IN9800964
IGC-151 T- a> 1993
Kon-Destructive Assay of Glove Box Solid Waste By Gamma Spectrometry
N.P. Seshadreesan, K. Swaminathan, R. Kumar and C.R. Venkata Subramani
GOVERNMENT. OF INDIA. DEPARTMENT OF ATOMC ENERGY INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH KALPAKKAM IGC - 151 1993
GOVERNMENT OF INDIA DEPARTMENT OF ATOMIC ENERGY
NON-DESTRUCTIVE ASSAY OF GLOVE BOX SOLID WASTE BY GAMMA SPECTROMETRY
N.RSESHADREESAN, K.SWAMINATHAN, R.KUMAR AND C.R.VENKATA SUBRAMANI
INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH KALPAKKAM 603102 TAMIL NADU INDIA ABSTRACT
This report presents the details of the system set up at Radiochemistry Laboratory for the assay of glove box solid waste for special nuclear materials. Results of trial runs carried out to establish the minimum detection limits for the assay of Plutonium and U-233 have been presented. Correction for matrix effects have been described. Design details of an improved system have been oresented. NON-DESTRUCTIVE ASSAY OF GLOVE BOX SOLID WASTE BY GAMMA SPECTROMETRY
* * * N.P.Seshadreesan , K.Swaminathan , R.Kumar and C.R.Venkata Subramani* l^ INTRODUCTION
Waste containing special nuclear materials is generated in the nuclear industry in a number of different processes and in different chemical forms. The term special nuclear materials refers to materials like plutonium and uranium ( both natural and enriched ) which are of interest in the nuclear industry and for which material accounting is essential.
The solid waste generated in the glove boxes of fuel reprocessing and fuel preparation laboratories may carry substantial quantities of special nuclear materials (SNM). The amount of SNM present may vary considerably depending on the nature of operations in the glove box, and in the boxes where powders are handled or solutions of high concentration are used, the amount of SNM in waste could be substantial. Hence, from the point of material balance, nuclear material accounting and also waste management, it is desirable to monitor the waste before disposal.
* Fuel Chemistry Division, Indira Gandhi Centre for Atomic research, Kalpakkam 603 102, Tamilnadu. The nature of operations being carried out in the box influence the assay method used. In the waste where the species of interest are present along with very high concentrations of fission products, assay can be carried out only by active neutron interrogation or by passive neutron assay. A typical example of this mixture is the hull waste produced at the front end of reprocessing schemes. However, where the SNM is substantially free from other gamma emitting species - a condition prevalent in most laboratory glove box waste - assay of the solid waste can be carried out by gamma spectrometry.
The present report discusses the assay of solid waste from glove boxes by high resolution gamma spectrometry. A detailed review of such systems is given in Ref 1. This report gives the performance characteristics of the system set up at Radiochemistry Laboratory. These wastes generated at RCL mainly plutonium bearing wastes - will not contain a very high level of beta-gamma activity and will be heterogeneous in nature requiring attenuation correction using transmission techniques. A similar study for U-233 bearing wastes was carried out at the request of the Reprocessing Programme at IGCAR and the results of this study are also presented. As assay of wastes containing enriched uranium is also of interest, a theoretical computation of the minimum detection limit for this purpose was carried out and these results are also presented. This report also discusses the effect of matrix gamma activity on the minimum detection limit. Design details of the system planned to be used on a routine basis for this work are also presented. Some preliminary results have been presented earlier.[2]
II. PRINCIPLE
The various isotopes of plutonium emit a host of gamma radiation whose intensities vary from a high of 10 % to a low of about 10~ %. The gamma signatures used for the assay of plutonium depend on the isotopic composition of the plutonium being assayed. The isotopic composition of the plutonium used in our laboratory is shown in Table 1. As can be seen from this table/ the major species present is Pu-239. The energy and intensity data on the gamma rays of this isotope are presented in Table 2 wherein the gamma rays used for this work are highlighted. In a similar fashion Table 3 gives the data on gamma rays of U-233 with the lines used being highlighted.[3]
The solid waste from the glove box are usually removed in a plastic container 6" in diameter and about 10" in height. The samples are expected to be nondescript in nature and may consist of tissue paper/ absorbant sheets/ neoprene gauntlets/ surgical gloves/ pipetting tips/ glass and metal scraps/ etc. Hence/ it is necessary to compensate for the non-homogeneous nature of the contents. The sample will be rotated during assay to minimize the effect of non-homogenity. To account for the attenuation inside the sample/ attenuation correction methods will be used. In the correction method used/ the attenuation coefficient of the sample is determined by placing the sample between an external source of gamma radiation and the detector and by measuring the count rate with and without the sample in position. The average attenuation
coefficient of the sample is obtained using the equation
1 = 1^ exp ( -ut) where
t is the diameter of the sample container
1^ is the count rate without the sample container/ and I is the count rate with the sample in position.
