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Fort St Vrain Plateout Analysis for Decommissioning Study.
	ML19332E821
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Site:	Fort Saint Vrain
Issue date:	02/23/1989
From:	Jovanovic V; GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
To:	;
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ML19332E816	List: ML19332E815; ML19332E818; ML19332E819; ML19332E821;
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	909658, NUDOCS 8912130022
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SUMMARY

TITLE FSV IATDCUT NALYSIS FOR DIXDt4ISSIC1mU &D APPROVAL LEVE'l 2 9y g ,_, C OESIGN DISCIPLINE SYSTEM DOC. TYPE PROJECT 00CUMENI NO. ISSUE NOJLT R. N 18 IC 1900 909658 A QUALITY ASSURANCE LEVEL SAFETY CLASSIFICATION SEISMIC CATEGORY ELECTRICAL CLASSIFICATION QAL - I FSV - I FSV - I N/A N APPROVAL ISSUE PREPARED SSUE DATE DE PTION/ DG APPLICABLE l =* N/C APR 15 Ote V. ovancuii: !,

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A. paz r ). Alberste i . CN005790 qu\ g 2970624 AsM l CONTINUE ON GA FORM 14851 NEXT INDENTURED DOCUMENTS P.O. 28235 P G A PROPRIETARY 'IN FO RM ATION THIS DOCUMENT IS THE PROPERTY OF G A TECHNOLOGIES INC. ANY TRANSMITTAL OF THIS DOCUMENT OUTSIDE GA WILL BE IN CONFIDENCE. EXCEPT WITH THE WRITTEN CONSENT OF GA,(1) THIS DOCUMENT MAY NOT BE COPIE0 IN WHOLE OR IN PART AND WILL BE RETURNED UPON REQUEST OR WHEN NO LONGER NEEDE0 BY RECIPIENT AND (2)INFORMATION CONTAINED HEREIN MAY NOT BE COMMUNICATED TO OTHERS AND MAY BE USED BY RECIPIENT ONLY FOR THE PURPOSE FOR WHICH IT WAS TRANSMITTED. NO GA PROPRIETARY INFORMATION PAGE OF - A . . -. - - _ xx-

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909658/A CONTENTS

SUMMARY

.................................................................... 5

1. INTRODUCTION ........................................................... 7

2. METHODOLOGY AND ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.1 Core Power Distribution Analysis ...............................+.. 8 2.2 Thermal Analysis .................................................. 8 2.3 Fuel Performance .................................................. 9 2.3.1 Fuel Particle Failure ..................................... 11 2.3.2 Gaseous Fission Product Release ........................... 11 2.3.3 Metallic Fission Product Release ........................... 12 2.4 Radionuclide Invento ry Analys is . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 i 2.4.1 Circulating Inventories ................................... 13 2.4.2 Plateout Inventory ......................................... 14 2.5 Plateout Distribution Analysis ................................... 14 2.5.1 Methodology for Plateout Analysis ......................... 14 2.5.2 Model for Plateout Analysis ............................... 15 2.5.3 Fission Product Source Terma ............................... 15 2.5.4 S o rp t i on I s o t he rms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

3. RES ULTS AND D I SCUS S I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 3.1 Fuel and Graphite Temperatures .................................... 20 3.2 Burnup and Fluence Distributions ................................. 21 3.3 Fuel Particle Failure Predictions ................................. 21 l 3.4 Gaseous Fission Product Release Predictions ....................... 21 3.5 Metallic Fission Product Release Predictions ..................... 22 l 3.6 Radionuclide Inventory Predictions ..................... ......... 22

! 3.7 Plateout-Distribution Predictions ................................. 23 l c

4. COMPARISON OF PREDICTED AND MEASURED FISSION PRODUCT RELEASE AND CIRCULATOR PLATE 0UT ACTIVITIES ......................................... 25 i

5. REFERENCES ............................................................. 28 i TABLES ....................................... ............................. 33 FIGURES .................................................................... 46 APPENDIX ................................................................... 87 l

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909658/A t CONTENTS (CONT'D.) l PAGE TABLES I

1. TSV Predicted Core-Averaged Fission Gas Releases at EOC5 ............. 33

2. FSV Predicted Cs-137 Release at EOC5 ................................. 34 -

3. FSV Predicted Sr-90 Release at EOCS .................................. 35

4. FSV Predicted Circulating and Plateout Inventories at EOC5 ........... 36

5. FSV Plateout Model Parameters ........................................ 44

6. FSV Integrated Plateout in Each Primary Circuit Component; at E005 .... 45 '

FIGURES

1. FSV Core Region Identification Numbers .............................. 46

2. Peak Fuel Temperature Distributions for Segments 4 and 10 . . . . . . . . . . . 47

3. Time Averaged-Fuel Temperature Distributions for Segments 4 and 10 .. 48 l 4. Peak Graphite Temperature Distribution for Segments 4 and 10 ........ 49 l

5. Time Average Graphite Temperature Distributions for Segments 4 and 10 .............................................................. 50

6. Fissile Burnup Distributions for Segments 1 and 7 . . . . . . . . . . . . . . . . . . . 51

7. Fertile Burnup Distributions for Segments 1 and 7 . . . . . . . . . . . . . . . . . . . 52

8. Fissile Burnup Distributions for Segments 2 and 8 ................... 53

9. Fertile Burnup Distributions for Segments 2 and 8 ................... 54 i

10. Fissile Burnup Distributions for Segments 3 and 9 . . . . . . . . . . . . . . . . . . . 55 l 11. Fertile Burnup Distributions for Segments 3 and 9 ................... 56

12. Fissile Burnup Distributions for Segments 4 and 10 .................. 57

13. Fertile Burnup Distributions for Segments 4 and 10 . . . . . . . . . . . . . . . . . . 58

14. Fissile.Burnup Distribution for Segment 5 ........................... 59

15. Fertile Burnup Distribution for Segment 5 ........................... 60

16. Fissile Burnup Dist ribution f or Se gment 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . 61

17. Fertile Burnup Distribution for Segment 6 ........................... 62 1

18. Fast Neutron Fluence Distributions for Segments 1 and 7 ............. 63 )

19. Fast Neutron Fluence Distributions for Segeants 2 and 8 ............. 64 l

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4 l 909658/A CONTENTS (CONT'D.) PAGE 1

20. Fast Neutron Fluence Distributions for Segments 3 and 9 .............. 65

21. Fast Neutron Fluence Distributions for Segments 4 and 10 ..... ....... 66 1

22. Fast Neutron Fluence Distribution for Segment 5 ...................... 67

23. Fast Neutron Fluence Distribution for Segment 6 ...................... 68

24. Core Average Fissile Particle Failure History ........................ 69 '

25. Core Average Fertile Particle Failure History ........................ 70

26. FSV Predicted Core-Averaged Kr-85m Release ........................... 71

27. FSV Predicted Core-Averaged Xe-138 Release ........................... 72

28. P ADLOC Pl a t eou t Mo d e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3

29. Cs-134 Plateout Profiles at EOC5 ..................................... 74

30. Cs-137 Plateout Profiles at EOCS ..................................... 75

31. I-131 Plateout Profiles at EOC5 ...................................... 76

32. I-129 Plateout Profiles at EOC5 ...................................... 77

33. Sr-90 Plateout Profiles at EOC5 ...................................... 78

34. Te-127m Plateout Profiles at EOC5 .................................... 79

35. Comparison of Predicted and Measured Kr-85m Release .................. 80

36. Comparison of Predicted and Measured Xe-138 Release .................. 81

37. Noble Gas Half Life Dependence Including Short Lived Species ......... 82 l 38. Cs-134 Predicted Plateout Profiles and C2105 Data as of July 5, 1988 . 83 l

l

39. Cs-137 Predicted Plateout Profiles and C2105 Data as of July 5, 1988 . 84 404. I-131 Predicted Plateout Profiles and C2105 Data as of July 5, 1988 .. 85

41. Sr-90 Predicted Plateout Profiles and C2105 Data as of July 5, 1988 .. 86 i

i APPENDIX APPENDIX A "TSORT Data for FSV Cycles 1 - 5" by W. L. Lefier .......... 87 I l Page 4

          .          _             . _ . _ _                __   _       . ___ _         _ _ _ . . _ _ _      _ _ _ _ _ ~ .

i 909658/A l l

SUMMARY

A fuel performance and plateout distribution analysis for the Fort St. Vrain Nuclear Generating Station (FSV) was conducted for the entire operating life of the plant in order to support an FSV deconunissioning l study. Specifically, the purpose of this effort was to predict core power j distributions, fuel and graphite temperature distributions, fuel particle ' failure, gaseous and metallic fission product release, and plateout distributions in the primary coolant circuit. The analysis covered the i period from the beginning of cycle 1 (50C1) to the projected end of cycle 5 ' (EOC5) which is the assumed end-of-life in the decomunissioning study. This 1 analysis has been updated to include a comparison between analytical predictions and recent plateout data from FSV circulator C2105 which was i removed after the July 5, 1988 FSV shutdown and radiochemically examined at GA. ,. The analyses were based on the reference fuel performance and fission l- product transport models from Issue F of the Fuel Design Data Manual l' (FDDM/F) except for certain revisions in the fuel failure and fission gas

. release models. The most significant changes are in the fuel performance !

model for irradiation-induced OPyC failure for FSV fuel and in the diffusion parameter for 'the release of xenon from heavy-metal contamination. The revisions were based upon earlier FSV surveillance data and other experimental data for TRISO fuel particle.s. Irradiation-induced failure of ' the OPyC coating of the TRISO particles was predicted to be a major cause of , particle failure in the original FSV plateout analysis. Upon reexanination of the data base for fuel compacts made with FSV production materials, the ! original model for OPyC failure was shown to be excessively conservative for l FSV fuel, and the model was revised accordingly. The observation that the

FSV BOL Xe R/B data were underpredicted in the original analysis indicated ,

i that the Xe R/B from heavy-metal contamination - the dominant source of gas != release at BOL - was too small. Review of the available test data (primarily measurements in the GA TRIGA reactor) supported that conclusion, and the reference R/B correlation was revised accordingly. Detailed core release calculations were performed,for the key reference radionuclides: Kr-85m, Xe-138, Cs-137 and Sr-90. In addition, the circulating and plateout inventories of some 250 other radionuclides in the i primary coolant circuit were c&lculated by extrapolation of the calculated l' results for the aforementioned key nuclides. Only fission products and 10 activation products, such as Cs-134, Ag-110m and H-3, were included in the

       -analysis, while ex-core activation products were not considered.                                             The predicted release rates for the short-lived fission gases, including I-131, at the EOCS were lower than the " expected" values given in the FSV FSAR hence, the circulating and plateout inventories of these nuclides were also comparable to FSAR " Expected" inventories. The predicted release rates of the long-lived fission metals, including the cesium isotopes, were significantly below the FSAR " Expected" values.                                 This lower predicted l        release rate along with a shorter plant operating time resulted in predicted total cesiums inventories at EOC5 that are more than an order of magnitude less than FSAR " Expected" values.

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909658/A These total plateout inventories were then used as input in the plateout distribution analysis. Plateout distributions in primary coolant circuit of the FSV reactor were calculated for the key radionuclides: Cs-137, Cs-134, Sr-90, I-129, I-131 and Te-127m. A one-dimensional model of the primary circuit components was developed to simulate the flow and mass transfer characteristics of individual primary circuit components. The reference sorption isotherms from the FDDM/F were used to describe the sorptive capacity of the primary circuit materials of construction for the aforementioned radionuclides. The predicted plateout distributions are highly nonuniform, and they vary from nuclide to nuclide because of differences in the chemistries of the various fission products. However, the highest specific plateout concentrations (pCi/cm3) generally occur on the lower reflector elements and on the steam generator tube bundles. Because of their long half-lives, Cs-137 (30.1 yr) and Sr-90 (28.4 yr) are typically of particular importance in decommissioning evaluations. The maximum plateout concentration of Co-137 is predicted to be about 1 pCi/cm2 and to occur on the lower l reflector elements and in the reheater section of the steam generators. The maximum plateout concentration of St-90 is predicted to be about 0.08 pCi/cm2 and to occur on the steam generator reheater section, with

slightly lower concentrations predicted for the upper and lower reflector elements.

A comparison was also made between the predicted and the available measured fission gas release histories of Kr-85m and Xe-138 and between the predicted and measured plateout for the circulator C2105 as of the July 5, 1988 FSV shutdown. As a result of the use of the revised fuel failure and fission gas release models, better agreement with the measured fission gas release was obtained than in the original analysis. Also, better agreement with the data was obtained for xenon release at the beginning-of-life when the. release.is dominated'by the release from heavy-metal contamination, and the improvement is due to the revised value for the xenon diffusion parameter. A comparison with the plateout data indicates a good agreement was obtained between the predictions and data in case of Sr-90. However, i I the I-131 plateout data were overpredicted by a factor ranging from 3 to 17 depending upon the method used to calculate the I-131 release rate, while I the Co-137 data were underpredicted by a factor ranging from 1.4 to 9 ) depending upon the sorption isotherms used in the analysis. j i. Page 6 l

i s 909658/A

1. INTRODUCTION As described in Reference 1, Public Service Company of Colorado (PSC) has contracted with General Atomics (GA) to provide estimates of the amount and distribution of fission product plateout activity in the Fort St. Vrain (FSV) primary circuit at the End-of-Cycle 5 (EOCS). The EOC5 is the assumed end-of-life (EOL) for an 75V deconsnissioning plan that PSC is developing for l submittal to the NRC.

In support of this deconunissioning study, core power distributions, cora temperature distributions, fuel failure distributions, fission product release rates, total EOL plateout inventories, and plateout distributions were predicted for the TSV plant. The analyses covered the period from the Beginning-of-Cycle 1 (BOC1) to the E005. For the period from BOC1 to approximately 25 ef fective full-power days (ETPD) of cycle 4, the actual i operating history was approximated in the analysis by 102 constant power intervals. For the remainder of cycle 4 and for cycle 5, it was assumed that the reactor would operate at 80% power for 80% of the time. In addition to this analytical effort, a comparison was made between the predicted and the available measured fission gas release histories of Kr-85m and Xe-138 in order to assess the validity of these predictions. The results of the analysis were documented in the original issue of this document (Issue N/C). As reported herein, this analysis has been revised to include a comparison of the predicted and measured plateout activities on TSV circulator C2105 which was removed during the July 5, 1988, shutdown and radiochemically examined at GA (Ref. 38) and to update the predicted plateout activities. s s Page 7

5, i 909658/A I

2. METHODOLOGY AND ASSUMPTIONS These analyses were conducted using methods that are advanced relative to those used in FSV licensing analyses. Comparisons of results obtained with values in FSV licensing documents will show significant differences.

