey Meter (Phase I) and Isotope Identifying Gamma Spectrometer (Phase II)

A. Wolf

FZD Research Center Dresden — Rossendorf, Germany During the Period from March 2003 through April 2006, FZR Research Center Rossendorf had a Cooperative Research Agreement (CRA) with the IAEA. The Cooperative Research Agreement on “Feasibility study of a wrist watch based radiation survey meter (Phase I) and isotope identifying gamma spectrometer (Phase II)” was intended to give an substantial effort in the supporting and improving of the technical capabilities to detect and respond to illicit trafficking of nuclear and other materials and the detection of radiological threat by studying the possibilities of a new class of detection devices, Spectroscopic Personal Radiation Detectors. This work was funded by FZR. The long-range objective of the effort is to contribute to the improvement of the technical means by combining radiation detection with categorization of the nuclides. This allows to reduce the response to innocent alarms caused often in populated areas by radio-pharmaceuticals in persons. This is achieved by continuous, automated measurement of gamma spectra and their evaluation with respect to the nuclide that had caused an alarm. During the first phase March 2003 to Dec 2003 we studied the possibility of the simulation and modeling of gamma spectra using a heuristic method. The results were reported by Jörg Brutscher on the Research Co-ordination & Technical Meeting IAEA Headquarter Vienna, 1–5 Dec 2003. As result an improved model for the Compton background was presented. This includes the efficiency-corrected Klein-Nishina distribution for scattering inside the detector as well as an approach to describe backscattering from surroundings. Based on modeling results and test measurements, a medium-size CZT detector was selected for use in this project. Beginning in March 2003 a feasibility study off an ultra miniaturized very low power radiation survey meter in shape of a wrist watch was started by component and detector selection, packaging studies, development of very low power high density electronic circuits and low power processor with enough power to allow measurement and control data processing (isotope identification). In addition an algorithm for a linearization of the dose rate indication based on the spectrometric information was developed and tested. A first prototype of a wrist watch based radiation survey meter with a room temperature planar solid state detector was completed in June 2004 and was presented at the Research Coordination Meeting in Sochi, Russian Federation, Oct 2004. During the Olympic Games in Athen 2004 and the Soccer World Championship Germany 2006 four field test samples of the radiation survey meter were made available to the respective support team of the IAEA. And a first prototype test was also done in cooperation with the Institute for Systems, Informatics and Safety of the Joint Research Centre; Ispra Italy in May 2005.

In the second phase starting in July 2005 the feasibility study of a wrist watch based Spectral Personal Radiation Detector with internal Isotope Identification has been performed. The selection of suitable spectroscopic solid state detectors was carried out and advanced pulse processing and data acquisition solutions were developed and implemented. In Cooperation with the Atom Institute of the Austrian Universities (ATI) in Vienna an isotope identification algorithm with low requirements to the numeric processing power and for use with spectral data with low statistics is under development. First results were presented at the third Co-ordination & Technical Meeting IAEA at Headquarter Vienna, 23–27April 2006, Vienna by Andreas Wolf and Martha Swoboda. In cooperation with the AIT and the IAEA a catalog of gamma spectra of single nuclides which are used for medical and industrial purposes, natural occurring nuclides and nuclear material as well as mixtures of this nuclide categories was established. It is used, supplemented by a catalog of simulated gamma spectra with low counting statistics for testing of the isotope identification algorithm. A first prototype of a wrist watch based Spectroscopic Personal Radiation Detector (SPRD) is expected at the end of 2006, based on following hardware/firmware features: - spectroscopic CZT detector 10x10x5 mm3 with resolution FWHM 25keV for Cs137 - low power main and shaping amplifier with peak detection and low power multi-channel

analyzer with 2 k channels - suppression of microphonic noise - spectrum memory for at least 10 spectra - battery life time about 24 hours with full measuring function - alarm types: radiation alarm and safety alarm with adjustable thresholds - working in automated SPRD (Spectral Personal Radiation Detector) or manual RID

(Radionuclide Identification Device) mode for expert users - indication of dose rate and accumulated dose - identification of single, unshielded nuclides with doserate 1µSv/h above background in

300 sec in energy range up to 800 keV, and in 600 sec in energy range to 1.6 MeV - identification of very weak sources in the expert mode with a further extension of

measurement time - drop test of 1m high in shipping case

Picture 1: First prototype of WatchMon, showing finder screen for industrial nuclides.

Picture 2: First prototype of WatchMon, showing identify screen for mix of industrial

nuclides.

Procedures for, and Testing of, Border Monitoring Equipment

R. Kouzes Pacific Northwest National Laboratory

During the period from March 2003 through April 2006, Richard Kouzes of Pacific Northwest National Laboratory had a Cooperative Research Agreement (CRA) with the IAEA. The Cooperative Research Agreement on “Procedures for and Testing of Border Monitoring Equipment” states the purpose of this activity was as follows:

The IAEA, the EC, and Member States purchase border monitoring equipment to support Member States in improving their capabilities to detect and respond to illicit trafficking of nuclear and other radioactive materials. To make sure that the correct equipment is purchased and deployed, specifications and associated test procedures are required. Assistance in developing such equipment specifications is included under this CPR. Validated test laboratories are needed to be available for performing the tests for equipment according to the equipment type and outlined test procedures. Under this agreement, PNNL will be assisting in drafting detailed, engineering level test procedures to be validated and accepted by the IAEA. This effort is contingent upon PNNL obtaining appropriate support.

The long-range objective of the effort is to contribute to the improvement of equipment to detect illicit trafficking of nuclear and other radioactive materials. The equipment is needed for use by law enforcement services to characterize such materials before and after seizure. The outcome of the work will help States in improving their capabilities to combat illicit trafficking through enhanced detection systems. At the Consultants Group Meeting, Vienna, 17–21 March 2003, the draft "Technical/Functional Specifications for Border Monitoring Equipment" was discussed. PNNL will attempt to participate in follow-up meetings on the further review of this specification.

The purpose of this Cooperative Research Agreement was to allow for the sharing of knowledge on the development and testing of border security equipment for the interdiction of illicit trafficking in nuclear and other radioactive material. The Radiation Portal Monitor Project at Pacific Northwest National Laboratory has been deploying equipment for the interdiction of illicit radioactive materials at the U.S border since 2002. This work in the U.S. has been of direct interest and value to the IAEA’s international programs in border security. The U.S. Customs and Border Protection, part of the Department of Homeland Security, approved of, and provided financial support for, this activity with the IAEA. During the time period of this CRA, PNNL wrote a number of open documents that were shared with the IAEA. A bibliography of relevant reports and a few key abstracts are provided below.

Bibliography of Related Reports Co-Authored by Richard Kouzes DISCRIMINATION OF NATURALLY OCCURRING RADIOACTIVE MATERIAL IN PLASTIC SCINTILLATOR MATERIAL James Ely, Richard Kouzes, Bruce Geelhood, John Schweppe, Ray Warner, IEEE Transactions on Nuclear Science NS 51 (August 2004), p 1672–1676.

Abstract – Plastic scintillator material is used in many applications for the detection of gamma rays from radioactive material, primarily due to the sensitivity per unit cost compared to other detection materials. However, the resolution and lack of full-energy peaks in the plastic scintillator material prohibits detailed spectroscopy. Therefore, other materials such as doped sodium iodide are used for spectroscopic applications. The limited spectroscopic information can, however, be exploited in plastic scintillator materials to provide some discrimination. The discrimination between man-made and naturally occurring sources would be useful in reducing alarm screening for radiation detection applications that target man-made sources. The results of applying the limited energy information from plastic scintillator material for radiation portal monitors are discussed.

HOMELAND SECURITY INSTRUMENTATION FOR RADIATION DETECTION AT BORDERS Richard Kouzes, James Ely, Randy Hansen, John Schweppe, Edward Siciliano, David Stromswold, Proceedings of the Fourth American Nuclear Society International Topical Meeting on Nuclear Plant Instrumentation, Controls and Human-Machine Interface Technologies (NPIC&HMIT 2004), Columbus, OH. ISBN 0-89448-688-8 (September 2004).

Abstract – Countries around the world are deploying radiation detection instrumentation to interdict the illegal shipment of radioactive material crossing international borders at land, rail, air, and sea ports of entry. These efforts include deployments in the US and a number of European and Asian countries by governments and international agencies. Items of concern include radiation dispersal devices (RDD), nuclear warheads, and special nuclear material (SNM). Radiation portal monitors (RPMs) are used as the main screening tool for vehicles and cargo at borders, supplemented by handheld detectors, personal radiation detectors, and x-ray imaging systems. Some cargo contains naturally occurring radioactive material (NORM) that triggers “nuisance” alarms in RPMs at these border crossings. Individuals treated with medical radiopharmaceuticals also produce nuisance alarms and can produce cross-talk between adjacent lanes of a multi-lane deployment. The operational impact of nuisance alarms can be significant at border crossings. Methods have been developed for reducing this impact without negatively affecting the requirements for interdiction of radioactive materials of interest. Plastic scintillator material is commonly used in RPMs for the detection of gamma rays from radioactive material, primarily due to the efficiency per unit cost compared to other detection materials. The resolution and lack of full-energy peaks in the plastic scintillator material prohibits detailed spectroscopy. However, the limited spectroscopic information from plastic scintillator can be exploited to provide some discrimination. Energy-based algorithms used in RPMs can effectively exploit the crude energy information available from a plastic scintillator to distinguish some NORM from items of concern. Whenever NORM cargo limits the level of the alarm threshold, energy-based algorithms can produce better detection probabilities for small SNM sources than gross-count algorithms. This paper discusses lessons learned from the use of plastic-based RPMs at borders. Results of observations of NORM and computations related to NORM characteristics are discussed as is the use of energy-based algorithms for NORM rejection.

COMPARISON OF PLASTIC AND NAI(TL) SCINTILLATORS FOR VEHICLE PORTAL MONITOR APPLICATIONS E.R. Siciliano, J.H. Ely, R.T. Kouzes, B.D. Milbrath, J.E. Schweppe, D.C. Stromswold, NIM-A 550, pp. 647–674 (September 21, 2005).

Abstract – The demand for radiation portal monitor (RPM) systems has increased, and their capabilities are being further scrutinized as they are being applied to the task of detecting nuclear weapons, special nuclear material, and radiation dispersal device materials that could appear at borders. The requirements and constraints on RPM systems deployed at high-volume border

crossings are significantly different from those at weapons facilities or steel recycling plants, where RPMs have been historically employed. In this new homeland security application, RPM systems must rapidly detect localized sources of radiation with a very high detection probability and low false-alarm rate, while screening all of the traffic without impeding the flow of commerce. In light of this new Department of Homeland Security application, the capabilities of two popular gamma-ray-detector materials as applied to these needs are re-examined. Both experimental data and computer simulations, together with practical deployment experience, are used to assess currently available polyvinyl-toluene and NaI(Tl) gamma-ray detectors for border applications.

DETECTING ILLICIT NUCLEAR MATERIALS R.T. Kouzes, PNNL-SA-43674, American Scientist 93, PP. 422–427, September–October 2005. THE USE OF ENERGY WINDOWING TO DISCRIMINATE SNM FROM NORM James Ely, Richard Kouzes, John Schweppe, Edward Siciliano, Denis Strachen, Dennis Weier, accepted to NIM-A, February 2006.