The average attenuation coefficient of the sample thus
calculated is used to correct the count rates obtained.
III. INSTRUMENTATION
The assay of gamma activity was carried out using a high purity coaxial germanium detector ( relative efficiency 30%/ resolution 1.8 keV ). The output of the detector was fed to a ORTEC model 572 spectroscopy amplifier and the amplifier output connected to a PC based multichannel analyser using a Tennelec/Nucleus 4K PCA card.
IV. EXPERIMENTAL
Working standards - GBWA standards - were prepared by adding a small aliquot of a plutonium standard solution to polythene bags containing tissue swipes and absorbent paper. The standards
were kept in the box for a day or two to allow the solution to evaporate. They were then removed and doubly sealed individually. Standards containing 4.2 mg , 10.52 mg / 21.04 mg and 31.56 mg of plutonium were prepared in this fashion. Standards of U-233 containing 7.39mg, 14.78mg/ 29.56mg and 44.34mg were prepared in a similar fashion.
A schematic of the system used is shown in Fig 1. The GBWA standards were kept in the sample container packed with tissue swipes. No collimation was used and the entire plastic container was seen by the detector. The container was placed on a motorised stand and rotated to minimize the effect of non-homogenity. The standards and various combination of these standards were counted. The experiment was repeated using different matrices like neoprene gauntlets, steel scrap and sodium. Matrix correction runs were carried out using a point source of Eu-152 alongwith point sources of Cs-137 and Na-22. The Eu-152 source was chosen because of its multiple gamma emission - enabling it to be used for attenuation correction for the different gammas of interest in one step. The effect of matrix gamma activity was determined by placing sources of Cs-137 and Co-60 at different distances from the detector.
V_;_ CALCULATIONS
The minimum detection limit for any gamma ray is a function of the background at the energy of interest. In the present work the minimum detection limit has been calculated in the following way. The average background at the gamma ray energy of interest is obtained from the spectrum. Assuming a maximum variation of 2 f i a fresh background is generated using pseudo random numbers. On this background a gaussian peak is superimposed. Tha area under the peak above background gives the peak area that would be obtained. Different values of peak area are used in increasing order till the peak is visible above the background. The value of the peak area when the peak is first visible above background is taken as the minimum detectable area and the corresponding concentration is defined as the minimum detection limit.
VI. RESULTS AND DISCUSSION
The results obtained for the assay of plutonium are presented in Tables 4-7 for different matrices such as tissue swipes/ neoprene gauntlets/ steel scrap and sodium. As can be 2 seen from the values of R obtained/ there exists a high degree of linearity between the peak area obtained and the concentration of plutonium. In a similar fashion/ Tables 8-10 present the results obtained for the assay of U-233 in tissue swipes/ neoprene gauntlets and steel scrap matrices. As can be seen from the results of regression analysis/ these also show a very good correlation between peak area and concentration.
In the energy range of interest the variation of the attenuation coefficient as a function of energy is expected to be linear on a log-log scale. In the present work/ attenuation correction was carried out by using a source of Eu-152 alongwith point sources of Cs-137 and Na-22. The Eu-152 source was chosen for its multiple gamma lines as can be seen from Table 11. The attenuation factor for the various lines measured using a matrix of neoprene gauntlets is shown in Table 12 and Fig 2 plots the attenuation factor as a function of energy in log-log scale.
Using the attenuation factors obtained experimentally for the various matrices/ the corrected peak area was obtained for the three gamma lines of plutonium and for one gamma line of U- 233. The results of the attenuation correction studies are presented in Table 13. The corrected peak area - the last column in the table - is expected to be equal within normal statistical error limits. As has been mentioned in the table/ the geometry used for the tissue and neoprene matrices was different from the geometry used for the sodium and steel scrap matrix. However/ the agreement within samples in different matrices in the same geometry is quite good.
The results of the calculations on the minimum detection limits are presented in Tables 14 and 15. Table 14 presents the background observed and the minimum peak area required for assay of plutonium as a function of counting time. As would be expected the ratio of peak area to background decreases as counting time increases. Table 15 presents the results of the computation of minimum detection limits for plutonium in tissue matrix as a function of counting time. The minimum detection limit decreases with an increase in counting time. A counting time above 5000s per sample is not practical and hence these calculations have assumed a maximum counting time of 5000s.
