*Corresponding author. Tel.: #34 1 346 6232; fax: #34 1346 6576; e-mail: espartero@ciemat.es.

Nuclear Instruments and Methods in Physics Research A 422 (1999) 790—794

Development and application of a radioactivity characterizationsystem for low-level radioactive waste

A.G. Espartero*, G. Pin8 a, J.A. Suarez

Ana& lisis y Caracterizacio& n de Residuos Radiactivos, Departamento de Fisio& n Nuclear, CIEMAT, Avda. Complutense, 22. 28040 Madrid, Spain

Abstract

Low-level technological radioactive wastes, in Spain, are commonly produced in research and medical centers. Thesewastes must be characterized before conditioning in order to determine their radioactive content for inventory purposes.A prototype has been designed for beta—gamma radiological characterization of standardized 25 l bags containingheterogeneous low-density technological radioactive wastes within the density range 0.05—0.6 g/cm3. The system consistsof an iron shielding box with three NaI(Tl) and a silicon implanted detectors for gamma and gross beta activitydeterminations, respectively. The study of the measurement method, carried out with rotating scanning, included theoptimization of the detection solid angle to minimize the uncertainties and the influence of the relative position ofthe radioactive material. Several materials and densities, in the range aforementioned, were considered to obtain theexperimental attenuation factors, used for fitting a correction algorithm in function of density and c-emission energy. Thesensitivity of this method, calculated for the most frequent average density of this kind of waste (0.1 g/cm3), is lower than50 Bq/kg for the main b—c emitters (137Cs and 60Co) and lower than 480 Bq/kg for gross beta activity. ( 1999 ElsevierScience B.V. All rights reserved.

Keywords: Low-level radioactive wastes; HPGe detectors; Silicon implanted detectors; Raw waste

1. Introduction

The radioactive solid waste management of therejected material produced in medical and researchcenters in Spain means a little percentage (2%) inthe forecast of the total volume of the low andmedium level radioactive waste to manage in thenear future by current national radioactive wasteplan [1]. This figure does not take into account the

solid heterogeneous radioactive waste generatedinto the nuclear cycle facilities with the same char-acteristics than those generated in the research andmedical centers (called maintenance waste) [2],nevertheless the geographic scattering of the facili-ties, the significant volume of waste produced(416 m3 in 1993) and the diversity of the wasteactivity, nuclides and matrix nature [3] leads toa very difficult management of this type of radio-active waste.

The main characteristics of this type of solidwaste are the heterogeneity of the material nature

0168-9002/99/$ — see front matter ( 1999 Elsevier Science B.V. All rights reserved.PII: S 0 1 6 8 - 9 0 0 2 ( 9 8 ) 0 0 9 9 8 - X

and the random distribution of the radioactivity inthe waste matrix. For this reason, the difficulties tofind out a sampling procedure to obtain a represen-tative samples of the solid raw waste [4], requirea specific characterization method to determine theradioactive content in 100% inspection of the pro-duced waste container units (standardized plasticbags) by a non destructive method, employing anexperimental procedure to obtain a correction al-gorithm in function of easily to measure physicalparameters of the waste unit such as average den-sity instead of using the percentage composition ofthe waste matrix [5].

The objective of the methodology developedis to minimize the uncertainty and the minimumdetectable activity for gamma and gross beta deter-minations of standardized solid waste containers(25 l plastic bags) and its possible application tocharacterize solid dismantling waste [6] and forexemption purposes in waste recycling or free re-lease [7].

2. Measurement system description

2.1. Shielding cell

The expected activity level of this type of radio-active waste is very low, so it is necessary to isolatethe measurement set, that means detector andsample, from the environmental activity to opti-mize the detection limits. The measurement set isplaced into a shielding cell whose dimensions are500]500]500 mm. This cell was made in low60Co content iron, whose walls have 150 mm thick-ness to obtain a mass energy absorption coefficientof 118 g/cm2, equivalent to 104 mm thickness leadwall.

The system consists of a maximum of three 3]3inches INa(Tl) or HPGe detectors arranged in onewall of the cell and a Si implanted detector in theopposite wall. The detectors are separated by ironpieces whose size is variable in order to changetheir relative heights, and the number of detectorsemployed in the measurements (Fig. 1).