Abstract – Energy windowing is an algorithmic alarm method that can be applied to plastic scintillator-based radiation portal monitor (RPM) systems to improve operational sensitivity to certain threat sources while reducing the alarm rates from naturally occurring radioactive material. Various implementations of energy windowing have been tested and documented by industry and at Pacific Northwest National Laboratory, and are available in commercial RPMs built by several manufacturers. Moreover, energy windowing is being used in many deployed RPMs to reduce nuisance alarms and improve operational sensitivity during the screening of cargo. This paper describes energy windowing algorithms and demonstrates how these algorithms succeed when applied to “controlled” experimental measurements and “real world” vehicle traffic data.

THE RESPONSE OF RADIATION PORTAL MONITORS TO MEDICAL RADIONUCLIDES IN BORDER APPLICATIONS Richard Kouzes, Edward Siciliano, April 25, 2005; accepted by Radiation Measurements.

Abstract – Radiopharmaceuticals are administered to a large number of individuals each year, and at any given time can be found at detectable levels in about one in 2600 Americans. Such individuals are among the population crossing international borders where equipment has been deployed for radiation detection, and the resulting “nuisance” alarms produced by these individuals must be resolved by further inspection like all other alarms. During 2001, a total of approximately 14.4 million medical procedures using radionuclides were performed in the U.S. This number rose to over 16.5 million in 2004. Of this total number of procedures, approximately 98% were diagnostic procedures and 2% were therapeutic procedures. Although there were over 45 different commercially-available products used in over 75 different types of medical procedures, only 17 different radioisotopes comprise the complete set of active radionuclides in the commercially-available radiopharmaceuticals in the U.S. There are other experimental radiopharmaceuticals in use, but their presence in the population is rare. Of these 17 radioisotopes, nine are customarily administered to outpatients and produce sufficiently energetic photons that can readily be detected. The radioisotope 99mTc is administered in about 90% of the procedures and is the one most likely to be detected by radiation screening equipment. This paper reports on various impacts of radiopharmaceuticals observed in people at border crossings. Calculations were performed to simulate the response of multi-lane portal radiation monitor deployments to these types of sources. Results are given for the time period during which radiopharmaceuticals might be detectable by radiation screening equipment at border crossings.

NATURALLY OCCURRING RADIOACTIVE MATERIALS IN CARGO AT U.S. BORDERS Richard Kouzes, James Ely, John Evans, Walt Hensley, Elwood Lepel, Joseph McDonald, John Schweppe, Edward Siciliano, Dan Strom, Mitch Woodring, Packaging, Transport, Storage & Security of Radioactive Material, Vol. 17, No. 1 pp. 11–17 (January 2006).

Abstract – In the U.S. and other countries, large numbers of vehicles pass through border crossings each day. The illicit movement of radioactive sources is a concern that has resulted in

the installation of radiation detection and identification instruments at border crossing points. This activity is judged to be necessary because of the possibility of an act of terrorism involving a radioactive source that may include any number of dangerous radionuclides. The problem of detecting, identifying, and interdicting illicit radioactive sources is complicated by the fact that many materials present in cargo are somewhat radioactive. Some cargo contains naturally occurring radioactive material that may trigger radiation portal monitor alarms. Such nuisance alarms can be an operational limiting factor for screening of cargo at border crossings. Information about the nature of the radioactive materials in cargo that can interfere with the detection of radionuclides of concern is necessary to help anticipate and recognize likely sources of these nuisance alarms.

BASELINE SUPPRESSION OF PORTAL MONITOR VEHICLE GAMMA COUNT PROFILES: A CHARACTERIZATION STUDY Charles A. Lo Presti, Dennis R. Weier, Richard T. Kouzes, John Schweppe, September 2005. Accepted by NIM-A.

Abstract – Radiation portal monitor (RPM) systems based upon polyvinyl toluene scintillator (PVT) gamma ray detectors have been deployed to detect illicit trafficking in radioactive materials at border crossings. This report sets forth a characterization of the baseline suppression effect in gross-count gamma ray profiles due to shadow shielding by vehicles entering RPMs. Shadow shielding is of interest because it reduces the alarm sensitivity of RPMs. This observational study investigated three types of PVT based commercial RPM systems currently deployed at selected ports of entry in terms of spatial effects relative to detector panel positioning. Radiation portal monitor sites were characterized by driver versus passenger side, top versus bottom panel, and narrow lanes versus wide lanes as observed for a large number of vehicles. Each portal site appears to have a distinctive baseline suppression signature, based on percent maximum suppression relative to measured background. Results suggest that alarm algorithms based on gross-counts may be further refined through attention to individual site characteristics. In addition, longer vehicle transit times were often correlated with stronger baseline suppression, suggesting that baseline suppression studies should take transit time into account.

KEY ELEMENTS OF PREPARING EMERGENCY RESPONDERS FOR NUCLEAR AND RADIOLOGICAL TERRORISM John W. Poston, Sr., Steven M. Becker, Brooke Buddemeier, Jerrold T. Bushberg, John J. Cardarelli, W. Craig Conklin, Brian Dodd, John R. Frazier, Fun H. Fong, Jr., Ronald E. Goans, Ian S. Hamilton, Richard T. Kouzes, Jonathan M. Links, Phil L. Liotta, Fred A. Mettler, Jr., Terry C. Pellmar, Leticia S. Pibida, Michael J. Puzziferri, Carson A. Riland, Joseph P. Ring, Thomas M. Seed, James M. Smith, Robert C. Whitcomb, National Council on Radiation Protection and Measurements Commentary SC2-1, December 31, 2005. COMPARISON OF PLASTIC AND NAI(TL) SCINTILLATORS FOR VEHICLE PORTAL MONITOR APPLICATIONS D.C. Stromswold, E.R. Siciliano, J.E. Schweppe, J.H. Ely, B.D. Milbrath, R.T. Kouzes, B.D. Geelhood, W.K. Hensley, Nuclear Science Symposium Conference Record, 2003 IEEE Volume 2, 19–25 Oct. 2003 Page(s): 1065–1069 Vol. 2. DISCRIMINATION OF NATURALLY OCCURRING RADIOACTIVE MATERIAL IN PLASTIC SCINTILLATOR MATERIAL (also listed with refereed journals) James Ely, Richard Kouzes, Bruce Geelhood, John Schweppe, Ray Warner, Nuclear Science Symposium Conference Record, 2003 IEEE Volume 2, 19–25 Oct. 2003 Page(s): 1453–1457 Vol. 2. IEEE NSS Portland.

NATURALLY OCCURRING RADIOACTIVE MATERIALS AND MEDICAL ISOTOPES AT BORDER CROSSINGS Richard Kouzes, James Ely, Bruce Geelhood, Randy Hansen, Elwood Lepel, John Schweppe, Edward Siciliano, Dan Strom, Ray Warner, Nuclear Science Symposium Conference Record, 2003 IEEE Volume 2, 19–25 Oct. 2003 Page(s):1448–1452 Vol. 2. OVERVIEW OF PORTAL MONITORING AT BORDER CROSSINGS Bruce Geelhood, James Ely, Randy Hansen, Dick Kouzes, John Schweppe, Ray Warner, Nuclear Science Symposium Conference Record, 2003 IEEE Volume 1, 19–25 Oct. 2003 Page(s): 513–517 Vol. 1. HOMELAND SECURITY INSTUMENTATION FOR RADIATION DETECTION AT BORDERS Richard Kouzes, James Ely, Randy Hansen, John Schweppe, Edward Siciliano, David Stromswold, American Nuclear Society, Columbus, OH (September 19, 2004). DATA-BASED CONSIDERATIONS IN PORTAL RADIATION MONITORING OF CARGO VEHICLES Dennis Weier, James Ely, Robert O’Brien, Richard Kouzes Institute of Nuclear Materials Management, Orlando, FL (July 2004). FIELD TEST OF A NAI(TL)-BASED VEHICLE PORTAL MONITOR AT A BORDER CROSSING D.C. Stromswold, B.D. Milbrath, D.L. Stephens, L.C. Todd, R.R. Hansen, R.T. Kouzes, IEEE NSS Conference Record (2004). Nuclear Science Symposium Conference Record, 2004 IEEE Volume 1, 16–22 Oct. 2004 Page(s): 196–200 PIET-43741-TM-187, COMPARISON OF NAI(TL) SCINTILLATORS AND HIGH PURITY GERMANIUM FOR VEHICLE PORTAL MONITOR APPLICATIONS J.H. Ely, E.R. Siciliano, and R.T. Kouzes, IEEE NSS Conference Record (2004). Nuclear Science Symposium Conference Record, 2004 IEEE Volume 3, 16–22 Oct. 2004 Page(s): 1584–1587. RADIATION DETECTION AT BORDERS FOR HOMELAND SECURITY Richard Kouzes, PNNL-SA-40234, Physics and Society 33, p. 13, July 2004. TECHNOLOGIES FOR DETECTION OF CONTRABAND Daniel J. Strom, Richard Kouzes, Presented at the 38th Health Physics Society Midyear Meeting on Material Control and Security: Risk Assessment, Handling and Detection, New Orleans, 13–16 February 2005. SPECTROSCOPIC AND NON-SPECTROSCOPIC RADIATION PORTAL APPLICATIONS TO BORDER SECURITY Richard Kouzes, James Ely, Brian Milbrath, John Schweppe, Edward Siciliano, David Stromswold, Presented at the 2005 IEEE – DHS R&D Conference, Boston, MA, 27–28 April 2005. RADIATION PORTAL MONITOR INTELLIGENCE GATHERING AND PROCESSING Mitchell Woodring, James Ely, Edward Ellis, Robert Burnett, Terri Mercier, Presented at the 2005 IEEE – DHS R&D Conference, Boston, MA, 27–28 April 2005. INVESTIGATIONS OF ENERGY WINDOWING FOR PLASTIC-BASED RADIATION PORTAL MONITORS James Ely, Richard Kouzes, Dennis Weier, Mitch Woodring, John Schweppe, Presented at the 2005 IEEE – DHS R&D Conference, Boston, MA, 27–28 April 2005.