The results of the influence of external gamma activity on the minimum detection limit are presented in Table 16. As the external source strength increases/ the peak area above background progressively decreases leading to an increased minimum detection limit; the minimum detection limit increases from 1.4 mg to 13.8 mg of plutonium as shown.
Assuming a counting period of 5000s / the minimum detection limits for assay of both plutonium and U-233 are presented in Table 17. From this table it can be seen that the minimum detection limit for the assay of plutonium is around 0.5 mg of plutonium in tissue matrix for a counting time of 5000s and around 2 mg of U-233 under identical conditions.
One of the major problems in applying attenuation corrections in the present experimental set up, is the non- uniformity of the packing. As the entire container is being seen by the detector and since the species of interest could be spread over the container, use of point sources to determine the attenuation correction is obviously incorrect. Initial trial runs showed a marked variation in attenuation coefficient as a function of energy due to the packing density. Only those trials where the packing density was fairly high gave satisfactory results. The only reasonable technique to overcome this problem is to collimate the area being seen by the detector. This implies that segmented gamma scanning is necessary. Design details of such a system are presented later.
a However, even the present system with all its limitations can be successfully used for the assay of waste from glove boxes as long as the packing density is uniform or at least fairly uniform.
VII. ESTIMATION OF MINIMUM DETECTION LIMIT FOR U-235 ASSAY
The assay of wastes bearing enriched uranium is of interest and hence an estimate of the minimum detection limit obtainable with this system for the assay of such wastes was carried out. The most intense gamma ray of U-235 has an energy of 185 keV and the 188 keV gamma line from U-233 was taken as the reference gamma for efficiency calculations. Assuming a counting time of 5000s, the minimum detection limit was calculated as 6.2 mg of U- 235 which corresponds to about 246 mg of Uranium enriched to 2.5%. The walls of the building contain Thorium ore and this is reflected in the background spectrum. Thorium ore has a peak at 185 keV due to Ra-226 and Th-230 and this peak is seen in our background spectrum when the counting time is around 40000s This peak is not visible at low counting times and hence the above calculations for a counting period of 5000s assumed a straight line background. However, for long counting times, the existence of this peak should be kept in mind and appropriate corrections made. VIII._ DESIGN OF A NEW SYSTEM
The system used for the above work was made by modifying the system used for the analysis of primary sodium samples from FBTR. Design of a system exclusive for this work has been recently completed and is described below.
The proposed system will incorporate segmented gamma scanning using a lead collimator instead of scanning the entire vessel as at present. This will enable us to locate the position of the activity inside the container. The container will be rotated around its vertical axis (as at present) for uniform attenuation due to the matrix. The vertical movement of the sample in front of the collimator will be carried out using a stepper motor. The system will be a PC-controlled unit and the same PC will be used to acquire/ analyse and quantify the spectra at different vertical locations. This will ensure minimum of manual operations and once a container has been placed in position/ all control passes to the PC.
IX. CONCLUSIONS
The present system has a minimum detection limit of 0.5 mg of plutonium and 2 mg of U-233 in a tissue paper matrix for a counting period of 5000s assuming normal background gamma activity with little or no contribution from the sample. Theoretical computations indicate a minimum detection limit of 6.2 mg for U-235 corresponding to 246 mg of 2.5 % enriched
10 Uranium under above conditions. The matrix weight being 500 g, this works out to a minimum detection limit of 62 nCi/g for
Plutonium and 38 nCi/g of U-233 both of which are well under the limit of 100 nCi/g for storage of alpha bearing waste above ground.
X_;_ ACKNOWLEDGEMENTS The authors thank Shri T.G.Srinivasan for assistance in preparation of the glove box waste assay standards.
XI. REFERENCES 1. Reilly, T.D. and Parker, J.L. A guide to gamma-ray assay for Nuclear Material Accountability. LA-5794-M, 1975. 2. N.P.Seshadreesan, K.Swaminathan, R.Kumar and C.R.Venkata Subramani/ Proc. Symposium on Management of Radioactive and Toxic Wastes/ Kalpakkam/ March, 1993.
