ML100900227
| ML100900227 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, Three Mile Island  |
| Issue date: | 02/28/2010 |
| From: | Aissa M, Difilippo F, Emmett M, Gauld I, Ilas G Office of Nuclear Regulatory Research, Oak Ridge |
| To: | |
| References | |
| Job Code Y6685, ORNLITM-2008/071 NUREG/CR-6968 | |
| Download: ML100900227 (140) | |
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- U.S.NRC United States Nuclear Regulatory Commission NUREG/GR-6968 ORNLITM-2008/071 ProtectingPeople and the Environment Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation-Calvert Cliffs, Takahama, and Three Mile Island Reactors Office of Nuclear Regulatory Research
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U.S.NRC #US.NRCNUREG/CR-6968 United States Nuclear Regulatory Commission oTM-2008/071 ProtectingPeople andthe Environment Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation-Calvert Cliffs, Takahama, and Three Mile Island Reactors Manuscript Completed: June 2009 Date Published: February 2010 Prepared by G. Ilas, I.C. Gauld, F.C. Difilippo, M.B. Emmett Oak Ridge National Laboratory Managed by UT-Battelle, LLC Oak Ridge, TN 37831-6170 M. Aissa, NRC Project Manager NRC Job Code Y6685 Office of Nuclear Regulatory Research
ABSTRACT This report is part of a report series designed to document benchmark-quality radiochemical isotopic assay data against which computer code accuracy can be quantified to establish the uncertainty and bias associated with the code predictions. The experimental data included in the report series were acquired from domestic and international programs and include spent fuel samples that cover a large burnup range.
The measurements analyzed in the current report, for which experimental data is publicly available, include 38 spent fuel samples selected from fuel rods with a 2.6 to 4.7 wt % 235U initial enrichment, which were irradiated in three pressurized water reactors operated in the United States and Japan and achieved burnup values from 14to 56 GWd/MTU. The analysis of the measurements was performed by employing the two-dimensional depletion sequence of the TRITON module in the SCALE code system.
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TABLE OF CONTENTS A BSTRA CT .............................................................................................................................................. iii LIST OF FIGURES ..................................................... ...... vii LIST OF TABLES ...................................................................................................................................... ix A CKN O WLED GM EN TS ............................................................................................................................ xi A CRON YM S .................................................................................................................................... Ii.......
xi I INTRO DUCTION ............................................................................................................................ 1 2 EXPERIM EN TA L PROGRA M S .............................................................................................. 5 2.1 Dom estic Program s ....................................................................................................... 5 2.1.1 TM I-1 ..................................................................................................................... 5 2.1.2 Calvert Cliffs ............................................... 5 2.2 International Program s .................................................................................................. 6 2.2.1 JAERI (Takaham a-3) ....................................................................................... 6 3 ISOTOPIC M EA SU REM ENTS ................................................................................................ 7 3.1 TM I-1 Sam ples ........................................................................................................... 7 3.1.1 AN L M easurem ents ......................................................................................... 7 3.1.2 GE-VN C Measurem ents .................................................................................. 8 3.2 Calvert Cliffs Sam ples .................................................................................................. 17 3.2.1 PNN L M easurem ents ...................................................................................... 17 3.2.2 K RI M easurem ents .......................................................................................... 18 3.2.3 Experim ental Data U sed for Code Validation ................................................. 19 3.3 Takaham a-3 Sam ples .................................................................................................. 28 4 ASSEMBLY AND IRRADIATION HISTORY DATA .......................................................... 33 4.1 TM I-1 Sam ples .................................................................................................................. 33 4.2 Calvert Cliffs Sam ples ................................................................................................ 41 4.3 Takaham a-3 Sam ples .................................................................................................. 44 5 COMPU TATION AL M OD ELS .............................................................................................. 49 5.1 Com putational Tools ..................................................................................................... 49 5.2 TM I-1 Sam ples ........................................................................................... ................... 49 5.3 Calvert Cliffs Sam ples .................................................................................................. 54 5.4 Takaham a-3 Sam ples .................................................................................................. 55 6 RESU LTS ...................................................................................................................................... 59 6.1 TM I-1 Sam ples .................................................................................................................. 59 6.2 Calvert Cliffs Sam ples .................................................................................................. 67 6.3 Takaham a-3 Sam ples .................................................................................................. 70 7 SUMM ARY ................................................................................................................................. 77 v
TABLE OF CONTENTS (continued)
Pagie 8 RE FER EN C E S ............................................................................................................................... 79 APPENDIX A: EFFECT OF MODELING DETAILS ON PREDICTED NUCLIDES FOR TM I-I SA M PLES ............................................................................................. A -1 APPENDIX B: SELECTED TRITON INPUT FILES ....................................................................... B-1 vi
LIST OF FIGURES Page Figure 4.1 Assembly layout for TMI-I samples-NJ05YU .......................................................... 35 Figure 4.2 Assembly layout for TMI-1 samples-NJ070G ........................................................ 36 Figure 4.3 Assem blies surrounding assem bly NJ070G 41 4.............................
Figure 4.4 Assembly layout for Calvert Cliffs samples ................................................................. 42 Figure 4.5 Assembly layout for Takahama-3 samples .................................................................. 45 Figure 5.1 TRITON assembly model for TMI-1 samples in assembly NJ05YU .......................... 51 Figure 5.2 TRITON assembly model for TMI-1 samples in rod 012 of assembly NJ070G ...... 52 Figure 5.3 TRITON assembly model for TMI-1 samples in rod 01 of assembly NJ070G ....... 53 Figure 5.4 TRITON assembly model for Calvert Cliffs samples ................................................ 54 Figure 5.5 TRITON assembly model for Takahama-3 SF95 samples ......................................... 56 Figure 5.6 TRITON assembly model for Takahama-3 SF96 samples ......................................... 57 Figure 5.7 TRITON assembly model for Takahama-3 SF97 samples ......................................... 58 Figure 6.1 TMI-I samples from assembly NJ070G-major actinides ............................................... 60 Figure 6.2 TMI- 1 samples from assembly NJ070G-minor actinides .......................................... 60 Figure 6.3 TMI-1 samples from assembly NJ070G-fission products (Nd) ................................ 61 Figure 6.4 TMI-I samples from assembly NJ070G-fission products (Cs, Sm, Eu, Gd) ............. 61 Figure 6.5 TMI-1 samples from assembly NJ05YU-major actinides .......................................... 62 Figure 6.6 TMI-1 samples from assembly NJ05YU-minor actinides .................... 62 Figure 6.7 TMI-1 samples from assembly NJ05YU-fission products (metallics) ..................... 63 Figure 6.8 TMI-1 samples from assembly NJ05YU-fission products (Nd, Cs, Sm, Eu, Gd) .......... 63 Figure 6.9 Calvert Cliffs sam ples- actinides ................................................................................ 67 Figure 6.10 Calvert Cliffs samples-fission products (Nd, Cs) ..................................................... 68 Figure 6.11 Calvert Cliffs samples-fission products (Sm, Eu, Gd, Sr, Tc) ................................... 68 Figure 6.12 Takahama-3 samples- uranium nuclides ..................................................................... 73 Figure 6.13 Takahama-3 samples-plutonium nuclides ................................................................ 73 vii
LIST OF FIGURES (continued)
Page Figure 6.14 Takahama-3 samples-minor actinides (Np, Am) ....................................................... 74 Figure 6.15 Takahama-3 samples-minor actinides (Cm) .............................. 74 Figure 6.16 Takahama-3 samples-fission products (Nd) .............................................................. 75 Figure 6.17 Takahama-3 samples-fission products (Cs, Ce, Eu) ................................................. 75 Figure 6.18 Takahama-3 samples-fission products (metallics) ...................................................... 76 Figure 6.19 Takahama-3 samples-fission products (Sm) .............................. 76 Figure A. 1 Assemblies surrounding NJ05YU during cycle 9 .......................................................... A-2 Figure A.2 Assemblies surrounding NJ05YU during cycle 10 ......................................................... A-2 Figure A.3 TRITON model #2 for TMI-1 samples in assembly NJ05YU ........................................ A-3 Figure A.4 Effect of modeling assumptions on U and Pu-assembly NJ05YU ............................... A-6 Figure A.5 Initial layout- assem bly NJ070G ................................................................................... A -9 Figure A.6 Initial TRITON model for samples in corner rod 01 of assembly NJ070G ........ A-10 Figure A.7 Effect of modeling assumptions on 2. 5U and 23 9 Pu-assembly NJ070G ...................... A-12 235 239 Figure A.8 Effect of modeling assumptions on U, pu, and 148Nd for sample 01 S7 ................ A-16 viii
LIST OF TABLES Page Table 1.1 Summ ary of spent fuel measurements ........................................................................... 3 Table 3.1 Experimental techniques and uncertainties for TMI-l samples measurements at AN L ................................................................................................................................. 9 Table 3.2 Experimental results (g/g 238U) for TMI-I samples from assembly NJ05YU ............... 10 Table 3.3 Experimental results (gig Uinitial) for TMI-1 samples from assembly NJ05YU ............ 12 Table 3.4 Experimental techniques and uncertainties for TMI- 1 samples measurements at GE -V N C ........................................................................................................................ 14 Table 3.5 Experimental results (g/g 238U) for TMI-1 samples from assembly NJ070G .............. 15 Table 3.6 Experimental results (gig Uiniti.a) for TMI-1 samples from assembly NJ070G ........ 16 Table 3.7 Experimental techniques and uncertainties for Calvert Cliffs samples-PNNL data ..... 19 Table 3.8 Experimental results (g/g fuel) for Calvert Cliffs samples-PNNL data .................... I..... 20 Table 3.9 Experimental results (g/g Uinitiai) for Calvert Cliffs samples-PNNL data .................. 21 Table 3.10 Experimental techniques and uncertainties for Calvert Cliffs samples-KRI data .......... 22 Table 3.11 Experimental results for Calvert Cliffs samples-KRI data ............................................. 23 Table 3.12 Comparison of PNNL and KRI data (relative to 145Nd) .............................................. 24 Table 3.13 Experimental results (g/g Uinitial) for Calvert Cliffs samples-KRI data ........ ....... 25 Table 3.14 Experimental results (g/g Uinitial) for Calvert Cliffs samples used for code validation' .... 26 Table 3.15 Experimental techniques and uncertainties for Takahama-3 samples .......................... 29 Table 3.16 Experimental results (gig Uinitial) for Takahama-3 samples from rod SF95 ................. 30 Table 3.17 Experimental results (g/g Uiniti.a) for Takahama-3 samples from rod SF96 ................. 31 Table 3.18 Experimental results (gig Uinitiai) for Takahama-3 samples from rod SF97 ................. 32 Table 4.1 Assembly design data for TMI-1 samples ................................................................... 37 Table 4.2 Burnup, power and moderator density data for TMI-I samples .................................. 38 Table 4.3 Fuel temperature and concentration of soluble boron in moderator for TMI-1 sam ples from assem bly NJ05YU ................................................................................ 39 Table 4.4 Fuel temperature and concentration of soluble boron in moderator for TMI-1 sam ples from assem bly N J070G .................................................................................. 40 ix
LIST OF TABLES (continued)
Page Table 4.5 Cooling time at measurement date for TMI-1 samples .............................................. 40 Table 4.6 Assembly design data for Calvert Cliffs samples ....................................................... 43 Table 4.7 Burnup history data for Calvert Cliffs samples ............................................................ 44 Table 4.8 Moderator, fuel temperature, and cooling time data for Calvert Cliffs samples ....... 44 Table 4.9 Assembly design data for Takahama-3 samples ......................................................... 46 Table 4.10 Burnup, power, sample location, and moderator data for Takahama-3 samples ....... 47 Table 4.11 Operation history data for Takahama-3 samples .......................................................... 48 Table 4.12 Soluble boron concentration in moderator forTakahama-3 samples ............................ 48 Table 6.1 C/E-i (%) for TMI-I samples from assembly NJ070G ............................................... 64 Table 6.2 C/E-i (%) for TMI-1 samples from assembly NJ05YU ............................................ 65 Table 6.3 C/E-I (%) for Calvert Cliffs samples ......................................................................... 69 Table 6.4 C/E- I (%) for Takahama-3 samples ............................................................................ 71 Table A. 1 Effect of modeling assumptions on C/E-1 (%) for samples from assembly N J0 5Y U ........................................................................................................................... A -5 Table A.2 C/E-I (%) for samples in assembly NJ05YU - computational model #1 ....................... A-7 Table A.3 C/E- 1 (%) for samples in assembly NJ05YU - computational model #2 ....................... A-8 Table A.4 C/E -I (%) for samples in assembly NJ070G - computational models # I an d # 2 ............................................................................................................................ A -13 Table A.5 Effect of modeling assumptions on C/E-I (%) for sample 013S7 from assem bly NJ070G .......................................................................................................... A -15 x
ACKNOWLEDGMENTS This work was supported by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, under Project JCN Y6685, ExperimentalDatafor High Burnup Spent Fuel Validation. The authors acknowledge the review and helpful comments of R. Y. Lee and D. E. Carlson of the Office of Nuclear Regulatory Research, and C. J. Withee of the Spent Fuel Project Office. Review of the manuscript by our colleague at Oak Ridge National Laboratory, G. Radulescu, and the careful formatting of this document by D. J. Weaver is very much appreciated and acknowledged.