FIELD TESTS OF A NAI(TL)-BASED VEHICLE PORTAL MONITOR AT BORDER CROSSINGS B.D. Milbrath, D.C. Stromswold, J.W. Darkoch, J.H. Ely, R.R. Hansen, R.T. Kouzes, R.C Runkle, W.A. Sliger, J.E. Smart, D.L. Stephens, L.C. Todd, M.L Woodring, Presented at the 2005 IEEE – DHS R&D Conference, Boston, MA, 27–28 April 2005. PERFORMANCE OF ENERGY WINDOW RATIO CRITERIA AT RADIATION PORTAL MONITORING SITES Dennis Weier, Richard Kouzes, James Ely, Mitch Woodring, Derrick Bates, Charles LoPresti, July 2005 INMM, Phoenix, AZ. CREATION OF REALISTIC MODELS FOR HOMELAND SECURITY PURPOSES S.M. Robinson, R. Kouzes, R.J. McConn Jr., R. Pagh, J.E. Schweppe, and E.R. Siciliano, April 28, 2005. Proceedings of DHS conference Research & Development Partnerships in Homeland Security, Boston, April 27–28, 2005. COMPUTER MODELING OF RADIATION PORTAL MONITORS AT THE PACIFIC NORTHWEST NATIONAL LABORATORY R.T. Pagh, R.T. Kouzes, R.J. McConn, Jr., S.M. Robinson, J.E. Schweppe, and E.R. Siciliano, Proceedings of American Nuclear Society Conference, Washington, DC, November 13–17, 2005. SPECTROSCOPIC AND NON-SPECTROSCOPIC RADIATION PORTAL APPLICATIONS TO BORDER SECURITY R.T. Kouzes, J.H. Ely, B.D. Milbrath, J.E. Schweppe, E.R. Siciliano, D.C. Stromswold IEEE Transactions on Nuclear Science NSS San Juan Conference Record N14-12 321-325 (2005). SPECTROSCOPIC PORTAL MONITOR PROTOTYPE Kathy McCormick, David Stromswold, James Ely, John Schweppe and Richard Kouzes, IEEE Transactions on Nuclear Science NSS San Juan Conference Record N14-4 292-296 (2005). VALIDATION OF COMPUTER MODELS FOR HOMELAND SECURITY PURPOSES J.E. Schweppe, J.H. Ely, R.T. Kouzes, R.J. McConn, Jr., R.T. Pagh, S.M. Robinson, and E.R. Siciliano, IEEE Transactions on Nuclear Science NSS San Juan Conference Record N14-5 297-301 (2005). COMPARISON OF LaBr3:Ce And NaI(Tl) SCINTILLATORS FOR RADIO-ISOTOPE IDENTIFICATION DEVICES Brian D. Milbrath, Bethany J. Choate, James E. Fast, Richard T. Kouzes, John E. Schweppe, IEEE Transactions on Nuclear Science NSS San Juan Conference Record N14-1 283-287 (2005). COMPUTER MODELING OF RADIATION PORTAL MONITORS FOR HOMELAND SECURITY APPLICATIONS R.T. Pagh, R.T. Kouzes, R.J. McConn, Jr., S.M. Robinson, J.E. Schweppe, and E.R. Siciliano, Proceedings of ANS Conference, July 2005. THE ROLE OF SPECTROSCOPY VERSUS DETECTION FOR BORDER SECURITY AND SAFEGUARDS R. Kouzes, J. Ely, MARC VII Conference Record, Kona, HI, April 3–7, 2006. THE USE OF ENERGY INFORMATION IN PLASTIC SCINTILLATOR MATERIAL James Ely, Richard Kouzes, Denny Weier, MARC VII Conference Record, Kona, HI, April 3–7, 2006. IMPROVED PERFORMANCE OF ENERGY WINDOW RATIO CRITERIA OBTAINED USING MULTIPLE WINDOWS AT RADIATION PORTAL MONITORING SITES Dennis Weier, Charles LoPresti, James Ely, Derrick Bates, Richard Kouzes, July 2006 INMM, Nashville, TN.

Testing and Analyzing Instrumentation and Procedures for Radiation Detection and Identification

P. Beck, G. Sdouz

Austrian Research Centers GmbH – ARC Starting on November 2003 the Austrian Research Centers GmbH – ARC (formerly ARC Seibersdorf research GmbH) had a Cooperative Research Agreement (CRA) with the IAEA. The topic of the Cooperative Research Agreement was “Testing and analyzing instrumentation and procedures for radiation detection and identification”. The scope of the participation on the CRP depended on the available resources of ARC. Amongst others following research subject were recommended: - Detection limits for SNM detection in presence of medical and other isotopes at a field

test side: Analysis of the alarm frequency for medical isotopes - Model border crossing point at the Vienna Airport: Present and workout strategies

together with the IAEA to implement the ITRAP-Airport project as a Model border crossing point study.

- Contribute to development and test of IAEA specification document - Develop and test specifications for the new generation of spectral pedestrian monitors

to be incorporated into the IAEA specification document. - Analysis of Nuclear Material and other radioactive materials from outside of

standardized nuclear container, An overview on ARC’s activities in the area of Illicit Trafficking was presented during IAEA’s Research Coordination Meeting by G. Sdouz in April 2006. During the time period of this CRA, a number of papers and documents to this topic were published. A list of relevant documents are provided below.

Bibliography of Related Reports Co-Authored by Peter Beck Arlt R., Swoboda M., Boeck H., Schrenk, Beck P., et al., Automated Flagging of „Innocent Radiation Alarms at Airport Pedestrian Monitors“, Paper presented in Budapest, 22.–24.9.2003 Beck P., HiSEM – High Sensitive Monitoring, Paper presented at University Vienna, FONAS Conference, Vienna, 17.9.2003 Beck P., International Nuclear Threat and Radiation Detection in the Field, Paper presented at Federal Ministry of Defence, ROTA Training, Korneuburg, 14.3.2003

Beck P., Border Monitoring Experience, Austria, Paper presented at IAEA Meeting “Technical Specifications of Border Monitoring Equipment”, VIC, Vienna, 17.3.2003 Beck P., ITRAP-Airport, Paper presented at SIAGIT Meeting, Ministry of Trade and Labour, Vienna, 16.1.2004 Beck P., Illicit Trafficking nuklearer Stoffe – Internationale Bedeutung und Maßnahmen der Kontrolle, Paper presented at Ministry of Interior, Vienna, 17.2.2004 Beck P., Reichart F., Klösch W., Stehno G., ITRAP-Airport, Illicit Trafficking Radiation Detection Assessment Programme, ARCS Report ARC Sr – G0001, April 2004 Beck P., International Nuclear Threat and Radiation Detection in the Field, Paper presented at Federal Ministry of Defence, ROTA Training, Korneuburg, 10.5.2004 Beck P., Illicit Trafficking of Radioactive Material, Paper presented at IAEATraining Course, Vienna, 5.7.2004 Beck P., Aktueller Status und Zukunft von ITRAP in Österreich, Paper presented at SIAGIT Meeting, Federal Ministry of Trade and Labour, Vienna, 29.10.2004 Beck P., Prevention of Radiation and Nuclear Terrorism, Paper presented at NATO-Advanced Study Institute, Geilo, Norway, 14.–17.April 2005 Schwaiger M., Nuclear and Radiological Detection Equipment – Technology Review and Requirements, Paper presented at IMPACT WP300 Meeting, TNO, Rijswijk, Netherlands, 13.9.2005 Schwaiger M., Beck P., International Nuclear Threat and Radiation Detection in the Field, Paper presented at Federal Ministry of Defence, ROTA Training, Korneuburg, 28.11.2006

Capacity Building for Detection and Response to Illicit Trafficking of Radioactive Materials

M. Melich

HUMA-LAB APEKO Ltd.,Kosice, Slovakia Under the Research Agreement a “National Intercepting System for radioactive contaminated Materials” was developed. It consists of the following elements: The real time information system — ILTRAM, detection equipment and detection response procedures. To guarantee smooth operation, the responsibilities of different state organizations needed to be defined and the information exchange was initiated. The emerging system was discussed in several national and international workshops and seminars. Participants came from the Public Health Authority of the Slovak Republic (SR), the Customs Directorate, Ministry of Economy, Slovenske Elektrarne, the Nuclear Regulatory Authority, the State Fund for Decommissioning of NPP, the Office of Civil Protection, the police and other. Development and implementation were complemented by training courses with participation of customs coordinators and officers from the Customs Criminal Unit. The training covered radiation protection, response plans, detection equipment and its technical parameters, practical exercises with hand-held instruments, practical operation of the information system ILTRAM, and training of the response to different scenarios. Special effort was paid also to the training of the coordination of the cooperation between different parties involved: Customs Directorate, Customs Criminal Unit, Presidium of the Police Force, Office of Civil Protection of the Ministry of Interior, Laboratory of Radiochemical Analyses (SLRA), Slovenske Elektrarne, APEKO. In the training the functioning of the whole system was demonstrated, including the response to illicit trafficking.

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Detection of Nuclear Materials, Including Concealed Highly Enriched Uranium, using Enhanced Detection Methods

J.L. Jones, D.R. Norman

Idaho National Laboratory

Idaho National Laboratory (INL) had a representative at all three Research Coordination Meetings (RCMs) of the Coordinated Research Project (CRP) on Improvement of Technical Measures to Detect and Respond to Illicit Trafficking of Nuclear and Radioactive Materials. During the RCMs the challenges associated with detecting masked or shielded highly enriched uranium (HEU) using passive detection systems were highlighted. Active interrogation methods were considered to address some of these detection challenges. One of the active methods considered was the Pulsed Photonuclear Assessment (PPA) method being developed at INL in collaboration with Idaho State University’s Idaho Accelerator Center (IAC) and Los Alamos National Laboratory (LANL) [1–3]. The PPA method relies on unique signatures generated from photon-induced fission emissions from nuclear materials. An illustration of a basic PPA inspection system is shown in Figure 1. The major components of the system include a pulsed high energy (> 8 MeV) electron accelerator beam producing bremsstrahlung x-rays, delayed neutron and gamma-ray detectors, and a gray scale mapping system.

Figure 1. An illustration of a basic Pulsed Photonuclear Assessment System.

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During this CRP a committee member, Dr. Rolf Arlt, visited the INL for a demonstration of the PPA method at the IAC. During this demonstration 5 kg of depleted uranium was detected inside 5 cm (2 in) of lead at a standoff distance of 8 meters (26 ft) within 120 seconds. The CRP committee invited INL to provide a presentation on this PPA method during a 2004 RCM meeting in Sochi Russia. As requested, a presentation “PPAT for the Detection of Nuclear Materials, including Shielded Highly Enriched Uranium” (presentation available in RCM minutes) was given at the October 2004 RCM meeting highlighting the basic technology and the current status of the PPA Cargo Inspection prototype system. During the question and answer section after the presentation, Dr. Maxim Karetnikov from the Russian Kurchatov Institute commented that they were working on photonuclear detection techniques also, and had very positive results. Also during the Sochi meeting, contacts with several other detector experts and others that use active methods were made. During the third and final CRP meeting held April 2006 at the IAEA headquarters in Vienna Austria, it was highlighted on multiple occasions including the meeting summary that many of the challenges of detecting nuclear materials could be addressed with active interrogation methods. Also during this meeting Dr. Karetnikov and Daren Norman identified areas of common research interests and initiated a follow on meeting that included a tour of the INL and IAC facilities in the United States. A prototype PPA Cargo Inspection System, shown in Figure 2, was assembled and characterized [4–6]. It measures delayed neutron and gamma-ray emissions for nuclear material detection, and utilizes transmission gray scale mapping to identify any significant shielding. This prototype detection system has demonstrated the ability to detect nuclear materials concealed in several INL Cargo “Calibration” Pallets. In Figure 2, a forklift has just set a metric-ton polyethylene “calibration” pallet with concealed nuclear material into the PPA inspection system. Other “calibration” pallets include borated polyethylene, lead, iron, wood, celotex, water and a 4.5 metric-ton nested lead in borated polyethylene pallet. Current areas of research and development for the PPA method include prompt emission and standoff detection to improve nuclear material detection capabilities. Some PPA performance improvements have been identified, such as using higher energies [7]. When considering the PPA methods for cargo inspection systems it is important to remember that the World Health Organization (WHO) would not exclude higher photon energies (> 10 MeV) inspections. The WHO document for photon inspection of foods states; [8]

However, this conclusion is not intended to preclude other safe surveillance systems designed to operate at higher energy levels or dose. In such cases, assurance should be provided that, at the point of consumption, food would not contain a measurable detectable amount of induced radioactivity.

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Figure 2. INL Prototype PPA Cargo Inspection System.

The PPA method can provide efficient nuclear material detection and identification even in heavily shielded or masked configurations. As PPA methods continue to be developed they could address primary screening needs as well as secondary screening for alarm resolution. This CRP provided the opportunity to address the importance of shielded and masked nuclear material detection on an international basis. Collaborations with international colleagues are beginning to address common concerns and interests. New CRPs in nuclear material detection related areas could provide additional national and international benefits.

