safety analysis report aea technology qsa, inc. model 770 type … · 2012. 11. 18. · aea...

Safety Analysis Report

AEA Technology QSA, Inc.

Model 770 Type B (U)-85

Transport Package

September 2002

Revision 2

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SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 Burlington, Massachusetts Page ii

CONTENTS

SECTION 1: GENERAL INFORMATION ............................................................. 1

1.1 Introduction .................................................................................................................................... 1

1.2 Package D escription ....................................................................................................................... 1 1.2.1 D escription of the 770 Transport Package ......................................................................... 1 1.2.2 Technical Data ......................................................................................................................... 4

1.3 Contents of Packaging .................................................................................................................... 4

1.4 Containm ent Boundary .......................................................................................................... 4 1.4.1 Containm ent V essel ........................................................................................................ 4

SECTION 2: STRUCTURAL EVALUATION .................................................... 5

2.1 Structural D esign ............................................................................................................................ 5 2.1.1 Overview .................................................................................................................................. 5 2.1.2 Design Criteria ......................................................................................................................... 5

2.2 W eights and Centers of Gravity ................................................................................................ 5

2.3 M echanical Properties of M aterials ......................................................................................... 6

2.4 General Standards for All Packages ......................................................................................... 6 2.4.1 M inim um Package Size ....................................................................................................... 6 2.4.2 Tamper Indicating Feature .................................................................................................. 6 2.4.3 Containm ent System ........................................................................................................ 7 2.4.4 Chem ical and Galvanic Reactions ....................................................................................... 7

2.5 Lifting and Tie-down Standards for All Packages ................................................................... 7 2.5.1 Lifting D evices .................................................................................................................... 7

2.5.2 Tie-down Devices .................................................................................................................... 8

2.6 N orm al Conditions of Transport ................................................................................................ 8 2.6.1 H eat .......................................................................................................................................... 8

2.6.1.1 Engineering Analysis ....................................................................................................... 8 2.6.2 Cold ....................................................................................................................................... 11 2.6.3 Reduced External Pressure ................................................................................................ 12 2.6.4 Increased External Pressure .............................................................................................. 12 2.6.5 Vibration ................................................................................................................................ 12 2.6.6 W ater Spray ........................................................................................................................... 12 2.6.7 Free Drop ............................................................................................................................... 13 2.6.8 Com er drop ............................................................................................................................ 13 2.6.9 Com pression .......................................................................................................................... 13 2.6.10 Penetration ............................................................................................................................. 13 2.6.11 Sum m ary ................................................................................................................................ 13

2.7 Hypothetical Accident Conditions of Transport ..................................................................... 14 2.7.1 Free D rop ............................................................................................................................... 14 2.7.2 Puncture ................................................................................................................................. 14

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2.7.3 Crush ...................................................................................................................................... 14 2.7.4 Thermal .................................................................................................................................. 14 2.7.5 Immersion - Fissile M aterial .............................................................................................. 15 2.7.6 Immersion - All Packages .................................................................................................. 15 2.7.7 Summary of Damage ............................................................................................................. 16

2.8 Special Form ................................................................................................................................. 17

2.9 Fuel Rods ...................................................................................................................................... 17

SECTION 3 - THERMAL EVALUATION ........................................................ 18

3.1 Description of Thermal Design Characteristics ..................................................................... 18

3.2 Summary of Thermal Properties of M aterials ....................................................................... 18

3.3 Technical Specifications of Components ................................................................................ 18

3.4 Thermal Evaluation for Normal Conditions of Transport ..................................................... 19 3.4.1 Thermal M odels ..................................................................................................................... 19 3.4.2 M aximum Temperatures .................................................................................................. 19 3.4.3 M inimum Temperatures ..................................................................................................... 19 3.4.4 M aximum Internal Pressures ............................................................................................. 19 3.4.5 M aximum Thermal Stresses ............................................................................................. 19 3.4.6 Evaluation of Package Performance for Normal Conditions of Transport ........................ 20

3.5 Thermal Evaluation for Hypothetical Accident Conditions of Transport ............................ 20 3.5.1 Thermal M odel ...................................................................................................................... 20 3.5.2 M aximum Internal Pressure .............................................................................................. 20 3.5.3 M aximum Thermal Stresses ............................................................................................. 20 3.5.4 Evaluation of Package Perform ance for Thermal Test ..................................................... 20

3.6 Thermal Analysis Details ...................................................................................................... 21

3.6.1 Surface Temperature Analysis ........................................................................................... 21 3.6.2 Model 770 Transport Container Type B(U) Source Capsule Thermal Analysis .............. 21

3.6.3 Source Assembly Thermal Analysis ..................................................................................... 22

SECTIO N 4: C O NTA IN M ENT ......................................................................... 24

4.1 Containment Boundary ................................................................................................................ 24 4.1.1 Containment Vessel ............................................................................................................... 24 4.1.2 Containment Penetrations ................................................................................................ 24 4.1.3 Seals and W elds ..................................................................................................................... 24 4.1.4 Closure ................................................................................................................................... 24

4.2 Requirements for Normal Conditions of Transport ................................................................. 24 4.2.1 Containment of Radioactive M aterial .............................................................................. 24 4.2.2 Pressurisation of the Containm ent Vessel .......................................................................... 24 4.2.3 Containment Criterion ....................................................................................................... 24

4.3 Contaitment Requirements for Hypothetical Accident Conditions ........................................ 25

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4.3.1 Containment of Radioactive Material .............................................................................. 25 4.3.2 Containment Criterion ...................................................................................................... 25

4.4 Special Requirements .................................................................................................................... 25

SECTION 5: SHIELDING EVALUATION ...................................................... 26

5.1 Design Features ............................................................................................................................ 26

5.2 Source Specification ..................................................................................................................... 26 5.2.1 G am m a Source ....................................................................................................................... 26

5.3 Model Specification ...................................................................................................................... 26

5.4 Shielding Evaluation .................................................................................................................... 26

SECTION 6: CRITICALITY EVALUATION .................................................... 31

SECTION 7: OPERATING PROCEDURE ...................................................... 32

7.1 Procedure for loading the package ......................................................................................... 32

7.2 Procedure for unloading the package ..................................................................................... 33

7.3 Preparation of an emptypackagefor transport ..................................................................... 34

SECTION 8: ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ..... 35

8.1 Acceptance test .............................................................................................................................. 35 8.1.1 V isual Inspection ................................................................................................................... 35 8.1.2 Structural and Pressure Tests ........................................................................................... 35 The swage coupling between the source capsule and the cable of the source assembly is subjected to a static tensile test with a load of 445 N (100 lbs) .............................................................................. 35 8.1.3 Leak Testing .......................................................................................................................... 35 8.1.4 Com ponent Tests ................................................................................................................... 35 8.1.5 Tests for Shielding Integrity ............................................................................................. 35 8.1.6 Thermal Acceptance Tests ................................................................................................ 36

8.2 Maintenance program .................................................................................................................. 36

APPENDIX A .............................................................................. Drawings

APPENDIX B ......................................................................... Test Plan 114

APPENDIX C .................. : ............................................. Test Plan 114 Report

APPENDIX D .......................................... Microshield Transmission Calculations

APPENDIX E ...................................................... Test Plan 88 Summary Results

AEA Technology QSA Inc Burlington, Massachusetts

SAR Model 770 Transport Package Revision 2 - September 2002

Page v

List of Figures and Tables

Figure 1-1: Figure 1-2: Figure 5-1:

Model 770 Source Changer Cutaway View of 770 Source Changer (from Top) Shield Assessment Configuration

Table 1-1: Radionuclide Capacities for the 770 Table 1-2: Radionuclide Decay Energy Table 2-1: Mechanical Properties of Principal Package Materials Table 2-2: Insolation Data Table 2-3: Summary of Damages During Performance of TP 114 Table 3-1: Thermal Properties of Principal Package Materials Table 5-1: Co-60 Normal and Hypothetical Accident Dose Rate for 770 Table 5-2: Transmission Shielding Assessment for 770 Table 5-3: Shielding Transmission Parameters Table 5-4: Ir-192 Normal and Hypothetical Accident Assessed Dose Rates for

770 Table 5-5: Cs-137 Normal and Hypothetical Accident Assessed Dose Rates

for 770 Table 5-6: Sc-46 Normal and Hypothetical Accident Assessed Dose Rates for

770

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 Burlington, Massachusetts Page 1 of 37

SECTION 1: GENERAL INFORMATION

1.1 Introduction

This document is issued in support of the application by AEA Technology QSA Inc. for full renewal of the Certificate of Compliance for the Model 770 Type B(U) Radiographic Source Changer. The container has been assigned USNRC Certificate of Compliance Number 9148 for domestic shipments and IAEA Certificate of Competent Authority No. USA/9148/B1(U) for international shipments.