Results presented in this report are believed to be conservative. I 2.1 Core Power Distribution Analysis i The radial power distributions were calculated with the GAUGE computer l code (Ref. 2) for all of the regions and fuel columns with seven I calculational points per. column, as shown in Figure 1. These results were synthesized with the GATT code (Ref. 3) axial power flux and burnup distributions using the TSORT code (Ref. 4). The number of time points used in the analysin is given in Tables 1 through 5 of Ref. 5, which is included here as Appendix A for easy reference. As discussed in more detail in Ref. 5, a different procedure was used for cycles 4 and 5 core power distribution calculations regarding the treatment of the partially buffered fuel elements on the periphery of the l active core than has been done in previous FSV fuel performance analyses. For the calculation of the fission rates in the partially buf fered fuel columns, which is done in the thermal and fuel performance computer code SURVEY (Ref. 6), the initial atom densities, as calculated with the physics . codes, were used directly without the adjustments made previously for cycles 1, 2 and 3; this procedure is more consistent with the assumptions used in core physics analyses. Details are provided in Appendix A. The results of the core physics calculations were used as input for the thermal and fuel performance analyses. ; 2.2 Thermal Analysis Nominal thermal and flow parameters and nominal values of material l properties were used in the analysis. The material properties of the fuel element graphite (Ref. 7) account for thermal expansion and for the effects of fluence and temperature on thermal conductivity and irradiation-induced shrinkage. The model for the thermal conductivity of the fuel rods, given in Ref. 8, was used in the analysis; this model accounts for the effects of irradiation, temperature and shim content (if any) on the fuel rod thermal conductivity. The thermal conductivity and dimensional change models for I graphite and fuel rods (Ref. 7 and 8) have been programmed into the l SURVEY / THERM computer code (Ref. 6) which was used to calculate the fuel and graphite tamperatures as a function of location and time for the first five : cycles. The code was modified in order to be able to account for the variation of the shim content with the fuel reload segment; no shim o particles were used in the FSV fuel until segment 8. l The properties of H-327 graphite were used for the entire analysis although half of the segment 9 and all of the segment 10 fuel elements are made of H-451 graphite. This assumption results in slightly conservative L temperatures as the effects of slightly smaller fuel hole gap for the H-327 graphite at elevated temperatures (as a result of less irradiation-induced shrinkage) are more than compensated by the lower thermal conductivity of H-327 compared to H-451. Page 8

I i l 909658/A ; The core coolant inlet temperature and the total core coolant flow rate l were input as a function of time. These values were taken from Tables 1 ' through 5:(for cycles 1 through 5) of Ref. 5 (Appendix A). The exit coolant i temperature deviation, due to mismatch, thermocouple error, etc., was varied , with time for each region from BOC1 to 25 EFPD's of cycle 4. These values l were taken from the data logger tapes for the remainder of cycle 4 and for cycle 5, the values corresponding to the assumed 801 power were used. The analysis was performed at the bottom points of all six axial layers of fuel elements. 2 . 3. Fuel Performance Typically, the two dominant sources of fission product release from the , core are as-manufactured, heavy metal contamination (i.e., heavy metal outside the coated particles) and particles whose coatings fail in service. In. addition, the volatile metals (Cs and Sr) can, at sufficiently high , temperatures and long times, diffuse through the sic coating and be rolessed from intact TRISO particles. There are multiple barriers to the release of fission products from a HTGR core the fuel kernel, the particle coatings, the fuel rod matrix, and the fuel element graphite. The effectiveness cf the individual barriers to fission product release may depend upon a number of factors including the chemistry and half-lifes of the various fission products, temperature, and irradiation effects. These barriers are described briefly below. The first barrier to fission product release is the fuel kernel itself. The kernel of a failed fuel particle retains > 95% of the radiologically important, short-lived fission gases such as Kr-85m and I-131; however, the

   . effectiveness of the fuel kernel for retaining gases can be reduced if the exposed kernel is hydrolyzed by reaction with trace amounts of water vapor which may be present in the helium coolant.                               The retentivity of fuel                         ;

kernels for long-lived, volatile fission metals such as Cs and Sr is strongly dependent upon the temperature and the burnup. The primary barrier to fission product release from the core is the silicon carbide and/or pyrocarbon coatings of each fuel particle. Both the sic and outer pyrocarbon (OPyC) coatings provide a barrier to the release of ' fission gases. The Sic coating acts as the primary barrier to the release of metallic fission products because of the low diffusion coef ficient of fission metals in sic the OPyc coating is partially retentive of Cs at lower temperatures but provides little holdup of Sr. The fuel rod matrix is rather porous and provides little holdup of the fission gases which are released from the fuel particles. However, the ' matrix is a composite material which has a high content of amorphous carbon, and this constituent of the matrix is highly sorptive of metallic fission products, especially Sr. While the matrix is highly sorptive of metals, it provides little diffusional resistance to the release of fission metals because of its high interconnecte( porosity. The fuel element graphite, which is denser and has a more ordered structure than the fuel rod matrix, is somewhat less sorptive of the fireion Page 9

       ..                                                                              j l

l 909658/A 1 1 metals than the matrix, but it is much more effective as a diffusion barrier ! than the latter. The ef fectiveness of the graphite as a release. barrier ; decreases as the temperature increares. Under typical FSV core conditions, the fuel element graphite attenuates the release of Cs and from the core by , more that an. order of magnitude, and the Sr is essentially quantitatively ; retained. The above discussion applies to the transport of fission products that are produced in the kernels of intact particles. Obviously, fiscion products resulting from fissions in heavy-metal contamination outside of the particles are not attenuated by the kernels or coatings, nor are the fission products produced in the kernels of failed particles appreciably attenuated by the failed coatings. . The performance of coated fuel particles is calculated by models ; defining several potential f ailure mechanisms. The HTGR fuel performance models calculate fission product release to the reactor coolant during normal operation f rom the following seven sources L. Coating damage during fuel manufacture, resulting in heavy metal contamination on coating eurfaces and in the fuel rod matrix. 2 .' Pressure vessel failure in particles with defective or missing coating layers. -

3. Pressure vessel failure in standard particles, i.e.,

particles without manufacturing defects.

4. Irradiation-induced failure of outer pyrocarbon coating as function of fluence in both standard and defective particles.

5. Failure of the sic coating caused by fission product / sic interaction.

6. Failure of the sic coeting by thermal decomposition.

[

7. Failure of the sic coating.due to kernel migration in the presence of a thermal gradient.

These failure mechanisms and the physical models that describe them are discussed in Ref. 8. l The models and material property data for predicting fission gas release from heavy-metal contamination and failed particles are described in

   .Refs. 8 and 9.        These models give the release rate-to-birth rate ratios (R/B) from failed particles and contamination as a function of chemical element, isotope half-life, temperature, and burnup.             In addition, the effect of fuel hydrolysis, or reaction of exposed fuel kernels with water, on gas release is included.       These fission gas release models are embodied in the SURVEY /PERFOR code (Ref. 6).

l The models for predicting fission metal release from fuel particles and fuel elements are described in Ref los the material property data are given Page 10 i i L_.

   ,                     ,                   .                              a.

_._.-_s.-. . __.1.- --- a ~ - ~ . - . . I t 909658/A in Ref. 8. The transport of fission metals through the kernel, coatings, fuel rod matrix, sind fuel element graphite is modeled as a transient diffusion process in the TRAFIC code (Ref. 11). The sorption isotherms , which are used in the calculation of the rate of evaporation of volatile ! metals.from graphite surfaces account for an increase in' graphite sorptivity l with increasing neutron fluence. - The methodology described here was used to predict the fuel performance and fission product release for the FSV core. The application of the fuel performance methodology is described in the following sections. 2.3.1 Fuel Particle Failure ! Using the calculated fuel temperature histories (Section 2.2), burnup , and fast fluence hist.ories, and fuel performance models, the fuel particle failures were calculated se a function of time using the fuel performance code SURVEY /PERFOR (Ref. 6). The particle failures from the aforementioned mechanisms were calculated at the 50% confidence level. As reconsnanded in Ref. 12, the fuel failure models given in the FDDM/F (Ref. 8) were used for ' the FSV fuel with the exception for (1) the kernel migration coefficient correlations which are not specified in the FDDM/F for the reference FSV - fuels the appropriate correlations for the FSV fuel, based on the data from ; Ref. 13 and 14, were used in this analysis, and (2) the revised OPyC failure model, as discussed in the following paragraph. The manuf acturing defect , fractions and irradiation induced maximum failure fractions for the different service-limit fuels and for different fuel segments up to fuel segment 10 were taken from Table 1 of Ref. 12. Of particular importance to the subject analysis is that fuel particles, especially the fertile particles, in segments 1-7 had OPyC coatings with high microporosities. Based upon the results of accelerated fuel irradiation capsules and fuel test elements in Peach Bottom 1, very high failure rates (>30%) for these OPyC coatings at a modest fast fluence of 2 x 1025 n/m2 were predicted in the original analysis. This original failure model was too conservative and its use resulted in an overprediction of the fission gas release data as discussed in the Issue N/C of this document and in Refs. 30 and 31; consequently, it was decided to reexamine the data base for the OPyc f ailure for the FSV fuel compacts. Since the conse rvatism in the model was considered to be due to the inclusion - of nonrepresentative irradiation data, the OPyC performance model was revised by including only the data for the fuel compacts made with FSV production materials (Ref. 34), and the revised mods 1 was used in this analysis. The SURVEY /PERFOR code had ,to be modified so that it has the capability to correctly account for the variation of the fuel quality with service limit, fuel-segment and axial location. The analysis was performed at six axial locations in order to account for the effects of the axial variation of fuel temperature and burnup on the fuel particle failure.

     -2.3.2     Gaseous Fission Product Release                                                                                                              .

1 The R/B models are included in the SURVEY /PERFOR code which was used to calculate the fission gas release at six axial locations. The results of the SURVEY code analysis were averaged radially and axially using the Page 11

909658/A post-processing codes TREAD and TFIN (Ref. 6) in order to obtain core-average results. The gaseous fission product releases were calculated for the two key isotopes Kr-85m and Xe-138. These isotopes were also selected for analysis because their predicted releases can be directly compared with measurements taken as part of the FSV radiochemistry surveillance program. The fission gas release measurements were obtained from Refs. 15, 16, 17 and 35: the latter reference contains the data up to the July 5, 1988 reactor shutdown. In calculating the fission gas release, it was assumed that the , performance models for the UC2 fuel particles apply to the (Th/U)C2 and ThC2 f fuel particles, as reconsnanded in Ref. 8. The fission gas release model was revised in that the temperature enhancement term (Ref. 8) was eliminated because it has asen shown to be excessively conservative for steady-state ' reactor operation. The diffusion parameter for the release of xenon f rom heavy-metal contamination was also revised (Ref. 36) because the use of the original value (Ref. 8) resulted in an underprediction of the xenon release for the FSV core at the beginning of cycle 1 when the entire release is from . the heavy-metal contamination as discussed in Ref. 30 the revised value is based upon the available experimenta1' data. For the calculation of the fission gas release due to as-manufactured heavy-metal contamination of the , fuel rods, the thorium and uranium contamination fractions were taken from the QC records for the initial core and reload segments (Refs. 18, 19 and

20). Nominal values for the material properties (Ref. 8) were used in the calculations. 3 The releases fer other isotopes can be obtained by assuming that the release rate-to-birth rate ratio (R/B) varies as the square root of isotope half-life. Moreover, it is assumed that bromine and selenium isotopes have the same release characteristics as krypton isotopes and that iodine and tellurium isotopes have the same release characteristics as xenon.

2.3.3 Metallic Fission Product Release The analysis was performed for the key nuclides Sr-90 and Cs-137 by using the TRAFIC computer code (Ref. 11) and the fuel and graphite l temperature and fuel particle failure histories described in Sections 2.2 and 2.3.1. Material property data were obtained from FDDM/F (Ref. 8) with the following exception. In the calculation of the diffusion of cesium l through the fuel element graphite, the dif fusivity from the Issue B of the FDDM were used since the use of the FDDM/F correlation resulted in a i significant underprediction of the FSV plateout probe data while the use of the FDDM/B values overpredicted the probe data by a factor of 6 (Ref. 31). The following assumptions were used in modeling fission metal release of cesiuse and strontium from the fuel particles:

1. No release from intact particles.

2. Complete release from failed particles.

The analysis was performed at six axial locations. In addition to the direct release of Sr-90 metal from failed particles and fuel-rod matrix Page 12

I J l 909658/A contamination, two additional sourcas of the St-90 release were considered:

1. Heavy-metal manufacturing contamination on coolant hole surfaces of fuel elemento. ;

2. Precursor (Kr-90) release and subsequent decay.

4 For Cs-137, the contribution from the precursor (Xe-137) decay was also accounted for in addition to the direct release; the Ce release due to

}-  contamination on coolant hole surfaces of fuel elements is negligible                    i g    compared to the release of Ce from failed fuel partic19s.

2.4 Radionuclide Inventory Analysis The circulating and plateout inventories of 250 fission and activation product nuclides were calculated; these do not include ex-core activation products. The fission product nuclides included can be circumscribed by two - general criteria: only nuclides with a yield from U-233 U-235. Pu-239, or Pu-241 greater than 0.1% and with a half-life greater than 1 min are included. Exceptions are the short-lived noble gases Kr-90, Kr-91, Xe-139. - and Xe-140 which have important daughters. Radionuclides with a half-life longer than ~220,000 yr are treated as stable nuclides (this includes the radiologically important nuclide I-129, which has a 1.6 x 107 yr half-life). Fission product yields, branching ratios, and decay constants were obtained from Ref. 21. ; The activation product nuclides H-3, C-14, and As-110m were also included in the analysis. The release of tritium f rom fuel elements and from the control rods into the helium coolant is treeted conservatively by - assuming that all the tritium from the failed fuel particles (ternary fission), from the graphite and matrix Li-6 impurity activation, and B-10 reactions in tho' lumped burnable poison and control rods ,is completely released into the helium coolant. The resulting circulating inventory of tritium was calculated with the TRITGO computer code (Ref. 23). The release of C-14 into the helium coolant is negligible and has been set to zero. The production modes of Ag-110m are (1) the low direct Yission yield and (2) the ' neutron activation of stable Ag-109, a higher yield fission product nuclide, resulting primarily from Pu fissions. The other activation product nuclides considered in this analysis are Cs-134, Cs-136, Pm-148m, Pm-148, Eu-152, Eu-154, and Eu-155. These nuclides are produced through the neutron activation of other fission products. The release of these nuclides from the core is determined on the same basis as , for the fission product isotopes of that element (e.g., Cs-134 and Cs-136 are assumed to have the same fractional releases as Cs-137, which was calculated with the detailed analysis described in Section 2.3.3). 2.4.1 Circulating Inventories l The release fractions (release rate into coolant / birth rate in fuel) ; for Kr-85m and Xe-138 were taken from the results of the detailed analysis ' , described in Section 2.3.2. These values define the release fractions for l other isotopes of krypton and xenon (by assuming that release fractions vary l Page 13

                  -                     . ._ -              .   .      .. .-_.. -             _    - . ~ . . . .