REFERENCES

[1] J.L. Jones, et al., Pulsed Photoneutron Interrogation: The GNT Demonstration System, INL Report WINCO-1225, 1994

[2] J.L. Jones, et al., Proof-of-Concept Assessment of a Photofission-Based Interrogation

System for the Detection of Shielded Nuclear Material, INL Report INEEL/EXT-2000-01523, 2000

[3] J.L. Jones, et al., Photofission-based, Nuclear Material Detection: Technology

Demonstration, INL Report INEEL/EXT-02-01406

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[4] J.L. Jones, et al., Detection of Shielded nuclear material in a cargo container, NIM A 562, pp 1085-1088, 2006

[5] J.L. Jones, et al., Pulsed Photonuclear Assessment (PPA) Technique: CY-05 Project

Summary Report, INL Report INL/EXT-05-01020, 2005 [6] J.L. Jones, et al., Pulsed Photonuclear Assessment (PPA) Technique: CY04 Year-end

Progress Report, INL Report INEEL/EXT-05-02583, 2005 [7] D.R. Norman, et al., Inspection applications with higher electron beam energies,

NIM B 241, pp 787–792, 2005 [8] Bulletin of the World Health Organization, Vol. 31, pp. 297–301, 1990

Inspection of Shipping Containers for Undisclosed Radioactive Materials

V. Valkovic, S. Blagus, D. Sudac

Department of Experimental Physics, Rudjer Boskovic Institute (RBI), Croatia

The objective of this project was to develop a conceptual design of a system for

inspection of shipping containers, in order to determine if they contain radioactive

materials, or other hazardous chemicals such as explosives (which are carbon-rich

materials).

A neutron beam interrogation system was set up in a laboratory. The neutron was

obtained by means of the d+t a + n reaction; the associated alpha particle was fixed in

such a way that the neutron beam was in the horizontal plane at 90o to the deuteron beam.

The neutron ‘tagged’ in this way was then expected to interact with the interrogated

object, and produce γ radiation by A(n, n’ γ)A processes on nuclei of hidden substances.

Test results indicated that a hidden carbon-rich source (e.g. TNT) could be localized and

identified even when hidden behind steel. These results serve as a “proof-of-principle”

that objects hidden inside a closed container can be located and identified by their

chemical composition by using such a “neutron sensor”. It is expected that similar

positive results could be demonstrated for shielded HEU.

Characterization of Various Survey Meters through Car-Borne Survey in Java Island

as a Basis Data for Searching Orphan Sources

U. Pande Made, A. Yus Rusdian Center for Nuclear Material Security

National Nuclear Energy Agency (BATAN), Center fir Nuclear Material Security Technology, Indonesia

An assessment of available technical means has been performed to select suitable

instruments to be used in car borne surveys for the detection and localization of orphan

sources. Criteria used for the selection included (i) that the instrument should be simple

to operate, (ii) that it should provide reliable results, and (ii) that little time for processing

or transferring raw data from the instrument to a desktop computer (for decision making

using statistical methods) is required. Based upon these criteria, four hand-held

instruments in the BATAN inventory were selected for further study.

Measurements data using these four instruments, taken during a 400 km car-borne survey

of western Java, was analyzed. The data was taken both in rainy and clear weather

conditions, thus recorded fluctuations from the instrument response (when the vehicle

was driven along the same paths in both weather conditions) included variations due

washout by the rain of Rn daughters. The data from all instruments was analyzed using

various statistical tests to determine the most sensitive instruments (with lowest number

of alarm fluctuations) and best statistical analysis technique for finding anomalous

radiation levels. Using this data, as well as data from taken during controlled field

measurements, it was determined that the best instrument could detect a 7.4 GBq 137Cs

source from as far as 145 m away, when traveling in a car at 50 kph.

Development and Utilization of Nuclear Analytical Techniques for Analysis of Nuclear Forensic Materials

H. Demirel, P. Arikan

Ankara Nuclear Research and Training Center, Turkish Atomic Energy Authority

When packages are seized at borders and have been identified as containing illicit nuclear

or other radiological materials, it is very important to determine the origin of the

materials, the last legal user of the materials, and the smuggling route. This is required to

(i) close the ‘hole’ that allowed that material to escape, and (ii) to prosecute the smuggler.

Both classic and nuclear forensics techniques are used to answer these questions.

Under this project, laboratories were identified in Turkey to receive suspect radiological

or nuclear samples. Considerable work was done to develop the methodology of activities

for these laboratories. Work is ongoing to improve (i) the methods of analysis, (ii) legal

‘traceability’ of the samples, (iii) interpretation of the results, and (iv) the confidential

handling of suspect materials.

Development of Recommendations, Guidelines and Methods for Coast-Guard Officers on Measurement

using Portable and Hand-Held Isotope Measurement Devices

M. Divizinvuk, Sevastopol National Institute of Nuclear Energy and Industry, Sevastopol, Ukraine

The goal of this contract was to examine the most effective procedures for Coast Guard

officers to use for the interception of illicitly smuggled nuclear and radiological

materials. Specific equipment tailored for marine use was to be tested. Studies to examine

the most efficient placement of radiation monitoring assets were to be made.

Theoretical analysis performed showed that when operative information (e.g. external

intelligence) is absent, all ships in a territorial sea must be supervised/examined for

possible trafficking issues. Radiation detection devises were placed upon a helicopter,

which then flew at 80 km/hr at 100m over land vehicles carrying radioactive sources.

Effective detection capability was demonstrated. Based upon this, a conceptual proposal

was developed to place such devices on buoys in a water channel to form a floating portal

monitoring system.

Research Co-ordination Meeting of the Agency’s Co-ordinated Research Project on “Improvement of Technical Measures to Detect and Respond to

Illicit Trafficking of Nuclear and Radioactive Materials” Sochi, Russia Federation,

4–8 October, 2004

Participants Overview Sheet

# COUNTRY Contr. # Participants Name

Address

1. AUSTRALIA 12516 Mr. Dimitri Alexiev

Australian Nuclear Science and Technology Organization (ANSTO); Lucas Heights Research Labs., New Illawarra Road, Lucas Heights 2234/ Locked Mailbag No.1, Menai NSW 2234, Australia Tel: +61 2 9717 3182 Fax: +61 2 9717 9265 E-mail: [email protected]

2. AUSTRIA 12585 Mr. Helmut Boeck

Atominstitut der Oesterreichische Universitaeten (ATI) Rektorabteilung Technische Universitaet Wien, Stadionallee 2, 1020 Vienna, Austria Tel: +43 1 58801 14168 Fax: +431 5880114199 E-mail: [email protected]

3. AUSTRIA 12586 Mr. Peter Beck

ARC Seibersdorf research GmBH, Health Physics Division A-2444 Seibersdorf, Austria Tel: +43 50550 2480 Fax: +43 50550 2502 E-mail: [email protected]

4. BELARUS 12587 Mr. Sergey Korneyev

Joint Institute for Energy and Nuclear Research 99, academic A. C. Krasin st. , Minsk, 220109, Belarus Tel: +375 17 2994448 Fax: +375 17 299 4712 E-mail: [email protected]

5. CHINA 12588 Mr. Qingli Zhang

China Institute for Radiation Protection (CIRP), P.O. Box 120, Taiyuan, Shanxi, 030006 Public Republic of China Tel: +86 351 2203461 Fax: +86 351 7020407 E-mail: [email protected] zhangqing‗[email protected]

6. CROATIA 12589 Mr. Vladivoj Valkovic

Ruder Boskovic Institute, Dept of Experimental Physics Rudjer Boskovic Institute, Bijenicka c.54, 10000 Zagreb, Croatia Tel: +385 1 468 0101 Fax: +385 1 468 0239 E-mail: [email protected]

7. FRANCE 12590 Mr. Jean-Louis Szabo

Commissariat à l’énergie atomique (CEA); Centre d’études nucléaires (CEN); Dépt. d’instrumentation et de métrologie des rayonnements ionisants (DIMRI), B.P. Fontenay aux Roses, France Tel: +33 1 46 547558 Fax: +33 1 46 549942 E-mail: [email protected]

State Military Scientific and Technical Center DELTA 73, Mnatoby st. 380002, Tbilisi, Georgia Tel: +995 32 955646 Fax: +995 32 956080 E-mail: [email protected]

8. GEORGIA 12591 Mr. Lerry Meskhi

Nuclear and Radioactive Safety Service 87 Paliashvilistr., Tbilisi Georgia Tel: +995 32 251632 E-mail: [email protected]

9. GERMANY 12517 Mr. Frank Gabriel

Forschungszentrum Rossendorf FZR Postfach 510119, 01314 Dresden, Germany Tel: +49 351 260 3109 Fax: +49 351 260 3110 E-mail: [email protected]

10. GERMANY 12592 Mr. Said Abousahl

European Commission Joint Research Centre, Institute for Transuranium Elements Postfach 2340, 76125 Karlsruhe, Germany Tel: +49 7247 951 259 Fax: +49 7247 951 99545 E-mail: [email protected]

11. GERMANY 12593 Mr. Dietrich E. Becker

Bundesamt für Strahlenschutz (BfS) Willy-Brandt-Strasse 5, 38226 Salzgitter, Germany Tel: +49 1888 3330 Fax: +49 1888 333 1858 E-mail: [email protected]

12. INDONESIA 12594 Mr. Yus Rusdian Akhmad

National Nuclear Energy Agency (BATAN) Kawasan Puspiptek Serpong Gedung 90, Serpong, Tangerang 15310 Indonesia Tel: +62 21 7560 183 Fax: +62 21 7560 895 E-mail: [email protected]

13. ITALY 12518 Mr. Richard Berndt

JRC-Joint Research Centre of the European Union, Institute for the Protection and the Security of Citizens Via Fermi 20, I-21020 Ispra (Va), Italy Tel: +39 (0332) 78 5317 Fax: +39 (0332) 785072 E-mail: [email protected]

14. ITALY 12595 Ms. Patricia Mortreau

JRC-Joint Research Centre of the European Union, Institute for the Protection and the Security of Citizens Via Fermi 20, I-21020 Ispra (Va), Italy Tel: +39 0332 78 9211 Fax: +39 0332 78 5072 E-mail: [email protected]

15. KOREA 12597 Mr. Jong Uk Lee

Korea Atomic Energy Research Institute, Technology Center for Nucelar Control, Dept. of Nuclear Security & Protection PO Box 105, Yuseong, Daejeon Republic of Korea 305-600 Tel: +82 42 868 8332 Fax: +82 42 861 8819 E-mail: [email protected]

16. POLAND 12596 Mr. Marek Moszynski

Andrzej Soltan Institute for Nuclear Studies, Dept. of Detectors and Nuclear Electronics PL 05-400 Otwock – Swierk Poland Tel: +48 22 718 0586 Fax: +48 22 779 3481 E-mail: [email protected]

17. RUSSIA 12444 Mr. Nikolay E. Kravchenko

Russian Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182, Russia Tel: +7 0951967265 Fax: +7 095 196 60 32 E-mail: [email protected]

18. RUSSIA 12598 Mr. Maxim Karetnikov

Russian Research Center “Kurchatov Institute”, Scientific and Technological Division “Electronics”, Kurchatov Sq. 1, 123182Moscow, Russia Tel: +7 095 196 7265 Fax: +7 095 196 6032 E-mail: [email protected]

19. RUSSIA 12599 Mr. Sergey Ullin

MEPhI Moscow Engineering and Physics Institute Kashirskoe shosse, 31, Moscow 115409, Russia Tel: +7 095 324 65 89 E-mail: [email protected]