The package is identified as the Model 770 and is to be approved as Type B(U)-85 in accordance with the Code of Federal Regulations, 10 CFR Part 71 and the IAEA Regulations for the Safe Transport of Radioactive Material, Safety Series No. 6, 1985 Edition (As Amended 1990). The Model 770 is manufactured under a NRC approved QA program 71-0040.

The Model 770 is designed to be used for the transport of special form sources contained within a locating assembly within the device shielding. Typical uses include source transport and transfers for sources used in industrial radiography applications.

The source assembly contains a source capsule that has met special form requirements. The Model 770 can contain a maximum of two source assemblies when fully loaded and can hold the following radionuclide activities:

Table 1-1: Radionuclide Capacities for the 770

Radionuclide Maximum Container Curies

Co-60 800

Ir-192 1,000

Cs-137 1,000

Sc-46 800

1.2 Package Description

1.2.1 Description of the 770 Transport Package

The Model 770 serves as the storage and transport packaging for radioactive sources and consists of the following major components:

"* Stainless steel external container weldment "* Stainless steel internal container weldment enclosing a depleted uranium shield "* Two lock assemblies "* Stainless steel shipping plates "* Stainless steel rectangular skid tubes


The device consists of a stainless steel housing (exterior weldment) which encloses an internal stainless steel weldment frame surrounding the depleted uranium shield. The shield is secured in position by use of stainless steel restraining blocks around the shield base and two semi-circular stainless steel restraints above the shield. The restraining blocks and upper restraints are welded to the inner weldment. Once assembled, the shield becomes fixed in place within the internal weldment. The shield is further restrained in the internal weldment by use of a two part polyurethane potting compound which solidifies to a density of 65 lb-ft3. All surfaces of the shield where it could contact the stainless steel are covered using a 0.01 in thick copper separator. This is done to prevent the formation of a eutectic between the shield and the steel surfaces.

The internal weldment frame incorporates an exo-skeleton comprised of stainless steel tubing. This tubing acts as a shock absorber and impact limiter for the device as well as adding significant strutural integrity. The internal weldment is then further enclosed within the final external weldment.

The depleted uranium shield provides the primary radiation protection for the Model 770. When the sources are in the shielded position, the shield limits the transmission of gamma rays to a maximum dose level of 200 mR/hr at the package surface and 10 mR/hr at one meter from the surface of the package. There is a crimp and stop plug installed at the center of the source tube, prior to casting, which prevents the sources from exiting the opposite side of the source changer during loading into the fully stored position.

The Model 770 radioactive source assembly contains a source capsule that is qualified as Special Form and is contained within a source assembly for location and securement within the device. A maximum of two such source assemblies can be loaded within the 770 device for shipment, one in each side of the source tube of the device.

The device lock assemblies are key operated to prevent unauthorised personnel from actuating the device. The lock assembly secures over the teleflex cable which is part of the source positioning assembly. Once the lock is engaged and the source cap installed, source movement within the device is prevented, keeping the source in the shielded storage position.

The lock assemblies are recessed into the 770 exterior weldment. During transportation the lock assemblies are protected by a stainless steel shipping plate which is bolted to the external weldment and fitted with seal wires for evidence of tampering. The bolted plate is also slightly recessed into the external structure which protects it from direct shear in a drop or puncture situation.

The Model 770 is fitted with three, rectangular, stainless steel tubes welded to the base of the device. This allows for lifting/movement using a forklift and acts as an impact limiter for device drops on the base. The device must be lifted/handled by mechanical means as there are no lifting handles or eyebolts on the device.



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Figure 1-1: Model 770 Source Changer

INNER CONTAINER ASSEMBLY.

SOURCE CHANGER ASSEMBLY

SHIELD ASSEMBLY

TUNNEL

BODY

Figure 1-2: Cutaway View of 770 Source Changer (from Top)



Page 4 of 37

1.2.2 Technical Data

Length: Height: Width: Shipping Weight: Capacity: Shielding: Transport Status: Sources: Descriptive Drawing:

61 cm (24 inches) 50 cm (192 inches) 57 cm (22Y2 inches) 437 kg (970 pounds) See Table 1-1 Depleted Uranium 191 kg (425 pounds) Type B(U) Package USA/9148/B(U) Special form R77090 Revision B

1.3 Contents of Packaging

The Model 770 Source Changer is designed for the use and transport of special form sources in activities listed in Table 1-1. The radioactive material is contained in metallic capsules, which are seal welded. The maximum activity, for industrial radiography sources, is defined as output Curies as required in ANSI N432 and 10 CFR 34.20. The heat output of the sources is shown in Table 2 below:

Table 1-2: Radionuclide Decay Energy

Radionuclide Package MeV/Decay Watts/Package Activity (Ci) I

Iridium-192 1,000 1.46 8.6 Co-60 800 2.82 14 Sc-46 800 2.73 13 Cs-137 1,000 1.18 7 Resource references:

Table of Isotopes, Volumes I & II, Eighth Edition. John Wiley & Sons, Inc., 1996.

1.4 Containment Boundary

1.4.1 Containment Vessel

The containment system for the Model 770 Source Changer is the special form radioactive source capsule referred to in Section 4.1.1 of this application. The capsules have been certified as Special Form Radioactive Material in accordance with 10 CFR Part 71, US Department of Transportation regulations (or equivalent), and IAEA Safety Series No. 6, 1985 (As Amended 1990).

1.5 Drawings

Descriptive assembly drawings for the 770 are included in Appendix A to this SAR.


SECTION 2: STRUCTURAL EVALUATION

This section identifies and describes the principal structural engineering design of the packaging, components and systems important to safety and to compliance with the performance requirements of 10 CFR Part 71.

2.1 Structural Design

2.1.1 Overview

The Model 770 has three rectangular steel tubes welded to the base of the unit which provide some impact limiting effects. In addition, the internal welded, stainless steel framework provides additional shock-absorbing benefits as well as limiting shield movement relative within the device. The principal shielding and security assemblies are protected by the projector weldment and shipping plates during transportation. The exterior shell of the package is made from grade 300 Series stainless steel, which is not susceptible to brittle fracture at reduced temperatures. There are no components of the package which may be considered brittle within the temperature range -40'C to +38°C (40'F to 100.4°F) other than the depleted uranium shield, which is brittle thoughout the entire range.

The Model 770 has no features suitable for tie-down, and lifting must be accomplished by mechanical means either from the base or from above using rigging.

The Model 770 is comprised of six (6) structural components: a source capsule, source holder assembly, shield assembly, inner housing weldment, outer housing weldment and locking assembly. The source capsule is the primary containment vessel for the radioactive material. The shield assembly provides shielding for the radioactive material and, together with the source holder assembly and locking assembly, assures proper positioning of the radioactive source during transport and storage.

2.1.2 Design Criteria

The Model 770 Source Changer is designed to comply with the requirements for Type B(U) packaging as prescribed by 10 CFR 71 and IAEA Safety Series No. 6, 1985 Edition (As Amended 1990). All design criteria are evaluated by a straightforward application of the appropriate section of 10 CFR 71 or IAEA Safety Series No. 6.

2.2 Weights and Centers of Gravity

The weight of the Model 770 Source Changer is 970 lbs (437 kg). The weight of the depleted uranium shield is 425 lbs (191 kg).

Based on the symmetry of the device, the center of gravity (C of G) is at the center of the device.



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2.3 Mechanical Properties of Materials

Table 2.1 lists the mechanical properties of the principal materials used in the package construction. The resources referred to in the last column of each are listed after the table.

Table 2-1: Mechanical Properties of Principal Package Materials

Material Tensile Yield Strength Elongation Resource SStrength I

Depleted Uranium 65 ksi 30 ksi 12% Ref: #1, p. 20-35

Stainless Steel gr304 gr303

75 ksi 85 ksi typical

30 ksi 35 ksi typical

40% typical 50% typical

Ref. #2, p. 368 Ref. #2, p3 6 8

Resource references:

1. Howard E. Boyer and Timothy L. Gall, Editors, Metals Handbook. Metals Park, Ohio: American Society for Metals, 1985.

2. J.R. Davis, Metals Handbook Desk Edition, 2nd Edition ASM International.

2.4 General Standards for All Packages

2.4.1 Minimnunm Package Size Reference: "* USNRC, 10 CFR 71.43(a) "* USDOT, 49 CFR 173A12(b) "* IAEA Safety Series No. 6, para. 525

The package is 24 inches (61 cm) long, 19½ inches (50 cm) high and 222 inches (57 cm) wide therefore exceeding the minimum package size requirements specified by 10 CFR 71.43(a) and IAEA Safety Series No. 6, para. 525.