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ACRONYMS ANL Argonne National Laboratory ARIANE Actinides Research In A Nuclear Element ATM Approved Testing Material BOC beginning of cycle BPR burnable poiSon-rod CE Combustion Engineering CEA Commisariat Al'Energie Atomique C/E calculated-to-experimental.
DOE U.S. Department of Energy EFPD effective full power days EOC end of cycle EPRI Electric Power Research Institute GE-VNC General Electric-Vallecitos Nuclear Center GKN II Gemeinschaftskernkraftwerk Unit II ICP-MS inductively coupled plasma mass spectrometry ID-MS isotope dilution mass spectrometry ITU Institute for Transuranium Elements JAERI Japanese Atomic Energy Research Institute KRI Khoplin Radium Institute LANL Los Alamos National Laboratory LA luminescent analysis LWR light water reactor MALIBU MOX and UOX LWR Fuels Irradiated to High Burnup MOX mixed oxide MS mass spectrometry MTU metric ton uranium (106 grams)
NEA Nuclear Energy Agency NRC U.S. Nuclear Regulatory Commission OECD Organization for Economic Cooperation and Development ORNL Oak Ridge National Laboratory PNNL Pacific Northwest National Laboratory PSI Paul Scherrer Institute PWR pressurized water reactor REBUS Reactivity Tests for a Direct Evaluation of the Burnup Credit on Selected irradiated LWR fuel bundles RSD relative standard deviation SCALE Standardized Computer Analyses for Licensing Evaluations SCK-CEN Studiecentrum voor Kernenergie - Centre d'6tude de lEnergie Nucl~aire SFCOMPO Spent Fuel Isotopic Composition Database TIMS thermal ionization mass spectrometry TMI Three Mile Island YMP. Yucca Mountain Project xiii
1 INTRODUCTION The current trend toward extended irradiation cycles and higher fuel enrichments of up to 5 wt % 235U has led to an increase of the bumup range for discharged nuclear fuel assemblies in the United States that is expected to exceed 60 GWd/MTU. Accurate analysis and evaluation of the uncertainties in the predicted isotopic composition for spent nuclear fuel in the high burnup regime requires rigorous computational tools and experimental data against which these tools can be benchmarked. However, the majority of isotopic assay measurements available to date involve spent fuel with bumups of less than 40 GWd/MTU 235 and enrichments below 4 wt % U, limiting the ability to directly validate computer code predictions and accurately quantify the uncertainties of isotopic analyses for modem, high-bumup fuel.
This report is part of a report series that documents high-quality radiochemical assay data against which computer code predictions of the isotopic composition in high bumup fuel can be validated. Quantifying and evaluating these uncertainties is fundamental for understanding and reducing the uncertainties associated with predicting the high burnup fuel characteristics for spent fuel transportation and storage applications involving decay heat, radiation sources, and criticality safety evaluations with bumup credit, as well as for reactor safety studies and accident consequence analysis. The report series presents a compilation of recently available isotopic measurements involving high bumup pressurized water reactor (PWR) fuel as well as older isotopic measurements for low- and medium-range bumup fuel that can be used for code validation purposes. Previous experiments were selected primarily on the basis of having extensive fission product measurements.
The experimental data included in the report series were compiled from domestic and international programs. The isotopic assay measurements include data for a total of 45 spent fuel samples selected from fuel rods enriched from 2.6 to 4.7 wt % 235U and irradiated in five different PWRs operated in Germany, Japan, Switzerland, and the United States. The samples cover a large bumup range, from 14 to 70 GWd/MTU. A summary of the experimental programs and measured fuel characteristics is listed in Table 1.1.
The current report includes the experimental data and analysis of measurements for which information is publicly available and was not obtained through multi-collaborative international programs. Data for 38 fuel sample measurements are presented in this report: 22 of domestic origin and 16 from experiments carried out in Japan. The bumup range for these samples is 14 to 56 GWd/MTU. The Japanese experimental data is publicly available in the Spent Fuel Isotopic Composition Database (SFCOMPO),
originally developed by the Japanese Atomic Energy Research Institute (JAERI) and now administered by the Nuclear Energy Agency (NEA), a specialized agency within the Organization for Economic Cooperation and Development (OECD). As indicated in Table 1.1, a second report documents the analysis of experimental data acquired by Oak Ridge National Laboratory (ORNL) through participation in two international programs: (1) "Actinides Research In A Nuclear Element" (ARIANE) and (2)
"Reactivity Tests for a Direct Evaluation of the Bumup Credit on Selected Irradiated LWR Fuel Bundles" (REBUS), both coordinated by the Belgian company Belgonucleaire. A third report presents the analysis of experimental data obtained by ORNL through participation in the MALIBU international program coordinated by Belgonucleaire. Each of the three reports mentioned in Table 1.1 present information on the radiochemical analysis methods and uncertainties, assembly design description and irradiation history, and computational models and results obtained using the "Standardized Computer Analyses for Licensing Evaluations" (SCALE) code system.'
Section 2 of the current report presents a summary of the experimental programs evaluated. The radiochemical methods employed and the associated experimental uncertainties are provided in Section 3.
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Information on the assembly design data and irradiation history is presented in Section 4, and details on the computational models developed and simulation methodology used are shown in Section 5. A comparison of the experimental results to the results obtained from code simulations with SCALE are presented in Section 6.
2
Table 1.1 Summary of spent fuel measurements Reactor Measurement Experimental Assembly Enrichment No. of Measurement Burnup(s)
(country) facility program name design (wt % 235U) samples methods (GWd/MTU)
TMI-1 a ANL YMP 15 x 15 4.013 11 ICP-MS, 44.8-55.7 (USA) (USA) a-spec, y-spec TMI-1 a GE-VNC YMP 15 x 15 4.657 8 TIMS, 22.8-29.9 (USA) (USA) a-spec, y-spec Calvert Cliffs a PNNL, KRI ATM 14 x 14 CE 3.038 3 ID-MS, LA, 27.4 -44.3 (USA) (USA, Russia) ca-spec, y-spec Takahama 3 JAERI JAERI 17 x 17 2.63,4.11 16 ID-MS, 14.3-47.3 (Japan) (Japan) a-spec, y-spec G6sgen b SCK-CEN, ITU ARIANE 15 x 15 3.5, 4.1 3 TIMS, ICP-MS, 29.1, 52.5, 59.7 (Switzerland) (Belgium, Germany) a-spec, P-spec, y-spec GKN 11 b SCK-CEN REBUS 18 x 18 3.8 1 TIMS, ICP-MS 54.0 (Germany) (Belgium) a-spec, y-spec Gosgen C CEA, PSI, SCK-CEN MALIBU 15 x 15 4.3 3 TIMS, ICP-MS, 46.0, 50.8, 70.4 (Switzerland) (France, Switzerland ,Belgium) a-spec, y-spec Documented in current report.
b Documented in G. Has, 1.C. Gauld, and B. D. Murphy, Analysis of ExperimentalDatafor High Burnup PWR Spent Fuel Isotopic Validation-ARIANE andREBUS Programs(U0 2
Fuel), NUREG/CR-6969 (ORNL/TM-2008/072), Oak Ridge National Laboratory, Oak Ridge, Tennessee (May 2008).
'Documented in G. Has, and I. C. Gauld, and B. D. Murphy, Analysis of Experimental Datafor High Burnup PWR Spent Fuel Isotopic Validation-MALIBU Program(U0 2 Fuel),
NUREG/CR-6970 (ORNL/TM-2008/13), Oak Ridge National Laboratory, Oak Ridge, Tennessee (May 2008).
2 EXPERIMENTAL PROGRAMS This section provides a brief overview of the measured isotopic assay data compiled in this report for code validation and a summary of the experimental programs from which they were acquired. A description of the measurement techniques and experimental data and uncertainties is provided in Section 3. -
2.1 DOMESTIC PROGRAMS 2.1.1 TMI-1 Measurements on 19 spent fuel samples from the Three Mile Island (TMI) Unit 1 reactor were performed under the auspices of the U.S. Department of Energy (DOE) Yucca Mountain Project (YMP). Fuel rods were obtained from two separate assemblies, identified as NJ05YU and NJ070G. Radiochemical analyses were performed at two independent experimental facilities: Argonne National Laboratory (ANL) and General Electric-Vallecitos Nuclear Center (GE-VNC). Measurements on 11 of the TMI-1 samples from rod H6 of assembly NJ05YU were performed in 1998 and 2000 at ANL; 2 whereas, the other eight 3 TMI-l samples, from rods 01, 012, and 013 of assembly NJ070G, were analyzed in 1999 at GE-VNC.
Fuel rod H6 had an initial enrichment of 4.013 wt % 235U and achieved local sample burnups from 45 to 56 GWd/MTU over two irradiation cycles (cycles 9 and 10). Rods 01, 012, and 013 had an initial enrichment of 4.657 wt % 235U and achieved burnups between 22 and 30 GWd/MTU in one irradiation cycle (cycle 10).
Previous benchmark calculations performed using these measurements have yielded uncharacteristically large deviations in isotopic results in comparison with past experience: YMP has published results indicating deviations in the predicted 2 3 9pu concentrations that ranged up to 30-40% higher than the measurements. 4 Past experience with other spent fuel samples evaluated by ORNL (and YMP) have yielded lower deviations as compared to measurements. 5 The large deviations obtained by YMP were inconsistent with the results observed for similar burnup samples from the Takahama-3 reactor 6 and also differed from literature results for spent fuel validation for the French Gravelines reactor 7 obtained using French codes and data.