20. RUSSIA 12600 Mr. Andrey Kuznetsov

V.G Khlopin Radium Institute, Applied Physics Laboratory 2nd Murinsky pr 28., 194021 St. Petersburg, Russia Tel: +7 812 2470 173 Fax: +7 812 2478 095 E-mail: [email protected]

21. RUSSIA 12601 Mr. Aleksandr Borisenko

Russian Customs Academy Vladivostok Branch, 16-V Strelkovaya Street, 690014, Vladivostok, Russia Tel: +7 4232 236 711 Fax: +7 4232 236 711 E-mail: [email protected]

22. RUSSIA 12602 Mr. Michael Majorov

Scientific Engineering Center Nuclear Physics Research, 2nd Murinsky pr. 28, Saint Petersburg, 194021, Russia Tel: +7 812 24739 24 Fax: +7 812 247 82 44 E-mail: [email protected]

23. SLOVAKIA 12603 Mr. Matous Melich

Huma-Lab Apeko Ltd Spectrometry Lab Letna 45 04001 Kosice, Slowakia Tel. and Fax: +42 1556829254 E-mail: [email protected]

24. TURKEY 12604 Mr. Pervin Arikan

Turkish Atomic Energy Authority Ankara Nuclear Research and Training Centre Besevler TR-06100, Ankara, Turkey Tel: +90 312 2126 230 Fax: +90 312 2234439 E-mail: [email protected] Gazi University, Ankara, Turkey E-mail 2: [email protected]

25. UKRAINE 12606 Mr. Mikhail Divizinyuk

Sevastopol National Institute of Nuclear Energy and Industry, Nuclear Regulatory and Physical Protection Service Kurchatova Str., 99033 Sevastopol, Ukraine Tel: +38 0692 71 0169; +38 0692 71 0046 Fax: +38 0692 71 0138 E-mail: [email protected]

26. USA 12519 Mr. Richard Kouzes

Pacific Northwest National Laboratory 920 Battelle Boulevard, P.O. Box 999, Richland, WA, USA Tel: +1 (509) 376 2320 Fax: +1 (509) 376 3868 E-mail: [email protected]

27. USA 12812 Mr. James Jones

Idaho National Engineering and Environmental Lab. (INEEL), Idaho Falls

28. UZBEKISTAN 12607 Mr. Vitaly Petrenko

Uzbekistan Academy of Sciences; Institute of Nuclear Physics Ulugbek Village, Tashkent 702132, Uzbekistan Tel: +10 998 712 6415 52 Fax: +10 998 712 64 25 90 E-mail: [email protected]

29. LATVIA 13096 Ms. Larisa Grigorjeva

Institute of Solid State Physics, Riga, Latvia E-mail: [email protected]

Usability Guide for

Manufacturers of Radiation Monitoring Devices

Document Version: 1.0

Document Number: 42_Useabilty guide IAEA RPT27.doc - 1.0

Date: 11 May 2007

Author: Mike Alexander Mike Alexander Usability Consultant

Himmelstrasse 58a A-1190 Vienna Austria Europe Tel: + 43 1 320 8091 Fax: + 43 1 320 8090 Mobile: + 43 664 320 8091 Email: [email protected]

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Contents 1. Introduction ....................................................................................................................................1 1.1. What is usability?...........................................................................................................................1 1.2. Why is usability important?............................................................................................................1 1.3. User Centred Design Principles ....................................................................................................2 1.3.1. Ongoing user-requirements monitoring .....................................................................................2 1.3.2. Focus group ...............................................................................................................................2 1.3.3. User representative in Development..........................................................................................2 1.4. What are the benefits?...................................................................................................................3 1.4.1. Understanding the Costs ...........................................................................................................3 1.4.2. Understanding the Benefits........................................................................................................3 1.4.3. Making the Decisions.................................................................................................................3 2. Managing the usability process .....................................................................................................4 2.1 Understanding the different usability activities ..............................................................................4 2.1.1. Usability Review.........................................................................................................................4 2.1.2. Usability Walkthrough ................................................................................................................5 2.1.3. Usability Test..............................................................................................................................6 2.2. Product Development ....................................................................................................................7 2.2.1. Development Process................................................................................................................7 2.3. Choosing the right process............................................................................................................8 2.4. Choosing the right usability activities.............................................................................................9 2.5. Getting the most benefit ..............................................................................................................11 2.5.1. Usability Report........................................................................................................................11 2.5.2. Using the Usability Report .......................................................................................................11 2.5.3. Action Document......................................................................................................................12 2.5.4. Using the Action Document .....................................................................................................13 3. Usability Principles.......................................................................................................................14 3.1. User Control.................................................................................................................................14 3.2. Consistency .................................................................................................................................16 3.3. Navigation Feedback...................................................................................................................22 3.4. Status Feedback..........................................................................................................................24 3.5. Progress Feedback......................................................................................................................25 3.6. Error Message Feedback ............................................................................................................27 3.7. Forgiveness .................................................................................................................................28 3.8. Naming ........................................................................................................................................29 3.9. Ease of use..................................................................................................................................30 4. Hardware Usability Guidelines ....................................................................................................32 4.1. Handling.......................................................................................................................................32 4.2. Controls .......................................................................................................................................33 4.3. User control .................................................................................................................................33 4.4. Use of Icons.................................................................................................................................34 4.5. Multi-language Support: ..............................................................................................................34 5. Software Usability Guidelines......................................................................................................35 5.1. Dialogs.........................................................................................................................................35 5.2. Menus ..........................................................................................................................................35 5.3. Controls .......................................................................................................................................35 5.4. Feedback .....................................................................................................................................35 5.5. Error Recovery Support...............................................................................................................35 5.6. Multi-language support ................................................................................................................35 5.7. Error handling ..............................................................................................................................36 5.8. Forgiveness .................................................................................................................................36 5.9. Accessibility .................................................................................................................................36 6. Reading Tips................................................................................................................................37 7. Checklists ....................................................................................................................................38 7.1. Planning Phase Checklist ............................................................................................................38 7.2. Prototyping Phase Checklist........................................................................................................39 7.3. Testing Phase Checklist ..............................................................................................................40 7.4. Building Phase Checklist .............................................................................................................41 7.5. Hardware Usability Checklist.......................................................................................................42 7.6 Software Usability Checklist ........................................................................................................43

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1. Introduction 1.1. What is usability? Usability is like air. Although you need it, you tend not to notice it most of the time. You only begin to notice when it is suddenly missing. The best measure of usability is when a user describes a device, or piece of software as “intuitive”. They might say things like, “I didn’t know how to use it, but I just tried, and it worked out right”. You know certain things about your users before you can build products that they find intuitive to use. You will need to know things like the following:

• What is a users’ motivation for using your product? • What sort of formal education can you expect them to have had? • What sort of job-specific training can you expect them to have? • What previous experience can you reasonably expect your users to have had? • What age-range are you addressing with your product? • What assort of language abilities will they have?

These factors, amongst others, will determine what you have to do to build a product that your users find intuitive to use. You must know the answers to these questions before you can build a product that suits them. 1.2. Why is usability important? It all boils down to a question of money. There are business reasons for investing in usability because bad usability costs money. Let’s have a look at some of the costs. Education Costs: If a product is not intuitive to use, it will be necessary for potential users to receive some kind of training, and training cost money in these ways:

• Instructor salaries • Lost productivity as students attend courses • Maybe even travel and/or hotel expenses

Turnover Costs: If your user population has a fast turnover, your educations costs will be ongoing over the life of the system. Error and Error-recovery Costs: If a product is not intuitive to use, users may make the wrong choices or get unexpected results as they work, and they will have to recover from the errors they have made. Error recovery can take different forms, but each form costs money. Here are some examples of where the money is lost:

• Calling a help desk • Asking a colleague • Trial and error

Whichever method is used, the incorrect work has to be undone, and then be done properly. The productivity of users is therefore unnecessarily low which, in itself costs money. If usability can be improved, these ongoing and hidden costs can reduced or avoided altogether.

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1.3. User Centred Design Principles The best complement a product can receive is when its users describe it using comments like, “intuitive”, “easy”, “friendly”, “simple-to-use”, and “helpful”. To build a product having these qualities, it is essential to involve people who can view problems from the user’s perspective and who understand the user’s point of view.

1.3.1. Ongoing user-requirements monitoring While a usability professional can do this to some extent, and it is recommended practise to involve usability professionals, the best people to represent the users’ interests are the users themselves. Design activities should therefore be centred around a group of representative users, and there should be permanent checking and re-checking with this group that the design continues to address their needs in a way that they find acceptable.

1.3.2. Focus group A group of vocal potential end users, sometimes known as a “focus group” should be given the opportunity of contributing to user requirements, criticising ongoing design works, and trying out prototypes.

1.3.3. User representative in Development Input from the focus group must be documented and a member of the development team should have the responsibility of ensuring that users’ input flows smoothly into the development process and is properly responded to. This person is also responsible to scheduling events such as walkthroughs1 and usability tests2 at the appropriate times.

1 Walkthroughs are explained in detail on page 5. 2 Usability tests are explained in detail on page 6.

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1.4. What are the benefits?

1.4.1. Understanding the Costs To perform usability activities and to follow up on their outcomes, certain resources will be needed that fall into the following categories:

Time resources (constrained by project deadlines) Personnel resources (constrained by staffing levels) Financial resources (constrained by available budget)

For any usability activity to be worthwhile there must be a return on this investment and it is important to understand where and when this return on investment will occur. The benefits will not be felt during project development. On the contrary, it is during development that usability investments are made. Summary: Usability activities cost money to perform.

1.4.2. Understanding the Benefits The benefits will be felt throughout the expected life of the product – perhaps a period of 5 or more years. Successful usability activities and appropriate follow-up during project development will produce a product that is easier to learn, more intuitive to use, and simpler to maintain. The savings during the product’s life will occur in the following ways: • Shortened user-training times • Reduced need for help desks and other support facilities • Less time wasted as users recover from errors they have made • Improved user confidence and, therefore, improved user productivity • Less maintenance required for correcting initial system design errors. Summary: Properly performed usability activities have the potential to make great savings over the

life of the product.

1.4.3. Making the Decisions The budget for product development and the budget for daily operations are often not managed by the same function. It is therefore important that the decision to perform usability activities is made by a function that is sensitive to the overall budget requirements and appreciates that a comparatively small investment during development will be saved many times over during ongoing daily operations.

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2. Managing the usability process 2.1. Understanding the different usability activities There are three basic usability activities that can help in building more easy to use instruments:

• Usability Review • Usability Walkthrough • Usability Test

It is important to understand the differences between the activities and when in the development process they can bring you the most advantage. The activities are described in detail below:

2.1.1. Usability Review This is an economical usability activity where the usability professional, assisted by a product specialist, examines design proposals, a prototype, or a previous version of the product, and reports on concerns and opportunities for improvement.

When to use: • At project initiation time • During design the stage • Before investment is made in construction • While designs are still flexible

Advantages:

• Quick and cheap to perform • Gives immediate results • Checklist helps track issues during development

Disadvantages:

• End users are not directly involved • Some end-user requirements may therefore be overlooked

Resources:

• Product specialist (2 days) • Usability professional (7 days) including report preparation

Deliverables:

• Hard- and softcopy report of all concerns found • Rationale explaining the reason for each concern • Suggested alternative implementations for discussion • Checklist for the on-going management of the usability concerns.