A turn-table is placed at the base of centralplacement to load the samples giving a maximumrotating speed of 18 r.p.m. and, in its base, the

turn-table has a weighting electronic cell to knowthe sample weight with 10 g accuracy and a 100 kgof maximum load to calculate the average densityof the materials inside the plastic bags.

2.2. Detection system

Gamma activity determination is performed byINa(Tl) detectors (3]3 in. crystal). The selection ofscintillation detectors was leading by the higherefficiency given by this kind of crystals opposite tosemiconductor detectors, in order to reduce themeasurement time and the detection limits andminimize the costs for the possible industrial ap-plication of this prototype.

A silicon implanted detector (900 mm2), witha nominal resolution of 25 keV, was used to deter-mine the gross beta activity.

Both INa(Tl) and Si implanted detectors are,respectively, connected to four MCA boards in-cluding amplifier, ADC and HV power supply,inside a PC.

2.3. Measurement procedure

Every measurement was performed using a chan-nel by channel sum of the individual gamma spec-trum, acquired with continuously rotation of thesample at 14 r.p.m.

3. Results and discussion

3.1. Detectors arrangement

The relative positions of the INa(Tl) detectorscan be changed in height in order to select the bestconfiguration which optimizes the detection solidangle. Nine configurations of 3 detectors and sixconfigurations of 2 detectors were evaluated. Thesedetectors were arranged within a height intervalbetween 35 and 5 cm measured from the base of theturn-table.

Average efficiency curves were calculated usinga 152Eu reference sample, placed at four differentdetector—sample distances (10, 15, 20 and 25 cm),and checked with known activity samples of 60Co

791A.G. Espartero et al. /Nucl. Instr. and Meth. in Phys. Res. A 422 (1999) 790—794

VII. APPLICATIONS

Fig. 1. General arrangement of developed system.

and 137Cs measured at the same radial distancesfrom the detectors.

In these conditions the relative errors were lessthan 18% for all configurations of 3 detectors andless than 23% for all configurations of 2 detectorsconsidered. These errors are assumed as acceptablefor this kind of non-destructive measurementmethodology.

The selection of the minimum uncertainty de-tector arrangement was carried out using a statis-tical function (F

S), which was defined considering

the experimental relative errors of the 60Co and137Cs determinations, and the resultant standarddeviation of the average efficiency calculation:

FS"+ Er#+ p

%&&.

The selection factor (FS) indicates that the lowest

uncertainty corresponds to the detectors layoutwhose heights are 35, 20 and 5 cm, since the over-lapping of the individual solid angles minimizes theinfluence of the radial position of the radioactivematerial into the plastic bag boundaries.

3.2. Influence of the sample size

To evaluate the influence of the sample size in theaccuracy of the activity determinations, referencesamples of 60Co, 137Cs and 152Eu with differentsizes were measured, varying their positionsin height from the base of the turn-table andthe detector-sample distances at 10, 15, 20 and25 cm.

Relative errors are lower than 30% for the de-tectors arrangement whose heights are 35, 20 and5 cm, respectively, since with this detector layoutthere is a maximum measurement volume becauseof the optimum intersection of the solid angles. Forthis reason, the mentioned configuration will be theonly one considered in the subsequent studies.

3.3. Average density influence

Data of the material contents and weights of 211plastic bags from different producers were used inorder to know the most frequent materials and

A.G. Espartero et al. /Nucl. Instr. and Meth. in Phys. Res. A 422 (1999) 790—794792

Fig. 2. Experimental efficiency attenuation factors and fitting curves for several matrices within the range of density 0.063—0.54 g cm~3.

associated average densities present in this type ofwastes. The obtained results indicate that the mostfrequent average density is 0.12 g cm~3 and themain materials present in these bags and their aver-age densities are such as plastic overshoes(0.063 g cm~3), polyethylene vials (0.10 g cm~3),glass vials (0.22 g cm~3), latex gloves (0.25 g cm ~3)and steel plates (0.54 g cm~3).

Average efficiency calibration curves were ob-tained using a reference sample of 152Eu, placed at10, 15, 20 and 25 cm from the detectors, ina simulated container full up of each materialsaforementioned.

Attenuation factors were calculated as a ratiobetween the efficiency corresponding to a certainaverage density and the efficiency obtained withoutany material into the simulated container. Theseattenuation factors, for each average density con-sidered, were fitted by a linear correlation analysisas shown in Fig. 2.