i 909658/A as the square root of the isotope half-life). Isotopes of bromine and selenium are assumed to have the same release characteristics as kryptons , isotopes of iodine and tellurium, the same as xenon. The coolant activities f are based on a 40% plateout per circuit loop for all condensable nuclides except iodine, for which 1% plateout/ pass was assumed. 2.4.2 Plateout Inventory The total plateout inventory is calculated independently by , conse rvatively assuming that all the condensable radionuclides that are i released f rom the feel elements plate out in the primary circuit. In other words, the calculation of the circulating and plateout inventories in the primary circuit is decoupled. The circulating activity (Section 2.4.1) is calculated using an appropriate plateout per pass. Consequently, an exact mass balance is not preserved: 1.e., the sum of the plateout inventory, the circulating inventory, and the purification system inventory slightly exceeds the total core release. However, even with a plateout per pass as small as 1% (the FSV FSAR value), >99% of the condensable activity is plated out so the degree of conservatism is small. The plateout inventories of the condensable fission gases, e.g., iodine and tellurium, were calculated with the same core release fractions as used to calculate their circulating activities (Section 2.4.1). The fractional l-release calculated for Cs-137 in the detailed analysis (Section 2 3.3) was used for the other cesium isotopes. Strontium isotopes were conservatively assumed to have a fractional release of 1 x 10-7 as a result of heavy-metal contamination on coolant hole surfaces (Sr released from fuel-rod matrix contaminatior. and failed fuel particles is completely retained by the fuel-element graphite). Based upon their relative volatilities (as inferred from their elemental boiling points), As, Rb, and Cd isotopes are assumed to have the same release fractions as Cs and Eu, Sb, Ba, and Sm isotopes, to have the same release fractions as Sr (Ref. 10). 2.5 Plateout Distribution Analysis The plateout distributions of condensable fission products in the primary coolant circuit are necessary to establish reactor decommissioning i requirements. Consequently, the distributions of the key radionuclides in i the primary circuit need to be calculated. Computation of fission product plateout profiles in the primary circuit requires details of the primary circuit geometry, reactor operating conditions, i.e., helium coolant flow rates and temperature profiles as a function of time along the primary circuit and finally, the fission product release rates as a function of time, i The methodology used in obtaining the fission product release rates is described in Section 2.3. The methodology and assumptions used in the fission product plateout analysis is described below. 2.5.1 Methodology for Plateout Analysis Plateout distributions are calculated using the PADLOC computer code (Ref. 26). This computer code performs a mass transfer calculation using Page 14 l

1 i l 909658/A J mass transfer correlations and sorption isotherms to determine the partitioning of condensable radionuclides between the flowing coolant and the fixed surfaces in a recirculating loop. The plateout model in PADLOC is limited to a one-dimensional cylindrical geometry, such that all components of. the primary circuit must be modeled as a equivalent series ' of coupled j sections of parallel banks of cylindrical tubes. i I The calculations were performed for the following key nuclides: 3r-90, l I-129, Cs 134 Co-137 and Te-127m. The multiple-species version of the code (PADLOC*INDIF) was used in the analysis to correctly calculate the plateout , distributions of nuclides such as Sr-90 and Cs-137 which have gaseous precursors. 2.5.2 Model for Plateout Analysis A block diagram of the primary circuit of the FSV reactor is shown in Fig. 28. This block diagram shows the helium coolant flow path in the primary circuit of the reactor. A majority of data for this one-dimensional model of the primary circuit components for the PADLOC analysis was formulated based on the FSV coolant circuit parameters given in Ref. 27. Additional dimensional datu ass needed to model the core inlet and outlet plenums, the core barrel / annulus and the core support blocks. This data was obtained from FSV drawings listed as Ref. 28. Since the PADLOC code can accommodate flow networks composed only of cylindrical tubes, modeling of each component was accomplished with due consideration for the preservation of the surface area, the flow area and mass transfer characteristics. In certain cases, it is impossible to niodel , i exactly both the correct surface area and the flow area. In such cases, l approximations are required. Some of the components are divided into various sections, or branches in the PADLOC terminology, such that the flow characteristics could be correctly modeled. The model used in the analysis consists of 17 branches with one or more parallel cylindrical channels: the diameter of the equivalent channel is taken as the hydraulic diameter of the actual geoinetry. The significant parameters of the model for each branch are listed in Table 5 and the branch connections are shown in Fig. 28. l (Details of the derivation are given in Ref. 29).

2.5.3 Fission Product Source Terms The source terms for fission product plateout analysis includes a direct release contribution and in some cases a precursor contribution as well. In the case of the cesium isotopes, there is direct release of both Cs-137 and Cs-134 metal f rom the core s Cs-137 plateout also results from releece and subsequent decay of its precursor contribution Xe-137, but l Cs-134 has no gaseous precursor. Similarly, for Sr-90, there is direct i Sr-90 metal release as well as the contributions f rom its Kr-90 precursor.

Only direct release contributions need to be considered for I-129, I-131 and Te-127m. The source terms for the key nuclides are obtained from the gaseous and metallic fission product release analysis. Time histories of the fractional releases for the gaseous precursors, i.e., Kr-85m and Xe-138 are available for 120 time points during cycles 1 through 5 of operation of the reactor. Page 15

909658/A These are shown in Figures 26 and 27. The precursors of interest to the plateout analysis are Kr.90 and Xe.137, which are precursors for Sr.90 and Cs.137, respectively. Time dependent fractional releases, or R/Bs, for Kr.90 are obtained from the R/B data for Kr.85m by the relation: . (a/B)Kr-85m (AKr.90)0.5 (R/B)Kr.90 (AKr.85m)0.5 where A is the decay constant. Similarly, time dependent R/Bs for Xe.137 are obtained by (R/B)Xe.137 (axe.138)0.5 (R/B)Xe-138 (axe.137)0.5 Once the time dependent values for the R/B are known, a trapezoidal l integration is performed to obtain time averaged values for each cycle of operation. This averaging technique is needed to obtain a correct mass balance of fission product release during the operation of the reactor. The averaging technique is also essential, because PADLOC computer coda can not handle the detailed operational history of the reactor, and it is not necessary to follow the detailed operation to obtain accurate plateout profiles. The time average R/B values for gaseous precursors are converted to l average core release rates by using the following relationship. R =PxFxY x R/B Where R = Release rate (atoms /s) P = Reactor operating power (watts) F = Fissions /(watt.s) = 3.15 x 1010 Y = fission yield (atoms / fission) R/B = Release Rate / Birth Rate of gaseous precursor The average calculated release rates for Kr.90 and Xe.137 during each cycle of operation are given in the table below. These release rates are in units of pg/cm 2.s for core coolant channels and tre the input for PADLOC plateout analysis. i I l Page 16

909658/A FSV Plateout Analysis Xe-137 and Kr-90 Release Rates Start End Average Average Xe-137 Average Kr-90 (day) (day) Power Xe-137 Release Kr-90 Release Fraction R/B Rate R/B Rate pg/cm2.s pg/cm2.s l j 0.00 491.81 0.37 2.554E-06 2.722E-12 7.002E-06 3.122E-13 491.81 842.31 0.54 2.382E-06 3.686E-12 9.226E-06 5.975E-13 H 842.31 1317.40 0.62 2.083E-06 3.7372-12 1.345E-05 1.010E-12 I l 1317.40 1765.70 0.65 2.042E-06 3.821E 12 1.736E-05 1.360E-12 l 1765.70 2130.70 0.80 1.870E-06 4.297E-12 1.776E-05 1.708E-12 i Metallic fission product release rates or source terms for PADLOC plateout analysis are computed in a slightly different manner. The total release of a given metallic fission product at the end of each cycle is available from the metallic fission product release analysis ! (Section 2.3.3). This release data can be converted to release rates for input to PADLOC. ; The release in Curies of Cs-137 during each cycle of operation and the , corresponding release rates in pg/(cm2.s) are listed below. The release i rates given here are three times the release for Cs-137. Using three times the actual value in the PADLOC analysis accounts for the other isotopes of cesium, such that the total mass cesium metal on the primary circuit is 3 properly accounted in the sorption calculations. The plateout profiles calculated using this method are then reduced by a factor of 3 to obtain the actual values. l l l Page 17

1 909058/A ; t FSV Plateout Analysis Cs-137 Release Rates Cycle Start End Release Release Number Days days Curies Rate pg/cm2 ,,- l 1 0.0 491.81 1.746E+00 7.295E-12 2 491.81 842.31 8.677E+00 5.076E-11 3 842.31 1317.41 2.265E+01 9.775E-11 4 1317.41 1765.66 4.325E+01 1.978E-10 ' 5 1765.66 2130.66 1.676E+01 9.413E-11 Cs-134 is an activation product of the fission product Cs-133. This activation product builds up slowly at the beginning but more rapidly at the end of life. The release rate of Cs-134 is then a function of time and is difficult to follow precisely in the plateout analysis. Consequently, it is assumed that the release of Cs-134 is proportional to Co-137 release, such that the release f ractions for Cs-134 and Cs-137 are equal. The predicted I-131 R/B at EOC5 is 6.1 x 10-05 Since I-131 is a short lived isotope, the plateout profile is determined by the last 100 days of ' operation of the reactor, During this time the reactor is assumed to - operate at 80% of full power. From the R/B value of 6.1 x 10-05, th. release rate is computed to be 1.3 x 10-10 pg/(cm2 .s). It should be emphasised that the above I-131 R/B was calculated from ~ the Xe-138 R/B by I using the reference theoretical 0.5-power half-life dependence and that this procedure is conservative. The observed power dependence in FSV is considerably lower, as can be seen in Figure 17 (" Cycle 3 line, 70% power") of Reference 37, which is included as Figure 37 of this report. Also shown in Figure 37 is the recent FSV release data for short-lived nuclides taken prior to the July 5,1988 FSV shutdown (Ref. 35) and these data fall right on the " Cycle 3 line", thus, confirming the lower power dependence. Because ' the use of the lower power dependence will result in a more realistic I-131 source, the plateout distribution analysis was also performed for this I condition. For the 1.6 x 107 -yr half-life I-129, the end-of-life release fraction is 1.17 x 10-03, which is equivalent to a release rate of 5.97 x 10-10 ,' pg/(cm2.s) at 80% of full power. Similarly, I-127 release rate at 80% of full power is 1.47 x 10-10 pg/(cm2 ,,), ; page 18

r) 909658/A The release f raction for Te-127m is 2.14 x 10-04, which is equivalent to a release rate of 7.273 x 10-12 pg/(cm2 ,,), 2.5.4 Sorption Isotheram The sorptivities of graphite and metal surfaces of the primary circuit for the condensable fission products were calculated using the sorption isotherms given in FDDM/F (Ref. 8). For the sorption isotherms on graphite as given in FDDM/F, a conversion had to be made for compatibility with the PADLOC code input requirements since the concentration is given in amoles /g of graphite in the FDDM/F, and the PADLOC code requires concentration based ' on the geometric surface area. In making this conversion, it was assumed that the entire mass of graphite (i.e., the entire internal surface area) is available for sorption. Sorption data for Sr-90 and Te-127m on metals are not available, and they are assumed to exhibit perfect sink behavior, i.e., negligible desorption pressure for all surface concentrations and temperatures. There l are alec no sorption data for the silica brick which lines the core exit plenu . Since certain silica-alumina compounds (e.g., mullite) have very high affinities for cesium, the silica brick was conservatively assumed to be a perfect sink for cesium. Since the iodine sorptivity of most ceramic materials is very low at high temperatures, the silica brick was assumed to have the same iodine sorptivity as graphite. In any case, the surface to volume ratio in the core exit plenum is low resulting in unf avorable mass transfer conditions so that little deposition is predicted there even with

   .2    the assumption of perfect sink plateout.

Page 19

            =                . _ . _       . - _                   - .  . - .       - . . - - -.- -                         - .. .     . - .

909658/A

3. RESULTS AND DISCUSSION '

3.1 Fuel and Graphite Temperatures Although the analysis was performed for the entire active core, the fuel and graphite temperature predictions are presented only for the y refueling segments 4 and los the latter was chosen because the maximum fuel and graphite temperatures were predicted to occur . in a fuel element , belonging to this segment, and the former represents a " typical" initial core fuel segment regarding the fuel and graphite temperatures. The volume ; distributions of the predicted fuel and graphite temperatures are presented in Figures 2 through 5 for tie fuel reload segments 4 and 10sthe figures show the distributions of the peak and time-averaged temperatures. , The largest temperatures were predicted to occur at a few locations on the outer boundary of the active core at the corners of the fuel elements surrounded on both sides by reflector elements and in every case in reloaded fuel elements. At these peripheral locations very large point power factors and power tilts were calculated in with the GAUGE code for both partially buf fered and fully buffered fuel elements. In case of partially buffered fuel elements, the GAUGE code can overestimate such point power factors and tilts by some 151. For the fully buffered fuel elements, the GAUGE code with the seven-group cross-sections used in this analysis for cycles 3, 4 and 5 (Ref. 5) calculates such point power factors that are some 20% higher than with the four-group cross-sections the use of the latter is considered to give more realistic representation at such points at the reflector boundary interface. Also, the SURVEY code overestimates graphite temperatures at such points since it does not account for convective losses to the gap flow and for the radiative losses to the adjacent reflector element. A previous detailed thermal analysis of such fuel elements (Ref. 24) indicated that, due to physics modeling difficulties, the predicted intercolumn power tilts in partially buffered fuel elements can be excessively conservative, resulting in significant overpredictions of fuel p and graphite temperatures, as was the case in this analysis. Even though ' these temperatures are excessive, they were included in the analysis since l they are the result of the current core physics methodology and since their l use results in clearly conservative predictions of temperature-induced fuel failure and fission product release. For the entire core, the peak and time-averaged maximum fuel tempera-tures of 147500 (26870F) and 10760C (19680F) were predicted, as shown in Figures 2 and 3. These maximum temperatures were predicted to occur in segment 10, region 26, column 1, local point 5, which is a point in a fully buffered fuel element (see Fig. 1 for the region, column and local point designation). The peak graphite temperature was predicted to be 14170C (25830F) and 10460C (19140F) on time-average basis, as shown in Figures 4 and 5. The next highest fuel temperature was predicted to occur in region 28, c'olumn 5, local point 5, which is a point in a partially buffered fuel element of segment 7. Somewhat lower fuel and graphite temperatures were predicted for segments 8 and 9, again at the outer radial boundary and the bottom of the active core and several days after the reload. It should be emphasized that only very small fractions of fuel and graphite in the Page 20 __._..__._._.-_,_______a . . _ . . . . _ _ _ _ _ ~