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2.1.2. Usability Walkthrough

This is a professional usability activity where a product specialist presents design proposals or a prototype to a group of potential users. The usability professional takes note of user concerns, and reports on opportunities for improvement.

When to use:

• When a paper or physical prototype of key functions is available • During design the stage • Before major investment is made in construction • While designs are still flexible

Advantages:

• User feedback goes directly to the development team • Gives immediate results • Checklist helps track issues during development

Disadvantages:

• Presentation requires preparation and rehearsal • Planning and coordination of attendees necessary • Hosting of attendees necessary

Resources:

• Product specialist (2 days) includes rehearsal • Potential end users (1 day each) • Usability professional (7 days) including report preparation

Deliverables:


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2.1.3. Usability Test This is a professional usability activity where potential users individually perform key tasks while being observed by a usability professional and a product specialist. The usability professional takes note of difficulties and user concerns, and reports on opportunities for improvement.

When to use:

• When design is relatively firm • Before system is made generally available • While there is still the possibility to make system changes

Advantages:

• You can see how the system will perform under real conditions • Checklist helps you track issues up to delivery and on to next release

Disadvantages:

• Is relative expensive to perform • Requires careful preparation • Planning and coordination of attendees necessary • Hosting of attendees necessary

Resources:

• Product specialist (5 days) • Potential end users (½ day each) • Usability professional (25 days) including planning, set-up and report preparation • Two adjacent offices where TV equipment can be installed

Deliverables:


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2.2. Product Development

2.2.1. Development Process While development processes may differ from organisation to organisation, they can generally be divided into phases, as follows:

1. Planning 2. Prototyping 3. Testing 4. Building

The phases are described below: Planning Phase (1): In the planning phase, product plans are drawn up and information of the following kind will be gathered:

• What are we going to build? • What will the main functions be? • What kinds of users are we aiming at? • How much is this project going to cost? • How much profit are we likely to make?

Prototyping Phase (2): In the prototyping phase, product ideas will be tried out using paper prototypes or inexpensive software prototyping tools. Testing Phase (3): In the testing phase, the designs will be checked until a stable design that can be put into production is achieved. Building Phase (4): The building phase is the most expensive phase. Project staffing levels increase as programmers and/or engineers join the project to create the finished product. Here is a diagram that shows how the phases generally differ in duration.

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2.3. Choosing the right process There have been different philosophies as to the best development process, and the process any particular organisation uses will be influenced, in part, by the type of product they build. In the “waterfall process”, which was widely used in the early days of the computing industry for building software, each phase had to be successfully completed before the next phase could begin, as shown below:

This process has the disadvantage that, once the design has been “frozen”, it is difficult to respond to requirements that may become apparent only in the later phases. The “iterative design process” was developed to combat this. In the iterative design process, progress is generally in a forward direction, but iterations into previous phases can be made as the need becomes apparent, as shown below:

From the usability point of view, this process has the great advantage that it allows the development team to respond to design flaws and also user-requirements that become apparent later in the process. The disadvantage is that it needs careful management to ensure that, despite the iterations, the planned schedules are still maintained.

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2.4. Choosing the right usability activities In the diagram below, two curves are plotted against the duration of a given project to show the following two parameters:

• Ability to accommodate change • Cost of making changes

As can be seen above, as projects progress, the ability to accommodate change diminishes, and the cost of making changes to work that is already in-place, increases. So, to get the best benefit out of any usability work, it is important to schedule the right activity at the right time. In the Planning Phase (1), the ability to make change is high, and the costs of making changes are low. This is therefore a good time to do a usability review of a previous version or of the present designs. This activity will deliver requests for change, but at a time when they can be accommodated easily, so your final product benefit greatly. In the Prototyping Phase (2), there is still flexibility to make change and the cost of change, although rising, is not prohibitively high. This is therefore a good time to do a usability walkthrough with a paper or real prototype to get user feedback on your designs. This activity will generate requirements and user comments that will influence product design and will enable you to enter the next phase with a relatively stable design. In the Testing Phase (3), the ability to make changes is becoming limited and the impact of any change is also rising. If the previous usability work was performed properly, it should have revealed the major problems and fixes should now be in place. This is a good time to do a usability test to confirm that your design is correct before entering the final stage where your resources are concentrated on building the final product. In the Building Phase (4), the ability to make change is severely limited and the costs of making change are generally prohibitive. Usability activities this late will, therefore, not deliver data that can be used in the current development. Such data are only useful for a subsequent releases or later modifications to the product.

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In summary, here are the suggested timings to get the best advantage from usability activities:

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2.5. Getting the most benefit Each usability activity will produce documentation showing the current usability status. To get the most benefit from any usability activity, it is important to understand what the documentation offers and how you can best use it to improve the product.

2.5.1. Usability Report The usability report gives detailed information on the completion of a usability activity and contains the following major items: Activity description: A description of the usability activity that gave rise to the report Management Summary: A management-level summary of the main items resulting from the activity Observations: Details of each usability item found A sample observation is shown below:

2.5.2. Using the Usability Report The report will be used by the following people for the following purposes: Management: To understand the current usability status and to allocate resources Project management: To allocate resources effectively Developers: To develop appropriate solutions to users’ problems

5.1.3 (1) Possible Confusion Between Full-charge and Failure

See also 5.1.24 (3) Charge Not Shown When Instrument Turned Off Observation: When the instrument is fully charged, the “Charging” LED switches off. If the charger circuit failed the LED would also be off, and a user may think an instrument is fully charged when, in fact, it is defect. Rationale: If equipment can have different states and these states influence the way a user may use it, it is good practice to provide users with feedback of the current state. Certainly users would need to know the difference between a fully charged instrument and a defect one. Suggested Alternative Implementation: Consider implementing the LED to show the all the possible states of the charging function, as follows:

LED permanently on Instrument is fully charged LED blinking Instrument is charging LED off Instrument is not charging

Severity indicator

Identification number Problem summary

Cross-references where appropriate

Background information

Problem details

Suggested improvement

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2.5.3. Action Document On the completion of a usability activity, the action document lists the problems found, as shown in the excerpt below: The columns (A, B, C, etc) are used to document the intended actions to rectify the reported problem. Since the possible actions are project-dependent, the columns are initially blank. A round-table meeting (usually involving at least, project management, development, and a usability professional) is held to agree on appropriate actions to address each of the usability items reported. The result of the meeting will be a completed action document that could look like this:

Column Action or comment A Has already been addressed B Will be addressed in next prototype planned for week 24 C Alternative solutions will be prototyped and shown to user group for comment D Needs investigating – Chris will report back in next week’s progress meeting E Not yet feasible – postponed to a future release

Item Description Page A B C D E 5.1.1 (1) Dose Rate Display Is Difficult To Interpret 16 5.1.2 (1) Warning And Alarm Settings Not Shown 19 5.1.3 (1) Possible Confusion Between Full-charge and Failure 20 5.1.4 (1) Warning And Alarm Values Difficult To Set 21 5.1.5 (1) Invalid Dates Can Be Set 22 5.1.6 (1) Some Dates Are More Difficult To Set… 23 5.1.7 (1) Date Format Cannot Be Changed 24 5.1.8 (1) Some Times Are More Difficult To Set… 25

Item Description Page A B C D F 5.1.1 (1) Dose Rate Display Is Difficult To Interpret 16 5.1.2 (1) Warning And Alarm Settings Not Shown 19 5.1.3 (1) Possible Confusion Between Full-charge and Failure 20 5.1.4 (1) Warning And Alarm Values Difficult To Set 21 5.1.5 (1) Invalid Dates Can Be Set 22 5.1.6 (1) Some Dates Are More Difficult To Set… 23 5.1.7 (1) Date Format Cannot Be Changed 24 5.1.8 (1) Some Times Are More Difficult To Set… 25

and so on…

Severity indicator

Identification number Problem summary

Intended action to address the problem.

(explained below)

Cross reference to details in usability report

These project-specific actions are developed in a

round-table meeting

Each usability item is allocated an appropriate action during

the round-table meeting

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2.5.4. Using the Action Document The action document will be used by the following people for the following purposes: Management: To understand the current usability status and to allocate resources Project management: For on-going monitoring of the usability status Developers: To agree on strategies for developing solutions to users’ problems As usability items are addressed, the action document has to be updated and re-circulated to its users. This is usually done by the project manager who uses the Action Document to communicate the current usability status to all concerned. Summary: The Action Document is a living document that shows the usability status of the product at all times. Efficient use of the Action document will enable management, project management, and development personnel to be informed of usability progress as the project advances towards final completion.

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3. Usability Principles There are various usability principles that, if adhered to, will help you produce a product whose users describe it as “easy-to-use”, intuitive, and “helpful”. As we have seen, a usable product is more economical to operate so, paying attention to these principles while designing your product will not only please your users, but will help them do their work more economically. Everything we do should have the underlying aim of using the available resources efficiently to provide users with the best tool possible. The principles detailed in this section will guide you in this aim. The usability principles can be categorised into the following groups:

• User control • Consistency • Feedback • Forgiveness • Visual Presentation • Naming

The principles are described in detail below. There are also some sample implementations, which have then been re-engineered to illustrate the usability principle. 3.1. User Control It is important that users feel they are in control at all times. This means that the interface operates as a result of user requests. Users should never feel the interface is driving them and the system should never perform an operation without the expressed wish of the user. A sample dialogue:

• Users are offered no choice • They must wait for the installation process to complete • They have apparently no control over the system for the moment An example of improved usability: Step 1:

• Users have a choice • If they select “Yes” the process continues

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Step2:

• Even though the process has started, users can interrupt

it if they need to. For example because a higher priority task has to be done, or it is just time for them to go home

• If the activity is expected to take a long time, foe example more than 4 seconds, a progress indicator should also be used. See details in “3.5 Progress Feedback” on page 25.

Advantages of improved user control: If users retain control at all times, the system is more likely to be doing what they want. This means that users are less likely to waste time in interrupting and recovering from the consequences of unwanted operations.

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3.2. Consistency “Be consistent!” It sounds very simple but, as a product grows, it is very easy for functions to be given names by one development group, and similar functions, developed by a different group, to be given a different name. This would explain why, for example, we find applications that have functions like these:

• Erase • Delete • Purge

Each of these functions causes some object or other to exist no more. They all do the same thing and in they do it in the same way but are called by different names. They could all be called “Delete”, because that is what they do. However, because of failing communication between the development teams, the differing terms have evolved and acquired meaning with each team. The usability principle is as follows: • Things that are the same should be named the same, look the same, and behave the same.

• Things that are intrinsically different should be named differently, and also look different to reduce the

possibility of confusion.

Note: There is no need to make things that are intrinsically different behave differently. In fact it is good to keep interactions as similar as possible because, having learnt one interaction, users can easily go on to use other similar functions without any training.

The advantage of consistency is that the education overhead is reduced. Once users have learned to use a particular function, they are able to use that knowledge wherever they encounter the function in the product. If the functions were named differently, users may assume there is a difference and will waste time in trying to understand the differences. As in most usability matters, it comes down to a question of saving time and money. A sample dialogue: Let us assume an instrument employs a system of menus to allow users to navigate around its functions and to select the various functions. To navigate the hierarchy and to choose a function, the following navigation choices will be necessary:

• Up • Down • Select

Let us also assume the instrument has a display and is provided with three function buttons, as follows:

The function buttons “A”, “B”, and “C” are used to access all the functions that the instrument offers. Since there are considerably more than three functions, the current use of each button is shown in the respective dialogue. This gives total flexibility.