3. Virginia S Shirley (Ed), Table of Radioactive Isotopes, Wiley-Interscience Publication, 1986.
11 Table 1. Iaotpic compost ion of Plutonium Isotope atom% Half life (yes) Pu-238 0.0179 8.77 E+01 Pu-239 93.3550 2.415 E + 04 Pu-240 6.2099 6.537 E+03 Pu-241 0.3940 1.45 E+01 Pu-242 0.023 3.8 E+05
Table 2 . Gamma signature of Pu-239_ Energy Intensity (keV) (%)
52 2.08 E-2 99 1.3 E-3 129* 6.2 E-3 146 1.13 E-4 171 1.09 E-4 196 1.07 E-4 204 5.6 E-4 341 6.63 E-5 375* 1-58 E-3 406, 1.0 E-5 413" • 1.51 E-3 451 1.92 E-4
Table 3. Gamma £igjia.tu_£e. o_f U-233
_ s_ as ••» «• «™ ™* —m ™* ^ "• — — — ^ "^ ^ •"" """ "^ ^ ^ ^ """ "*"' ~* ~~ ^ ^ ™" ~~ ~" "~ "^ "™ "™" ^ «— — — •— — Energy Intensity (keV) (%)
97* 2.2 E-2 146* 6.3 E-3 164* 6.6 E-3 188 2.12 E-3 248 4.5 E-3 291 5.2 E-3 317* 8.8 E-3
* Gamma rays used in the assay
12 Table 4. of Pu in tissue matrix Peak Area (cts/lOOOs) Pu Qty 129 kev 375 kev 414 kev (mg)
4.2 1820 243 201 10.5 4210 580 520 14.7 5905 815 695 21.0 8178 1118 1029 25.2 9552 1328 1192 31.6 12361 1720 1551 35.8 13868 1966 1771 42.1 16038 2319 1996 52.6 18907 2752 2445
Regression results (constrained to zero intercept) Slope 377.2 53.8 47.7 Err.slope 4.8 0.5 0.4 R2 0.994 0.998 0.997
Table 5. Assay of Pu in Neoprene matrix Peak Area (cts/lOOOs) Pu Qty 129 kev 375 kev 414 kev (mg)
10.5 3607 523 377 14.7 4906 753 552 21.0 7202 1047 826 25.2 8547 1231 1014 31.6 10143 1579 1392 35.8 11784 1725 1392 42.1 13396 2002 1714 52.6 17021 2476 2121 63.1 19391 2837 2315
Regression results (constrained to zero intercept) Slope 319.7 47.1 39.2 Err.slope 3.7 0.6 0.8 R2 0.994 0.992 0.983
13 Table 6. Assay of Pu in Sodium Peak Area (ct3/1000a) Pu Qty 129 kev 375 kev 414 kev (mg)
4.2 1349 166 34 10.5 2834 436 226 14.7 3845 639 406 21.0 5746 911 617 25.2 6943 1063 655 31.6 8308 1275 962 35.8 8780 14.13 1126 42.1 10812 1729 1385 52.6 13679 2127 1721
Regression results (constrained t;o zero intercept)
Slope 259.64 40.79 31.24 Err.slope 2.99 0.35 0.91 R2 0.995 0.997 0.978
Table 7JL Assay o£ P_u jun Steel pieces
Peak Area (cts/lOOOs) Pu Qty 129 kev 375 kev 414 kev
4.2 1724 *** ** * 10.5 4346 638 283 31.6 12308 *** *** 35.8 13892 2051 1407 42.1 16587 2543 1801 52.6 20417 3082 2317
Regression results (constrained to zero intercept) Slope 390.45 58.91 42.34 Err.slope 1.76 0.67 1.60 R2 0.999 0.997 0.979
14 Table 8. Assay of U-233 in Tissue swipes U Qty Peak area (cts/lOOOs) (mg) 97.2 kev 146 kev 164 kev 317 kev
7.4 979 560 468 282 14.8 1821 1127 1049 579 22.3 2565 1565 1597 940 29.6 3349 2108 2091 1229 37.0 4286 2439 2575 1348 44.3 5237 3208 3018 1793 51.7 5764 3442 3623 1787 59.1 6582 3809 4007 2173 73.9 7467 4525 4498 2390
Regression results (constrained to zero intecept)
slope 109.61 65.52 66.37 35.82 Err.slope 2.32 1.4 1.4 1.1 R2 0.982 0.980 0.982 0.958
Table 9. Assay of U-233 in Neoprene gauntlets
U Qty Peak area (cts/lOOOs) (mg) 97.2 kev 146 kev 164 kev 317 kev 7.4 496 400 337 192 22.2 1458 1101 1088 700 29.6 2132 1372 1518 1033 37.0 2645 2052 2054 1053 44.3 3126 2006 2335 1424 51.7 3668 2582 2716 1674 59.1 4023 2877 2843 1755 73.9 4880 3622 3730 2461 Regression results (constrained to zero intecept) slope 68.57 49.07 50.97 31.84 Err.slope 0.86 0.92 0.81 0.69 R2 0.994 0.987 0.990 0.984