2.4.2 Tamper Indicating Feature

Reference: "* USNRC, 10 CFR 71.43(b) "* USDOT, 49 CFR 173.412(a) "• IAEA Safety Series No. 6, para. 526

The shipping cover fixing bolts are fitted with a seal wire which provide a tamper indicating feature in accordance with 10 CFR 71.43(b) and IAEA Safety Series No. 6, para. 526.

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 1 Burlington, Massachusetts Page 7 of 37

2.4.3 Containment System

Reference "* USNRC, 10 CFR 71.43(c) "* USDOT, 49 CFR 173 A.12(d) "* IAEA Safety Series No. 6, para. 530

The containment system is the special form capsule referred to in Section 1.4.1 and 4.1.1 of this application.

2.4.4 Chemical and Galvanic Reactions Reference: "* USNRC, 10 CFR 71.43(d) "* USDOT, 49 CFR 173.410 (g) "* IAEA Safety Series No. 6, para. 512

The materials used in the construction of the Model 770 transport package are depleted uranium metal, stainless steel, polyurethane potting material and foam, brass and copper. To prevent the possible formation of a eutectic alloy from steel and depleted uranium during the Hypothetical Accident Conditions thermal scenario, defined by 10 CFR 71.73 (c)(4), the copper is used as a separator for all steel-uranium interfaces. With this construction there will be no significant chemical or galvanic reaction between package components during normal or hypothetical accident conditions of transport.

2.5 Lifting and Tie-down Standards for All Packages

2.5.1 Lifting Devices Reference: "* USNRC, 10 CFR 71.45(a) "* USDOT, 49 CFR 173.410 (b) "* IAEA Safety Series No. 6, para. 506 & 507

The Model 770 is designed to be lifted by the base using a fork lift. For this analysis, the base is assumed to be a flat, rectangular plate 24 inches (61 cm) long, 22½ inches (57 cm) wide, and 0.25 inches (6.4 mm) thick, simply supported along two sides, with a concentrated load at the centerline. Assuming 12.7 cm (5") forks and taking the support point at the center of the forks gives a support length of 35.6 cm (14"). The maximum stress on the base is:

a = PLc/4I

Where: P = The weight of the transport package 437 kg (970 lb) L = The length of the base between forks 35.6 cm (14 inches) c = Half the thickness of the base 0.32 cm (0.125 inches) I = The bending moment of inertia of the base 1.2 cm4 (0.029 in4)


Therefore, the stress generated in the skid is 14,634 psi. With a Safety Factor of 3 applied, the maximum stress in the skid is 43,901 psi. This is below the yield strength of the stainless steel base, 45,000 psi. Therefore, the lifting device is capable of supporting more than three times the weight of the transport package as required by 10 CFR 71.45(a).

2.5.2 Tie-down Devices Reference: "* USNRC, 10 CFR 71.45(b) (1) (2) (3) "• USDOT, 49 CFR 173.412 (I) "* IAEA Safety Series No. 6, para. 527

The Model 770 has no tie down attachments. The package can be blocked and braced according to standard transportation practices.

2.6 Normal Conditions of Transport

2.6.1 Heat Reference: "* USNRC, 1OCFR71.71(c)(1) "* IAEA Safety Series No. 6, para. 543

Table 2-2: Insolation Data

Surface Insolation for a 12 hour period (g cal/cm2) Horizontal base None

Other horizontal flat surfaces 800 Non-horizontal flat surfaces 200

Curved surfaces 400

The heat source in the Model 770 Source Changer is a maximum of 800 Curies of Cobalt-60. Cobalt-60 decays with a total energy liberation of 2.82 MeV per disintegration or 17.5 milliwatts per curie. Assuming all the decay energy is transformed into heat, the heat generation rate for the 800 Curies of Cobalt-60 would be approximately 14 Watts.

2.6.1.1 Engineering Analysis

This analysis determines the maximum surface temperature produced by solar heating of the package surface in accordance with 10 CFR 71.71(c)(1) and Table XII of IAEA Safety Series No. 6 (as amended 1990).

The model consists of taking a steady state heat balance over the surface of the package. In order to assure conservatism, the following assumptions are made:


"* The package is assumed to undergo free convective heat transfer and radiative heat transfer from the top and four sides.

"* The inside package faces are considered perfectly insulated so there is no conduction into the package. The faces are considered to be sufficiently thin so that no temperature gradients exist in the faces.

"* The package is approximated as a rectangular solid, 24" (610mm) long, 22 1/2" (572mm) wide and 19 1/2" (495mm) high.

"* The decay heat load (13.6 Watts) is added to the solar heat input load.

"* The steel surface of the package is silver stainless steel and therefore the emissivity coefficient is taken to be 0.6.

The maximum surface temperature is computed using the steady state heat balance relationship; heat input (Q,) equals heat output (Qut).

Q~n =Qout

Heat Input:

The total heat input is the sum of the solar heat input and decay heat.

Where,

Solar heat input:

The solar heat input is the combined solar heating of the top horizontal surface and four vertical side surfaces multiplied by the absorptive constant (V) for the material. The insolation data, provided in 10 CFR 71.71(c)(1), is found in Table 2-2, from which:

Top surface heat input: Qrr = 800 W/m2 x 0.35 meter2 = 280 W

Side surface heat input:: Qis = 200 W/m2 x 1.17m2 = 234W

Decay heat input: QDT = 13.6W

Absorptive constant V = 1.0 (most conservative)

Total heat input: Qi = V (Qrr + QIS )+ QDT = 528 W

Heat Output: The total heat output is the sum of the radiation and convection heat transfer (Reference: Heat Transfer, J.P. Holman, 4th Edition, 1976, p.253).

Radiation heat transfer (QR):

QR= B x E x ATs x {(Tw + 273)4 - (TA + 273)4}


Where:


Page 10 of 37

Stefan Boltzmann Constant, B = 5.669 x 10-' W/m2 OK4

Emissivity, E = 0.6 The top and side surface area of the package, ATs = 1.52m 2

The maximum surface temperature of the package, Tw 'C The ambient temperature, TA = 38°C

QR= 5.17 x 10.8 x {(Tw + 273)4 - (TA + 273)4)

Top surface convection (QT):

QT = Hr X AT x (Tw - TA)

Where:

..... equation 1.

..... equation 2.

The top surface area, AT = 0.35m2

The free convection coefficient for a flat horizontal surface is HT

From; Engineering Thermodynamics, Work and Heat Transfer. 4th Edition, Rogers and Mayhew, page 585.

For a heated plate facing up, HT = 1.32 {(O/I)/[K/m]}" 4

Where: and

Therefore:

o = Tw -TA I=LT LT is the average length of the top surface = (L + W)/2 = 0.59m

HT = 1.32 x {(1/LT)0 25 X [(Tw -TA)° 25 }

HT = 1.32 x{(1/0.59)0 25 x [(Tw -TA) 0 25]} HT = 1.51 x [rw -TA)°O2 1]

Substituting into equation 2. QT = 0.527 (Tw -TA)'

25..... equation 3.

Side surface convection (Qs):

Qs = Hs x As x (Ts - TA) ..... equation 4.

Where: As is the total side surface area , (1.17m2) Hs is the free convection coefficient for a flat vertical surface

From; Engineering Thermodynamics, Work and Heat Transfer. 4th Edition, Rogers and Mayhew, page 585.

For a vertical plate, Hs = 1.42 {(0/1)/[K/m]} "4

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Where: 0 = TW -TA and I = Ls

Ls is the average length of the side surfaces = (LT + H)/2 = 0.55m

Therefore: Hs= 1.42 * {(l/Ls ) 0 25 * [(Tw-TA)0 25 ]} Hs= 1.65 * [(Tw--TA)0 2

5]

Substituting into equation 4.

Qs = 1.9 (Tw -TA) 1 •25 ..... equation 5. Total heat output:

QoUr = QR + QT + Qs

and

QN= QR + QT + Qs = 520 W

Substituting for QR from equation 1, QT from equation 3 and Qs from equation 5

526 = 5.17 x 10"8 {(Tw + 273)4 - (TA + 273)4) + 0.527 (Tw -TA) 2 5 + 1.9 (Tw -TA)1. 25

Iteration of this relationship yields a maximum wall temperature (Tw) of 77'C (171'F).