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Sample menus: In this example, the “Down” and “Select” functions are allocated to the function buttons as shown:

• Because “Down” and “Select” functions are located on different keys in the two menus, users must refer

to each menu every time they use it to find out which function is allocated to which key. • Also, if this instrument is to be used by an international audience, the letters “A”, “B”, “C”, of the Latin

alphabet may not be familiar to them so, while they would learn it in time, they would have difficulty in communicating this information amongst their peers.

• To illustrate the problem more graphically, here is the same instrument with a keyboard that is probably

foreign to you:

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Sample of improved usability:

• The up function is on the “+” button in all menus

• The down function is on the “-” button in all menus

• The select function is on the “*” button on all menus

• The habits of use developed while using one menu are applicable to all menus

• The “+” and “-“ symbols are more likely to be intuitively understood by an international audience

• The function “+” and “-” buttons can also be used to increment values up and down, and to move

pointers etc in an intuitive way. Any choices that cannot be intuitively associated with a “+” or a “-“ are offered as menu choices that are selected with the ”*” button.

• Whatever the native language of users, they will have names for the plus, minus, and star symbols and

so will be able to talk about them effectively in their own language. • Screen space is not used to explain the function of the buttons so more space is available for other

uses.

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Sample Dialogue: Let us assume an instrument that has a function to show the ambient radiation as follows:

(A) As the dose rate increases, the number increases and the sliding scale moves further to the right. If the instrument is moved, the display updates and might look as shown below:

(B) Comparing examples (A) and (B): Although the radiation has increased from 83 n Sv/h to 132 n Sv/h, the slider actually moved to the left because the slider range was dynamically changed by the instrument from 100 to 1000.

Numerical value shown in

easy-to- read numbers

Units of measure clearly shown here

Scale gives visual indication of numerical

value

Scale gives visual indication of the new reading

Units of measure again shown clearly here

Numerical value again shown in easy-to-read

numbers

Maximum value clearly shown here

Maximum value revised and clearly shown here

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If the instrument is again moved, the display might look as shown below:

(C) Comparing examples (A) and (C): Although the radiation has increased by a factor of 300.000, the slider actually moves back to the left because the value changed from 83 to 26. However, the units of measure were dynamically changed by the instrument from nSv/h to mSv/h. The subtle change of “n” to “m” could be overlooked by users.

Scale gives visual indication of the new reading

Units of measure revised and again shown clearly

here Numerical value again shown in easy-to- read

numbers

Maximum value clearly shown here

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Sample of improved usability:

• All increases in radiation produce increases in the same direction • Small increases are shown by vertical movement • Cumulative vertical increases are shown as horizontal movement • Shorthand terminology has been introduced for communication between non-technical users

For example, “A reading of C1 or more indicates a life-threatening dosage”. • Units are written out in words nano, micro, milli to allow non-technical users to communicate with

technical users. Advantages of improved consistency: • Users are less likely to overlook potentially dangerous situations • Users can develop habits of use that allow them to become expert in operating the equipment. • As their expertise increases, they will need less time to perform a given task. • They also gain confidence, which allows them to concentrate their attention on their job and less on the

instrument they are using. Consistency therefore contributes to improved cost/benefit and reduces system running costs as more work can be done with a given manpower resource or, less manpower resource will be necessary to complete a given amount of work.

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3.3. Navigation Feedback It is often necessary to arrange functions in a hierarchy of menus that users navigate. If users know where they are at any given time, they are more likely to be able to navigate more efficiently. Sample menus I: Here is an example of a menu hierarchy that gives users access to various functions:

• The disadvantage of the above example is that on any given menu, users have no indication of where

they are in the menu hierarchy. An example of improved usability: In the top right-hand corner, there is an indication of the menu’s position in the hierarchy.

• There is a clear indication of where users are in the hierarchy, and how many steps are necessary to

return to the main menu.

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Sample Menus II:

• Assume that privileged users are given the left-hand menu, and have access to the “Software” choice. • Ordinary users are prevented from accessing the “Software” choice and it is removed from their menus

as shown on the right. An example of improved usability:

• As above, privileged users are given the left-hand menu, and have access to the “Software” choice. • Ordinary users are given the right-hand menu and do not have access to the “Software” choice, but the

choice has not been removed, it has been greyed out. Note On LCD displays that do not support greying out, the choice can be retained but be marked as unavailable with, for example, an asterisk, an “X”, or the text “unavailable”.

• The advantage of leaving a choice in the menu is that any numbering users may have learned or are told by peers (on the ‘phone for example) remains valid no matter whether they are privileged or non-privileged users.

• Users can also develop interaction habits that remain valid as they progress from ordinary to privileged users.

Advantages of improved navigation feedback: • Over time, users will remember the IDs of frequently used or otherwise useful functions, and will be able

to navigate to them without needing to carefully read the menu prompts. • This increased efficiency will, over time, reduce the running costs of the system.

“Software” choice removed from this

menu

“Software” choice retained, but greyed out

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3.4. Status Feedback Instruments are sometimes designed so that they can perform certain operations only in a given state or “mode”. Users find modal-operation limiting and sometimes confusing, so it is a good design philosophy to eliminate modes as far as practicable. However, modes are sometimes unavoidable, for example to protect certain classes of user from sensitive areas of operation. In such cases users must be made aware of the capabilities of the current mode, so that they appreciate what the instrument is currently able to do, and know what might have to be done to get to the other modes of operation if necessary. Consider an instrument that protects users by providing a visual and acoustical feedback when it detects dangerous levels of ambient radiation. The dialogs for changing the levels at which the alarm sounds should be protected from inadvertent changes that might prevent the instrument from warning of dangerous levels of radiation. The best way to protect this feature is to introduce a mode that is outside the normal hierarchy and is accessed perhaps with a password, or special key sequence known only to appropriate users. Although only certain users will be able to change the settings, there must be feedback of the instrument’s current settings to all users, so they are fully aware of the amount of protection a given instrument is currently giving them. Sample dialogs:

• An advanced dialog gives only privileged users the ability to change alarm settings. • The disadvantage is that ordinary users are not shown the settings that have been made, and so are

unaware of the level of protection the instrument is giving them. An example of improved usability:

• The same advanced dialog gives only privileged users the ability to change alarm settings • However, ordinary users can see the settings that have been made, as reflected by the arrows in the

middle and rightmost diagram. • Users will immediately notice an instrument that has settings outside the normally expected ranges, as

shown by the arrows on the rightmost diagram. • This is particularly important when an instrument may be shared among a group of users.

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3.5. Progress Feedback If an operation is likely to take some time to complete (more than 4 seconds for example), it is important to give users feedback such as: • Expected duration • Progress so far This will give users feedback to enable them to plan the use of their time and make changes to their plans if necessary. It is useful to provide a “Cancel” function so that users can stop an operation if it takes too long. Sample dialog I:

• Users have to wait for the operation to complete

• They cannot cancel the operation

• They have no indication of the possible duration.

An example of improved usability:

Users still have to wait of course, however: • They can cancel if the process takes inconveniently long to complete

• The movement of the progress indicator gives an indication of the possible duration Note: It is important that the progress indicator gives a smooth operation so that users can more accurately predict the amount of time needed for completion. For example, if some spectra had to be downloaded and the indicator shows the progress, there would be a smooth operation if the number of bytes transferred was monitored. The operation would be less smooth if the number of spectra transferred was monitored because the spectra data could be of differing sizes. Complicated spectra would take longer than small ones to download but, irrespective of size, would be shown as a single increment on the progress indicator.

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Sample dialog II: In this example, the operation is assumed to consist of individual steps (downloading spectra 1, 2, and 3). The progress indicator is used to show the progress within each step. When one step completes, the indicator returns to zero again, as shown:

• The progress indicator shows the progress within each step

• When a step completes the indicator shows 100%

• When the next step starts, the indicator starts again from zero

• There is no indication of the total number of steps to complete.

An example of improved usability:

• The progress indicator shows the progress over the whole process

• When one step completes, the indicator continues in a forward direction

• Users are shown the total number of steps to complete

• When it the progress indicator reaches 100% the whole process is complete

Spectrum 1: 89% complete



Overall process 21% complete



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3.6. Error Message Feedback Error messages are used to communicate out-of-line situations to users. Error messages are sometimes written by developers to explain to users what they have done wrong. This may seem appropriate from the developer’s point of view, but what users really needs is not an explanation of what they have done wrong; they need to know what has to be done to get out of the current situation and continue with productive work. To get the best benefit from error messages it is important that they help users to know how to proceed. Here are some real error messages that initially tell users what has gone wrong, and have been improved to help users recover more quickly:

Sample messages: Examples of improved usability: Timeout has occurred – you have been logged off Timeout has occurred – please log on again Insufficient storage to complete this operation Increase the size of the paging area and try again This operation cannot be performed here Switch to DEFINE mode to perform a DELETE

operation Required field cannot be blank Enter a name Invalid date Start date cannot be later than finish date Invalid date Month value is outside acceptable range <1–12> File not found Check that diskette “ABCD” is properly inserted

When it becomes necessary to issue an error message, users may well be in difficulties. It is important to review error messages to make sure they are formulated to help users, not to scold them. Writing error messages is a professional job that is best undertaken by someone with appropriate communication skills – a technical writer. Error message writing should not be considered as a sideline of the developer. Developers should work closely with their technical writer to come up with effective formulations.

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3.7. Forgiveness An interface that is forgiving will anticipate errors, and protect users against them. A forgiving interface also gives users the confidence they need to explore its functions and allow them to learn quickly through trial-and-error. Let us assume a hand-held device that works from an internal battery that has to be recharged. Sample design:

• Plug in low-voltage power lead to charge the instrument

• Device must be turned OFF to charge

• State-of-charge can only be seen when the instrument is turned ON

• User must remember to plug in the charging cable. An example of improved usability:

• Bays are provided to store the instruments (unit is possibly wall-mounting)

• As soon as an instrument is dropped into a charging bay, it starts charging

• Instruments charge independently of whether they are switched ON or OFF

• The state-of-charge is shown by coloured indicators (LEDs)

• If an instrument that is not fully charged is removed from its charging bay,

an acoustic beep and screen message warns users that the instrument may not be sufficiently charged for fault free operation.

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Further samples of improved usability: • Require user confirmation before letting the interface perform any actions that destroy data

• If defaults are offered, make sure the choice is to save data rather than destroy it

• If data can be destroyed, provide an UNDO function to recover it when necessary

• Support successive UNDOs as opposed to one UNDO and then a REDO

• Remind users to save any data they may have changed but have not yet stored

• Warn users well in advance of any low-battery condition

• Deactivate currently invalid choices to prevent users from inadvertently selecting them.

3.8. Naming The task of “naming” items is best given to a professional “wordsmith”; a technical writer who understands the language of the end users and also understands the functions of the instrument. The following things should be borne in mind:

• Things that are the same should be given the same name

• Things that are different should be named differently

• Names should be put into families, so that similar things are similarly named

• Wherever possible choices on menus should be action words (verbs)

• All menus should be titled

• The title of a menu should be the same as the choice that was used to access it.

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3.9. Ease of use While the initial “thrust” of development must be to make an instrument that functions well, the later stages of development will refine the design for practicability. Sample design:

• For keyboard operation users will need to use two-hands: − One hand to hold the instrument by the carrying handle − The other to press the keyboard buttons

• If the instrument will be used together with a hand-held remote sensor,

keyboard operation becomes very difficult: − One hand to hold the instrument − The other to hold the remote sensor and press the keyboard buttons

• Leftmost and rightmost function buttons are so far from the handle that thumb operation is not

possible

• Users have to learn the function of several keys (keyboard + function keys).