Investigations have been performed by Electric Power Research Institute (EPRI), following indications of fuel leakage during TMI-l cycle 10, to examine the causes of fuel rod failure, as it is mentioned in the abstract for the TR-108784-V 1 report. 8 As the fuel failure phenomena introduce additional uncertainties related to the actual operating conditions of the fuel that may affect the accuracy of code predictions, these type of uncertainties and their importance to the fuel simulations would need to be accounted for in a comparison of predicted and experimental isotopic assay data. Because no details are publicly available on the actual location or description of the failed fuel rods in cycle 10 and their relationship with the fuel rods for which measured isotopic assay data is presented here, no assessment of the impact of the failed fuel on the calculated results can be made based on the currently available unrestricted information.
2.1.2 Calvert Cliffs The measurements on three spent fuel samples from Calvert Cliffs Unit I reactor considered in this report were carried out at the Material Characterization Center at Pacific Northwest National Laboratory (PNNL) for the Approved Testing Material (ATM) Program 9 designed to characterize medium-burnup spent fuel representative of reactors operating in the United States. Lanthanide measurements for the same three samples have been also performed at Khoplin Radium Institute (KRI) in Russia. 10 These three 5
samples were selected from rod MKP-109 of assembly D047 that was irradiated in the reactor for four consecutive cycles. The assembly had an initial fuel enrichment of 3.038 wt % 235U, and the samples under consideration covered a burnup range from 27 to 44 GWd/MTU.
The PNNL data served as the basis of a benchmark for validating irradiated fuel used in criticality calculations13 11,12 and was used for the OECDiNEA burnup credit criticality safety calculation benchmark Phase I-B.
2.2 INTERNATIONAL PROGRAMS 2.2.1 JAERI (Takahama-3)
From 1990 to 1999, JAERI carried out a series of projects focused on obtaining high-quality experimental isotopic assay and criticality data to support the development of burnup credit for storage and transportation of spent fuel. The measurements included destructive radiochemical analyses of spent fuel samples, axial gamma scanning of spent fuel rods, and exponential experiments on spent fuel assemblies.
The measured data were used by JAERI for evaluating the accuracy of depletion or criticality computational tools.
Sixteen samples selected from three fuel rods irradiated in assemblies NT3G23 and NT3G24 of the Takahama-3 reactor were included for destructive isotopic analyses. Five of these samples belonged to a U0 2-Gd 2O 3 fuel rod with a 2.63 wt % 235U initial enrichment; whereas, the other 11 samples were from two U0 2 fuel rods with an initial enrichment of 4.11 wt % 235U. The burnup of these samples was between 14 and 47 GWd/MTU.
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3 ISOTOPIC MEASUREMENTS 3.1 TMI-1 SAMPLES 3.1.1 ANL Measurements The radiochemical analysis at ANL considered 11 samples from fuel rod H6 of TMI-I assembly identified as NJ05YU, cut from rod segments provided by GE-VNC. The samples for analysis were prepared by dissolution of an approximately 0.1-0.2 g aliquot of homogenized fuel sample powder.
Analyses were carried out by using inductively coupled plasma mass spectrometry (ICP-MS),
,y-spectrometry, and c-spectrometry to determine the isotopic mass of 31 nuclides. The results were reported relative to the measured 21BU content in the sample, as g/g 211U. Two measures of the experimental uncertainty, a within-sample precision and a bias uncertainty, were provided by ANL. The within-sample precision was estimated by ANL as one standard deviation through repeated measurements of samples, whereas the bias uncertainty was estimated from deviations of quality control standard solutions measured before and after fuel samples; the bias uncertainty included the propagation of error for normalization to 238U.2 The main experimental techniques used for each nuclide and the reported corresponding experimental uncertainties 2 are presented in Table 3.1. In addition to the bias values shown in the table, a bias uncertainty of 3.8% was reported for 238U; but no explanation was provided on the significance of this value; it is assumed here that it refers to the 238U concentration measured directly. The within-sample precision shown in the fifth column of the table was calculated so that it accounted for error propagation due to normalization of the concentration to the 238U content, as:
= r(,_,mpoe ))2 + smlrepo,, (3-1)
Ori,within-sainpl ]* i'iwithinsape
- 2 U,within-saraple
... (3 1 where i identifies the nuclide. The total uncertainty for the measured concentration of a nuclide i expressed relative to the 23 8U content is shown in the sixth column of Table 3.1 and was obtained by combining the within-sample uncertainty, calculated as in Eq. 3-1, and the reported bias, as:
- (7" 2[reported )2(32 ieto,Ial = V(iwihhin-sample Y + ([7t,;b;e. ) (3-2) 23 The total uncertainty is 3.7% for 1U, in the range 5-8% for plutonium nuclides, and about 5 - 7% for neodymium isotopes.
The reported results of the radiochemical analyses performed on TMI-1 samples at ANL 2 are shown in Table 3.2. In order to be compared with measured data obtained from other experimental programs, the experimental results were also expressed in units of g/g Uinitial, as shown in Table 3.3, using the initial uranium content in the sample as a basis. The concentration in g/g Uinitiai of nuclide i was determined as:1 a m, (3-3) mU + m'1,,+ Zm",' + Zm.,+238 m148Nd k M m 148Y 7
where m, is the mass of isotope i as reported in g/g 238U measured. The denominator in Eq. (3-3) is an estimate of the initial uranium content as a sum of the actinide (uranium, plutonium, americium, curium) weights in the measured sample and the weight loss in initial uranium due to burnup. The reduction in heavy metal mass due to burnup is approximated by 238 m,,*_, where Y is the average fission yield of 148Y 148Nd. A value Y = 0.0176 is recommended for PWR U0 2 fuel.14 Note that m 1U2
= I in Eq. (3-3). The relative standard deviations associated to the nuclide concentrations shown in Table 3.3 are assumed to be similar to the total uncertainty values in Table 3.1. No error propagation was carried out on the ratio in Eq. (3-3).
3.1.2 GE-VNC Measurements The measurements performed at GE-VNC 3 considered eight samples selected from three fuel rods from assembly NJ070G. Most of the 32 nuclides for which isotopic concentrations were measured at GE-VNC were determined by using thermal ionization mass spectrometry (TIMS) and some through y- or ca-spectrometry. The nuclide concentrations in the samples measured by TIMS were determined from measurements of spiked and unspiked samples. The nuclide content was reported as g/g 238U. The main experimental techniques used for each nuclide and the corresponding experimental uncertainty as reported are presented in Table 3.4. The experimental errors, reported by GE-VNC as relative uncertainty at a 95% confidence level, are shown in the third column of the table. The relative standard deviation (RSD) shown in the fourth column of the table was obtained as half of the reported uncertainty at a 95%
confidence level. The RSD for the GE-VNC measurements is 0.6% for all plutonium nuclides except for 238Pu, 0.5% for 235U, and 0.8% for neodymium isotopes.
The reported results of the radiochemical analyses performed on TMI-I samples at GE-VNC 3 are shown in Table 3.5. The measured results, expressed using the initial content of uranium in the sample as a basis, are presented in Table 3.6. The unit conversion was done by using Eq. (3-3).
8
Table 3.1 Experimental techniques and uncertainties for TMI-1 samples measurements at ANL Reported Reported Within-samote wihnsml isprecision Total'*
Nuclide ID Method a within-sample bias accounting for uncertainty precision uncertainty normalization to 23SU (%)
()(%) nomlzainto) %
U-234 ICP-MS 3.0 2.7 3.4 4.4 U-235 ICP-MS 1.5 2.9 2.3 3.7 U-236 ICP-MS 4.6 3.1 4.9 5.8 U-238 ICP-MS 1.7 4.2 Np-237 ICP-MS 4.1 3.4 4.4 5.6 Pu-238 a-spec 6.8 3.6 7.0 7.9 Pu-239 ICP-MS 4.3 3.3 4.6 5.7 Pu-240 ICP-MS 5.1 3.1 5.4 6.2 Pu-241 ICP-MS 3.2 2.9 3.6 4.6 Pu-242 ICP-MS 5.9 2.8 6.1 6.7 Am-241 y-spec 6.1 3.1 6.3 7.1 Am-242m ICP-MS NA 3.1 3.1 Am-243 ICP-MS 4.2 3.8 4.5 5.9 Mo-95 ICP-MS 1.7 3.4 2.4 4.2 Tc-99 ICP-MS 2.7 7.3 3.2 8.0 Ru-101 ICP-MS 1.6 5.3 2.3 5.8 Rh-103 ICP-MS 1.5 3.1 2.3 3.8 Ag- 109 ICP-MS 4.7 3.1 5.0 5.9 Cs-137 y-spec 3.6 2.7 4.0 4.8 Nd-143 ICP-MS 3.5 3.9 3.9 5.5 Nd-145 ICP-MS 4.8 3.5 5.1 6.2 Nd-148 ICP-MS 4.2 5.5 4.5 7.1 Sm-147 ICP-MS 3.3 9.4 3.7 10.1 Sm-149 ICP-MS 7.1 3.5 7.3 8.1 Sm-150 ICP-MS 3.5 3.2 3.9 5.0 Sm- 151 ICP-MS 6.1 3.2 6.3 7.1 Sm-152 ICP-MS 2.7 3.2 3.2 4.5 Eu-151 ICP-MS 12.0 2.9 12.1 12.5 Eu-153 ICP-MS 3.9 3.0 4.3 5.2 Eu-155 y-spec 6.4 2.7 6.6 7.2 Gd-155 ICP-MS 6.8 3.8 7.0 8.0 "Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
C b Calculated as shown in Eq. 3-1.
'Calculated as shown in Eq. 3-2.