This temperature would not adversely affect the package during normal transport since the melting temperatures of all safety critical components are well above this temperature. Additionally charring of the polyurethane foam will not begin to occur at such low temperatures. Therefore the package satisfies the requirements of 10 CFR 71.71(c)(1) and paragraph 543 of the IAEA Safety Series No. 6, 1985 Edition (as amended 1990).

Since each isotope loaded into the Model 770 (Section 1.1, Table 1-1) will be less than 800 Curies and generate less than 13.6 Watts as shown in Table 1-2, it can be assumed that no part of the package will be greater than 171"F or significantly effected by heating effects.

2.6.2 Cold

Reference: "* USNRC, 10 CFR 71.71 (c)(2) "* IAEA Safety Series No. 6, para. 516, 528 & 556

There are no components of the Model 770 that have increased susceptibility to failure by any mechanism at ambient temperatures of -40 0C.


2.6.3 Reduced External Pressure

Reference: "* USNRC, 10 CFR 71.71 (c)(3) "* USDOT, 49 CFR 173.412(f) "* IAEA Safety Series No. 6, para. 534

The Model 770 is not air tight. Although the shield is potted, the source tube is open to ambient pressures as there is no seal between the shipping covers and the outer weldment nor the inner weldment and the lock assemblies. Further, the the foam-potted space between the inner and outer weldments is also vented to ambient through three (3) holes in the top of the outer. As such, the entire Model 770 will always equalize to ambient pressures and will not be affected by pressure reductions.

The capsules' ability to withstand a reduced external pressure of 3.5 psi is defined in 3.6.2.

2.6.4 Increased External Pressure

Reference: * USNRC, 10 CFR 71.71(c)(4)

See 2.6.3. No differential pressures can build up by the same argument. None of the solid components are detrimentally affected by this magnitude of pressure. As such, the entire Model 770 will always equalize to ambient pressures and will not be affected by pressure reductions.

2.6.5 Vibration

Reference: "* USNRC, 10 CFR 71.71 (c) (5) "* USDOT, 49 CFR 173.410(0) "* IAEA Safety Series No. 6, para. 511

In the 19 years that the old Model 770 transport package has been in use, no transport packages have failed due to vibration. It is also well documented that fully welded structures will not be affected by normal incident vibration. It is therefore concluded that the revised Model 770 will withstand vibration normally incident to transport.

2.6.6 Water Spray Reference: "* USNRC, 10 CFR 71.71 (c) (6) "* USDOT, 49 CFR 173.465(b) "* 1AEA Safety Series No. 6, para. 621

Water spray preconditioning of the package was not performed. The Model 770 is constructed from waterproof materials throughout. Therefore, the water spray test would not reduce the shielding effectiveness or structural integrity of the package.


2.6.7 Free Drop

Reference: "* USNRC, 10 CFR 71.71(c) (7) "* USDOT, 49 CFR 173.465 "* IAEA Safety Series No. 6, para. 622

The 1.2 m (4 ft) free drop test was previously performed on the original version (carbon steel) of the Model 770 (TP 88 - see Appendix E) with the net result being an approximately 2.54 cm (1 in) triangle of deformation on the comer of the unit. This damage had no measurable effect on the dose profile for the unit. The current version of the Model 770 weighs approximately 45 kg (100 lbs) more than the previous version. The deformation to the current Model 770 will therefore be slightly greater. However, as the damage originally incurred was insignificant to the unit and, as such to the radiological safety of the package, the current version of the 770 will also meet this requirement.

2.6.8 Corner drop Reference: "* USNRC, 10 CFR 71.71 (c) (8) "* IAEA Safety Series No 6, para. 622

Not applicable, as the package does not transport fissile material nor is the exterior of the package made from either fibreboard or wood.

2.6.9 Compression Reference: "* USNRC, 10 CFR 71.71(c)(9) "* USDOT, 49 CFR 173.465(d) "* 1AEA Safety Series No. 6, para. 623

The compression test is assessed as part of the Test Plan Report 114. This assessment concludes that the 770 will meet the compression requirements of these references.

2.6.10 Penetration

Reference: "* USNRC, 10 CFR 71.71(c)(10) "* USDOT, 49 CFR 173.410(e) "* IAEA Safety Series No. 6V, para. 624

The penetration test is assessed as part of the Test Plan Report 114. This assessment concludes that the 770 will meet the penetration requirements of these references.

2.6.11 Summary Based on the assessment of the Model 770 package, it is concluded that the Model 770 transport package meets the Normal Conditions of Transport requirements.

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 Burlington, Massachusetts Page 14of37

2.7 Hypothetical Accident Conditions of Transport

2.7.1 Free Drop Reference: "* USNRC, 10 CFR 71.73(c)(1) "* USDOT, 49 CFR 173.467 "* IAEA Safety Series No. 6, para. 62 7(a)

Two Model 770 test packages were subjected to the 9 m free drop test in accordance with Test Plan 114. The results of these tests are documented in Test Report 114 and demonstrate that the Model 770 maintains its structural integrity and shielding effectiveness under the Accident Conditions of Transport free drop test.

2.7.2 Puncture Reference: "• USNRC, 10 CFR 71. 73(c)(3) "* USDOT, 49 CFR 173.467 "* IAEA Safety Series No. 6 Section VI, para. 627(b)

Two Model 770 test packages were subjected to the puncture test in accordance with Test Plan 114. The results of these tests are documented in Test Report 114 and demonstrate that the Model 770 maintains its structural integrity and shielding effectiveness under the Accident Conditions of Transport puncture test.

2. 7.3 Crush Reference: "* USNRC, 10 CFR 71.73 (c) (2) "* USDOT, 49 CFR 173.467 "* IAEA Safety Series No. 6, para. 62 7(c)

Since the activity shipped is less than 1,000A 2, and the source material is contained in special form sources, this test is not applicable.

2.7.4 Thermal Reference: "* USNRC, 10 CFR 71.73 (c) (4) "* USDOT, 49 CFR 173.467 "* IAEA Safety Series No. 6, para. 628

An assessment was performed to determine the ability of the Model 770 to pass the fire test based on previous satisfactory thermal tests performed on the Model 680 (CoC 9035) and 650L (CoC 9269) containers. (Reference Test Report TP72-S2 submitted with the SAR for the 680-OP and Test Report TPWO submitted with the SAR for the 650L). These reports detail thermal tests performed on the 680 projector with and without the overpack and on a cracked 650L respectively.


The 770 device is less susceptible to the thermal test than either unit by virtue of: "* The double shell construction, which acts as a thermal insolator in the fire test and

impact absorber. "* The relatively small distance between the inner container walls and the shield,

limiting its movement. "* A substantially more robust (welded verus bolted) shield support structure than the

Model 680. "* The polyurethane potting surrounding the 770 shield is more substantially robust and

compliant than the 680's foam potting. "* The welded construction of the 770 inner shell is more secure than the bolted end

plate design employed on the 680 and the 650L. "* Stainless steel, relative to the 650L inner container and 680 shell is substantially less

susceptible to impact fractures at temperatures within the testing range.

Failure as a result of pressure build up within the assemblies containing a trapped volume of air is not probable. The projector weldment has holes which allow venting of any trapped gases caused by degradation of the polyurethane foam surrounding the shield and inner structure.

Based on the minimal damage caused by the preceding destructive testing, no breach of the weldment occurred, therefore no air path to the shield was created which could cause uranium oxidation and resultant loss of effective shielding around the sources. It is concluded that the 770 package therefore meets the requirements of the fire test based on analysis of construction and condition of the units after the preceding destructive testing.

2.7.5 Imnmersion - Fissile Material

Not applicable.

2.7.6 Immersion - All Packages

Reference: "* USNRC, 10 CFR 71.73 (c)(6) "* IAEA TS-R-1, paragraph 729

The package would be subjected to an increased external pressure of 21.7 psig. The unit will withstand this pressure without loss of structural integrity.

The cylindrical special form source (primary containment) will be vulnerable to collapse due to the required assumed pressure increase of 21.7 psig. The source capsules are fabricated from Type 304 or 303 stainless steel. This analysis bounds any special form source capsule with a maximum inside diameter of 0.317 inch (8.05 mm) and a minimum wall thickness and weld penetration of 0.016 inch (0.41 mm). From Reference 1, the external collapsing pressure for a thin walled cylinder is:

Pcollapse = (t / R)(ay / (1 + (4ay/ E)(R / t)2))

Where: t 0.0 16 in (Weld Thickness)


R

E


Page 16 of 37

0.317 in (Inside Radius) 30,000 psi (Yield Strength) (Table 1) 28,000 ksi (Young's Modulus) (Reference 2)

From this relationship, the minimum collapsing pressure of the source capsule is 565 psi, which exceeds the required external pressure.