Keyboard

Buttons for most frequently used

functions

Carrying handle

Optional remote sensor

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An example of improved usability: All functions have been incorporated into the menus that can be operated by three buttons conveniently-located at the front end of the instrument’s carrying handle, as shown:

• All functions can be accessed using the three buttons

• Button positioning is suitable for left-handed or right-handed users

• All user controls are concentrated in a single place

• Once user understand the three buttons they can access all functions

• Single-hand operation leaves the user’s other hand free for other tasks: − Opening and closing doors, car compartment lids etc. − Holding a remote sensor − Making written notes

Navigation buttons:

+ Up - Down * Select

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4. Hardware Usability Guidelines 4.1. Handling Consider carefully the conditions under which instruments will be used:

• Ambient lighting • Air temperatures • Relative humidity • Noise levels • Available space • Amount of time the instrument will be carried • Usage Scenarios

All of these items will affect the instrument design, as follows: Ambient lighting: If the ambient lighting can be expected to be low the following should be considered.

• Backlighting of screen • If there are several buttons, they may need to be provided with internal lighting also • Larger battery capacity may be necessary to accommodate the backlighting power requirements • Backlighting should turn off after a time of inactivity to conserve battery power.

Air temperatures:

• If the ambient temperature can be expected to be high, users’ hands may become sweaty so hand-held instruments should be provided with a secure grip to prevent it being dropped.

• If the ambient temperature can be expected to be low, buttons should be designed to allow users to operate them with gloves on.

Relative humidity:

• If the relative humidity can be expected to be high, as with high temperatures, users’ hands may become sweaty so hand-held instruments should be provided with a secure grip.

Noise Levels:

• If the noise levels can be expected to be high, any acoustic alarms should be accompanied by a visual indication.

Available space:

• If the area to be searched is small or cramped (the corners of a car luggage compartment for example) the instrument and/or the instrument sensor must be sufficiently small to allow easy access.

Amount of time the instrument will be carried:

• If users can be expected to keep the instrument with them for long periods, it will require various qualities including: − Being as light as possible (so the user’s hands do not become tired) − Being finely balanced (so the instrument is pleasant to hold) − Capable of being carried on the belt (so users can perform other tasks) − Indicator of state-of-charge (so batteries do not run out in the middle of a task)

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Usage scenarios: Consider exactly how the instrument will be used and what ancillary tasks users will perform. This will affect the instrument design as shown below:

• User needs one hand free to open and close compartments and doors etc. − Single handed operation is a requirement

• Users wear protective clothing:

− Display must be clearly readable even if viewed through a visor − User must be able to operate device even when wearing protective gloves

• Users work standing and need to make written notes or fill out forms etc.

− Instrument needs some kind of holster so user release it to use both hands to hold pen and clip-board

• Users generally perform tasks in a pre-defined sequence − The required functions could be grouped together to provide easier and faster access − Some authorities work differently from others, a customisation function may be necessary to let

them group together the functions that they most frequently use

4.2. Controls • Reduce controls (buttons) to the absolute minimum • Suggested minimum:

− Up − Down − Select

• Locate controls so that one-hand (thumb) operation is possible 4.3. User control As discussed previously, it is important that users feel they are in control at all times so they can explore the instrument’s features and functions and quickly become productive with it. The basic requirements for this are:

• Provide quick and simple access to frequently used functions

• Provide an “undo” function if data can be irrevocably destroyed

• Provide a “Cancel” option to allow unwanted functions to be terminated

• Give feedback on long running processes

• Give feedback on position in menu hierarchy

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4.4. Use of Icons While icons can communicate complicated messages very economically, great care must be exercised in their use as some icons do not necessarily cross cultural borders, or speak to end users in a language they understand. Motifs: In some cultures, body parts including hands, fingers, head, ears, lips, nose, etc are considered profane, and should therefore be avoided in designs. Content: Search for simplicity in icons, as, on interfaces and in the documentation, icons may need to be reduced to just a few lines. If they are over complicated, they may become a source of confusion for users. Differentiation: It is important that icons can be differentiated easily. Avoid making icons look too much alike, for example, by putting a box around them. Since users tend differentiate icons by their shape or silhouette, the box would make differentiation difficult. Colour: This is probably not going to be an issue with the first generation of radiation monitors, but as screen technology improves, colour may be introduced to improve functionality. Be aware that 10% of the male population is colour-blind and that the most frequent colour-blindness is red/green. Use colour to help differentiate between items, but make sure that colour is not the only indicator as follows: Sample Icons:

The above icons use colour to differentiate them; so red/green colour-blind users will have difficulty in distinguishing one from the other. Example of improved usability

The above icons use colour and shape to differentiate them; so colour-blind users will be able to distinguish one from the other as easily as colour-sighted users. 4.5. Multi-language Support: Language-dependent items should be replaced by icons or removed from the hardware interface and be taken into the software interface where multiple language versions can be offered..

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5. Software Usability Guidelines 5.1. Dialogs

• Provide every dialog with a title

• Where appropriate provide the following standard buttons: − Cancel − Help

• Provide an “Undo” facility if appropriate

5.2. Menus • Provide every menu with a title

• Number menus to show their relative position in the overall hierarchy

• Number menu choices so users can remember the IDs of frequently used functions

• Provide a “Back” function so that, with repeated pushes, users are brought back to the main menu

• Display the main menu by default when the instrument is switched on and is ready for use

• All features, functions, and submenus must be accessible from the main menu

• Grey out, or leave empty, any choices that are currently unavailable 5.3. Controls • Keep the number of controls to a minimum

• Ensure the same name is used for controls throughout the interface 5.4. Feedback • Provide progress indicators for tasks expected to take more than 4 seconds • Show relative position in menu hierarchy 5.5. Error Recovery Support Ensure that, as far as possible, error messages don’t tell users what the problem is, but tell them what to do to remedy the situation. 5.6. Multi-language support • Allow users to set the language of the device to their preference

• Language-dependent items should be brought into the software interface and be removed from the

hardware interface • Architect software so that there is a single “engine” that provides the instrument’s functionality, and then

create the various language modules that can be attached to it. This allows for the simple addition of languages as needed, and also assures that the software functionality is independent of language.

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5.7. Error handling Ensure the version and level information is easily accessible, so that users can determine which software they are using. Ensure, as far as possible, that error messages tell users what to do, rather than telling them what they have done wrong. 5.8. Forgiveness Users make mistakes. The function of the interface is not just to offer functionality, but to inform and protect users so that any mistakes they may make do not have catastrophic consequences. 5.9. Accessibility If you provide supporting software that runs on computers, many government agencies require that the software is enabled for handicapped (physically challenged) users. The following will help you do this: Ensure all choices that are accessed by mouse interaction can also be accessed using keyboard interaction and/or shortcuts:

• Press the Tab key to move to (and automatically select) the desired function

• Press Alt-key together with a keyboard key to select the desired function

• Press ENTER to launch the function selected by either of the preceding methods

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6. Reading Tips There is a lot of literature to help you design better interfaces. Here are some reading suggestions to get you started: Reading Tips for Usability People:

Title: Essentials of Psychology Author: Dennis Coon Reference Number: ISBN: 0-314-02768-8 This book will give you a very good grounding in understanding people; what makes them do what they do; and how are they likely to respond to various stimuli. Title: The Psychology of Everyday Things Author: Donald A. Norman Reference Number: ISBN 0-456-06709 In this classic treatise, Donald Norman helps the reader understand how things are designed and how they ought to be designed.

Reading Tip for Designers:

Title: Designing the User Interface Author: Ben Shneiderman Reference Number: ISBN: 0-201-69497-2 This book by the prolific author Ben Shneiderman is a must on the bookshelves of all interface designers.

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7. Checklists 7.1. Planning Phase Checklist

Gather user requirements

Consider the following useful sources: • Usability review of a similar system or currently available version • Responses to questionnaires:

− Answered by users of the currently available version − Answered by potential users of the new product

• Difficulties with similar equipment or currently available version: − Problems frequently handled by help desks − Problems encountered by teachers and instructors

Define potential users

You need the following information to be able to design a product that suits your users: • Expected age range • Minimum formal education • Minimum previous experience • Motivation for using the product

Define user scenarios

Consider the types of tasks users will want perform and the most unfavourable conditions that theyare likely to encounter, including the following:

• Ambient lighting • Air temperatures • Relative humidity • Available space • Length of time the instrument will be carried

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7.2. Prototyping Phase Checklist

Design the product Develop product designs, while constantly referring back to the following:

• Requirements definitions • User definitions • Scenario definitions

Use the checklist (Usability guidelines) to aid in developing the design

Prototype the design Design should be documented in the form of a prototype that can take any of the following forms:

• Flipchart presentation • Slide show • Power-point presentation • Hardware and/or software demonstration

Check the design Plan and execute appropriate usability activities to check the design:

• Usability Review (with a single usability professional) • Usability Walkthrough (with a group end users)

Implement changes Review the design and, as appropriate, make necessary changes to the following:

• Requirements definitions • User definitions • Scenario definitions • Product design

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7.3. Testing Phase Checklist

Involve Users in design testing Plan and execute appropriate usability activities to check the design:

• Usability Walkthrough (with a group end users – takes 1 day) • Usability Test (with a group of users – takes 1 week)

Use the following checklists to aid in developing the design • Hardware Usability Guidelines

• Software Usability Guidelines

Implement changes Revise, as appropriate, any or all of the following:

• Requirements definitions • User definitions • Scenario definitions • Product designs

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7.4. Building Phase Checklist

Involve professionals during building Plan and execute appropriate usability activities to check the design:

• Usability professional (on-going monitoring of usability status) • Technical writer (terminology and documentation)

Use the following checklists to aid in building the product: • Hardware Usability Guidelines

• Software Usability Guidelines

Implement changes Revise, as appropriate, any or all of the following:

• Requirements definitions • User definitions • Scenario definitions • Product designs

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7.5. Hardware Usability Checklist

Handling In formulating your design, consider the following:

Ambient lighting Air temperature Relative humidity Noise levels Available space Time instrument will be carried

Controls Reduce controls to an absoluter minimum

User control Provide simple access to frequently used functions

Provide “Cancel” function Provide progress indicators for tasks longer than 4 seconds Give feedback on position in menu hierarchy Provide “Undo” function where appropriate

Icons Keep icons simple (use services of graphics artist)

Icons Easily differentiated? No colour-only differences Review motifs for cultural acceptability (use real end users)

Multilanguage support Multiple languages supported?

Language-dependent code is outside of core code?

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7.6. Software Usability Checklist

All Dialogs Have title

“Cancel” button “Help” button “Undo” button if appropriate

Menus Indication of relative position in hierarchy

All choices numbered Repeated “Back” function brings user to Main Menu Main menu displays when instrument becomes ready All features, functions and sub-menus accessible from main menu Unavailable choices greyed out or otherwise disabled

Controls Number of controls kept to minimum

Names used consistently

Feedback Indication of relative position in hierarchies

Forgiveness Interface protects users against errors

Error recovery Error messages help resolve the problem

Accessibility for PC components All choices accessible without mouse

Readers’ Comment Form Manuals like this are used by different people for different purposes. We are interested in any feedback you have about how useful you found it, and what additions, deletions, or other changes that would make it more useful to you. You may use this page for your comments, or e-mail your comments, quoting document number 42_Useabilty guide IAEA RPT27.doc - 1.0, to the following address: [email protected]

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