9
Table 3.2 Experimental results (g/g 238U) for TMI-1 samples from assembly NJ05YU Sample ID AIB D2 B2 C DIMA4 A2 C3 C2B B3J BIB DIA2 Burnup a (GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 U-234 2.21E-04 2.07E-04 2.02E-04 2.14E-04 2.14E-04 2.07E-04 2.OOE-04 1.96E-04 1.99E-04 2.04E-04 2.1OE-04 U-235 9.26E-03 7.94E-03 6.71E-03 7.13E-03 8.1 1E-03 6.84E-03 6.77E-03 6.75E-03 6.63E-03 6.94E-03 7.59E-03 U-236 5.50E-03 5.74E-03 5.84E-03 5.92E-03 5.81E-03 5.95E-03 5.77E-03 5.62E-03 5.92E-03 5.87E-03 5.94E-03 Pu-238 4.34E-04 3.50E-04 3.40E-04 3.57E-04 4.06E-04 3.83E-04 2.72E-04 4.97E-04 4.32E-04 4.69E-04 4.15E-04 Pu-239 5.45E-03 5.84E-03 5.72E-03 5.85E-03 5.85E-03 5.78E-03 5.97E-03 5.41E-03 5.52E-03 5.55E-03 5.94E-03 Pu-240 2.52E-03 2.87E-03 2.95E-03 2.98E-03 2.84E-03 3.01E-03 3.08E-03 2.76E-03 2.88E-03 2.86E-03 2.95E-03 Pu-241 1.30E-03 1.47E-03 1.50E-03 1.54E-03 1.55E-03 1.47E-03 1.52E-03 1.44E-03 1.48E-03 1.48E-03 1.60E-03 Pu-242 7.31E-04 8.55E-04 9.89E-04 9.74E-04 1.02E-03 9.99E-04 1OOE-03 1-OIE-03 1.20E-03 1.04E-03 1.05E-03 Np-237 6.50E-04 7.27E-04 7.48E-04 7.62E-04 7.42E-04 7.5 1E-04 7.39E-04 7.44E-04 7.66E-04 7.62E-04 7.69E-04 Am-241 3.73E-04 3.72E-04 3.69E-04 4.08E-04 5.70E-04 3.27E-04 3.28E-04 5.50E-04 5.49E-04 3.13E-04 3.65E-04 Am-242m 1.OOE-05 LOOE-05 1.OOE-05 1.OOE-05 9.09E-07 1.OOE-05 1.OOE-05 1.82E-06 1.35E-06 1.12E-06 6.63E-07 0
Am-243 1.34E-04 2.07E-04 2.76E-04 2.66E-04 2.OOE-04 2.75E-04 2.67E-04 2.12E-04 2.29E-04 2.22E-04 2.24E-04 Nd-143 1.06E-03 9.83E-04 1.08E-03 1.06E-03 1.17E-03 1.03E-03 1.03E-03 1.12E-03 1.15E-03 1.18E-03 1.21E-03 Nd-145 9.17E-04 8.92E-04 9.80E-04 9.71E-04 1.04E-03 9.50E-04 9.71E-04 1.02E-03 1.06E-03 1.07E-03 1.09E-03 Nd-148 5.24E-04 5.24E-04 5.89E-04 5.90E-04 5.94E-04 5.96E-04 6.04E-04 6.20E-04 6.25E-04 6.44E-04 6.60E-04 Cs-137 1.81E-03 1.74E-03 1.89E-03 1.96E-03 1.79E-03 1.91E-03 1.84E-03 1.91E-03 1.88E-03 1.91E-03 1.67E-03 Sm-147 2.43E-04 1.96E-04 2.01E-04 2.02E-04 2.55E-04 2.13E-04 1.97E-04 2.48E-04 2.69E-04 2.77E-04 2.74E-04 Sm-149 3.35E-06 3.33E-06 3.53E-06 3.45E-06 3.90E-06 4.13E-06 3.14E-06 3.64E-06 3.46E-06 3.72E-06 4.20E-06 Sm-150 3.85E-04 3.75E-04 4.06E-04 4.15E-04 4.47E-04 4.05E-04 3.92E-04 4.54E-04 4.91E-04 5.08E-04 4.93E-04 Sm-151 1.39E-05 1.36E--05 1.45E-05 1.35E-05 1.53E-05 1.36E-05 1.36E-05 1.44E-05 1.60E-05 1.63E-05 1.69E-05 Sm-152 1.31E-04 1.30E-04 1.40E-04 1.37E-04 1.45E-04 1.43E-04 1.36E-04 1.41E-04 1.54E-04 1.56E-04 1.55E-04 Eu-151 7.98E-07 7.57E-07 8.58E-07 7.42E-07 7.23E-07 9.56E-07 9.18E-07 7.62E-07 8.t1E-07 6.19E-07 7.21E-07 Eu-153 1.58E-04 1.68E-04 1.81E-04 1.81E-04 1.89E-04 1.85E-04 1.74E-04' 1.87E-04 1.99E--04 2.02E-04 2.06E-04 Eu-155 1.08E-05 1.32E-05 1.42E-05 1.55E-05 1.37E-05 1.39E-05 1.38E-05 1.08E-05 1.12E-05 1.68E-05 1.07E-05
Table 3.2 Experimental results (g/g 238U) for TMI-1 samples from assembly NJ05YU (continued)
Sample ID AIB D2 B2 CI D1A4 A2 C3 C2B B3J BIB D1A2 Burnup a (GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 Gd-155 8.85E-06 6.02E-06 7.08E-06 6.88E-06 1.51E-05 6.56E-06 7.22E-06 1.02E-05 1.13E-05 1.09E-05 I.IE-05 Mo-95 I. 12E-03 9.90E-04 1.22E-03 1.19E-03 1.18E-03 1.21E-03 1.09E-03 1.19E-03 1.22E-03 1.25E-03 1.21E-03 Tc-99 1.53E-03 1.05E-03 1.18E-03 1.17E-03 1.29E-03 1.17E-03 1.12E-03 1.47E-03 1.35E-03 1.43E-03 1.24E-03 Ru-101 1.20E-03 1.02E-03 1.30E-03 1.26E-03 1.19E-03 1.25E-03 1.1 E-03 1.27E-03 1.27E-03 1.29E-03 1.23E-03 Rh-103 6.41E-04 5.55E-04 6.80E-04 6.69E-04 6.53E-04 6.70E-04 5.93E-04 6.66E-04 6.73E-04 6.81E-04 6.72E-04 Ag-109 5.50E-05 5.01E-05 5.71E-05 5.80E-05 9.17E-05 6.46E-05 1.OOE-04 7.08E-05 8.45E-05 4.78E-05 5.02E-05 a IA ..... LII J. IV? i1 ir- iv - ÷. - ,1. . . -- ' T 0 0Rev.
0O00011I, ep00 A 1 (April A A rJ. 2002).
. ).lurid 0.2IgIIeII rtre jyme Unt 1 ix*xiuutuL,uuwlO i Iuy ouMnpartlOtS Il ai/zLfl --LlcuJUiIOt, I uLca IViountdiun FlIoUt JA %pUILt
- ,i-LJ.IN U-
Table 3.3 Experimental results (g/g Uinitial) for TMI-1 samples from assembly NJ05YU Sample ID AIB D2 B2 cI DIA4 A2 C3 C2B B3J BIB DIA2 BurnupM 4 (GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 U-234 2.06E-04 1.93E-04 1.87E-04 1.98E-04 1.98E-04 1.92E-04 1.85E-04 1.81E-04 1.84E-04 1.88E-04 1.93E-04 U-235 8.62E-03 7.39E-03 6.22E-03 6.60E-03 7.50E-03 6.34E-03 6.27E-03 6.24E-03 6.13E-03 6.40E-03 6.99E-03 U-236 5.12E-03 5.35E-03 5.41 E-03 5.48E-03 5.38E-03 5.5 1E-03 5.34E-03 5.20E-03 5.47E-03 5.42E-03 5.47E-03 Pu-238 4.04E-04 3.26E-04 3.15E-04 3.3 1E-04 3.76E-04 -3.55E-04 2.52E-04 4.60E-04 3.99E-04 4.33E-04 3.82E-04 Pu-239 5.08E-03 5.44E-03 5.30E-03 5.42E-03 5.41E-03 5.35E-03 5.53E-03 5.OOE-03 5.1 OE-03 5.12E-03 5.47E-03 Pu-240 2.35E-03 2.67E-03 2.73E-03 2.76E-03 2.63E-03 2.79E-03 2.85E-03 2.55E-03 2.66E-03 2.64E-03 2.71E-03 Pu-241 1.21E-03 1.37E-03 1.39E-03 1.43E-03 1.43E-03 1.36E-03 1.41 E-03 1.33E-03 1.37E-03 1.37E-03 1.47E-03 Pu-242 6.8 1E-04 7.96E-04 9.17E-04 9.02E-04 9.44E-04 9.25E-04 9.26E-04 9.34E-04 1.11 E-03 9.60E-04 9.66E-04 Np-237 6.05E-04 6.77E-04 6.93E-04 7.06E-04 6.87E-04 6.96E-04 6.84E-04 6.88E-04 7.08E-04 7.03E-04 7.08E-04 Am-241 3.47E-04 3.46E-04 3.42E-04 3.78E-04 5.27E-04 3.03E-04 3.04E-04 5.09E-04 5.07E-04 2.89E-04 3.36E-04 Am-242m 9.3 1E-06 9.3 1E-06 9.27E-06 9.26E-06 8.41E-07 9.26E-06 9.26E-06 1.68E-06 1.25E-06 1.03E-06 6.10E-07 Am-243 1.25E-04 1.93E-04 2.56E-04 2.46E-04 I 1.85E-04 I 2.55E-04 I 2.47E-04 1.96E-04 2.12E-04 2.05E-04 2.06E-04 Nd-143 9.87E-04 9.15E-04 1.OOE-03 9.82E-04 1.08E-03 9.54E-04 9.53E-04 1.04E-03 1.06E-03 1.09E-03 1.11 E-03 Nd-145 8.54E-04 8.3 1E-04 9.08E-04 8.99E-04 9.62E-04 8.80E-04 8.99E-04 9.43E-04 9.80E-04 9.87E-04 1.OOE-03 Nd-148 4.88E-04 4.88E-04 5.46E-04 5.47E-04 5.50E-04 5.52E-04 5.59E-04 5.73E-04 5.78E-04 5.94E-04 6.07E-04 Cs-137 1.69E-03 1.62E-03 I 1.75E-03 1.82E-03 I 1.66E-03 1.77E-03 1.70E-03 1.77E-03 1.74E-03 1.76E-03 1.54E-03 Sm-147 2.26E-04 1.83E-04 1.86E-04 1.87E-04 2.36E-04 1.97E-04 1.82E-04 2.29E-04 2.49E-04 2.56E-04 2.52E-04 Sm-149 3.12E-06 3. 1OE-06 3.27E-06 3.20E-06 3.61E-06 3.83E-06 2.91E-06 3.37E-06 3.20E-06 3.43E-06 3.87E-06 Sm-150 3.59E-04 3.49E-04 3.76E-04 3.84E-04 4.14E-04 3.75E-04 3.63E-04 4.20E-04 4.54E-04 4.69E-04 4.54E-04 Sm-151 1.29E-05 1.27E-05 1.34E-05 1.25E-05 1.42E-05 1.26E-05 1.26E-05 1.33E-05 1.48E-05 :1.50E-05 1.56E-05 Sm-152 1.22E-04 1.21E-04 1.30E-04 1.27E-04 1.34E-04 1.32E-04 1.26E-04 1.30E-04 1.42E-04 1.44E-04 1.43E-04 Eu-151 6.95E-07 7.05E-07 7.95E-07 6.87E-07 6.69E-07 8.85E-07 8.50E-07 7.05E-07 7.49E-07 5.7 1E-07 6.64E-07 Eu-153 1.47E-04 1.56E-04 1.68E-04 1.68E-04 1.75E-04 1.71E-04 1.61 E-04 1.73E-04 1.84E-04 1.86E-04 1.90E-04 Eu-155 1.OIE-05 1.23E-05 1.32E-05 1.44E-05 1.27E-05 1.29E-05 1.28E-05 9.99E-06 1.03E-05 1.55E-05 9.85E-06
Table 3.3 Experimental results (g/g Uinitiai) for TMiI-1 samples from assembly NJ05YU (continued)
Sample ID A1B D2 B2 C1 DIA4 A2 C3 C2B B3J BIB D1A2 Burnup a (GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 Gd-155 8.24E-06 5.61E-06 6.56E-06 6.37E-06 1.40E-05 5.23E-06 6.68E-06 9.43E-06. 1.04E-05 1.O1E-05 1.02E-05 Mo-95 1.04E-03 9.22E-04 1.13E-03 1.I OE-03 1.09E-03 1.12E-03 L.O1E-03 1.1OE-03 1.13E-03 .1.15E-03 1.I1E-03 Tc-99 1.42E-03 9.78E-04 1.09E-03 1.08E-03 1.19E-03 1.08E-03 1.04E-03 1.36E-03 1.25E-03 1.32E-03 1.14E-03 Ru-101 1.12E-03 9.50E-04 1.21E-03 1.17E-03 L.IOE-03 1.16E-03 1.03E-03 1.17E-03 1.17E-03 1.19E-03 1.13E-03 Rh-103 5.97E-04 5.17E-04 6.30E-04 6.20E-04 6.04E-04 6.21E-04 5.49E-04 6.16E-04 6.22E-04 6.28E-04 6.18E-04 Ag-109 5.12E-05 4.67E-05 5.29E-05 5.37E-05 8.48E-05 5.98E-05 9.26E-05 6.55E-05 7.81E-05 4.41E-05 4.62E-05 a As reported in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisonsto SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 1, Rev. A (April 2002).