909658/A as can be l reload segments were predicted to reach such high temperatures, seen in the figures. 3.2 Burnup and Fluence Distributions Tuel volume distributions of fissile and fertile particle burnup for At the EOC5, the segments 1 through 6 are shown in Figures 6 through 17. maximum fissile and fertile burnups were less than 19% and 5.1% FIMA, respectively. The maximum burnups occur in fuel segments 5 and 6 because these segments are not reloaded during the first five cycles, and thus their fuel has a longer residence time than other fuel segments. These values are lower than the design values ef the fissile and fertile burnup of 20% and 7% FIMA, respectively. Fuel volume distributions of fast neutron fluence for segments 1 and 6 are shown in Figures 18 through 23. Again, the maximum fluence is reached in fuel segments 5 and 6. The maximum value was less than 6.5 x 1025 n/m2, which is lower than the design value of 8 x 1025, 3.3 Fuel Particle Failure Predictions Time hiotories of the core-averaged fuel particle failures are shown These in Fig. 24 and 25 for the fissile and fertile fuels, respectively. failures are based on the FDDM/F fuel particle failure models and revisions thereof, s, discussed in Section 2.3. These failure models account for partially failed fuel particles, i.e. particles with f ailed SIC but intact OPyC coatings, which are assumed to retain fission gases but to release fission metals. Accordingly, fissiletwo sets of f ailure fractions are shown in particles, the maximum predicted Fig. 24 and 25. For the core-averaged fuel particle failures were 0.046% and 0.350% for the particles with exposed kernels (resulting in fission gas release) andFor failed the SIC coatings (resulting in fission metal release), respectively. fertile particles, the maximum prod.icted fuel particae failures were 0.086% and 1.25% for the particles with exposed kernels and f ailed Sic coatings, respectively. A comparison of these results with those of the original analysis indicates that the predicted fuel failures for the particles with exposed kernels were reduced as a result of the The use ofreduction largest the revised wasmodel in caseforof the the OPyC irradiation-induced failure. fertile particles, where the maximum predicted f ailure was reduced by a factor of 3.0. The large predicted f ailures for particles with f ailed SIC The coatings are due almost entirely to the manufacturing defects. decreases in the fuel particle failure at the end of each cycle are due to the replacement of a fuel segment (approximately one-sixth of the core)As canwithbe a higher-quality fresh fuel with lower manuf acturing defects. seen in Fig. 24, the fissile fuel failures increase somewhat during each cycle due to the increasing inservice f ailure. 3.4 Gaseous Fission Product Release Predictions The predicted release rate-to-birth rate ratios (R/Bs) or fractional releases for the two reference fission gases Kr-85m and Xe-138 are shown in Fig. 26 and 27 at a function of time. The predicted fission gas releases at Page 21

i 909658/A I EOL (EOC5) are shown in Table 1 along with the contributions from failed

fissile and fertile particles and the as-manufactured heavy-metal contamination to the total release. The predicted EOL R/Bs. for Kr-85m and Xe-138 were 2.0 x 10-5 and 2.1 x 10-6, respectively. A comparison with the ,

results of the original analysis indicates that the predicted EOL R/B for ' Kr-85m has been reduced by a factor of 1.6 as a result of the lower ; predicted fuel f ailure, as discussed in Sect. 3.3, and of the use of the revised R/B model for the failed fuel, as discussed in Sect. 2.3.2. However, there was virtually no reduction in the predicted EOL R/B for Xe-138, as the reductions in the predicted release from the f ailed fuel were just about compensated by the increase in the predicted release from heavy-metal contamination as a result of the enhanced value for the diffusion parameter, as discussed in Sect. 2.3.2. 3.5 Metallic Fiasion Product Release Predictions i The Cs-137 and Sr-90 predicted releases are shown in Tables 2 and 3. As can be seen in Table 2, the axial distribution of cesium release is highly nonlinear. Near the top of the core, the cesium release is very small since the fuel alement graphite is very effective at low temperatures in retaining coeium released from the fuel rods. The maximum cesium release was predicted to occur at the bottom of the axial layer 6 (the bottom of the active core) where maximum fuel particle f ailures and graphite temperatures were predicted to occurs as a result of high graphite temperatures, graphite attenuation of cesium is substantially reduced compared to the top of the core. This is particularly true for the buf fer points where the largest - fuel and graphite temperatures were predicted to occur, as discussed in Se: tion 3.1. 3ince these temperatures were overpredicted, the cesium release from these buffer points is conservative. The predicted EOL (EOC5) Cs-137 plateout inventory of 32.9 Ci (which includes a 2.9 Ci contribution from the Xe-137 decay) is well below the FSV FSAR values of 667 Ci for

                         " Expected" 30-yr activity and 5220 Ci for " Design" activity, as given in Table 3.7-2 of Ref. 25.

The direct Sr-90 release was negligibly small because of essentially complete retention by the fuel element graphite even at the hottesc t locations. The dominant contribution to the strottium release was from the Kr-90 decay, which accounted for 88% of total release at EOC5; the remaining contribution was from heavy-metal manufacturing contamination of the coolant 2 hole surfaces (a minimum fractional release of 1 x 10-7 is assumed for all fission products to account for this contamination). .The predicted EOL (EOCS) Sr-90 plateout inventory was 2.4 C1. This value is lower than the FSV FSAR values of 6.6 Ci for " Expected" 30-yr activity and 561 Ci for I " Design" activity, as given in Table 3.7 2 of Ref. 25. The above predicted inventory is a f actor of 1.5 lower than the original prediction (Issue N/C of this document) as a result of the lower Kr-90 predicted release. 3.6 Radionuclide Inventory Predictions The steady-state circulating and EOL plateout activities of some 250 fission products and activation products are given in Table 4. The coolant j activities are based on 40% plateout per circuit loop for all condensable j nuclides except iodine, for which 1% plateout/ pass was assumed. Column 3 of i Table 4 lists the values of the fission product and activation product Page 22 l i-l

   +
                                                                                                           )

909658/A 9 nuclides in the primary coolant with normal operation of the purification

               . system.      Column 4 of Table 4 lists the cumulative activities in the helium purification system during the plant lifetime (E005).

The total primary circuit plateout inventories at EOL (FOCS) are given p in column 5 of ' Table 4. Columns 6 through 11 give corresponding values after the following decay times: 10 d, 100 d, 1 yr, 3 yr, 10 yr and 30 yr. 3.7 Plateout Distribution Predictions The results of the analysis are presented in graphical form in Fige. 28 through 36 which show the plateout distributions of the key fission product - nuclides at EOC5. Essential features of the PADLOC model of the FSV primary circuit are ' summarized in Table 5. Integrated values of the activity of various nuclides in all the branches of the primary circuit at the EOC5 are given in Table 6. The plateout distributions for Cs-134 and Co-137 are shown in~ Figures 29 and 30. Two profiles are shown in each figure corresponding to the sorption isotherms for the oxidized and unoxidized alloy steel surf aces. While the surfaces of the FSV primary circuit components have been repeatedly exposed to an oxidizing environment, the oxidation state of the majority of the surfaces is not known. Thus, it was decided to perform the plateort distribution analysis for the two bounding cases. - These cesium

             ' plateout distributions show sorption limited plateout on the hottest metallic components, viz., the steam generator inlet ducts and the superheater and evaporator.            In the rest of the primary circuit, mass transfer control or perfect sink plateout behavior is predicted.                     The integrated plateout activity in each branch is _ tabulated in Table 6.              The differences = between the Cs-137 and Cs-134 relative plateout distributions can be attributed to the effects of the Xe-137 precursor -(Cs-134 has no gaseous precursor) and the differences in half lives (30.1 yr and 2.06 yr,
              . respectively).

A total of 32.9 Curies of Cs-137 and 35.3 Curies of Cs-134 are j' predicted to be plated out in the primary circuit of the reactor at the E005. For the case with the soprtion isotherms for unoxidized steel surfaces, the specific activities (activity per unit area) peak in the steam generator at the inlet of economizer section for both of the cesium isotopes, with the maximum surface activity 0.7 pC1/cm 2. The only L significant difference between the plateout behavior of these two isotopes

is caused by the precursor effect for Cs-137. For the case with the sorption isotherms for oxidized steel, the peak occurs further upstream at the exit - of the reheater as a result of higher sorption for the oxidized cases the maximum surf ace activity of 1.2 pCi/cm 2 was predicted at this location. While only about 10% of the total Cs-137 is produced from precursor l decay, this source' dominates the plateout in the downstream sections of the i primary circuit of the reactor. This can be observed in the circulator

             . outlet plenum and the core inlet plenum, which represent large volume sections.       In these sections the rate of production of Cs-137 in the coolant l               by decay of Xe-137 exceeds the rate of removal by plateout resulting in a L               net increase of the Co-137 concentration in the coolant with increasing l

Page 23

r - . J: 909658/A distance from the cores as a result the plateout activity, which is proportional to the coolant concentration, also increases with distance in these sections.

  .             Iodine is the most volatile of the condensable fission products released from the core, and the I-131 plateout distribution reflects this                ,

high volatility. A total of 1237 Curies of 1-131 and 1.1 x 10-03 Curies of I-129 are predicted to be plated out in the primary circuit of the reactor at the end of cycle 5 operation for the iodine source based upon the reference 0.5-power half-life dependence. The I-131 distributions are shown in Figure 31, for the reference . and the revised source half-life power ! dependence, as discussed in Sect. 2.5 3. The I-129 plateout distribution is shown in Figure 32. These iodine distributions exhibit sorption control, i.e., plateout increasing with decreasing surface temperatures throughout practically the entire primary circuit. The maximum I-131 surface activity i is observed at the outlet end of the steam generator economiser which is the i coldest surface in the grimary circuits the maximum predicted values were ! approximately 200 pCi/ca' and 30 / sci /cm 2 , for the two sources considered. ' Similarly, the maximum surface activity of I-129 is 5.0 x 10-05 pCi/cm2 1, also observed at the same location. The reason for the low activity of i I-129 is its long half life (16 million years). The plateout distribution of Sr-90 is shown in Figure 33. A large ! l fraction of the Sr-90 plateout on the primary circuit surfaces is , l transported .f rom the core via its precursor Kr-90.. The direct metallic 1 release contribution to - Sr-90 plateout is small. Due to lack of data on strontium sorption on metallic surfaces and its high affinity for oxygen, a ! perfect sink assumption was used for strontium in the analysis in this case, the deposition process is controlled by convective mass transfer from the coolant to the surface, .and every strontium atom that strikes the i

        -surface sticks. As a consequence, the plateout continuously decreases along the primary circuit, due to fission product depletion in the coolant.

The only deviation observed is related to the contribution to the St-90 e plateout from the decay of its Kr-90 precursor in the circulating coolant, , l which leads to increasing plateout from the inlet to the outlet in some sections of the primary circuit as discussed above for Xe-Cs-137. These i sections or branches have large coolant volumes associated with small plateout surfaces. Kr-90 decay in these branches increases the coolant concentrations such that increased plateout is observed. Large amounts of Sr-90 are sorbed by the graphite. reflectors. The integrated plateout activities in each branch is given in Table 6. The plateout distribution of Te-127m is shown in Figure 34 and the l integrated activity in each branch is given in Table 6. Te-127m depletion is observed throughout the primary circuit because of the assumption of l' perfect sink plateout. 1 l l [ I- Page 24 l-

1 l i 909658/A l

4. COMPARISON OF PREDICTED AND MEASURED FISSION PRODUCT RELEASE AND l CIRCULATOR PLATEOUT ACTIVITIES J

In order to assess the accuracy of the predictions reported above, a ; comparison was made between the predicted and the available measured fission ' gas release histories of Kr-85m and Xe-138, as shown in Fig. 35 and 36, and between the' predicted and measured plateout in the circulator C2105 as of the July 5,'1988 reactor shutdown, as shown in Fig. 38 through 41. l The comparison between the predicted and measured fission gaf release indicated that the Kr-85m measured data were somewhat overpredicted af ter cycle 2, and that the overprediction was approximately by a f actor of 2 at the last . data point in cycle 4. The agreement between the predicted and measured Xe-138 release was slightly better than for Kr-85m release. These i results indicate that there was still some conservatism in the fission gas release predictions, and that this was due to the overprediction of the total coating f ailure (e.g. , exposed kernels) and, to a lesser degree, to the overprediction of fuel temperatures. Nevertheless, the agreement between the fission gas predictions and the measured data is considered to ! be good because it is well within the factor of 4 margin which is the required predictive accuracy for HTGR fission gas release calculations. Also included in Fig. 35 and 36 are the FSV FSAR " Expected" values which indicate that both the Kr-85m and Xe-138 releases were well below the

       " Expected" values.

The plateout predictions shown in Fig. 38 through 41 show.the values

     ' for both a circulator which has been in service since the BOC1 and the circulator C2105 which has been in service since BOC3 and which remained in service for 450 EFPD's. After the July 5, 1988 reactor shutdown, the C2105 circulator was removed . and radiochemically examined at GA          the measured plateout activities were analyzed and back decayed to the July 5, 1988 shutdown, and the results were reported in Ref. 38.

As can be seen in Fig. 38,-the predicted plateout activities for Cs-134 bracket the data. The value based upon the sorption isotherms for unexidized steel alloys overpredicted the data by a factor of 2.0, while the value based upon the sorption for oxidized' steel underpredicted the data by a factor of 5.6. The Cs-137 data were underpredicted by a factor ranging from 1.4 to 8.9 depending upon the - sorption isotherms for unoxidized or oxidized steel were used in the analysis, as can be sean in Fig. 39. Therefore, better agreement with the cesium plateout data has been obtained when the. sorption isotherms for unoxidized surfaces were used in the l' analysis, as has been previously concluded in case of the CPL 2 inpile loop plateout data (Ref. 39). However, based upon visual observation of the C2105 circulator L (Ref. 38), one might expect that the ferritic surfaces in the primary circuit, including the evaporator / economizer sections of the steam generator, would be oxidized. In that event, the implication would be that the cesium release rates from the core had been underpredicted by about an order of magnitude. The predicted plateout distributions and the circulator C2105 data for I-131 are shown in Fig. 40. As discussed in Sect. 2.5.3, the I-131 source Page 25 L l c .

909658/A was calculated from the Xe-138 R/B daca by two different methods: (1) by

         .using the reference 0.5-power half-life R/B dependence, which is known to be
         . conservative, and (2) by using the observed power dependence based upon the                                                                                 l FSV release data for short-lived nuclides. By using the source calculated:

with the first method, the circulator C2105 data were overpredicted by a factor of 17. This overprediction is due mainly to the conservatism in the i I-131 source calculation, and also to the overprediction of the Xe-138 data by a factor of approximately 2, and, possibly, to the uncertainty in the iodine sorption' data. By using a more realistic source calculation with the second method,

t. _ the circulator data were overpredicted by a factor of 3. It should be emphasized that the above discussion is based upon the circulator C2105 reported plateout data of 387 nanocuries/cm2 (Ref. 38), and that these data

         -appear to be inconsistent with the data for the FSV circulators C2104 and C2101 for which the plateout activities of 1383 and-1360 nanocuries/cm2 have previously been reported (Ref. 38), and that the latter values would be in                                                                                    ,

better agreement with the predictions. This inconsistency in plateout data l ^,' cannot. be explained by the differences in power history during the short l period, - prior to the removal of circulstors, that is significant to the I-131 release. Because no explanation has been given for this inconsistency, and because there is no apparent reason for it, it should be investigated further. Very good agreement has been obtained between the predicted and measured Sr-90 plateout, as can be seen in Fig. 41 where the measured data were overpredicted by a factor of 1.8. This good agreement reflects the good agreement with the data for the main source (the precursor Kr-90 decay). Also, the assumption of perfect sink sorption behavior for strontium on steel appears to be correct. However, it must be emphasized that the Sr-90 plateout data for the circulator C2105 is based on measurements of extremely low levels of activity and that such measurements could be in error, as discussed in Ref. 38. Considering that the required accuracy in HTGR fission metal release calculations is that the predictions be accurate within a factor of 10, and that .there are additional uncertainties 'n i the plateout distribution calculations, the agreement between the predicted and measured plateout activities for the circulator C2105 is, in general, good. The best agreement ~ was obtained for Sr-90 plateout, for which the data were overpredicted by a factor of 1.8. However, the I-131 plateout data were overpredicted by a factor ranging from 3 to 17 depending upon the method used to calculate the I-131 release rate, while the Cs-137 data were underpredicted by a factor ranging from 1.4 to 9 depending upon the sorption L isotherms used in the analysis. The comparison of predicted and measured plateout activities could have been presented in tabular form. The reason for presenting the comparison graphically in Figs. 38 through 41 is to illustrate that the circulators represent a small fraction of the total plateout surface area in the FSV primary circuit. In order to make a definitive judgment regarding the accuracy of the plateout predictions presented in this report, additional data are needed. Ideally, these additional data would include measurements of the plateout activity on the steam generators where the highest specific Page 26

909658/A I activities (pCi/cm2) are predicted to occur. However, given the relative ' inaccessibility.~of _the steam generators within the -PCRV, in situ measurements are impractical. Consequently, one must utilize more indirect data sources, in particular, the plateout probes. , Regarding the cesium plateout, it would be informative to recalculate the ' activities for the plateout probe with che revised fuel performance models used in this ' analysis; the previous analysis, which was based upon

more conservative fuel failure models and upon the release from the maximum releasing part of the core, overpredicted the plateout probe data for.Cs-134 by a factor of a factor of 6.8 (Ref. 31). This tyoe of calculation was beyond the scope of this effort. l As reconsnended in the original issue of this document, the accuracy of the FSV end-of-life plateout predictions could be improved by a systematic L monitoring of the fission gas release rates between the present and l end-of-life, periodic examination of plateout probes, and radiochemical I analysis of components removed from the primary circuit, as has been done in-this analysis.- Measurements should - also be made to assess the ease of l decontamination of components removed from the primary circuit. If decontamination were to prove practical, the ultimate decommissioning of FSV could be greatly facilitated. Recent experience with decontaminating circulator C2105 is encouraging (Ref. 38); however, European experience implies that those components which had higher service temperatures may be the more difficult to decontaminate.- l :. l l-l L l l' l l L l L. Page 27

i - ,

      ,  4 f

909658/A

5. REFERENCES-3

1. ' Warembourg, D. . W., " Decommissioning Support, FSV Plateout Analysis," Public Service Company of Colorado Letter PG-1694, -

                        ' February 4, 1988.