Resource references:

1. Young, Warren C. Roark's Formulas for Stress & Strain, Sixth Edition. McGrawHill: New York, 1989, p. 634.

2. Hibbeler, R.C. Mechanics of Materials. 2nd Edition, 1991.

2.7.7 Summary of Damage Table 2-3 summarizes the results of the Normal Conditions of Transport and Hypothetical Accident testing performed on the Model 770, in the sequence that the tests were completed.

Table 2-3: Summary of Damages During Performance of TP114

Specimen I Test Performed Test ResultsSn#10 Compression test Assessed No damage possible

1 meter (40 inch) penetration bar Assessed No damage possible

1.2 meter (4 foot) drop, top, long 0 Under TP88, a 1" triangular dent in edge down the comer.

0 Weight Differential assessment (current Model 770 is 100 lbs heavier) shows no significant additional damage possible.

Post-Drop Inspection * N/A

9 meter (30 foot) drop 0 Slight denting of impact surface Flat on side * Slight dent in top side from

secondary impact (rebound landing)

1 meter (40 inch) puncture Flat on same side as 9m drop

* Puncture bar indent in side.



Page 17 of 37

Specimen Test Performed Test Results

Post-Drop Inspection 0 All covers and lock assemblies attached.

* Brass lock in upper assembly (away from impact) dislodged from housing.

* Brass lock in lower assembly (closest to impact) ¾ out of assembly.

* Upper source tube slightly crimped.

• Lower source tube minimally crimped

0 Sources held secure, no movement. Sn#1 1 9 meter (30 foot) drop 0 Impact edge bent in approximately

Vertical edge down 2-inches. * Slight bowing of shipping plate

boss closest to impact. 1 meter (40 inch) puncture 0 Slight impact witness from Target bent vertical edge Puncttire bar.

Post-Drop Inspection 0 All covers and lock assemblies attached.

* Brass locks in assemblies remained within housings.

* Difficult to remove one (1) of the shipping cover bolts (closest to impact)

* Source held secure, no movement.

Based on these results and the discussions preceding, it is concluded that the Model 770 Transport package maintained its structural integrity and shielding effectiveness under the Hypothetical Accident Conditions of Transport Tests.

2.8 Special Form

The Model 770 projector is designed to transport AEA Technology QSA Inc. source capsules that have been certified as special form radioactive material by U.S. Department of Transportation (or equivalent). The minimum wall thickness/weld penetration for a special form capsule to meet the SAR requirements is 0.016 inches (0.41 mm) as specified in Section 2.6.4.

2.9 Fuel Rods

Not applicable.



Page 18 of 37 I

Section 3 - THERMAL EVALUATION

3.1 Description of Thermal Design Characteristics

The Model 770 transport package is a passive thermal device having no mechanical cooling or relief valves. The exterior surface finish of the package is light silvery stainless steel having an absorptivity of about 0.44, or a reflectivity of 0.56. Cooling of the package is through free convection and radiation. There are no specific cooling or insulating design features.

Pressure relief of the projector weldment is only necessary during the thermal test and is provided by the vent holes in the device weldment, see para. 2.7.4 of this SAR.

The maximum contents of the package is 800 curies of Co-60 which will give a decay

heat generation rate of approximately 14 watts.

3.2 Summary of Thermal Properties of Materials

Table 3-1 lists the thermal properties of the materials in the package. The resource referred to in the last column is listed below the table.

Table 3-1: Thermal Properties of Principle Packa ,e Materials Material Density Melting Point Linear Resource

(g/cm 3) (0C) Expansion (tim/mi•l

Depleted Uranium 18.6 1,135 12 Ref.#l

Stainless Steel gr 304 7.9 1,400-1,450 17 Ref. #1 gr. 303 7.9

Polyurethane Potting Material 1.0 Does not melt. NA Manufacturer's (Surrounding Shield) Char point Data Sheet

-200 Polyurethane Foam (Surrounding 0.13 Does not melt. NA Manufacturer's inner weldment) Char point Data Sheet

-200

Resource reference:

1. Metals Handbook. American Society for Metals, 8th Edition.

3.3 Technical Specifications of Components

Not applicable.


3.4 Thermal Evaluation for Normal Conditions of Transport

3.4.1 Thernial Models

The heat analysis in Section 2.6.1 demonstrated that under the conditions described in 10CFR 71.71(c)(1) the surface temperature of the transport container will be around 770C. At this temperature there are no components within the package which will suffer a reduction in effectiveness. The surface temperature analysis in Section 3.6.1 demonstrates that no accessible surface of the package will reach the maximum allowable temperature of 50° C (122*F) specified in 10 CFR 71.43 (g) in still air at 38°C (100'F).

3.4.2 Maximum Temperatures

The maximum temperatures encountered under Normal Conditions of Transport will have no adverse effect on the structural integrity or shielding efficiency of the package.

As shown in Section 3.6.1, the maximum surface temperature can not exceed 1 180 F (48°C) with the package in the shade and at an ambient temperature of 100'F (38'C). Therefore, the package meets the requirements of 10 CFR 71.43(g).

The maximum surface temperature generated by solar heating is 171 °F (77°C) as computed in Section 2.6.1.

A review of the temperature operating ranges, melting points and an assessment of the relative expansion coefficients of the structural components of the package (Table 3-1) indicates that there will be no reduction in the structural effectiveness of the package.

3.4.3 Mininmum Temperatures

The are no components of the Model 770 which would suffer a reduction in effectiveness at -40'C. Furthermore, Table 3-1 indicates that no significant stresses will be developed as a result of differential thermal contraction of the components of the package.

3.4.4 Maximumn Internal Pressures

The pressure generated within the projector weldment dunng normal operating conditions is significantly below that which would be produced during the Hypothetical Accident Conditions thermal test, which is shown in section 2.7.4 to be within the design limits of the package and to meet the performance criteria for hypothetical accident conditions.

3.4.5 Maxim um Thermal Stresses

A review of the materials of manufacture and their relationship within the package suggests that there will be no significant thermal stresses induced in any component part of the Model 770. In addition design clearances between fitted components are sufficient to allow for thermal expansion at the maximum temperature of 171 *F (77°C) and thermal contraction at the minimum temperature of -40'F (-40'C).


3.4.6 Evaluation of Package Performance for Normal Conditions of Transport

A review of the operating ranges and thermal properties of the structural components of the package shows that there will be no reduction in structural effectiveness, as the component temperature ranges are in excess of those calculated from the thermal models. Analysis of the materials of construction and the package design indicate that there is sufficient clearance between components to allow for thermal expansion and contraction.

The normal transport conditions will have no adverse effect on the structural integrity or shielding efficiency of the package.

3.5 Thermal Evaluation for Hypothetical Accident Conditions of Transport

3.5.1 Thermal Model

The analysis performed in Section 2.7.4 demonstrates that the Model 770 would withstand the temperatures and pressures associated with the thermal test specified in 10CFR 71.73(c)(4), and meet the performance criteria for Hypothetical Accident Conditions stated in 1OCFR 71.51 (a)(2).

3.5.2 Maximum Internal Pressure

Section 3.6.3 provides an analysis of the source capsule, which serves as the primary containment under thermal test conditions. This analysis demonstrates that the maximum internal gas pressure at 1,475°F (800'C) would be 54 psi (370 kN/m2).

Therefore, if the source capsule were to reach a temperature of 1,475°F (800°C), the maximum stress in the capsule would be only 2.6% of the yield strength of the material.

3.5.3 Maximum Thermal Stresses

A review of the materials of manufacture and their relationship within the package suggest that there would be no significant thermal stresses induced in any component part of the Model 770. In addition design clearances between fitted components are sufficient to allow for thermal expansion at the maximum temperature of 1,475°F (800'C).

3.5.4 Evaluation of Package Performance for Thermal Test

The analysis performed in Section 3.6.3 proves that the containment capsule would withstand the temperatures and pressures generated in the thermal test.

In addition, the analysis presented in Section 2.7.4 demonstrates that the metallic structure of the Model 770 would retain its structural integrity, holding the shield and source in position, therefore the shielding effectiveness would remain during and after the thermal test.

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 [ Burlington, Massachusetts Page 21 of 37 I

3.6 Thermal Analysis Details

3.6.1 Surface Temperature Analysis

Reference: * 10 CFR 71.43(g) * 1AEA Safety Series No. 6, 1985 Edition (as amended 1990), para 543.

This analysis demonstrates that the maximum surface temperature of the Model 770 Transport package will not exceed 50'C (122*F) with the package in the shade and an ambient temperature of 38°C (100'F).