The resultant coefficients for each density werefitted as well by linear correlation analysis, obtain-ing an algorithm to calculate the loss of efficiencyfor any energy and any density within the range

studied, from the reference efficiency (Eff0):

Log EffX"Log Eff

0![(A

0#A

1o (g/cm3))

#(B0#B

1o (g/cm3)) E (keV)].

A simulated waste bag, containing a matrix ofpolyethylene vials and latex gloves with an averagedensity of 0.12 g cm~3, with a random distributedreference filter papers of 60Co and 137Cs, was mea-sured to evaluate the accuracy of the results ob-tained applying this algorithm.

The relative errors obtained in the activity deter-minations and the minimum detectable activity(MDA) calculated with 15 min of counting time areshown in Table 1. These relative errors are inagreement with those obtained in the study of theinfluence of the sample size. Both, the relative er-rors and the MDA values, are acceptable for non-destructive analysis.

3.4. Gross beta activity determination

A silicon implanted detector was placed in a ver-tical wall of the shielding cell at 15 cm from the baseof the turn-table to detect high energy b particles and

793A.G. Espartero et al. /Nucl. Instr. and Meth. in Phys. Res. A 422 (1999) 790—794

VII. APPLICATIONS

Table 1Relative errors and minimum detectable activity (¹

C"900 s)

obtained in the characterization of a 0.12 g cm~3 simulatedwaste bag, using the experimental algorithm to correct theaverage density influence

Nuclide Error % MDA (Bq/kg)

137Cs 13 47.860Co 3.3 37.1

Table 2Average beta efficiency and minimum detectable activity(¹

C"900 s) with different materials and densities

Average density(g cm~3)

Material Averageefficiency

MDA(Bq/kg)

0.063 Plastic overshoes 1.04]10~4 2.39]102

0.10 Polyethylene vials 3.24]10~5 4.80]102

0.22 Glass vials 1.54]10~5 4.85]102

determine the gross beta activity content in thesewastes. The study on the average density influencewas carried out with a reference sample of 90Sr/90Yplaced at four different detector-sample distances(10, 15, 20 and 25 cm), inside a simulated container.

Data of average beta efficiency and minimum de-tectable activity were obtained full up the simulatedcontainer of different materials such as plastic over-shoes (0.063 g cm~3), polyethylene (0.10 g cm~3) andglass (0.22 g cm~3) vials and are shown in Table 2.

In spite of the low efficiency obtained, due to thebeta particle self-absorption in the matrix and thedetector—sample distance, the MDA values are inan acceptable range for this methodology.

4. Conclusions

In order to minimize the measurement uncer-tainty, the detectors arrangement should provide

a maximum measurement volume to minimize theinfluence of the sample radial position. The lowestuncertainties are gotten with the detectors layout35, 20 and 5 cm in height.

The obtained results using the experimental effi-ciency correction algorithm indicate that it is notnecessary to have standardized bags of referencewith different densities and materials to calibratethe system. The experimental algorithm allows tocorrect the efficiency curve, obtained without anymatrix, in function of the average density of thesample for any emission energy into the calibrationrange.

The beta efficiency and the minimum detectableactivity obtained with a Si implanted detector, inthe mentioned experimental conditions with thenon destructive methodology developed, allows thegross beta activity determination with appropriatesensitivity.

Acknowledgements

The authors wish to thank ENRESA for thefinancial support of this work under the Associ-ation Contract CIEMAT-ENRESA.

References

[1] Cuarto Plan General de Residuos Radiactivos. Ministeriode Industria y Energıa, 1994.

[2] J.P. Ghysels, Proc. 5th Radioactive Waste Managementand Environmental Remediation, vol. 2, ASME, New York,1995, p. 889.

[3] M.T. Ortız, in: Senda (Ed.), Generacion y gestion de re-siduos de baja actividad, Ch. 1, SNE, 1994, p. 66.

[4] J.L. Pettier, Nucl. Technol. 115 (1996) 178.[5] B. Chabalier, Nucl. Technol. 115 (1996) 162.[6] P. Carboneras, M.T. Ortız, Estratos 35 (1995) 30.[7] IAEA Safety Guide No. 111-P.1.1.1, 1992.

A.G. Espartero et al. /Nucl. Instr. and Meth. in Phys. Res. A 422 (1999) 790—794794