Table 3.4 Experimental techniques and uncertainties for TMI-1 samples measurements at GE-VNC Reported uncertainty RSD b Nuclide ID________ Method"
_____________(%)
at 95% confidence (%)
(%)__
U-234 TIMS 1.0 0.5 U-235 TIMS 1.0 0.5 U-236 TIMS 1.0 0.5 U-238 TIMS 1.0 0.5 Np-237 cu-spec 5.8 2.9 Pu-238 cl-spec 5.0 2.5 Pu-239 TIMS 1.2 0.6 Pu-240 TIMS 1.2 0.6 Pu-241 TIMS 1.2 0.6 Pu-242 TIMS 1.2 0.6 Am-241 TIMS, cl-spec 7.0 3.5 Am-242m TIMS, cl-spec 7.0 3.5 Am-243 TIMS, cl-spec 7.0 3.5 Cm-242 TIMS, cl-spec 20.0 10.0 Cm-243 TIMS, cl-spec 5.5 2.75 Cm-244 TIMS, cl-spec 5.5 2.75 Cm-245 TIMS, cl-spec 5.5 2.75 Cs-134 y-spec 3.5 1.75 Cs-137 Y-spec 3.5 1.75 Nd-143 TIMS 1.5 0.75 Nd-145 TIMS 1.5 0.75 Nd-146 TIMS 1.5 0.75 Nd-148 TIMS 1.5 0.75 Nd-150 TIMS 1.5 0.75 Sm- 147 TIMS 1.7 0.85 Sm- 149 TIMS 1.8 0.9 Sm-150 TIMS 1.7 0.85 Smi-151 TIMS 1.7 0.85 Sm-152 TIMS 1.7 0.85 Eu- 151 TIMS 1.7 0.85 Eu-153 TIMS 1.8 0.9 Gd-155 TIMS 2.7 1.35
' Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
6 Relative standard deviation; calculated here as half of the uncertainty reported at a 95% confidence level.
14
23 8 Table 3.5 Experimental results (g/g U) for TMI-I samples from assembly NJ070G Sample ID O13S7 012S4 012S6 01SI O13S8 012S5 OIS3 O1S2 Burnup a (GWd/MTU) 22.8 23.7 24.0 25.8 26.3 26.5 26.7 29.9 U-234 3.65E-04 3.55E-04 3.48E-04 3.48E-04 3.40E-04 3.34E-04 3.35E-04 3.25E-04 U-235 2.53E-02 2.5 1E-02 2.55E-02 2.35E-02 2,34E-02 2.33E-02 2.32E-02 2.05E-02 U-236 4.49E-03 4.58E-03 -4.68E-03 4.83E-03 4.89E-03 -4.93E-03 4.99E-03 5.34E-403 Pu-238 6.411E-05 6.6813-05 8.29E--05 7.67E-05 9.29E-05 9.40E-05 1.00E-04 116E-04 Pu-239 5.77E-03 5.79E-03 6.60E-03 5.811E-03 6.28E-03 6.411E-03 6.44E-03 5.98E-03 Pu-240 1.46E-03 1.48E-03 1.61 E-03 1.62E-03 1.73E-03 1,76E-03 1.83E-03 1,98E-03 Pu-241 7.04E-04 7.34E-04 8.54E-04 8.04E-04 8.79E-04 8.97E-04 9.56E-04 9.79E-04 Pu-242 1.54E-04 1.58E-04 1.76E-04 1.92E-04 2.16E-04 2.20E-04 2.36E-04 3.04E-04 Np-237 3.01E-04 3.23E-04 3.50E-04 3.24E-04 3.71E-04 3.72E-04 3.89E-04 4.23E-04 Am-241 1.73E-04 1.62E-04 1.47E-04 1.22E-04 2.16E-04 2.22E-04 1.83E-04 2.12E-04 Am-242m 3.36E-07 3.77E-07 3.97E-07 2.93E-07 4.99E-07 5.18E-07 4.50E-07 4,53E-07 Am-243 1.71E-05 1.80E-05 1.7613-05 1.60E-05 2.85E-05 2.96E-05 2.74E-05 3.75E-05 Cm-242' 7.45E-09 1.97E-08 2.OOE-08 1.89E-08 1.25E-08 1.20E-08 2.90E1-08 1.75E-08 Cm-243 5.97E-08 6.36E-08 6.99E-08 5.50E-08 101IE-07 1.07E-07 1.04E-07 1.25E-07 Cm-244 2.62E-06 2.89E-06 3.22E-06 2.66E-06 5.23E-06 5.511E-06 5.32E-06 7.68E-06 Cm-245 1.14E-07 1.24E-07 1.67E-07 1.19E-07 2.74E-07 2.90E-07 2.81E-07 4.02E-07 Nd- 143 7.411E-04 7.5 1E-04 7.66E-04 7.95E-04 8.11E--04 8.16E-04 8.28E-04 8.92E-04 Nd-145 5.5 1E-04 5.59E-04 5.64E-04 6.OOE-04 6.08E-04 6.11 E-04 6.2 1E-04 6.87E-04 Nd-146 5.04E--04 5.12E-04 5.26E-04 5.56E-04 5.72E-04 5.76E-04 5.87E-04 6.58E-04 Nd-148 2.77E-04 2.8 11E-04 2.88E-04 3.05E-04 3.12E-04 3.14E-04 3.2 1E--04 3.58E-04 Nd-150 1.25E-04 1.26E-04 1.3 11E-04 1.3813-04 1.42E-04 1.43E-04 1.47E-04 1.6413-04 Cs-134 1.76E-05 2.22E-05 2.44E-05 2.51E-05 2.27E-05 2.276-05 2.90E-05 2.76E-05 Cs- 137 8.92E-04 9.05E-04 9.18E-04 9.71E-04 1.01E-03 I.OOE-03 1.031E-03 1.17E-03 Sm-147 1.86E-04 1.81E--04 1.79E-04 1.91E-04 1.99E-04 2.01E-04 1.94E-04 2.20E-04 Sm-149 4.23E-06 4.32E-06 4.73E-06 4.32E-06 4.42E-06 4.44E-06 4.72E-06 4.3613-06 Sm-I150 2.06E-04 2.11E-04 2.17E-04 2.30E-04 2.38E-04 2.41 E-04 2.47E-04 2.78E-04 Smi-151 1.35E-05 1.38E-05 1.58E-05 1.36E-05 1.51E-05 1.51E-05 1,53E-05 1.47E-05 Sm- 152 8.47E-05 8.62E-05 8.41E-05 9,23E-05 9.19E-05 9.27E-05 9.54E-05 1.07E-04 Eu- 151 4.48E-07 4.29E-07 4.89E-07 4,15E-07 4.99E-07 5.02E-07 4.61 E-07 4.74E-07 Eu-153 7.13E--05 7.37E-05 7.69E-05 8.05E--05 8.6 11E-05 8.65E-05 8.80E-05 1.01E-04 Gd-155 2. 1OE-06 2.03E-06 2.33E-06 2.46E-06 2.70E-06 2.68E-06 2.82E-06 3.09E-06 a As reported in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisons to SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 I, Rev. A (April 2002).
h Average of the two values measured by TIMS and y-spectrometry.
15
Table 3.6 Experimental results (g/g Uinitial) for TMI-1 samples from assembly NJ070G Sample ID O13S7 012S4 012S6 01SI 013S8 012S5 O1S3 OIS2 Burnupa (GWd/MTU) 22.8 23.7 24.0 1 25.8 26.3 26.5 26.7 29.9 U-234 3.43E-04 3.34E-04 3.26E-04 3.27E-04 3.19E-04 3.13E-04 3.14E-04 3.04E-04 U-235 2.38E-02 2.36E-02 2.39E-02 2.2 !E-02 2.19E-02 2.18E-02 2.17E-02 1.92E-02 U-236 4.22E-03 4.30E-03 4.39E-03 4.53E-03 4.58E-03 4.62E-03 4.67E-03 5.OOE-03 Pu-238 6.03E-05 6.28E-05 7.77E-05 7.20E-05 8.71E-05 8.8 1E-05 9.37E-05 1.09E-04 Pu-239 5.42E-03 5.44E-03 6.19E-03 5.45E-03 5.89E-03 6.0 1E-03 6.03E-03 5.60E-03 Pu-240 1.37E-03 1.39E-03 1.51E-03 1.52E-03 1.62E-03 1.65E-03 1.71E-03 1.85E-03 Pu-241 6.62E-04 6.90E-04 8.0 !E-04 7.55E-04 8.24E-04 8.41 E-04 8.95E-04 9.16E-04 Pu-242 1.45E-04 1.48E-04 1.65E-04 1.80E-04 2.03E-04 2.06E-04 2.21 E-04 2.85E-04 Np-237 2.83E-04 3.04E-04 3.28E-04 3.04E-04 3.48E-04 3.49E-04 3.64E-04 3.96E-04 Am-241 1.63E-04 1.52E-04 1.38E-04 1.1 5E-04 2.03E-04 2.08E-04 1.71E-04 1.98E-04 Am-242m 3.16E-07 3.54E-07 3.72E-07 2.75E-07 4.68E-07 4.86E-07 4.21 E-07 4.24E-07 Am-243 1.61E-05 1.69E-05 1.65E-05 1.50E-05 2.67E-05 2.77E-05 2.57E-05 3.5 1E-05 Cm-242' 7.OOE-09 1.85E-08 1.88E-08 1.77E-08 0.OOE+00 1.12E-08 2.72E-08 i .64E-08 Cm-243 5.61E-08 5.98E-08 6.56E-08 5.16E-08 9.47E-08 1.OOE-07 9.74E-08 1.17E-07 Cm-244 2.46E-06 2.72E-06 3.02E-06 2.50E-05 4.90E-06 5.16E-06 4.98E-06 7.19E-06 Cm-245 1.07E-07 1.17E-07 1.57E-07 1.12E-07 2.57E-07 2.72E-07 2.63E-07 3.76E-07 Nd- 143 6.97E-04 7.06E-04 7.18E-04 7.46E-04 7.60E-04 7.65E-04 7.76E-04 8.35E-04 Nd-145 5.18E-04 5.25E-04 5.29E-04 5.63E-04 5.70E-04 5.73E-04 5.82E-04 6.43E-04 Nd-146 4.74E-04 4.8 1E-04 4.93E-04 5.22E-04 5.36E-04 5.40E-04 5.50E-04 6.16E-04 Nd-148 2.60E-04 2.64E-04 2.70E-04 2.86E-04 2.93E-04 2.94E-04 3.01 E-04 3.35E-04 Nd-150 I. 18E-04 1.1 8E-04 1.23E-04 1.30E-04 1.33E-04 1.34E-04 1.38E-04 1.53E-04 Cs-134 1.65E-05 2.09E-05 2.29E-05 2.36E-05 2.13E-05 2.13E-05 2.72E-05 2.58E-05 Cs-137 8.38E-04 8.50E-04 8.61 E-04 9.12E-04 9.47E-04 9.37E-04 9.65E-04 1.IOE-03 Sm-147 1.75E-04 1.70E-04 1.68E-04 I .79E-04 1.87E-04 1.88E-04 1.82E-04 2.06E-04 Sm-149 3.98E-06 4.06E-06 4.44E-06 4.06E-06 4.14E-06 4.16E-06 4.42E-06 4.08E-06 Sm-150 1.94E-04 1.98E-04 2.04E-04 2.16E-04 2.23 E-04 2.07E-04 2.3 1E-04 2.60E-04 Sm-151 1.27E-05 1.30E-05 1.48E-05 1.28E-05 1.42E-05 1.42E-05 1.43E-05 1.38E-05 Sm-152 7.96E-05 8. 1E-05 7.89E-05 8.66E-05 8.62E-05 8.69E-05 8.94E-05 1.OOE-04 Eu-151 4.2 1E-07 4.03E-07 4.59E-07 3 .90E-07 4.68E-07 4.71 E-07 4.32E-07 4.44E-07 Eu-153 6.70E-05 6.93E-05 7.21 E-05 7.56E-05 8.07E-05 8.11E-05 8.24E-05 9.45E-05 Gd-155 1.97E-06 1.91E-06 2.19E-06 2.3 1E-06 2.53E-06 2.5 1E-06 2.64E-06 2.89E-06 a As reported in3. M. Scaglione, Three Mile Island Unit! RadiochemicalAssay Comparisonsto SAS2H Calculations, Yucca
" As reported in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisonsto SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 I, Rev. A (April 2002).
h Average of the two values measured by TIMS and y-spectrometry.