To assure conservatism, the following assumptions are used:

"* The entire decay heat (13.6 watts) is deposited in the exterior surfaces of the package.

"* The interior of the package is perfectly insulated and heat transfer occurs only from the exterior surface to the environment.

"* For conservatism, it is assumed that 100% of the total heat is deposited in one side face.

"* The only heat transfer mechanism is free convection.

"* The side face undergoes one-dimensional convective heat transfer.

Using these assumptions, the maximum wall temperature (Tw) is found from

Tw = (q/hA) + TA

where

q is the heat deposited per unit time on the face, 13.6 watts h is the free convection heat transfer coefficient for air: 5 watts/m2

A is the surface area of a side face, 0.283 mn2

TA is the ambient air temperature, 38°C (311 k)

From this relationship, the maximum temperature of the surface is 47.6°C (1 18'F).

3.6.2 Model 770 Transport Container Type B(U) Source Capsule Thermal Analysis

Reference: e 1AEA Safety Series No. 6, para 547

This analysis demonstrates that the pressure inside the Model 770 source capsule, when subjected to the Hypothetical Accident Conditions of Transport thermal test, does not exceed the pressure which corresponds to the minimum yield strength at the thermal test temperature.


The source capsules used in the 770 are all special form tested and approved. The thermal test for special form capsules involves heating the capsules at 800'C for at least 10 minutes and allowing the capsules to cool afterwards. Test capsules are tested for leak tightness after this test and must pass intact in order to achieve special form status.

The thermal test from the hypothetical accident conditions of transport requires heating the package to a temperature of 800'C for a period of 30 minutes. Special form capsules are also brought up to the 800'C temperature and allowed to cool prior to integrity testing. The container is a stainless steel construct and from the material properties of stainless steel can withstand this temperature without significant loss of structural integrity. The special form capsules also demonstrate their ability to retain integrity at 800'C. Therefore it is concluded that the container and contents meet the requirements of this section.

3.6.3 Source Assembly Thermal Analysis

Reference: * 1AEA Safety Series No. 6, para 553

This analysis demonstrates that the pressure inside the source capsule used in conjunction with the Model 770, when subjected to the Hypothetical Accident Conditions of Transport thermal test, does not exceed the pressure which corresponds to the minimum yield strength at the thermal test temperature.

The internal volume of the source capsule contains only the radioactive material, spacers and air. It is assumed at the time of loading that the entrapped air is at ambient temperature and pressure of 68°F (20'C) and 14.7 psi (101.3 kN/m2), respectively. This is a conservative assumption because, during the welding process, the internal air is heated, causing some of the air mass to escape before the capsule is sealed. When the welded capsule returns to ambient temperature, the internal pressure would be somewhat reduced.

Under the conditions of paragraph 553 of IAEA Safety Series No. 6, it is assumed that the capsule could reach a temperature of 1,475°F (800'C). Using the ideal gas law and requiring the air to occupy a constant volume, the internal gas pressure could reach 54 psi (370 kN/m2).

The capsule is assumed to be a thin walled cylindrical pressure vessel with the wall thickness equal to the depth of weld penetration.

The maximum longitudinal stress is calculated from:

aLA = PAp where

Stress Area, A = 7r/4x(8.865 2- 8.039 2)x10"6 = 1.1xl0"sm2

Absolute Pressure, P = 370-101.3 = 268.7 kN/m2

Pressure Area, Ap = ir/4x(8.039)2xl0"6 = 5.lxlO5 m2


From this relationship, the maximum longitudinal stress is calculated to be 180 psi (1246 kN/m 2).

The hoop stress is calculated from: 2cht = Pd

where Thickness of the cylinder, t = 0.41 mm - 0.0 16 inches

Inside diameter, d = 8.039 mm Hoop stress, cyl = (268.7x8.039)/(2x0.41) = 1.77 MN/mn2 = 257 psi

At a temperature of 1,598°F (870'C), the yield strength of type 304 stainless steel is 10,000 psi (69 MN/m2). Therefore, under the conditions of paragraph 553 of IAEA Safety Series No. 6, the stress generated is approximately 2.6% of the yield strength of the material.


SECTION 4: CONTAINMENT

4.1 Containment Boundary

4.1.1 Containment Vessel

The containment system for the Model 770 Transport Package is the radioactive source capsule. The source capsule is certified as special form radioactive material under 10 CFR part 71 and IAEA Safety Series No.6.

4.1.2 Containment Penetrations

There are no penetrations of the containment.

•4.1.3 Seals and Welds

The containment is seal-welded in accordance with AEA Technology QSA, Inc. procedures. Leak testing in accordance with ISO 9978 of the source has demonstrated compliance with special form requirements.

4.1.4 Closure

The closure device is the welded special form source capsule. The lock assembly and shipping cap maintain the source in the shielded position as described in Section 1.2.1.

4.2 Requirements for Normal Conditions of Transport

4.2.1 Containment of Radioactive Material

The source capsules used in conjunction with the package have satisfied the requirements for Special Form radioactive material as prescribed in 10 CFR 71 and IAEA Safety Series No. 6. There will be no release of radioactive material under Normal Conditions of Transport.

4.2.2 Pressurisation of the Containment Vessel

Pressurisation of the source capsule under the conditions of the Hypothetical Accident thermal test was demonstrated in Section 3.6.3 to generate stresses well below the yield strength of the capsule material. Therefore, the containment will withstand the less onerous pressure requirements of Normal Transport.

4.2.3 Containment Criterion

See para 2.6.1 -2.6.10 in which it is shown that the containment requirements of 10 CFR 71.51(a)(1) are met.



Page 25 of 37 I

4.3 Containment Requirements for Hypothetical Accident Conditions

4.3.1 Containment of Radioactive Material

The hypothetical accident conditions of 10 CFR 71.73 will result in no loss of package containment. This conclusion follows from the test and analytical information presented in Section 2.7.

4.3.2 Containment Criterion

See Section 2.7 and Section 3.5 which show that the package meets the containment requirements of 10 CFR 71.51 (a)(2).

4.4 Special Requirements

Not applicable.



Page 26 of 37

SECTION 5: SHIELDING EVALUATION

5.1 Design Features

Shielding of the radioactive source in a 770 device is provided by the depleted uranium shield assembly. The mass of the depleted uranium shield is approximately 191 kg (425 lbs).

5.2 Source Specification

5.2.1 Gaminma Source

The gamma sources are described in Table 1-1. The maximum activities are defined in output Curies as required by ANSI N432 (1980) and 10 CFR 34.20.

5.3 Model Specification

Not applicable, as actual radiation profiles were performed to demonstrate compliance.

5.4 Shielding Evaluation

Radiation profiling, with Co-60, of the specimens following Type B testing showed effectively no increase in the radiation levels measured prior to testing and therefore the dose remained within regulatory limits. Results of radiation profiles before and after testing are contained in the TP 114 report. These profiles were performed with two sources loaded in the device. A summary of the dose rates obtained using Co-60 (extrapolated to the maximum capacity of 800 Ci) are shown below. These dose rates were measured after both normal and hypothetical accident condition testing and are therefore worst case representations for normal conditions of transport:

Table 5-1: Co-60 Normal and Hypothetical Accident Dose Rates for 770 Surface One Meter

mR/hr pSv/hr mR/hr [tSv/hr

Top 46 460 1.7 17

Right 42 420 1.8 18

Front 138 1,380 5.2 52

Left 47 470 2.0 20

Rear 149 1,490 5.1 51 Bottom 52 520 2.21 22

Dose rate at one meter from the bottom extrapolated based on other profile readings due to difficulty involved with obtaining a 1 meter reading from the bottom of the 970 lb device.



Page 27 of 37

Based on the profile results using Co-60, the device shielding capacities for the other radionuclides can be assessed. Table 5-1 shows general information for each nuclide as well as a calculated shielding transmission, at capacity, for each nuclide through the depleted uranium shield and titanium source tube.

Table 5-2: Transmission Shielding Assessment for 770

Nuclide 770 Capacity F (R m2/hr Ci) Transmission Transmission Ratio (mR/hr) Relative to Co-60

Co-60 800 1.3 279 1

Ir-192 1,000 0.48 1.1 E-4 3.94 E-7

Cs-137 1,000 0.32 5 E-5 1.79 E-7

Sc-46 800 1.17 24.6 0.088

Transmission exposure rates were determined using Microshield V5 (see Appendix D). Transmission was calculated through the shortest portion of the depleted uranium shield and include attenuation through the titanium source tube -wall. The source was assumed to be a point source located at the center of the shield. (Note that the effective shielding between a source located at the center of the shield and one off-set slightly due to tube positioning in the 770 is the same due to the curved shield shape in this area.) Benefits from shielding/dose reduction of the steel in the 770 and the additional distance to the surface of the device were not taken into account in the assessments. Therefore the source transmission/dose rate values subsequently obtained are conservative.