2. .W agner, M. R., " GAUGE," - A Two-Dimensional Few Group Neutron Diffusion-Depletion Program for a Uniform Triangular Mesh," General

                       ' Atomics Report GA-8307, March 15, 1968.                                                                                                  !

L 3. Kraetsch, H. and Wagner, M. R., "GATT, A Three-Dimensional i

                       -Few-Group Neutron Diffusion Theory Program for a Hexagonal-Z Mesh,"

General Atomics Report GA-8547, 1967.

4. Archibald, R., "TSORT - A Computer Program to Process Nuclear Design Data for Use in Core Performance and Mechanical Design

. Calculations with the SURVEY Program " General Atomics Report l'

GA-A14524, June 1977.

5. Lefler, W. L., "TSORT Data for FSV Cycles 1 - 5," General Atomics Internal Correspondence CED:110 WLL:88, March 1, 1988. I

             - 6.        Hudritsch, W. W.,               V. Jovanovic, and D. L. Georghiou, " SURVEY, A Computer-Code for the Thermal and Fuel Performance Analysis of                                                                           !

High-Temperature Gas-Cooled Reactors," General Atomics Report GA-C17554, March 1984, (GA Proprietary Information). l

             ~ 7.        Price,       R. J.,         Graphite Design Data Manual," General Atomics Document 906374/A, September 27, 1984.

8. Myers, B. F., " Fuel Design Data Manual," Issue F, General Atomics Document 901866/F, August, 1987, (GA Proprietary Information).

9. Haire, M. J., and McEachern, D. W., " Gaseous Radioactivity in the Primary Coolant of an HTGR," General Atomics Report GA-A12946, October, 1974.

Page 28

 .j ?

909658/A 10.- Alberstein, D., P. D. Smith, and Haire, M. J., " Metallic Fission Product Release from the . HTGR Core," General Atomics Report GA-A13258, May 15, 1975.

11. Smith, P. D., "TRAFIC, A Computer Program for Calculating ~ the ;

Release of Metallic Fission Products from an HTGR Core," General Atomics Report GA-A14721, February, 1978. f

12. - Goodin, D. T., "FSV Fuel Performance," General Atomics Document 18-R-59/A,-to be issued.

13. Sims, J. R., Smith, C. L.-and Scott, C. B., " Migration of (Th,U)C2 Fuel Kernels," General Atomics Report GA-A13825, May 3, 1976.

14. Smith, C. L. , " Fuel Performance Models for Service Limits H, A, AA,

                'B, and D FSV TRISO (Th,U)C2 Fuel," General Atomics Internal Correspondence CLS:017:PMB:76, April 15, 1976.

115. Montgomery, F. C. and - R. D. Burnette, " FSV Cycle 4 Coolant i Chemistry and Tritium Sorption on Moderator . Graphite," DOE Report HTGR-86-079, September 1986.

16. Burnette, R. D., " Test Status Report on Fort St. Vrain HTGR Coolant and ' . Radiochemistry: July 1983 Through June 1984," General Atomics Document 907650/1, September 28,-1984.

17. Stansfield, O. M., "FSV Cycle 4 R/B Data," General Atomics Internal

                -Correspondence, CED 079:0MS 88, February 19, 1988.

18. Disselhorst, B. F., et al, " Fort St. Vrain Fuel Element Quality Control," General Atomic Report GA-D13772, December 31, 1975, (GA Proprietary Information).

l Page 29 a-

y W 909658/A' D , 19. -Scheffel, W. J., " Technical Support Document for Issue E of the HTGR Fuel Product Specification," General ' Atomics Document 203728/C, September, 1986.

20. . Disselhorst, B. F., " Comparison of Segment 8, 9, and 10 Fuel Rod Attributes," General Atomics Internal Correspondence BDF:025:85, May 6, 1985.

21. Vanslager, F. E., " RAD 2 A Computer Program for Calculating Fission .

Product Radioactivities," General Atomics Report GAMD-6519, July 23, 1965. 22.-. Rider, B. F. and M. E. Meek, " Compilation of Fission Product Yields," General Electric Co. Report No. NEDO-12154-2(E), June 30, 1978.

23. Compere, E. L., Fried, S. H. and Nestor, C. W., " Distribution and Release of Tritium in High-Temperature Gas-Cooled Reactors as a Function of Design, Operation, and Material Parameters," Oak Ridge

..                 National Laboratory Report ORNL-TM-4303, June, 1974.

24. Shamasundar, B. I., "A Two-Dimensional Thermal Analysis of Fort ;

St. Vrain Fuel Blocks," General Atomic Report GA-D14339, March 3,

                  '1977, (GA Proprietary Informatiori.

l.

25. Public Se rvice Company of Colorado, " Fort St. Vrain Nuclear ;

Generating Station FSAR," Section 3.7. l.

26. Hudritsch, W. W., "PADLOC, A One-Dimensional Computer Program for Calculating Coolant and Plateout Fission Product Concentrations,"

General Atomic Report GA-A14401, September 1981. l

27. Strong, D., "Cs & I Plateout in FSV at END OF CYCLE 1," General

[ Atomics Document 904427, Aug 13, 1980. p l 1' l r Page 30 l l I.

y l a.1 909658/A ) 28.; General Atomics FSV: Drawings, 90R1100-100, 90R1100-500,- 90R1100-700, 90R1100-702, 90R1105-701, and 90R1105-505. t' i 1

          -29. Acharya, R., "FSV Decom Support, Calculational File", General Atomics Document 909663, Mar 1988.                                                                                    .

30. ;Jovanovic, V., " Transport - Methods Validation Using FSV Data",

General Atomics Document 909411/0, June 19, 1987.

31. Jovanovic, V., "FSV Fuel Performance Analysis for the First Three Cy'cles," DOE Report HTGR-85-059, June 1905.

32. Warembourg, D. W., " Validation of FSV Plateout Analysis," Public Service Company of Colorado Letter PG-1759, August 2, 1988.

33. Jovanovic', V., " Preliminary Assessment of FSV Fuel Performance for Stretched Cycle 4," General Atomics Internal Correspondence CED 572 VJ 88, October 21, 1988.

3 4. - Scheffel, W. J., " Revised OPyc Failure Models for Fort St. Vrain (FSV)," General Atomics Internal Correspondence CED 461:WJSt88, , August 27, 1988. k

35. Dolphin, M. to V. Malakhof, " Data on R/B," Public Service Company of Colorado Telefax No. FT6-15986, August 22, 1988.

36. Myers, B. F., " Epsilon Value for Xenon Release from Heavy Metal Contamination," General Atomics Internal Correspondence CED:495:BFM 88, September 9, 1988.

3 7 '. Burnette, R., Hightower, G. R. and Montgomery, F. C., " Test Status Report on Fort St. Vrain HTGR Coolant and Radiochemistry: July 1981 through June 1982. Page 31

    , .             ,                                                                                    1 o                                                                                                       I m                                                                                                         j 909658/A
          -38. .Baldwin,;N., " Radiochemical Examination of Fort St. Vrain Helium l

Circulator C2105," . General. Atomics Decument No. 909776,

                 . November 11, 1988.

1

39. Jovanovic, V., " Reanalysis of Cesium Plateout in CPL 2 ' Inpile Loop," General Atomics Document No. 909620/0, February 23, 1988, (GA Proprietary Information).

9 Page 32

F= ;;;[;;il: i:- 909658/A j l 3-4 l Table 1 FSV PREDICTED CORE-AVERAGED FISSION GAS RELEASES AT EOC5 1 l 4 Kr-85m

                  . Total Predicted R/B, Hydrolyzed Fuel                            2.0 x 10-5
                     % R/B Due to failed fertile particles                             57.0
                     % R/B Due to failed fissile particles                             15.9
                     % R/B Due to as-manufactured heavy-metal contamination            27.1

[ Xe-138 Total Predicted R/B, Hydrolyzed Fuel 2.1 x 10-6 l -% R/B Due to failed fertile particles 23.9 l % R/B Due to failed fissile particles 16.1 L- % R/B Due to as-manufactured heavy-metal contamination 60.0 i l Page 33

c - fQ:n ,

  .          4 e,"        ,

57 , 909653/A fi Table 2-FSV PREDICTED Cs-137 RELEASE AT E005

Axial' Layer Release (C1) 1 1.9 x.10-5 2' 6.1 x 10-2

~'

3 3.11. 4 6.00.

                                 '5                                                           10.1 6                                                         10.7 Total predicted direct   release from entire core                30.0 Predicted' release from Xe-137 decay                              2.9 i
                            . Total predicted plateout inventory-                             32.9 i

a-e f Page 34 -G

                   $ y       -*-       v-                 y                                    -

                                                                        ^            "

p;, . <<

             ,,              Lv                   +                                                                                  ,
           ' V_

l 1 e g r_ - '909658/A- ] 1 I

 !L'~

Table 3 l FSV PREDICTED Sr-90 RELEASE AT EOC5 . PLATEOUT INVENTORY (C1)

               -il:                                                                                                                 ;

Predicted' direct release negligible Predicted direct release from as-manufactured heavy-metal contamination on coolant hole surfaces 0.3 Predicted release from Kr-90 decay 2.1

                          ; Total predicted plateout inventory                                            2.4                       ,

k t i) 1 1; l l Page 35 l. l-

                        ~       s                    .    ,     -

7 . _ TABLE 4 FSV PREDICTED CIRCULATING AND PLATEOUT INVENTORIES AT END OF CYCLE 5 Nuclide Halflife Circulat. Pur. Syst. < . ==-----------

                                                                                           ------- Plateout (Curies) ------------------------->

(Curies) (Curies) Initial 10 Day 100 Day 1 Year- 3 Yeare 10 Years 30 years Decay Decay Decay Decay Decay Decay H3 12.3-Y 1.51-01 7.63+02 0.00 0.00 0.00 0.00 0.00 0.00 0.00 C14 5730-Y 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00' O.00 GE79 43.0-S 9.97-03 2.10-05 4.61-02. 0.00 0.00 0.00 0.00 0.00 0.00 AS79 9.0-M 1.35-01 3.58-03 6.29+00 0.00 0.00 0.00 0.00 0.00 'O.00 SE79M 3.89-M- 6.56-02 4.26-03 7.52+00 0.00 0.00 0.00 0.00 0.00 0.00

              *SE79    STABLE      4.94-08        1.80-04      3.14-01             3.14-01  3.14-01    3.14-01                   3.14-01        3.14-01      3.14-01 SE80    STABLE      9.86-08        3.60-04      6.18-01             6.18-01  6.18-01    6.18-01                   6.18-01        6.18-01      6.18-01 SE81    18.5-M      7.43-02        4.05-03      7.03+00             0.00     0.00       0.00                      0.00           0.00         0.G0 BR81    STABLE      1.34-07        4.89-04      8.41-01             8.41-01  8.41-01    8.41-01                   8.41-01        8.41-01      8.41-01 SE82    STABLE      2.42-07        8.81-04      1.51+00            .1.51+00  1.51+00    1.51+00                   1.51+00        1.51+00      1.51+00 SE83M   70.-S       3.99-01        1.37-03      2.75+00             0.00     0.00       0.00                      0.00           0.00         0.00 SE83    22.5-M      9.86-02        6.53     1.13+01             0.00     0.00       0.00                      0.00           0.00         0.00 BRB3    2.4-H       8.46-02        4.35-02      7.53+01             0.00     0.00       0.00                      0.00           0.00         0.00 KR83M   1.86-H      9. 73 +01      1.33+01      0.00                0.00     0.00       0.00                      0.00           0.00         0.00
        ,      KR83    STABLE      7.50-04        2. 69+ 00    0.00                0.00     0.00       0.00-                     0.00           0.00         0.00 f$      SE84    3.3-M       8.48-01        8.23-03      1.50+01             0.00     0.00       0.00                      0.00           0.00         0.00

31.8-M 3.05-01 3.62-02 6.34+01 BR84 0.00 0.00 0.00 0.00 0.00 0.00 g KR84 STABLE 1.31-03 4. 76+00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 BR85 2.87-M 1.26+00 1.06-02 1.95+01 0.00 0.00 0.00 0.00 0.00 0.00 KR8SM 4.48-H 1.16+02 8.32+01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 KR85 10.73Y 1.52-01 4.97+02 0.00 0.00 0.00 0.00 0.00 0.00 0.00 RB85 STABLE 1.68-08 3.83-05 1.05-01 1.05-01 1.05-01 1.05-01 1.05-01 1.05-01 1.05-01 KR86 STABLE 2.42-03 8.82+00 0.00 0.00 0.00 0.00 0.00 0.00 0.00-KR87 76.-M 1.55+02 3.46+01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 RB87 STABLE 3.44-08 1.99-02 2.16-01 2.16-01 2.16-01 2.16-01 2.16-01 2.16-01 2.16-01 KRB8 2.8-H 2.58+02 1.27+02 0.00 0.00 0.00 0.00 0.00 0.00 0.00 RB88 17.7-M 2.87+00 1.27+02 2.60+02 0.00 0.00 0.00 0.00 0.00 0.00 SR88 STABLE 4.98-10 7.39-02 1.52-01 1.52-01 1 52-01 1.52-01 1.52-01 1.52-01 1.52-01' KRS9 3.16-M 6.03+01 5.61-01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 RB89 15.2-M 8.14-01 5.63-01 6.34+01 0.00 0.00 0.00 0.00 0.00 0.00 SR89 50.5-D 1.07-05 5.65-01 6.65+01 5.80+01 1.69+01 4.42-01 1.95-05 0.00 0.00 YB9 STABLE 2.92-10 1.06-06 4.09-02 4.12-02 4.26-02 4.31-02 4.32-02 4.31-02 4.31-02 ,

KR90 32.3-S 2.45+01 3.88-02 0.00 0.00 0.00 0.00 0.00 0.00 0.00 o 3 2

3. -

4

TABLE 4 FSV PREDICTED CIRCULATING AND PLATEOUT INVENTORIES AT END OF CYCLE 5 Nuclide Halflife Circulat. Pur. Syst. <-- --- -