16
3.2 CALVERT CLIFFS SAMPLES The three samples from the Calvert Cliffs reactor considered in this report belonged to a fuel rod of a 14 x 14 fuel assembly of Combustion Engineering (CE) design. The burnup of the samples covers the range 27 to 44 GWd/MTU. The samples were identified as 87-81, 87-72, and 87-63.
3.2.1 PNNL Measurements The measurements at PNNL were performed by using the following main spectrometric methods:
" y-spectrometry for 137 Cs;
" ca-spectrometry for 24 1Am and 237Np;
" P3-spectrometry for 99 Tc and 90Sr;
" isotope dilution mass spectrometry (ID-MS) for neodymium, uranium, and plutonium nuclides, using a calibrated triple spike of '5 0Nd, 233U, and 242pu;
" mass spectrometry (MS) after elemental separation of cesium for 133Cs and 135Cs;
" ICP-MS measurements relative to 143Nd and 145Nd for lanthanide elements: samarium, europium, gadolinium.
The lanthanide measurements were carried out by ICP-MS in general without previous chemical separation into individual elements. Therefore, there was an interference issue for data corresponding to nuclides with mass numbers 147 (Pm, Sm), 150 (Nd, Sm), 151 (Sm, Eu), and 155 (Eu, Gd). The data corresponding to these four mass numbers were adjusted by PNNL based on calculations in order to infer information for individual isotopes. The PNNL lanthanide data are not considered in this report for code validation purposes because of the large dependence of the reported measurement data on additional calculated results.
Isotopic measured concentrations were reported as g/g fuel, g/MTU, or Ci/g fuel depending on the reporting reference and the nuclide under consideration. 0'""'1 3' 5 A summary of the measured nuclides, methods used, and reported measurement uncertainties are summarized in Table 3.7. The magnitude of the experimental 145- errors varies with the method and the nuclide. For example, it is less than 1% for 143Nd and Nd and 1.6% for uranium and plutonium isotopes. These uncertainties, except for lanthanides, represent one relative standard deviation that is based on experience at the PNNL experimental facility.
For some isotopes, the measurement errors were not explicitly specified (1 33 Cs, 144Nd, 146Nd, and 14 8Nd).
It was stated though that the measurements for all neodymium nuclides provided very good quality data, as a chemical separation for neodymium was performed prior to ID-MS. The measurement errors reported for lanthanides were inferred by PNNL 11based on additional lanthanide measurement data on sample identified as 87-81: ICP-MS measurements by PNNL and Los Alarnos National Laboratory (LANL) and MS with luminescent analysis (LA) by KRI.
The measured nuclide concentrations in g/g fuel as reported in Refs. 9, 10, and 15 are presented in Table 3.8. Data for 13 4Cs was found only in Ref. 15. Data for 90Sr is shown in Table 3.8 both in Ci/g fuel, as reported in Ref. 9, and in g/g fuel, calculated as:
M90 s, (g / gfuet AM (3-4) 2 NA where A = 3.7 x 1010 ms*o(Ci/g fuel) reported radioactivity in units of Bq/s/g fuel
= 7.62759 x 1010 s-1 decay constant (half-life = 28.79 years) 17
M = 89.99 atomic mass NA = 6.022 x 1023 mol"1 Avogadro's number The PNNL data are also shown in units of g/g Uinitial in Table 3.9; the unit conversion was performed as:
m(g / g,,, )= 1.1345m(g / g,,,, (3-5) 3.2.2 KRI Measurements Additional lanthanide analysis was performed for the same three samples at KRI in St. Petersburg, Russia.10 The measurements included:
" Chemical separation of rare earth elements and transuranics followed by chemical separation of lanthanides into individual elements;
- ID-MS for neodymium and gadolinium isotopes using a spike of 42 ' Nd and "6'Gd;
" Luminescent analysis-laser-induced fluorometry for absolute measurement of europium and samarium content in the sample; the content was determined by comparison of the sample luminescence intensity with that of standard solutions containing known quantities of europium and samarium;
" MS for europium and samarium nuclides to determine relative isotope ratios;
- y-spectrometry for 154Eu and ...Eu.
As chemical separations were performed, the KRI measurements were not subject to mass interference from different elements, as it was the case with the PNNL measurements. The experimental results were reported by KRI as the ratio of nuclide mass to "45Nd mass or as the nuclide mass percentage relative to the corresponding element total mass. The measured nuclides and corresponding measurement error range are shown in Table 3.10. The reported results of the radiochemical analyses performed by KRI, as well as the reported experimental errors for each nuclide and sample are presented in Table 3.11. The concentration values shown in Table 3.11 for ... Sm, 'Gd, 'Gd, 'Gd, and 118Gd for sample 87-81 in 45 g/g1 Nd units were taken from Ref. 16. The values provided in Ref. 10 for these isotopes were different, as follows: 0.115 +/- 0.005, 0.0108 +/- 0.0004, 0.110 +/- 0.0002, <0.00007, and 0.0236 +/- 0.0005, respectively.
The values shown in italics in Table 3.11 for nuclide concentrations in g/g145 Nd units were derived based on the available data. For example, concentration for each of the measured gadolinium nuclides except for 115Gd, for which concentration relative to 145Nd was available, was calculated as:
c,,a~(g/gl45Nd)= mrGa(g/gGd) d(gig aNd) (3-6)
M 155a tg / gt~r)
In the current report, the experimental data reported by KRI are also expressed in units of g/g Uinitial (see Table 3.13) in order to be used in a consistent comparison with other sets of data from different experiments. As no absolute concentration values were reported by KRI, the 145Nd values provided by PNNL were used to renormalize the KRI experimental results. A study of the KRI and PNNL data, both expressed relative to "5Nd concentration, showed that the difference in data for those neodymium nuclides measured in all three samples at both experimental facilities were within one standard deviation, as shown in Table 3.12. Therefore, the use of the 145Nd concentrations determined at PNNL to renormalize the KRI data would not introduce additional large uncertainties. The uncertainties shown in Tables 3.12 and 3.13 include the error propagation due to renormalization.
18
3.2.3 Experimental Data Used for Code Validation The two sets of PNNL and KRI measured data are combined into one set for code validation purposes.
The combined set of experimental data is presented in Table 3.14. As previously mentioned, the samarium, europium, and gadolinium data reported by PNNL are not included in this set in order to minimize the associated uncertainties because these data were derived by adjusting the measured isotope ratios using calculated values; for these nuclides, the KRI data are used. The neodymium data in Table 3.14 correspond to PNNL measurements. The measured concentrations for the 154Eu and 55 ' Eu isotopes shown in Table 3.14 were obtained by combining the two values shown in Table 3.13, obtained by ID-MS and y-spectrometry, respectively, at KRI.
Table 3.7 Experimental techniques and uncertainties for Calvert Cliffs samples-PNNL data Nuclide ID Method" RSD U-234 ID-MS 1.6 U-235 ID-MS 1.6 U-236 ID-MS 1.6 U-238 ID-MS 1.6 Pu-238 ID-MS 1.6 Pu-239 ID-MS 1.6 Pu-240 ID-MS 1.6 Pu-241 ID-MS 1.6 Pu-242 ID-MS 1.6 Np-237 ca-spec 1.9 Am-241 a-spec 4.9 Cs-133 MS NA Cs-134c NA NA Cs-135 MS 14.0 Cs- 137 y-spec 3.5 Nd- 143 ICP-MS < 1.0 Nd- 144 ICP-MS NA Nd- 145 ICP-MS < 1.0 Nd-146 ICP-MS NA Nd- 148 ICP-MS NA Nd-150 ICP-MS NA Sm-147 ICP-MS 4.0 Sm-149 ICP-MS 18.0 Sm-150 ICP-MS 2.0 Sm-151 ICP-MS 7.0 Sm-152 ICP-MS 3.0 Eu-151 ICP-MS NA Eu-153 ICP-MS 2.0 Eu-155 ICP-MS 29.0 Gd-155 ICP-MS 29.0 Sr-90 0-spec 5.7 Tc-99 13-spec 3.5
' Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
6Relative standard deviation.
Measured value only reported in 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic CompositionAnalyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
19
Table 3.8 Experimental results (g/g fuel) for Calvert Cliffs samples-PNNL data Sample ID 87-81 87-72 87-63 BurnupI (GWd/MTU) 27.35 37.12 44.34 U-234 1.60E-04 1.40E-04 1.20E-04 U-235 8.47E-03 5.17E-03 3.54E-03 U-236 3.14E-03 3.53E-03 3,69E-03 U-238 8.43E-01 8.33E-0 I 8.25E-01 Pu-238 1.01E-04 1.89E-04 2.69E-04 Pu-239 4.26E-03 4.36E-03 4,36E-03 Pu-240 1.72E-03 2.24E-03 2.54E-03 Pu-241 6.8 1E-04 9.03E-04 1.02E-03 Pu-242 2.89E-04 5.76E-04 840E-04 Np-237 2.68E-04 3.56E-04 4.68E-04 Am-241 2.49E-04 3.43E-04 3.8 1E-04 Cs-133 8.50E-04 1.09E-03 1.24E-03 Cs-134 1.00E-05 2.OOE-05 3.OOE-05 Cs-135 3.60E-04 4.OOE-04 4.30E-04 Cs-137 7.70E-04 1.04E-03 1.25E-03 Nd-143 6.13E-04 7.16E-04 7.63E-04 Nd-144 9.43E-04 1.34E-03 1.64E-03 Nd-145 5. 1OE-04 6.53E-04 7.44E-04 Nd- 146 4.90E-04 6.82E-04 8.30E-04 Nd-I148 2.65E-04 3.59E-04 4.28E-04 Nd-150 1.24E-04 1.72E-04 2.08E-04 Sm- 147 1.90E-04 2.18E-04 2.30E-04 Sm-148 1.06E-04 1.64E-04 2.22E-04 Sm-149 2.90E-06 3.OOE-06 4.70E-06 Sm-150 2.07E-04 2.71 E-04 3.61 E-04 Sm-151 8.60E-06 8.60E-06 9.OOE-06 Sm-152 8.70E-06 1.04E-04 1.21 E-04 Eu-151 7.OOE-07 7.OOE-07 8.OOE-07 Eu- 153 7.90E-05 1.09E-04 1.48E-04 Eu-155 2. 1OE-06 3.30E-06 4.50E-06 Gd-155 2.50E-06 3.90E-06 5.30E-06 Tc-99 5.60E-04 7.20E-04 7.80E-04 Sr-90' 4.59E-02 5.90E-02 6.58E-02 Sr-90 3.33E-04 4.28E-04 4.77E-04
'As reported in 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE Systemfor PWR Spent FuelIsotopic CompositionAnalyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
"In Ci/g fuel, as reported in R.J Guenther et al. Characterizationof LWR Spent FuelMCC-Approved Testing Material TM- 104, PN L-5109-104 (1991).