Shielding material thicknesses and densities used in the Microshield calculations are

shown in Table 5-3.

Table 5-3: Shielding Transmission Parameters

Material Density (gicm3) Shielding Thickness (cm)

Depleted Uranium 18.75 12.1

Titanium 4.5 0.1

I


Dimensions were based on the following configuration (dimensions on picture are in inches - * indicates source position).


'p. IsS O'Ofl strsr�

\\

Figure 5-1: Shielding Assessment Configuration

Based on the Microshield transmission calculations and the physical radiation profiles performed using Co-60, the dose rates for the other three nuclides can be conservatively assessed by taking the Co-60 results and muliplying those values by the respective transmission ratio from Table 52. In the case of Ir-192 and Cs-137 the shield attenuation essentially prevents photon transmission. In these cases the dose reported for Normal and Hypothetical Accident conditions are based on the radiation transmitted by the depleted uranium shield itself. The worst case estimated values are shown in Tables 5-4 thru 5-6.

Table 5-4: Ir-192 Normal and Hypothetical Accident Assessed Dose Rates for 770 Surface* One Meter*

mR/hr pSv/hr mR/hr .iSv/hr

Top 0.4 4



Page 29 of 37

Table 5-5: Cs-137 Normal and Hypothetical Accident Assessed Dose Rates for 770 Surface* One Meter*

mR/hr pSv/hr mR/hr pSv/hr

Top 0.4 4

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 I Burlington, Massachusetts Page 30 of 37 I

Additionally there is a significant amount of steel, as well as distance, between the shielded sources and the external surface of the Model 770 device. Since the measured dose rates were from the unit which had undergone the normal and hypothetical transport testing, values provided for Co-60 are worst case. The assessed dose rates were based on values calculated at the surface and one meter from the surface of the shield. Since the damage incurred by the 770 during the testing did not alter the available depleted uranium shielding, the values provided for the other nuclides are conservative estimates. From these measurements and assessments it is concluded that the 770, when transporting any of the listed nuclides will not produce dose rates on the surface which exceed 200 mR/hr or 10 mR/hr at 1 meter from the surface in normal or hypothetical accident conditions of transport.



Page 31 of 37 I

SECTION 6: CRITICALITY EVALUATION

Not applicable.

SAR Model 770 Transport Package AEA Technology QSA Inc Revision 2 - September 2002 Burlington, Massachusetts Page 32 of 37 I

SECTION 7: OPERATING PROCEDURE

7.1 Procedure for loading the package

1. Prior to transportation, assure the package and its contents meet the following requirements:

a) The contents are authorised for use in the package.

b) The package is in good physical condition for transport.

c) All conditions of the certificate of compliance are met.

d) If more than one nuclide is being transported in the package (e.g., two sources, different nuclides), then ensure that the sum of the fractional source activities to the unit capacity does not exceed one (see example below):

For 300 Ci Co- 60 and 500 Cilr-192 300 Ci 0 Ci I -(0.375 + 0.5) = 0.875 which is less than 1 800Ci 1,00Ci)

2. Ensure that the source(s) are locked into place in the storage positions. The source assemblies must be fully inserted into the source tubes, lock the selector ring in the locked position, the plunger lock depressed and the key removed. Install shipping caps over the ends of the locked source assemblies. Place the cover plate over the lock assemblies and secure in place with sixteen 1/2-13 x 1" drilled head screws (eight for each plate). Attach a tamper indicator security seal between two bolts on each cover plate.

3. If the exposure device is to be shipped inside an outer packaging, mark the outside package "INSIDE PACKAGE COMPLIES WITH PRESCRIBED SPECIFICATIONS USA/9148/B(U)-85 Type B".

4. Perform a contamination wipe of the outside surface of the package and ensure removable contamination does not exceed 0.0001 jiCi when averaged over a wipe area of 300 cmr.

5. Survey all exterior surfaces of the package to assure that the radiation level does not exceed 200 mR/hr at the surface. Measure the radiation level at one meter from all exterior surfaces to assure that the radiation level is less than 10 mR/hr.


NOTE: If the device is to be shipped without an overpack, the radiation survey should be made of the outer surfaces of the device. If the device is to be shipped inside an overpack, the radiation survey should be from the outer surfaces of the overpack with the device packaged for shipment inside the overpack. Package labelling for transport is based on the radiation surveys from the outer surfaces of the overpack with the device packaged for shipment inside the overpack.

6. Ship the container according to the procedure for transporting radioactive material as established in 49 CFR 171-178.

NOTE: The US Department of Transportation, in 49 CFR 173.22(c), requires each shipper of Type B quantities of radioactive material to provide prior notification to the consignee of the dates of shipment and expected arrival.

7.2 Procedure for unloading the package

1. The consignee of a package of radioactive material must make arrangements to receive the package upon delivery. If the package is picked up at the carrier's terminal, 10 CFR Part 20.1906 requires that this be done expeditiously upon notification of its arrival.

2. Inspect the external unit for signs of damage. If damage is evident, the carrier's agent should be present while unpacking. Survey the exposure device with a survey meter as soon as possible, preferably at the time of pickup and no more than three hours later if it was received during working hours, or no more than 18 hours later if it was received after normal working hours.

3. Radiation levels should not exceed 200 rnR/hr at the surface of the exposure device nor 10 rnR/hr at a distance of 1 meter (40 inches) from the surface.

4. Actual radiation levels should be recorded on the receiving report. If the radiation levels exceed these limits, the container should be secured in a Restricted Area, and the appropriate personnel notified.

5. Visually inspect the Model 770 for signs of damage and assure that the seal wire has not been tampered with.

6. The radioisotope, activity, model number and serial number of the source and the package model number and serial number should be recorded in the receiving report.

7. Unloading of the Model 770 must be in accordance with the instructions supplied with the package per 10 CFR 71.89.


7.3 Preparation of an empty package for transport

1. In the following instructions, an empty package refers to a Model 770 without an active source.

2. To confirm the package is empty, insert the physical verification probe into both of the source tubes. The probe should be able to be inserted fully up to the marked line on the probe. If full insertion of the probe in either source tube is not possible, contact AEA Technology QSA, Inc. as there may be a cropped source or other item inserted at the bottom of the source tube. Do not ship the package as empty until confirmation can be made that a cropped source is not present in the source tube(s).

3. To ship an empty package perform a radioactive contamination wipe test of the outer shipping package. This consists of rubbing filter paper or absorbent material, using heavy finger pressure, over an area of 300 cm2 (46.5 in 2) of the package surface. The activity on the filter paper should not exceed 0.00001 uCi/cm2 of removable contamination.

4. After the survey prepare the package depending upon the radiation levels obtained as prescribed in 49 CFR 173.


SECTION 8: ACCEPTANCE TESTS AND MAINTENANCE

PROGRAM

8.1 Acceptance test

8.1.1 Visual Inspection

Visually inspect each Model 770 exposure unit to be shipped to assure the following:

" The exposure unit was constructed properly in accordance with Drawing Number R77090.

" All fasteners (screws) as required by the above referenced drawings are installed and secured.

" The relevant labels are attached and contain the required information and are marked in accordance with:

10 CFR 20 10 CFR 40.13(c)(6)(i) 10 CFR 34 10 CFR 71

8.1.2 Structural and Pressure Tests

The swage coupling between the source capsule and the cable of the source assembly is subjected to a static tensile test with a load of 445 N (100 lbs). All welds are inspected per drawing requirements.

8.1.3 Leak Testing

The source assembly used in the Model 770 is leak tested prior to use following manufacture. It is also leak tested for removable radioactive contamination at intervals not to exceed six months.

8.1.4 Component Tests

The lock assembly of the package is tested to assure that the security of the source will be maintained. Failure of this test will prevent use of the package until the cause of the failure is corrected and the package is re-tested.

8.1.5 Tests for Shielding Integrity

With the package containing an approved source capsule, the radiation levels at the surface of the package and at 1 meter from the surface of the package must not exceed 200 mR/hr and 10 mR/hr respectively when extrapolated to the rated capacity for the package.


8.1.6 Thermal Acceptance Tests

Not applicable.

8.2 Maintenance program

8.2.1 Leak Tests

As described in Section 8.1.3, the radioactive source assembly is leak tested at manufacture and additionally must be leak tested every six months with the same acceptance criteria.