                                                                                                                                          --------- Plateout ( Cu rie s ) ------------- = - - - - - ---->

(Curies) (Curies) Initial 10 Day 100 Day 1 Year 3 Years 10 Years 30 years Decay Decay Decay Decay Decay Decay RB9Gd 4'.28-M 1.60-01 4.79-03 3.63+00 0.00 0.00 0.00 0.00 0.00 0.'00 RB90 2.7-M 1.67+00 3.52-02 2.46+01 0.00 0.00- 0.00 0.00 0.00 0.00 SR90 29.-Y 7.03-08 1.64-04 2.40+00 2.40+00 2.38+00 2.34+00 2.23+00 1.89+00- 1.18+00 Y90 64.-H 1.45-05 2.26-03 ~ 2.68+00 2.42+00 2.39+00 2.34+00 2.23+00 1.89+00 1.18+00 ZR90 STABLE 2.66-11 2.53-06 1.80-03 1.80-03 1.80 1.80-03 - 1.80-03 1.80-03 1.80-03 KR91 9.0-S 7.76+00 3.43-03 0.00 0.00 0.00 0.00 0.00 0.00 0.00 RB91 58.5-S 1.85+00 4.98-03 1.10+01 0.00 0.00 0.00 0.00 0.00 0.00 SR91 9.48-H 1.87-03 7.03-03 1.45+01 3.48-07 0.00 0.00 0.00. 0.00 - 0.00 Y91 58.6-D 8.26-06 2.05-03 1.80+01 1.61+01 5.55+00 2.41-01 4.26-05 0.00 0.00 ZR91 STABLE 3.37-10 2,38-06 1.29-02 1.30-02 1.34-02 1.36-02 1.37-02 1.37-02 1.37-02 SR92 2.71-H 4.28-03 2.05-03 3.52+00 -0.00 0.00 0.00 0.00 0.00 0.00 Y92 3.53-H 3.32-03 4.11-03 7.07+00 0.00 0.00 0.00 0.00 0.00 0.00 ZR92 STABLE 3.44-10 3.75-06 6.44-03 6.t;4-03 6.44-03 6.44-03 6.44-03 6.44-03 6.44-03 SR93 7.5-M 9.43-02 2.08-03 3.67+00 0.00 0.00 0.00 0.00 0.00 0.00 m Y93 10.2-H 1.25-03 4.29-03 7.46+00 6.18-07 0.00 0.00 0.00 0.00 0.00 i E ZR93 STABLE 3.71-10 3.98-06 6.90-03 6.90-03 6.90-03 6.90-03 6.90-03 6.90-03 6.90-03 SR94 1.29-M 4.53-01 1.72-03 3.41+00 0.00 0.00 0.00 0.00 0.00 0.00 > U Y94 19.0-M 4.33-02 3.88 03 7.15+00 0.00 0.00 0.00 0.00 0.00 0.00 ZR94 STABLE 3.77-10 3.76-06 6.77-03 6.77-03 6.77-03 6.77-03 6.77-03 6.71-03 6.77-03 Y95 10.5-M 6.58-02 2.03-03 3.56+00 0.00 0.00 0.00 0.00 0.00 0.00 ZR95 65.5-D 7.77-06 4.15-03 7.20+00 6.48+00 2.50+00 1.51-01 6.62-05 0.00 0.00 ? NB95H 3.61-D 1.38-06 6.27-05 1.08-01 7.33-02 2.64-02 1.60-03 7.00-07 0.00 0.00 NB95 35.1-D 1.42-05 2.12-03 1.09+01 1.01+01 4.71+00 3.18-01 1.41-04 0.00 0.00 , MD95 STABLE 3.63-10 2.60-06 9.09-03 9.11-03 9.25-03 9.36-03 9.37-03 9.37-03 9.36-03 Y96 6.0-S 2.10+00 6.18-04 3.16+00 0.00 0.00 0.00 0.00 0.00 0.00 ZR96 STABLE 5.60-10 1.66-06 4.18-03 4.18-03 4.18-03 4.18-03 4.18-03 4.18-03 4.18-03 ZR97 16.8-H 6.41-04 1.90-03 3.27+00 1.64-04 0.00 0.00 0.00 0.00 0.00 NB97 73.6-M 8.81-03 3.81-03 6.55+00 1.77-04 0.00 0.00 0.00 0.00 0.00 HD97 STABLE 3.36-10 3.66-06 6.29-03 6.29-03 6.29-03 6.29-03 6.29-03 6.29-03 6.29-03 i NB98 51.0-M 1.19-04 1.78-05 3.07-02 0.00 0.00 0.00 ' O.00 0.00 0.00 H098 STABLE 3.26-10 1.20-06 2.06-03 2.06-03 2.06-03 2.06-03 2.06-03 2.06-03 2.06-03 NB99M 2.5-M 8.22-02 6.05-04 1.12+00 0.00 0.00 0.00 0.00 0.00 0.00 . NB99 14.0-S 9.45-01 6.49-04 2.06+00 0.00 0.00 0.00 0.00 0.00 0.00 @ 0 4

                                                                                                                                                             ,,-m~

yr , r -- m _ e

, = - - - - g += a

'[

TABLE 4 FSV PREDICTED CIRCULATING AND PLATEOUT INVDf70 RIES AT DID OF CYCLE 5 Nuclide Halflife Circulat. Pur. Syst. <- - -----

                                                                 -------- Plateout (Curies) ----= --                          -

(Curies) (Curles) Initial 10 Day 100 Day 1 Year 3 Years 10 Years .30 years Decay Decay Decay Decay. Decay . Decay HD99 66.02H 2.11-04 3.12-03 6.39+00 5.15-01 0.00 0.00 0.00 0.00 0.00 TC99M 6.02-H 1.55-03 1.64-03 8.45+00 4.98-01 0.00 0.00 0.00 0.00 0.00 TC99 STABLE .3.34-10 2.29-06 8.11-03 8.11-03 8.11-03 8.11-03 8.11-03 8.11-03 8.11-03 , NB100H 7.0-S 9.71-01 3.33-04 1.54+00 0.00 0.00 0.00 0.00 0.00 0.00 NB100 2.9-M 9.86-02 G.41-04 1.54+00 0.00 0.00 0.00 0.00 0.00 0.00 MD100 STABLE 3.40-10 1.76-06 3.08-03 3.08-03 3.08-03 3.08-03 3.08-03 3.08-03 3.08-03 MD101 14.6-M 3.31-02 '1.42-03 2.47+00 0.00 0.00 0.00 0.00. 0.00 0.00 TC101 14.2-M 3.44-02 2.84-03 4.95+00 0.00 0.00 0.00 0.00 0.00 0.00 EU101 STABLE 2.66-10 2.85-06 4.94-03 4.94-03 4.94-03 4.94-03 4.94-03 4.94-03 4.94-03 MD102 11.1-M 3.52-02 1.15-03 2.01+00 0.00 0.00 0.00 0.00 0.00 0.00 TC102M 4.3-M 9.00-02 2.27-03 4.02+00 0.00 0.00 0.00 0.00 0.00 0.00 EU102 STABLE 2.26-10 2.31-06 4.06-03 4.06-03 4.06-03 4.06-03 4.06-03 4.06-03 4.06-03 HD103 60.-S 2.31-01 6.81-04 1.40+00 0.00 0.00 0.00 0.00 0.00 0.00 EU103 39.6-D 5.78-06 1.52-03 2.84+00 2.38+00 4.93-01 4.75 03 1.33-08 0.00 0.00 m RH103M 56.-M 5.00-03 2.33-03 4.23+00 2.36+00 4.88-01 4.70-03 1.32-08 0.00 0.00

   $*  RH103   STABLE   1.56-10    5.67-07    3.87-03   3.87-03    3.87-03     3=87-03     3.87-03      3.87-03                        3.87-03 HD104   1.6-M    9.27-02    4.37-04    8.43-01   0.00       0.00        0.00       .0.00         0.00                           0.00
    $  TC104   18.-M    1.06-02    9.44-04    1.72+00   0.00       0.00        0.00        0.00         0.00                           0.00 RU104   STABLE   9.75-11    9.98-07    1.79-03   1.79-03    1.79-03     1.79-03     1.79-03      1.79-03                        1.79-03 TC105   8.0-H    1.11-02    2.61-04    4.59-01   0.00       0.00        0.00        0.00         0.00                           0.00 RU105   4.44-H   3.51-04    5.29-04    9.21-01   0.00       0.00        0.00        0.00         0.00                           0.00 RH105   35.5-H   4.29-05    7.98-04    1.38+00   1.40-02    0.00        0.00        0.00         0.00                           0.00 PD105   STABLE   5.10-11    1.86-07    1.28-03   1.28-03    1.28-03     1.28-03     1.28-03      1.28-03                        1.28-03 RU106   369.-D   8.08-08    1.14-04    1.96-01   1.92-01    1.63-01     9.87-02     2.50-02     '2.06-04                        0.00 PD106   STABLE   2.19-11    1.42-07    2.74-04   2.75-04    2.84-04     3.03-04     3.25-04      3.33-04                        3.33-04 EU107   4.2-M    4.15-03    5.13-05    9.24-02   0.00       0.00        0.00        0.00         0.00                           0.00 RH107   21.7-M   8.74-04    1.05-04    1.85-01   0.00       0.00        0.00        0.00         0.00                           0.00 PD107   STABLE   1.05-11    1.12-07    1.96-04   1.96-04    1.96-04     1.96-04     1.96-04      1.96                       1.96-04 RU108   4.5-M    2.02   2.68-05    4.81-02   0.00       0.00        0.00        0.00         0.00                           0.00 PD108   STABLE   5.64-12    2.05-08    3.53-05   3.53-05    3.53-05     3.53-05     3.53-05      3.53-05                        3.53-05 RH109   1.5-M    3.63-03    1.60-05    3.12-02   0.00       0.00        0.00        0.00         0.00                          0.00 PD109H  4.69-M   7.07-04    1.68-05    3.13-02   0.00       0.00        0.00        0.00         0.00                          0.00                       e PD109   13.46H   8.28-06    4.30-05    7.82-02   3.38-07    0.00        0.00        0.00         0.00                          0.00                       3 0
                         ,-            ,                      -      -                           . ,       _ _ _ ~ _ _ _ _ _ _ _ - _ _ _ _ _ . _ - - . _ - -

                                                                                                                                                                                 =                            -

1 m I TAB E 4 FSV PREDICTED CIRCUIATING AND luTEOUT INVENTORIES AT END OF CYCE 5 Nuclide Halflife Circulat. Pu r. Sy s t . <--- = - - - ---=--

                                                                                                                                                   ------ Plateout (Curies) ------------------------->

(Curies) (Curies) Initial 10 Day 100 Day 1 Year 3 Years 10 Years 30 years Decay Decay Decay Decay Decay Decay AG109 STABLE 3.35-08 1.22-04 2.10-01 2.10-01 2.10-01 2.10-01 2.10-01 2.10-01 2.10-01 PD110 STABLE 2.29-12 8.34-09 1.43-05 -1.43-05 1.43-05 1.43-05 1.43-05 1.43-05 1.43-05 AG110M 252.-D 2.97-06. 3.06-03 5.27+00 5.12+00 4.00+00 1.93+00 2.58-01 2.23-04 0.00 RH111 63.-S 1.91-03 5.90-06 1.20-02 0.00 0.00 0.00 0.00 0.00 0.00 PD111 22.-M 1.31-04 1.33-05 2.48-02 0.00 0.00 0.00 0.00 0.00 0.00 AG111M 74.-S 1.79-03 1.96-05 3.75-02 0.00 0.00 0.00 0.00 0.00 0.00 AG111 7.47-D 2.22-03 7.03-02 1.21+02 4.78+01 1.13-02 0.00 0.00 0.00 0.00 CD111 STABLE 1.86-10 5.16-05 8.95-02 9.00-02 9.03-02 9.03-02 9.03-02 9.03-02 9.03-02 PD112 20.1-H 1.55-06 5.51-06 9.47-03 2.41-06 0.00 0.00 0.00 0.00 0.00 AG112 3.13-H 9.96-06 1.10-05 1.89-02 2.85-06 0.00 0.00 0.00 0.00 0.00 PD113 1.5-M cl.04-03 4.61-0C 8.96.03 0.00 0.00 0.00 0.00 0.00 0.00 AG113 5.3-H 5.60-06 8.84-06 1.61-02 0.00 0.00 0.00 0.00 0.00 0.00 CDll3 STABLE 1.29-10 4.78-07 8.23-04 8.23-04 8.23-04 8.23-04 8.23-04 8.23-04 8.23-04 SN119M 245.-D 5.91-11 5.95-08 1.02-04 9.94-05 7.71-05 3.64-05 4.60-06 3.32-09 0.00 3 SN119 STABLE 8.97-13 3.31-09 5.71-06 5.71-06 5.71-06 5.72-06 5.73-06 5.73-06 5.73-06 E SN123 129.-D 1.89-09 1.03-06 1.77-03 1.68-03 1.03-03 2.49-04 4.91-06 5.30-12 0.00

SB123 STABLE 1.31-12 5.49-09 9.66-06' 9.67-06 9.75-06 9.84-06 9.87-06 9.87-06 9.87-06 y SN125 9.65-D 1.92-07 7.86-06 1.35-02 6.58-03 1.03-05 0.00 0.00 0.00 0.00 SB125 2.73-Y 5.05-09 1.68-05 2.90-02 2.89-02 2.72-02 2.26-02 1.36-02 2.30-03 1.43-05 I TE125M 58.-D 1.80-05 7.66-03 7.62+00 6.76+00 2.31+00 1.02-01 3.26-03 5.49-04 3.41-06 i

TE125 STABLE 3.26-08 2.06-04 2.10-01 2.11-01 2.11-01 2.11-01 2.11-01 -2.11-01 2.11-01 l SN126 STABLE 1.01-11 3.69-08 6.34-05 6.34-05 6.34-05 6.34-05 6.34-05 6.34-05 6.34-05 l SB126M 19.0-M 7.82-04 4.44-05 7.59-02 8.19-07 8.19-07 8.19-07 8.19-07 8.19-07 8.19-07 SN127M 4.4-M 1.39-03 1.80-05 3.23-02 0.00 0.00 0.00 0.00 0.00 0.00 SN127 2.12-H 2.05-04 7.67-05 1.32-01 0.00 0.00 0.00 0.00 0.00 0.00 SB127 3.8-D 6.42-06 1.97-04 3.40-01 5.54-02 4.12-09 0.00 0.00 0.00 0.00 TE127M 109.-D 7.76-05 3.58-02 6.17+01- 5.79+01 3.26+01 6.04+00 5.81-02 0.00 0.00 TE127 9.4-H 8.13-03 4.85-02 8.37401 5.67+01 3.20+01 5.92+00 5.69-02 0.00 0.00 1 127 STABLE 1.36-05 4.96-02 1.80+00 1.80+00 1.80+00 1.80+00 1.80+00 1.80+00 1.80+00 SN128 59.-M 8.66-04 1.50-04 2.59-01 0.00 0.00 0.00 0.00 0.00 0.00 SB128M 10.4-M 5.16-03 3.07-04 5.34-01 0.00 0.00 0.00 0.00 0.00 0.00 l SB128 9.0-H 7.65-06 1.25-05 2.14-02 2.02-10 0.00 0.00 0.00 0.00 0.00 . j TE128 STABLE 4.69-07 1.71-03 2.94+00 2.94+00 2.94+00 2.94+00 2.94+00 2.94+00 2.94+00 g

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                                                                                -- ---- Plateout ( Cu rie s ) ------------------------- > '

(Curles) (Curles) . Initial 10 Day 100 Day 1 Year 3 Years 10 Years 30 years Decay- Decay Decay Decay Decay Decay

,      E134    STABE    1.17-02     4.25+01     0.00               0.00          0.00         0.00       0.00         0.00          0.00.