20
Table 3.9 Experimental results (g/g Uinitial) for Calvert Cliffs samples-PNNL data Sample ID 87-81 87-72 87-63 Burnup a (GWd/MTU) 27.35 37.12 44.34 U-234 1.82E-04 1.59E-04 1.36E-04 U-235 9.61E-03 5.87E-03 4.02E-03 U-236 3.56E-03 4.OOE-03 4.19E-03 U-238 9.56E-01 9.45E-01 9.36E-01 Pu-238 1.15E-04 2.14E-04 3.05E-04 Pu-239 4.83E-03 4.95E-03 4.95E-03 Pu-240 1.95E-03 2.54E-03 2.88E-03 Pu-241 7.73E-04 1.02E-03 1.16E-03 Pu-242 3.28E-04 6.53E-04 9.53E-04 Np-237 3.04E-04 4.04E-04 5.31E-04 Am-241 2.82E-04 3.89E-04 4.32E-04 Cs-133 9.64E-04 1.24E-03 1.41E-03 Cs-134 1.13E-05 2.27E-05 3.40E-05 Cs-135 4.08E-04 4.54E-04 4.88E-04 Cs-137 8.74E-04 1.1 8E-03 1.42E-03 Nd- 143 6.95E-04 8.12E-04 8.66E-04 Nd- 144 1.07E-03 1.52E-03 1.86E-03 Nd- 145 5.79E-04 7.41E-04 8.44E-04 Nd-146 5.56E-04 7.74E-04 9.42E-04 Nd-148 3.01E-04 4.07E-04 4.86E-04 Nd-150 1.41E-04 1.95E-04 2.36E-04 Sm- 147 2.16E-04 2.47E-04 2.61 E-04 Sm-148 1.20E-04 1.86E-04 2.52E-04 Sm- 149 3.29E-06 3.40E-06 5.33E-06 Sm-150 2.35E-04 3.07E-04 4.1OE-04 Sm-151 9.76E-06 9.76E-06 1.02E-05 Sm- 152 9.87E-06 1.18E-04 1.37E-04 Eu-151 7.94E-07 7.94E-07 9.08E-07 Eu-153 8.96E-05 1.24E-04 1.68E-04 Eu-155 2.38E-06 3.74E-06 5.11 E-06 Gd-155 2.84E-06 4.42E-06 6.01 E-06 Tc-99 6.35E-04 8.17E-04 8.85E-04 Sr-90 3.78E-04 4.86E-04 5.4 1E-04
'As reported in 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
21
Table 3.10 Experimental techniques and uncertainties for Calvert Cliffs samples-KRI data Method" RSD b Nuclide ID ________ (%)
Nd-143 ID-MS 0.7-1.9 Nd-145 ID-MS NA Sm-147 MS, LA 2.5-3.3 Sm-149 MS, LA 7.4-20.0 Sm-150 MS, LA 2.3-4.2 Sm- 151 MS, LA 3.2-4.7 Sm-152 MS, LA 2.7-4.4 Sm-154 MS, LA 5.7 Eu-151 MS, LA 9.7 Eu-154 MS, LA, y-spec 5.3-8.6 Eu-155 MS, LA, y-spec 2.7-16.7 Gd-155 ID-MS 0.2-3.3 Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
6 Relative standard deviation.
22
Table 3.11 Experimental results for Calvert Cliffs samples-KRI dataa Sample ID 87-81 87-72 87-63 Burnupb (GWd/MTU) 27.35 37.12 44.34 ID-MS data ng/g s gig element
- g/g element 145N sample (%) (%)
M (%) (%)(gig4Nd (%)
Nd-142 287 3 1.0 0.039 0.001 2.6 1.25 0.077 0.9 0.76 0.048 1.9 Nd-143 9100 40 0.4 1.218 0.008 0.7 18.15 1.120 0.010 0.9 16.37 1.040 0.020 1.9 Nd-144 14060 55 0.4 1.882 0.011 0.6 33.93 2.084 0.9 35.34 2.235 1.9 Nd-145 7470 32 0.4 1.000 16.28 1.000 15.81 1.000 Nd-146 7270 35 0.5 0.973 0.006 0.6 17.12 1.052 0.9 17.87 1.130 1.9 Nd-148 3910 24 0.6 0.523 0.004 0.8 8.96 0.550 0.9 9.24 0.584 1.9 Nd-150 1850 150 8.1 0.248 0.020 0.8 4.32 0.265 0.9 4.59 0.290 1.9 Sm-147 2975 100 3.4 0.398 0.013 3.3 30.57 0.365 0.009 2.5 28 0.365 0.012 3.3 Sm-148 1290 20 1.6 0.173 0.003 1.7 18.3 0.218 2.3 20.39 0.226 3.2 Sm-149 35 5 14.3 0.005 0.001 20.0 0.22 0.0025 0.0003 12.0 0.41 0.0054 0.0004 7.4 Sm-150 2696 110 4.1 0.361 0.015 4.2 32.89 0.391 0.009 2.3 33.06 0.431 0.014 3.2 Sm-151 96 35 36.5 0.013 0.005 38.5 1.08 0.0127 0.0004 3.1 0.97 0.0127 0.0006 4.7 Sm-152 1160 60 5.2 0.155 0.005 3.2 12.56 0.148 0.004 2.7 12.05 0.157 0.006 3.8 Sm-154 393 20 5.1 0.053 0.003 5.7 4.32 0.051 2.3 5.12 0.067 3.2 Eu-151 23 2 8.7 0.0031 0.0003 9.7 0.74 0.00141 2.7 1.91 0.00407 2.8 Eu-152 11 10 90.9 0.0002 0.0002 100.0 0.04 0.00008 2.7 0.25 0.00053 2.8 Eu-153 1100 20 1.8 0.1472 0.0027 1.8 91.98 0.17551 2.7 90.25 0.19212 2.8 Eu-154 79 7 8.9 0.0105 0.0009 8.6 6.26 0.01195 2.7 6.58 0.01401 2.8 Eu-155 13 2 15.4 0.0018 0.0003 16.7 0.98 0.00187 0.00005 2.7 1.01 0.00215 0.00006 2.8 Gd-154 176 4 2.3 0.0236 0.0005 2.1 13.28 0.0202 3.0 13.27 0.0237 3.4 Gd-155 81 2 2.5 0.0108 0.0004 3.7 6.58 0.0100 0.0003 3.0 6.62 0.0118 0.0004 3.4 Gd-156 825 16 1.9 0.1100 0.002 1.8 65.10 0.0989 3.0 63.20 0.1127 3.4 Gd-157 6 4 75.0 <0.00007 1.84 0.0028 3.0 3.24 0.0058 3.4 Gd-158 180 4 2.2 0.0241 0.0005 2.1 13.20 0.0201 3.0 13.70 0.0244 3.4 Gd-160 19 2 10.5 0.0025 0.0003 12.0 y-spec data Eu-154 0.0117 0.0006 5.1 0.0119 0.0007 5.9 Eu-155 0.00182 0.00009 4.9 0.00209 0.00012 5.7 Values shown in italics are inferred based on available data. All other values are as given in M. C. Brady-Raap and R. J. Talbert, Compilationof Radiochemical Analyses of Spent Nuclear Fuel Samples, PNNL-13677, Pacific Northwest National Laboratory, Richland, Washington (September 2001); A. A. Rimski-Korsakov, A. V. Stepanov, A. D. Kirikov, RadiochemicalAnalysis of Spent Reactor Fuel Sample-Report of Results, V.G.Khlopin Institute, St. Petersburg, Russia, Communication to PNNL (1993).
b As provided in 0. W. Herman, S. M. Bowman, M.C. Brady and C. V. Parks, Validation of the SCALESystemfor PWR Spent FuellsotopicComposition Analyses, ORNIJTM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
Table 3.12 Comparison of PNNL and KRI data (relative to 145Nd)
Sample Nuclide PNL data KRI data Difference ID ID g/g 14'Nd ar g/g 145Nd
- g/g 145Nd a 87-81 Nd-143 1.202 0.017 1.218 0.008 -0.016 0.019 87-72 Nd-143 1.096 0.017 1.120 0.010 0.024 0.020 87-63 Nd-143 1.026 0.015 1.040 0.020 -0.014 0.025 24
Table 3.13 Experimental results (g/g Uinitial) for Calvert Cliffs samples-KRI data Sample ID 87-81 87-72 87-63 Burnup '
(GWd/MTU)
ID-MS data
___D-MS___I daa g/g Uj,,j181
_/gUiita IRSD 27.35
(%)
b 37.12 RSD gI/g UinitiaI (%)
g gig Ui/
44.34 t1 RSD
(%)
Nd-142 2.26E-05 3.2 5.69E-05 2.1 4.06E-05 2.7 Nd- 143 7.05E-04 2.0 8.30E-04 2.1 8.78E-04 2.7 Nd- 144 I1.09E-03 2.0 1.54E-03 2.1 1.89E-03 2.7 Nd-145 5.79E-04 1.9 7.4 1E-04 1.9 8.44E-04 1.9 Nd-146 5.63E-04 2.0 7. 79E-04 2.1 9.54E-04 2.7 Nd- 148 3.03E-04 2.0 4. 08E-04 2.1 4.93E-04 2.7 Nd-150 1.44E-04 8.3 1.97E-04 2.1 2.45E-04 2.7 Sm-147 2.30E-04 3.8 2.70E-04 3.1 3.08E-04 3.8 Sm-148 1.00E-04 2.6 1.61E-04 3.0 2.24E-04 3.8 Sm-149 2.90E-06 20.1 1.85E-06 12.1 4.56E-06 7.6 Sm-150 2.09E-04 4.6 2.90E-04 3.0 3.64E-04 3.7 Sm- 151 7.53E-06 38.5 9.4 1E-06 3.6 1.07E-05 5.1 Sm-152 8.97E-05 3.7 1.1OE-04 3.3 1.33E-04 4.2 Sm-154 3.07E-05 6.0 3.81E-05 3.0 5.63E-05 3.8 Eu- 151 1.79E-06 9.9 1. 05E-06 3.3 3.43E-06 3.4 Eu- 152 1.16E-07 100.0 5. 66E-08 3.3 4.49E-07 3.4 Eu-153 8.52E-05 2.6 1.30E-04 3.3 1.62E-04 3.4 Eu-154 6.08E-06 8.8 8. 85E-06 3.3 1.18E-05 3.4 Eu-155 1.04E-06 16.8 1.39E-06 3.3 1.81E-06 3.4 Gd-154 1.37E-05 2.8 1.50E-05 3.6 2. OOE-05 3.9 Gd-155 6.25E-06 4.2 7.41E-06 3.6 9.96E-06 3.9 Gd-156 6.37E-05 2.6 7.33E-05 3.6 9.5JE-05 3.9 Gd-158 1.40E-05 2.8 1.49E-05 3.6 2. 06E-05 3.9 Gd-160 1.45E-06 12.1 y-spec data Eu-154 8.67E-06 f 5.4 O1.E-05 6.2 Eu-155 1.35E-06 5.3 1.76E-06 6.0
'As reported in 0. W.Hermann, S. M. Bowman, M. C. Brady and C. V.Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic CompositionAnalyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