15 Table 10. Assay of u-233 in Steel pieces U Qty Peak area (cts/lOOOs) (mg) 97.2 kev 146 kev 164 kev 317 kev
7.4 760 543 509 301 14.8 1543 1038 1012 612 22.2 2400 1493 1586 898 29.6 3544 2000 1953 1117 37.0 4122 2351 2393 1380 44.3 4529 2983 2908 1678 51.7 5455 3374 3500 2020 59.1 5918 3717 3820 2153 73.9 6861 4616 4792 2957 Regression results (constrained to zero intecept) slope 101.11 64.2 65.66 38.54 Err.slope 2.62 0.74 0.54 0.52 R2 0.973 0.995 0.998 0.993
Table 11. Gamma Sinature of Eu-152* Gamma Energy Absolute ( kev) Intensity (%)
121.8 28.4 244,.7 7.8 344,.3 26.6 411 2.23 444 3.1 778.9 12.96 964 14.62 1085.9 10.14 1112.1 13.54 1408 20.85
16 Table 12. Calculation of Attenuation factor jji neoprene matrix
Energy I Atten. factor (kev) (cps) (cps) (cm-1) 121.8 6.074 1.885 0.078 244.7 1.305 0.571 0.055 444 0.363 0.186 0.045 511 8.847 5.103 0.037 661.6 2.493 1.456 0.036 778.9 0.957 0.603 0.031 868 0.306 0.190 0.032 1112 0.789 0.529 0.027 1275 2.352 1.624 0.025 1408 1.022 0.709 0.024
Table 13. Results of the attenuation studies in different matrfces.*
Energy Matix Observed u corrected ( kev) cts/lOOOs cts/lOOOs
129 Tissue paper 6460 0.036 7871 (Pu-239) Neoprene 4901 0.078 8798 steel 8650 0.025 10847 sodium 5746 0.0733 10847
375 Tissue paper 974 0.0144 1086 (Pu-239) Neoprene 806 0.0454 1134 steel 1248 0.0122 1368 sodium 911 0.0428 1255 414 Tissue paper 672 0.0136 745 (Pu-239) Neoprene 560 0.0433 775 steel 779 0.0114 849 sodium 617 0.0408 837
note: steel and sodium geometries are different from the other two.
317 Tissue paper 1229 0.0145 1370 (U-233) NeopVene 1033 0.0417 1411 steel 1117 0.0151 1251
* Using the 20 mg plutonium and 29 mg U-233 standard
17 Table 14. Background observed and the minimum peak required in the" regions of interest* counting Energy ( kev) time(sec) 129 375 414
1000 370 ( 370) 56 (100) 47 (100) 2000 740 ( 560) 112 (150) 94 (150) 3000 1110 ( 600) 168 (200) 141 (200) 5000 1849 (1000) 280 (280) 235 (235)
values in brackets indicate minimum peak area required
Table 15. Computation of minimum detection limits for Pu assay. counting minimum detection limit (mg) time(sec) 129 kev 375 kev 414 kev
1000 0.981 1.859 2.095 2000 0.663 1.394 1.571 3000 0.530 1.227 1.397 5000 0.530 1.041 0.981
Table 16. Effect of external Gamma Activity on Pu assay (in sodium matrix). External Activity Peak Area* MDL in mg Pu (uCi) (cts/lOOOs) (CT=1000s) Cs-137 Co-60
0 0 8295 1.4 9.9 2.3 7513 3.8 18.23 4.2 7525 4.6 30.2 6.9 7146 5.8 51 11.4 5469 7.8 65 14 3470 13.5
using 129 kev of Pu-239
18 Table 17. Minimum Detection Limits in different matrices
(in mg for a counting time of 5000s) Isotope Tissue Neoprene Sodium Steel
Pu-239 0.53 0.63 0.77 0.51
U-233 1.96 2.5 1.98
19 Pu bearing Ge Detector waste
O
Motor
Fig.1. Schematic of Glove box solid waste assay system -2.5- o -3.5- 4.4 4.8 5.2 5.6 6 6.4 6.8 7.2 7.6 Ln(Energy) Fig.2. Variation of attenuation factor with energy 21