CS134 2.06-Y 9.45-06 2.05-02 3.53+01- 3.49+01 3.22+01- 2.52+01 1.28+01 1.22+00- 1.45-03 I 135 6.585H 8.75+00 1.02+01 3.63+02 0.00 0.00 0.00 0.00 0.00 0.00 E135M 15.3-M 6.65+01 2.19+00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 E135 9.17-H 2.27+02 1.60+02 0.00 0.00 0.00 0.00 'O.00 0.00 0.00' CS135 STABE 9.64-08 2.34-04' 6.04-01 6.04-01 6.04-01 6.04-01 6.04-01 6.04-01 6.04-01 1 136 85.-S 9.12+00 3.80-02 1.04+01 0.00 0.00 0.00 0.00 0.00 0.00 XE136 STABE 1.09-02 3.96+01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.17-04 6.46-03 CS136 13.0-D 1.11+01 6.51+00 5.37-02 3.87-08 0.00 0.00 0.00 XE137 3.84-M 3.7!r+01 4.23-01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 CS137 30.1-Y 5.45-06 4.93-02 3.28+01 3.28+01 3.26+01 3.21+01 3.06+01 2.61+01 1.64+01 BA137M 2.55-M 1.91-02 4.85-02 3.13+01 3.10+01 3.08+01 3.03+01 2.90+01 2.47+01 1.55+01 BA137 STABE 4.33-11 1.95-06 3.01-02 3.01-02 3.02-02 3.06-02 3.15-02 3.43-02 4.03-02 XE138 14.2-M 6.57+01 2.74+00 0.00 0.00 0.00 0.00 0.00 0.00 0.00

   ,   CS138M  2.9-M    1.23-02     1.05-04     1.93-01            0.00          0.00         0.00       0.00         0.00          0.00
  $'o  CS138   32.2-M   4.22-01     2.75+00     6.93+01            0.00          0.00         0.00       0.00         0.00          0.00 BA138   STABE    5.94-10     1.94-06     6.65-02            6.65-02       6.65-02      6.65-02    6.65-02      6.65-02       6.65-02 g   XE139   39.7-S   1.11+01     2.16-02     0.00.              0.00          0.00         0.00       0.00         0.00          0.00 CS139   9.3-M    3.03-01     2.36-02     1.46+01            0.00          0.00         0.00       0.00         0.00          0.00 BA139   83.3M    9.33-03     2.57-02   .1.82+01             0.00          0.00         0.00       0.00         0.00          0.00 IA139   STABE    5.33-10     1.94-06     2.00-02            2.00-02       2.00-02      2.00-02    2.00-02      2.00-02       2.00-02 XE140   13.6-S   4.25+00     2.84-03     0.00               0.00          0.00         0.00       0.00         0.00          0.00 CS140   63.8-S   1.14+00     4.32-03     7.28+00            0.00          0.00         0.00       0.00         0.00          0.00 BA140   12.79D   5.13-05     6.44-03     1.09+01            6.35+00       4.83-02      2.76-08    0.00         0.00          0.00 1A140   40.23H   2.98-04     2.12-03     1.46+01            7.34+00       5.56-02      3.17-08    0.00         0.00          0.00 CE140   ST/2LE   5.36-10     3.91-06     1.68-02            1.68-02       1.68-02      1.68-02    1.68-02     '1.68-02      1.68-02 BA141   18.3M    3.68-02     1.98-03     3.45+00            0.00          0.00         0.00       0.00         0.00         0.00

1A141 3.87-H 3.00-03 4.01-03 6.94+00 0.00 0.00 0.00 _0.00 0.00 0.00 CE141 32.53D 1.47-05 6.04-03 1.04+01 8.45+00 1.24+00 4.36-03 0.00 0.00 0.00
PR141 STABE 5.17-10 1.89-06 1.29-02 1.30-02 1.33-02 1.33-02 1.33-02 1.33-02 1.33-02

! BA142 10.7-M 6.15-02 1.94-03 .3.39+00 0.00 0.00 0.00 0.00 0.00 0.00 IA142 92.4-M '7.69-03 3.99-03 6.93+00 0.00 0.00 0.00 0.00 0.00 0.00 , 1 STABE 5.30-10 5.67-06 9.80-03 CE142 9.80 03 9.80-03 .9.80-03 9.80-03 9.80-03 9.80-03 o a a e ~ , < -

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TABE 4 FSV PREDICTED CIRCULATING AND PIATEDUT IWENTORIES AT EMD OF CYCE 5 Nuclide Halflife Circulat. Pur. Syst. <- ------ -- -

                                                                                           - Plateout (Curies) ----               -    --

(Curies) (Curies) Initial 10. Day 100 Day 1 Year 3 Years 10 Years 30 years - Decay Decay Decay Decay Decay Decay SM152 STABE 2.38-11 '2.36-07 4.14-04 4.14-04 4.14-04 4.14-04 4.14-04 4.14 4.14-04 EU152 13.-Y 3.96-12 1.32-08 2.27-05 2.27-05 2.24-05 2.15-05 1.93-05 1.33-05 4.58-06 ND153 67.5-S 9.85-03. 3.26-05 6.59-02 0.00 0.00 0.00 0.00 0.00 0.00 PM153 5.4-M 3.16-03 7.73-05' 1.46-01 'O.00 -0.00 0.00 0.00 -0.00 0.00 91153 45.5-H 5.91-06 1.24-04 2.26-01 6.31-03 0.00 0.00 0.00 0.00 0.00 EU153 STABE 1.29-11 4.70-08 3.08-04 3.09-04 3.09 3.09-04 3.09-04 3.09-04 3.09-04 ND154 7.73-D 4.20-07 1.38-05 2.36-02 9.65-03 3.01-06 0.00. 0.00 0.00 0.00 PM154 2.8-M 1.98-03 3.00-05 '5.36-02 9.65-03 3.01-06 0.00 0.00 -0.00 0.00 91154 STABE 6.28-12 5.21-08 9.18-05 9.18-05 9.18-05 9.18-05 9.18-05 9.18-05 9.18-05 EII154 8.6-Y 1.68-09 5.37-06 9.22-03 9.20-03 9.02-03 8.51-03 7.24-03 4.12-03 8.20-04 91155 22.2-M 1.47-04 9.59-06 1.66-02 0.00 0.00 0.00 - 0.00 0.00 0.00 EU155 4.8-Y 6.20-10 5.47-06 9.47-03 9.43-03 9.10-03 8.19-03 6.14-03 2.23-03 1.24-04 GD155 STABE 1.93-12 1.29-08. 2.17-05 2.18 2.25-05 2.44-05 2.86-05 3.67-05 4.10-05 91156 9.4-H 2.67-06 4.43-06 7.61-03 1.57-10 0.00 0.00 0.00 0.00 0.00 3 E1J156 15.2-D 6.94-08 8.90-06 1.53-02 9.82-03 1.62-04 9.05-10 0.00 0.00 0.00-E 'GD156 STABE 1.26-12 1.36-08 2.36 2.37-05 2.39-05 2.39-05 2.39-05 2.39-05 2.39-05 '

91157 83.-S 4.76-04 1.94-06 3.81-03 0.00 0.00 0.00 0.00 0.00 0.00

 $       EU157      15.2-H     9.62-07     4.24-06   7.77-03.         1.37-07        0.00                0.00      0.00         0.00      0.00 GD157      STABE      5.58-13     5.78-09   1.04-05          1.04-05        1.04-05             1.04-05   1.04-05      1.04-05   1.04-05 TOTALS     1.33+03     3.70+03   6.89+03          1.36+03        2.39+02             1.12+02   7.90+01      5.59+01   3.43+01-i                                                                                                                                                               ,

. Stable nuclides are given in grams. I

~ R o cr 77556 l5s l _ 6 7 7 7 6 56 6 67 74 _ 000C0 0 0000 00 0 0000 _ A + + + + + + + + + + + + + + + + + 51 550 1 2446' 27 2 1 1 29 _ h 35838 0 2883 94 g 26288 9 1 3567 577 3 83 nm ac 92227 1 2447 46 7 2 4650 1 277 46847 8 8557 27 7 0616 r _ B 61 752 6 1 1 1 1 51 5 U 51 1 8 l o 76677 6 6667 68 8 9663 V 00000 0' 0000 h

                       + + + + +              +       + + + ~     + 00 +      +

0

0000

                                                                                                + + + +                         _

30709 3 4662 89 0 3080 c 1 7 969 3 1 441 81 7 6767 n 0347 7 3 3007 1 6 9 4337 g 46532 9' 9441 _ rm 53731 96 3 8636 _ 1 21 1 5 91 1 1 321 B c . . . . . . 26368 1 5446 24 1 1 622 _ s l e n 0 0 2' 002 1 2 O G 0002 61 2 1 070 $ n 1 0001 1 2 038 882 6 522 833 a 77 5 622 761 h 22' 4 233 2 C s _ e d 00555 1 1 1 0 0005 75 0 5000 . 1 1 1 1 1 1 1 o N - s v r e 002* 4 1 01 1 4 25 3 501 3 t 00000 0 0000 e 00 0 0000 w

                  )    + + + + +               -      + - - +             + +           +       + + - -

m 2 d 22973 5 3227 1 0 r a mn cE 66281 4 8001 88 0 4 0267 8649 P ( 33525 8 6227 23 6 3363 2291 7 4 5999 81 7 1 235 9 9 0 8 0996 00 9 0958 l n 8.~ e o 1 1 211 3 1 881 47 4 71 47 d i o t M c 00254 1 01 1 4 25 3 501 3 e 00000 0 0000 00 0 0000

      -   5 t_   S     + + + ++               -    i - - +                + +          +       + + - -

e uo sn 22973 5 3227 1 0 0 0267 l 66281 4 8001 88 4 864 9 b e si 33525 8 6227 23 6 3363 a t og 2291 7 4 5999 81 7 r e 1 235 Tal CB 99870 3 0996 00 9 0958 P 1 1 21 1 0 1 881 47 4 71 47 n i a ) r m 55000 0 0000 00 0 V cd 66000 0 0500 ( n 677 0 00 0 0660 t 55200 7 1 000 80 6 057 1 S .E 1 1 977 1 1 1 7 25 9 51 m 1 61 4 24 7 4 t a 41 1 9 9 r i o D F 55000 0 0000 00 0 0500

                   .n ri 66000                0         677 0               00           0       0660 d g 55200                7         1 000               80           6       0571 ye  1 1 977                        1 1 1 7             25           9       51 HB           1 61                               4       24           7       4

_4 1 1 9 9

                 )               .

m 221 22 1 2222 22 3 3222 c 00000 0 0000 00 0 0000 ( + + + + + + + + + + + + + + + + + 00000 0 0000 00 0 0000 h 00000 0 0000 00 0 0000 t g 00200 0 0000 00 1 9000 59400 0 9330 84 4 8970 n 71 873 8 8442 59 0 61 7 0 e . L 41 536 6 1 1 1 3 45 1 1 1 72 s m m k u e c n t r m t om r e e e us s rl ur rl l n nrsy oB no or oP t i eorS t et t e t u L sl t o ctl a at r es Om/uP ct n at eerP rt rar ull eco rl o eeeeerrel rrneutl ei of pt nl nht oent ooernefl t C e pi eneeat zeut t l rnl ef a h e R SE uxGIGR eaiGOaaP h rm aAnR ll ec B I Ri c vr m m room uu r f ne ieeea a epna cc e eeei am ra t wrr e e paoe r r r r pd r cooot t uvct i i o o pi u BN ALCCS S SEES CC C CUSP l' i ,' f < ; ; ,

TABLE 6 Fort St. Vrain Integrated Plateout in Each Primary Circuit Component at EOC5 Branch Cs-134' Cs-137 I-131 I-129 Sr-90 Te-127m Name (Curies) (Curles) (Curies) (Curies) (Curies) (Curies) Active Core 0.000E+00- 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Imer Reflector 1.510E+01 1.326E+01 6.077E-03 1.037E-08 5.820E-01 3.196E+01 Core Support Blocks 3.668E-01 3.224E-01 2.923E-04 4.987E-10 1.028E-02 5.664E-01 Core Exit Plenum 8.166E-02 7.195E-02 2.028E-04 3.461E-10 1.937E-03 1.000E-01 Steam Generator Inlet 1.617E-03 1.443E-03 9.125E-03 1.557E-08 1.770E-02 8.049E-01 Steam Generator Reheater 1.029E+00 8.771E-01 6.272E-01 1.066E-06 4.527E-01 1.923E+01 Superheater , 2.699E+00 2.305E+00 1.478E+00 2.516E-06 1.544E-01 6.909E+00 Econcznirer 7.513E+00 6.339E+00 2.78BE+01 4.341E-05 2.244E-02 9.067E-01 Evaporator 7.398E+00 6.879E+00 1.120E+03 4.868E-04 8.131E-03 1.696E-01 Steam Generator Outlet Plenian 2.277E-02 2.272E-02 1.164E+01 2.531E-05 6.615E-04 4.924E-04 ! , Circulators 1.458E-01 1.388E-01 4.432E+00 7.501E-06 8.754E-03 3.305E-03 u Circulator Outlet Plenum 5.385E-03 6.882E-03 6.929E+00 2.507E-05 1.274E-03 1.161E-04 Core Barrel / Liner Annulus 1.196E-01 1.886E-01 4.172E+0) 8.176E-05 '5.435E-02 2.579E-03 Core Inlet Plenum 1.410E-02 3.302E-02 1.943E+01 7.128E-05 1.315E-02 3.041E-04 Upper Reflectors 8.109E-01 2.448E+00 2.702E-01 6.322E-07 1.069E+00 1.752E-02 Side Reflectors 1.053E-03 5.767E-03 1.761E-01 5.441E-07 1.838E-03 2.210E-05 Purification System 1.106E-03 2.902E-03 2.853E+00 3.542E-04 1.293E-03' 2.404E-05 TOTAL 3.530E+01 3.290E+01 1.237E+03 1.100E-03 2.400E+00 6.067E+01 e Based upon the scurce rate calculated from the xenon data using the square root of half-life dependence. ,o t ** 8 Plateout distribution based upon sorption isotherms for unoxidized alloy steel surfaces. h 5

q; m 1

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                                                                                         .                               8. S.G. Economizer                       .                                                         :

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10 ' 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 BRANCHES _ _ ~ . . . _ , . . .. . . . . _ . . . , . . _ . _ . . - . . . . - . _ _ _ _ _ _ _ _ _ . ___._m -

t Figure 30 Cs-137 PLATEOUT PROFILES AT EOC5

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(gWO/!D) N011V81N3DN00 30V.380S I l l l Page 76 1 l

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. Rgure 32 1-129 PLATEOUT PROFILES AT EOC5

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Figure 34 Te-127m PLATEOUT PRORLES AT EOC5

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Figure 35. . COMPARISON OF FSV PREDICTED AND MEASURED Kr-85m RELEASE-i 10 * . .