8.2.2 Subsystem Maintenance

The lock assembly is tested prior to each use of the package. Additionally the package is inspected for tightness of fasteners, proper seal wires, integrity of welds and general condition before each use.

8.2.3 Shielding

Prior to each use, a radiation survey of the package is made to assure that the radiation levels do not exceed 200 mR/hr at the surface nor 10 mR/hr at 1 m from the surface.

8.2.4 Miscellaneous

Inspections and tests designed for secondary users of this package are included in Section 8.2.5.

8.2.5 Inspections ant Tests for Secondary Users

8.2.5.1 Labels

1. Ensure labels are securely fastened to the container. If not securely fastened, reattach or replace with hardware specified by AEA Technology QSA, Inc.

2. Ensure that all labels required for the device are attached to the container.

3. Ensure labels are clean and easily legible. If labels are illegible or cannot be cleaned, replace the label. If any holes remain uncovered after replacement of the label, fill these holes with USM SSD42SSBS designation pop-rivets or Scott drives.

8.2.5.2 Device Housing

Examine the housing for signs of wear, damage or omission of any required safety wires.


1. Ensure that the housing integrity is secure and has not been breached by puncture, impacts, cracks, rust or welds that are no longer intact. Punctures, cracks, broken or incompletely fused welds and flaking or extensive pitted rust are criteria for rejection. These items indicate the need for repair or removal from use based on the conditions severity.

2. Ensure all safety wires are intact as required by the descriptive drawings for the Model 770. If a safety wire is missing, shows signs of damage or breaks or if the gauge wire used is of lower-grade, then contact AEA Technology QSA, Inc for guidance.

8.2.5.3 Lock Assembly

Ensure that the lock assembly is securely attached to the device housing. Operate the selector ring and assure that it rotates freely. Ensure that the lock plunger operates smoothly and that the shipping cap is firmly attached to the device and installs smoothly. If the device fails any of these checks, contact AEA Technology QSA, Inc for guidance.

8.2.5.4 Threaded Holes

Ensure that the threaded holes used to secure the shipping plates will accommodate and engage the required size bolt (or by using the appropriate thread gauge). Ensure the shipping plates fit snugly within the bosses and the bolts do not rub against the ID of the clearance holes during installation. If the bolts fail to clear properly, contact AEA Technology QSA, Inc for guidance

8.2.5.5 Source Tube

Ensure that the source tubes are clear from obstruction, snagging, jamming or hang-up by inserting the verification probe into the source tube. If any obstructions/snagging are encountered in any source tube, contact AEA Technology QSA, Inc.

SAR Model 770 Transport Package Revision 2 - September 2002AEA Technology QSA Inc

Burlington, Massachusetts

APPENDIX C

TESTPLAN 114 REPORT

(Less Appendices)

TEST PLAN 114 Report Rev I

MODEL 770 TYPE B

TRANSPORT PACKAGE

AEA Technology QSA Inc. 40 North Avenue

Burlington MA 01803

SEPTEMBER 2002

AEA TECHNOLOGY QSA INC. Test Plan 114 Report Rev I Burlington, Massachusetts SEP 02

Page 1 of 20# 9

CONTENTS

SECTION 1. PURPOSE ............................................................................................. 2

SECTION 2. SCOPE OF TESTING ........................................................................... 3

Section 2.1 Normal Conditions oftTransport ................................................................................................... 3

Section 2.2 Hypothetical Accident Conditions ................................................................................................ 3

SECTION 3. TEST UNIT DESCRIPTIONS ................................................................ 5

Section 3.1 Test Unit I - Serial Number 10 .......................................................................................................... 5

Section 3.2 Test Unit 2-Serial Number 11 .......................................................................................................... 5

Section 3.3 Projector version ................................................................................................................................. 6

SECTION 4. CHANGES TO TEST CONDITIONS OR ORIENTATIONS .................... 7

Section 4.1 Normal and Accident Conditions of Transport .......................................................................... 7

Section 4.2 Hypothetical Accident Conditions (71.51(a)) ............................................................................... 7

SECTION 5. TEST SPECIMEN RESULTS ................................................................ 8

Section 5.1 Test One - Sn #10 ................................................................................................................................ 8

Section 5.2 Test Two Sn#11 ................................................................................................................................ 12

,SECTION 6. SUMMARY AND CONCLUSIONS ..................................................... 14

SECTION 7. APPENDIX A - DRAW INGS .............................................................. 15

SECTION 8. APPENDIX B - CALCULATIONS ........................................................ 16

SECTION 9. APPENDIX C - TEST DATA SHEETS ................................................ 17

SECTION 10. APPENDIX D - MANUFACTURING RECORDS ............................... 18

SECTION 11. APPENDIX E - PROFILE SHEETS ................................................... 19

SECTION 12. APPENDIX F - ORIGINAL TEST PLAN ............................................. 20

" \$SS WORK ST If F\Eng inccii g\-ype B\1` 114 770 SSIFTP 114 Report ReN I.doc

AEA TECHNOLOGY QSA INC. Burlington, Massachusetts

Test Plan 114 Report Rev 1 SEP 02 Page 2 of 20# 9

Section 1. Purpose

The purpose of this test was to assess the ability of the modified Model 770 to withstand the most onerous portion of Hypothetical Accident Conditions, the drop sequence (9m (30-foot) and Im (-3-foot) puncture). The resultant damage from the drop was then used in conjunction with the Model 770's design, as the basis for assessment of the device's response to the thermal portion of the test sequence.

All other test comprising Normal Transport and Hypothetical Accident Conditions will be assessed based on either previous testing, or engineering analysis.

Additionally, an assessment is made that the Model 981 projector is sufficiently close in design to the Model 770 Transport Container that testing of the 770 will bound the 981 for transport compliance.

C.\S$S WORK STtl"A-ngi neeing\Type BVI P 114 770 SSI*FTIP1 14 Report Re\ I doc

AEA TECHNOLOGY QSA INC. Test Plan 114 Report Rev I Burlington, Massachusetts SEP 02

Page 3 of 20# 9

Section 2. Scope of Testing

Section 2.1 Normal Conditions of Transport

The tests for Normal Conditions of Transport described in 10 CFR 71.71 are the compression test, penetration test and

1.2m (4-foot) free drop test.

Compression Test The structure of the container is capable of supporting the required weight. If we use the Chicago short column

formula* and just take the four side plates, the buckling weight would be well over 300 000 lbs, which is 60 times

more than the required 5000lbs (rounding the weight of the container to 1000 lbs.). Not taken into account here are

some plate buckling concerns and that short, thin-walled columns are very difficult to examine mathematically.

However, even if the error were an order of magnitude, the factor of safety is still acceptable. This calculation also

does not take into account the remaining structure, which is immense compared to the just the side plates. As such, it

is deemed as to be acceptable. (See Appendix B Calculations) (*- Mark's Standard Handbook, 9h Ed., p. 5-43)

Penetration Test The effect of a 13-lbs. (5.9 kg) bar dropped from 1.2m feet onto an object constructed of ¼/ inch plate steel is

negligible. The force of the bar is also not sufficient to damage one of the bolts securing the shipping plates. Even if a

bolt were removed, the worst scenario from an impact, the remaining seven (7) would hold the plate securely.

1.2m (4-foot) Free Drop Test This test was previously performed on the original version (carbon steel) of the Model 770 (under Test Plan 88) with

the net result being an approximately 1 inch triangle of deformation on the comer of the unit. This damage did not

increase the dose profile of the package more than 20% (It actually had no measurable effect). The current version of

the Model 770 weighs approximately 12.5% more than the previous version. As such, the deformation to the current

Model 770 would be proportionally greater, or approximately a 1 1/8-inch deformation (Using a straight linear approximation). This amount of damage will in no way affect the radiological safety of the package, as at that depth

only the outer shell is affected, nor increase the does profile above the above stated values.

The Water Spray Preconditioning Not necessary as the 770 Transport Package is constructed of waterproof materials throughout. The water spray would

not contribute to any degradation in structural integrity.

# Section 2.2 Hypothetical Accident Conditions

The Hypothetical Accident Tests described in 10 CFR 71.73 are the 9m (30-foot) drop, the Im (-3-foot) puncture

drop and thermal tests.

The crush test described in 10 CFR 71.73(c)(2) will not be performed because the contents are qualified as Special

Form radioactive material.

9m (30-foot) Drop Test Two tests were performed. They are described in the following sections.

I m (-3-foot) Puncture Trop Tests Two tests were

safety analysis report aea technology qsa, inc. model 770 type … · 2012. 11. 18. · aea...

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