                                                                         .-           +

l CYCLE 1 . = = CYCLE 2 h CYCLE 3 : : : CYCLE 4' =d. CYCLE 5-* i : ~ i . . i : : : ! HFSV FSAR " EXPECTED" VALUEl! . . i . ? : . q-M m - - r k E l0-5 7 Ig ) : . I- . A ! g Q - li J r 31 's F

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R/B MARCH 8, 1983 DATA (REF. 35) 10-7 . d MAY 6, 1988 DATA (REF. 35) O JUNE 6, 1988 DATA (REF. 35) l 7 JULY 5, 1988 DATA (REF. 35) !

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 'E                                                                         Q CORRECTED DATA (0YCLE 1) i 10-*    -

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Sr-90 PREDICTED PLATEOUT PROFILES AND C2105 DATA '

AS.OF JULY 5,1988

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                                                                -i 1

i i 909658/A~ l 1 t:. 1 V 1

i1 1

1 l APPENDIX A TSORT DATA FOR FSV CYCLES 1 - 5 By W. L. Lefier t 4 i i s h

.5

( ', i 1 l Page 87 I: l- [' . L l)'. . - , ....

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                        ? INTERNAL CORRESPONDENCE                                                                                  '

G A 1076 909658/A-PROJECiT 1900' -

                                          .                                                              IN REPLY                  >

REFER TO-CED110iWLLi8s , DATE: 3/01/6s

                        ~ rROui             W. L.' Lefier TO -              D. L. ,Hanson susJacTi          TSORT Data For FSV Cycles 1 - 5 i
      ?3                  .   ' Power and flux tapes for FSV Cycles 1-5 have been produced with the TSORT (Ref.

y 1) computer program.~ These tapes will be used to carry out fuel performance calculations. ; The time history of power, fluence, and FIMA distributions used in TSORT for Cycles , 11-3 an'd Cycle 4 to 24.9 EFPD were taken from the fuel accountability GAUGE (Ref. 2) , calculations. Due to the large number of time points in these GAUGE calculations, fuel l performance calculations for all the time points would be unnecessarily expensive. Time , points with reactor power and control rod patterns (power distribution) roughly the same ' g ' as the preceding poirit(s) and those with very low reactor powers were eliminated. l Burnup calculations'with the GAUGE code were done for Cycle 4 from 24.9 to 292 i EFPD and for Cycle 5 from 0 to 292 EFPD, with a 60 day shutdown between cycles, during

                      . which segment 10 was loaded. These calculations were done at 80% power, as instructed by_

L PSC. c The final condensed TSORT history is shown in Tables 1 - 5 for Cycles 1 - 5, respectively. Tables 1,2, and 3 are reprinted from References 3,4, and 5, respectively. The first 24.9 EFPD of Table 4 is taken from Reference 6. The power / flow ratios and core inlet temperatures i

shown in these tables were obtained from the FSV Data Logger tapes for' Cycles 1 - 3 and' '

D the first 24.9 EFPD of Cycle 4. For the remainder of Cycle 4 and for Cycle 5, the power /fiow : ratios were best estimates provided by S. Mu5oz for 80% power conditions. l 'Both the-GAUGE and GATT (Ref. 7) computer programs are used to perform fuel !Y accountability calculations for FSV operation. The " official" fuel accountability data which

                        -are reported to the NRC are from the G,ATT results. In the first few cycles the agreement p                         between the GATT and GAUGE fuel accountability calculations was good, and no adjust.

L ments were needed. However, it was desirable.at the beginning of Cycle 4 to adjust the ? - GAUGE fuel loadings to the GATT fuel loadings, thus ensuring that the GAUGE model , L" would start the beginning of each cycle with the " official" fuel loadings. The same procedure will be followed in future cycles. However, since there are no GATT core loadings until the completion of each cycle, the loadings at the beginning of Cycle 5 are from the end of the Cycle 4 GAUGE depletion, except for the refueled segment. 1 Page 88 i

                                                                . - ~   .                             - --

909658/A ] CED:120:WLL:57 l l Two problems result from normalizing the GAUGE fuel loadings to the GATT fuel loadings. The first problem is that the loadings for the Fuel Test Elements (FTEs) are modeled explicitly in the G ATT model and are not accounted for in the pre-Cycle 4 G AUGE model. The result of this difference is that incorrect (negative) FIMAs are calculated for the columns which the FTEs are in by the TSORT program. The initial GAUGE fuel loadings do not account for the FTEs while the final GAUGE fuel loadings do account for FTEs, To correct this problem, the FIMAs for the columns with FTEs in them were set to the FIMA of the central column of the patch (region) which the FTE column is in. It should be noted that the FIMA for FTE columns was not strictly correct in previous, pre-Cycle 4, TSORT calculations since the FTE loadings were not included. The second problem is that, for some columns, there are small inconsistencies between the initial GAUGE and GATT fuel loadings. The result is that the FIMA for some columns at the beginning of Cycle 4, after the GAUGE fuel loadings have been adjusted to the GATT fuelloadings,is not the same as the FIMA at the end of Cycle 3. These differences are expected to become smaller with continuing depletion. The TSORT results tapes contain no buffered column data. The TSORT calculations were done with the six control columns on the core / reflector boundary defined as fully-

   . buffered columns. All other columns were defined as standard columns. The results tapes were then modified to remove all references to buffered data. TSORT gives erroneous results when all columns are defined as standard columns. This problem should be corrected.

All references to buffered data were removed because this is consistent with the modeling of buffer columns in the GAUGE model, which cannot properly model partially-buffered columns. Since G AUGE cannot properly model partially-buffered columns, SURVEY (Ref.

8) cannot properly calculate FIMAs for partially-buffered columns.

The axial power profiles used in TSORT were titose derived for Cycle 3 from GATT results (Ref. 5). These axial power profiles were used for all cycles because TSORT only allows one set of profiles. This limitation is conservative because the axial power distribution tends to flatten with burnup, i.e., the use of a flattened distribution should result in the overprediction of fuel temperatures. The GAUGE tapes for Cycles 4 and 5 used as input to TSORT and the power and flux tapes produced by TSORT have been archived. The archive number is RPSD5018. The G AUGE tapes for Cycles 1 and 2 and for Cycle 3 had been previously archived with archive numbers RPSD2614 and RPSD3894, respectively. 1 s Page 89 l

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        ,                                                                                                               909658/A    l CED110:%WU?

l

References:

l1. Archibald, R.,. "TSORT - A Computer Program to Process Nuclear Design Data for.Use in Core Performance and Mechanical Design Calculations with the SURVEY Program," GA-A14524, June 1977.

1

2. Wagner, M. R.,'" GAUGE - A Two-Dimensional Few Group Neutron Diffusion-y Depletion Program for a Uniform Triangular Mesh," G A-8307, March 15,1968.

V 3. Wan, M., "FSV Initial Core Fuel Performance Calculations," G A memo FFE:184:MW:79, May 22,1979.

4. Lefler, W. L., "TSORT Data for FSV Cycles 1 and 2," G A memo CNE:195:WLL:81, October 9,1981.

5. Lefler, W. L., "TSORT Data For FSV Cycle 3," Document No. 907736, May 10,. ,

1985. I 6. Lefler, W. L., "TSORT Data For FSV Cycle 4 To 24.9 EFPD," G A memo CED:261:WLL:87, April 2,1987.

7. Kraetsch,- H. and M. R. Wagner, "GATT, A Three-Dimensional Few-Group Neutron Diffusion Theory Program for a Hexagonal-Z Mesh," GA 8547,1967.

! 8. Georghious, D. L., " SURVEY - A Computer Code for the Thermal and Fuel Perfor - mance Analysis of High-Temperature Gas-Cooled Reactors," GA D14869, November 1978. 1-

          -DISTRIBUTION:

D. Alberstein g A. Baxter L V. Jovanovic

V. Malakhof l EDF E

1 i 1 Page 90 L l

N

1 909658/A .I

[ CED:110 WLL:87 ) 1 TABLE 1 TSORT DATA FOR FSV CYCLE 1 Time Thermal . Core Inlet Point EFPD Power (%) Power / Flow Temperature ('F) 1 0.0 5.0 0.568 550.0-2 3.0 - -23.5 0.553 593.2 3- 4.6 27.7 0.722 631.2 )

                '4      :12.8         29.7          0.803          640.3
               .5=       19.0          4.9          0.672 -        611.1 6       20.0         28.9          0.698-         616.6 7       26.0-        27.9          0.622        ~ 584,8 8       26.7         28.7          0.736          614.7 9       28.6         28.6          0.689          613.9 10        30.5         28.4          0.751          619.4 11        33.8         38.7          0.791          626.4 12        38.8         36.5          0.804          622.5                     ,

13 51.7 29.6 0.781' 636.9

              '14        53.0         54.6         0.870           659.1 15        57.4         25.4         0.868           636.2                    i 16        62.8         34.0         0.731           629.3                     '

17 68.9 51.1 0.842 651.4 18 72.9 65.3 0.915 659.9 '

              ~19       -77.0         30.5         0.782           633.3 20       '79.7         65.3         0.916           664.4 21        87.2         61.5         0.797           637.9 22 -      89.7         61.5         0.837           652.0 23        92.5         63.5         0.904           658.8 24       102.3         22.8         0.626           653.2 25       103.1         38.0         0.778           633.2 26       105.2         62.9'        O.870           669.4 27       113.7         51.2        .0.838         - 649.5

28 115.9 28.6 0.615 627.3 29 120.0 49.3 0.862 I 643.9-30 123.5 2.1 0.862 643.9 31 123.5 36.1 0.742 609.6 32 128.3 12.0 0.655 568.9 33 129.5 53.7 0.909 644.9

         . 34      136.6          53.3         1.002          654.3 35-      144.9         64.8         0.929          656.3 36      173.4          35.2         0.929          656.3 Full power = 841.7 MW(t) 8 Full flow = 3.49 x 10 lb/hr                                                    ,

Page 91

7-- 1 909658/A CED:110:WLL:87 TABLE 2 TSORT DATA FOR FSV CYCLE 2 I Time Thermal Core Inlet i Point EFPD Power (%)' Power / Flow Temperature (*F) 1- 0.0 0.0 - - 2 0.0 47.4 0.739 637.8 ~ 3 4.7 65.1 0.896 668.1 4 8.1 65.2 0.874 665.0 5 13.1- 40.4 0.706 637.8 6 15.9 20.2 0.483 530.0 7 16.9 32.1 0.793 637.7 ; 8 18.4 53.4 0.766 654.4 9 23.7 66.0 0.889 673.1 10 36.8 33.8 0.683 631.2 11 43.4 55.8 0.876 638,0 12 - 60.3 69.7 0.884 667.6 13 .70.1 52.5 0.830 635.9 14 81.2 43.1 0.746 621.5 15' 84.0 55.7 0.834 612.2 16- 90.2 44.6 0.782 647.6 17 95.3 59.0- 0.912 671.9

              .18          96.7           46.7          0.717                                                               677.5 19~        101.1           63.9          0.909                                                               674.9 20        108.9            68.9          0.961                                                               672.5 21        123.8            43.7          0.797                                                               643.6 l               22        126.8            69.2          0.888                                                               677.6
              .23        132.2            38.0          0.777                                                               642.7' 24        138.4            55.8          0.754                                                               678.1 25        141.0            68.3          0.892                                                               678.5 26        145.8            69.7          0.848                                                               690.2 27        157.3       . 39.5          0.738                                                               635.4 28        163.5            69.3          0.871                                                               686.1 29        168.4            47.2          0.770                                                               649.9 30        170.5            80.3          0.880                                                               695.1 g               31        175.0            70.0          0.859                                                               680.9 32       188.7             70.0          0.859                                                              680.9 L

p>_ Full power = 841.7 MW(t) 8 Full flow = 3.49 x 10 lb/hr l' l Page 92 l-

    , - o                                                                                 .

909658/A CED110iWLL:87

 -1
                                                                                           \

TABLE 3 TSORT DATA FOR FSV CYCLE 3 -l l Time Thermal Core Inlet Point EFPD Power (%) Power / Flow Temperature (*F) , 1 0.0 70.2 0.882- 680.2 2 19.0 69.4 0.907 671.0 3 39.7 68.3 0.916 653.7 7 4 44.5 34;6- 0.718 599.4 5 50,7 5.9 0.274 360.5 6 50.9 68.3 0.892 677.3 7 53.0 100.0 0.906 737.9 8 55.5 69.9 0.892 677.3 9 71.3 70.2 0.892 677.3 10 85.4 69.6 0.920 667.2 11 105.3 69.6 0.868 677.9 12 123.4 69.6 0.856 680.8 13 146.7 70.1 0.878 677.9 14 153.3 70.1 0.897 688.9

              ,15       165.2         26.1         0.528         617.3 16       167.5         26.1         0.528         617.3 17       170.4         42.4         0.828         625.0-18       187.5         61.8         0.992-        704.5 19'      207.7         65.9         0.868         684.2 20       228.3         69.4         0.884        684.0-21       247.5         32.8         0.558         653.7 22       251.0         70.7         0.874        690.9 23       268.2         70.0,        0.860         691.2 24       282.7              4 67.5         0.886        685.2 25       293.8'        30.3        0.662         639.6 26       294.6         30.3        0.662         639.6 Full power = 841.7 MW(t)

Full flow = 3.49 x 10e Ib/hr Page 93

[ , 909658/A CED:110 WLL:87 9 TABLE 4 TSORT DATA FOR FSV CYCLE 4 I Time

Thermal Core Inlet

              - Point     EFPD      Power (%)    Power / Flow   Temperature (*F) 1         0.0 -      28.5        0.628            648.0 l1 2         4.9         9.0        0.331            376.6 l                   3         8.9        26.6        0.532           639.1-4         13.0       33.9-        0.702           646.4 5         15.6       34.5         0.725           644.0 6         17.4       30.0         0.674           629.9-l-                                                                                              t 7        20.7        30.1         0.672           620.4

'. 8 24.9 80.0 0.92 680.0 l 9 29.0 80.0 0.92 680.0 l- 10 45.0 80.0 0.92 680.0 11 85.0 80.0 0.92 680.0-12 125.0 80.0- 0.92 680.0 13 165.0 80.0 0.92 680.0 1 14 205.0 80.0 0.92 680.0 L- 15 238.6' 80.0 0.92 680.0

                '16       265.3        80.0         0.92            680.0 l

r l 17 292.0 80.0 0.92 680.0 L L Full power = 841.7 MW(t) Full flow = 3.49 x 10e Ib/hr Page 94

909658/A CED:llo:WLL:si I r

    ,                                    TABLE 5 TSORT DATA FOR FSV CYCLE 5 Time                 Thermal Core Inlet -

Point EFPD Power (%) Power / Flow Temperature (*F) 1 0.0 80.0 - 0. 92 680.0 2 12.0 80.0 0. 92 680.0 4 3 52.0 80.0 0. 92 680.0 4 92.0 80.0 0. 92 680.0 i 5 132.0 80.0 0. 92 680.0

6 172.0 80.0 0. 92 680.0 7: 212.0 80.0 0. 92 680.0 8- 252.0 80.0 0. 92 680.0 9 "292.0 80.0 0. 92 680.0 , Full power = 841.7 MW(t) Full flow = 3.49 x 10e Ib/hr i s Page 95 A}}

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