6 Relative standard deviation.
25
Table 3.14 Experimental results (g/g Uinijtial) for Calvert Cliffs samples used for code validation a Sample ID 87-81 I 87-72 87-63 Burnup (GWd/MTU) 27.35 37.12 44.34 RD c ERSD RSD Nuclide ID g/g UinitaI (%) gig UinitiaI % ,/g Ulnitial U-234 1.82E-04 2.3 1.599-04 2.3 1.36E-04 2.3 U-235 9.61E-03 2.3 5.87E-03 2.3 4.02E-03 2.3 U-236 3.56E-03 2.3 4.OOE-03 2.3 4.19E-03 2.3 U-238 9.56E-0 1 2.3 9.45E-01 2.3 9.3 6E-0 1 2.3 Pu-238 1. 15E-04 2.3 2.14E-04 2.3 3.05E-04 2.3 Pu-239 4.831E-03 2.3 4.95E-03 2.3 4.95E-03 2.3 Pu-240 1.95E-03 2.3 2.54E-03 2.3 2.88E-03 2.3 Pu-241 7.73E-04 2.3 1.02E-03 2.3 1.16E-03 2.3 Pu-242 3.28E-04 2.3 6.53E-04 2.3 9.53E-04 2.3 Np-237 3.04E-04 2.5 4.04E-04 2.5 5.31E-04 2.5 Am-241 2.82E-04 5.2 3.89E-04 5.2 4.32E-04 5.2 Cs-133 9.64E-04 1.24E-03 1.41E-03 Cs- 134 1.13E-05 2.27E-05 3.40E-05 Cs-135 4.08E-04 14.1 4.54E-04 14.1 4.88E-04 14.1 Cs-137 8.74E-04 3.8 1.18E-03 3.8 1.42E-03 3.8 Nd- 143 6.95E-04 1.9 8.12E-04 1.9 8.66E-04 1.9 Nd-144 1.07E-03 NA 1.52E-03 NA 1.86E-03 NA Nd- 145 5.79E-04 1.9 7.41 E-04 1.9 8.44E-04 1.9 Nd-146 5.56E-04 NA 7.74E-04 NA 9.42E-04 NA Nd-148 3.01 E-04 NA 4.07E-04 NA 4.86E-04 NA Nd-150 1.4 1E-04 NA 1.95E-04 NA 2.36E-04 NA Sm-147 2.30E-04 3.8 2.70E-04 3.1 3.08E-04 3.8 Sm-148 1.OOE-04 2.6 1.61 E-04 3.0 2.24E-04 3.8 Sm-149 2.90E-06 20.1 1.85E-06 12.1 4.56E-06 7.6 Sm-150 2.09E-04 4.6 2.90E-04 3.0 3.64E-04 3.8 Sm- 151 7.53E-06 38.5 9.4 1E-06 3.7 1.07E-05 5.1 Sm-152 8.97E-05 3.7 1.1OE-04 3.3 1.33E-04 4.3 Sm-154 3.07E-05 6.0 3.8 1E-05 3.0 5.63E-05 3.8 Eu- 151 1.79E-06 9.9 1.051E-06 3.3 3.43E-06 3.4 Eu- 152 1.16E-07 100.0 5.66E-08 3.3 4.49E-07 3.4 Eu-153 8.52E-05 2.6 1.30E-04 3.3 1.62E-04 3.4 Eu-154 6.08E-06 8.8 8.76E-06 6.4 1.09E-05 7.0 Eu-155 1.04E-06 16.8 1.37E-06 6.2 1.79E-06 6.9 26
Table 3.14 Experimental results (g/g Uinitiai) for Calvert Cliffs samples used for code validation (continued)
Sample ID 87-81 87-72 87-63 Burnup (GWd/MTU) 27.35 37.12 44.34 Nuclide 1D g/g RSD RSD RSD Uint %Ual MOo gigginiti (%) g ig Gd-154 1.37E-05 2.8 1.50E-05 3.6 2.OOE-05 3.9 Gd-155 6.25E-06 4.2 7.41E-06 3.6 9.96E-06 3.9 Gd-156 6.37E-05 2.6 7.33E-05 3.6 9.51E-05 3.9 Gd-158 1.40E-05 2.8 1.49E-05 3.6 2.06E-05 3.9 Gd-160 1.45E-06 12.1 Tc-99 6.35E-04 5.9 8.17E-04 5.9 8.85E-04 5.9 Sr-90 3.78E-04 3.8 4.86E-04 3.8 5.41E-04 3.8
' Data for Sm, Eu, and Gd isotopes correspond to KRI measurements. The other isotope data correspond to PNNL measurements.
6 As reported in 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM- 12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
' Relative standard deviation.
27
3.3 TAKAHAMA-3 SAMPLES The 16 samples that were measured at JAERI were cut from three fuel rods irradiated in the Takahama-3 reactor operated in Japan. After sample cutting, the elements were separated by using exchange separation methods. The following experimental techniques were used to determine the nuclide concentrations:17
" ID-MS o major actinides: uranium, plutonium o lanthanides: neodymium, samarium
" a-spectrometry plus MS o americium, curium
" y-spectrometry 06 34 13 7 54 25 O l Ru, 1 Cs , Cs, 144Ce, 1 Eu, 1 Sb
" cc-spectrometry 0 237Np A summary of the nuclides measured, methods used and corresponding experimental uncertainties are presented in Table 3.15. The reported experimental uncertainties were not specific for each sample measurement, but were typical values based on previous measurement experience at JAERI. Not all nuclides shown in the table were measured in each of the samples. The reported experimental RSD is less than 0.5% for all measured plutonium, samarium, and neodymium isotopes, as well as for 2 35 U and 238U.
For minor actinides measured by MS and cc-spectrometry the experimental errors are larger, in the 2 to 10% range, The nuclides determined through y-spectrometry have measurement errors between 3 and 10%.
The experimental results of the radiochemical analyses for the 16 samples from fuel rods identified as SF95, SF96, and SF97 were reported as g/MTU initial. These data were reported at discharge time, except for samarium nuclides in samples from rod SF97 that were reported at 3.96 years after discharge.
The measured data are presented in Tables 3.16-3.18 in g/g Uinitial.
28
Table 3.15 Experimental techniques and uncertainties for Takahama-3 samples Method ' RSD b Nuclide ID (%)
U-234 ID-MS < 1.0 U-235 ID-MS < 0.1 U-236 ID-MS < 2.0 U-238 ID-MS <0.1.
Pu-238 ID-MS < 0.5 Pu-239 ID-MS < 0.3 Pu-240 ID-MS < 0.3 Pu-241 ID-MS < 0.3 Pu-242 ID-MS < 0.3 Np-237 a-spec < 10.0 Am-241 MS, a-spec < 2.0 Am-242m MS, a-spec < 10.0 Am-243 MS, a-spec < 5.0 Cm-242 MS, a-spec < 10.0 Cm-243 MS, a-spec < 2.0 Cm-244 MS, a-spec < 2.0 Cm-245 MS, a-spec < 2.0 Cm-246 MS, a-spec < 5.0 Cs-134 y-spec < 3.0 Cs-137 y-spec < 3.0 Ce-144 y-spec < 10.0 Nd-142 ID-MS < 0.1 Nd-143 ID-MS < 0.1 Nd-144 ID-MS < 0.1 Nd-145 ID-MS < 0.1 Nd-146 ID-MS < 0.1 Nd-148 ID-MS < 0.1 Nd-150 ID-MS < 0.1 Sm-147 ID-MS < 0.1 Sm-148 ID-MS < 0.1 Sm-149 ID-MS < 0.1 Sm- 150 ID-MS < 0.1 Sm-151 ID-MS < 0.1 Sm-152 ID-MS < 0.1 Sm-154 ID-MS < 0.1 Eu-154 y-spec < 3.0 Ru- 106 y-spec < 5.0 Sb-125 Y-spec < 10.0 Main technique is listed; some nuclides require multiple techniques to eliminate interferences.
Relative standard deviation. As reported (Ic) in Y.Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent L WR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
29
Table 3.16 Experimental results (g/g Uinitial) for Takahama-3 samples from rod SF95 Sample ID SF95-1 SF95-2 SF95-3 SF95-4 SF95-5 Burnup (GWd/MTU) 14.30 24.35 35.42 36.69 30.40 U-234 2.987E-04 2.850E-04 1.873E-04 1.870E-04 2.829E-04 U-235 2.674E-02 1.927E-02 1.326E-02 1.230E-02 1.544E-02 U-236 2.672E-03 4.024E-03 4.911E-03 4.999E-03 4.566E-03 U-238 9.499E-0 1 9.424E-0 I 9.338E-01 9.335E-01 9.388E-01 Pu-238 1.718E-05 7.102E-05 1.539E-04 1.588E-04 1.020E-04 Pu-239 4.227E-03 5.655E-03 6,194E-03 6.005E-03 5.635E-03 Pu-240 7.802E-04 1.539E-03 2.186E-03 2.207E-03 1.821E-03 Pu-241 3.690E-04 9.578E-04 1.486E-03 1.466E-03 1. 153E-03 Pu-242 3.790E-05 1.844E-04 4.516E-04 4.803E-04 2.976E-04 Am-241 1.378E-05 2.344E-05 3.3 1OE-05 2.351 E-05 2.840E-05 Am-242m 1.840E-07 5.201E-07 7.877E-07 7.282E-07 5.687E-07 Am-243 2.682E-06 2.289E-05 8.047E-05 8.472E-05 4.400E-05 Cm-242 1.510E-06 7.672E-06 1.964E-05 2.328E-05 1.006E-05 Cm-243 1.451E-08 1.240E-07 3.720E-07 3.976E-07 2.293E-07 Cm-244 2.712E-07 5.042E-06 2.562E-05 2.837E-05 1.064E-05 Cm-245 5.519E-09 1.962E-07 1.396E-06 1.587E-06 4.839E-07 Cm-246 2.560E-10 1.190E-08 1.049E-07 1.251E-07 1.952E-08 Nd-142 3.429E-06 8.887E-06 2.116E-05 2.222E-05 1.371E-05 Nd-143 4.63 1E-04 7.149E-04 9.299E-04 9.373E-04 8.303E-04 Nd-144 3.276E-04 6.046E-04 9.347E-04 1.024E-03 7.928E-04 Nd-145 3.328E-04 5.384E-04 7.392E-04 7.598E-04 6.518E-04 Nd-146 2.809E-04 4.925E-04 7.340E-04 7.624E-04 6.185E-04 Nd-148 1.592E-04 2.736E-04 3.979E-04 4.126E-04 3.40 1E-04 Nd-150 7.200E-05 1.258E-04 1.895E-04 1.959E-04 1.572E-04 Cs-134 2.343E-05 7.012E-05 1.404E-04 1.471E-04 1.014E-04 Cs-137 5.405E-04 9.336E-04 1.347E-03 1.400E-03 1. 148E-03 Ce-144 1.937E-04 3.160E-04 4.560E-04 4.301E-04 3.868E-04 Eu-154 4.093E-06 1.306E-05 2.525E-05 2.657E-05 1.817E-05 Ru- 106 4.447E-05 8.340E-05 1.360E-04 1.401 E-04 1.208E-04 Sb- 125 1.471E-06 2.900E-06 3.733E-06 3.169E-06 3.262E-06
'As reported in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Credit for Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
30
Table 3.17 Experimental results (g/g Uinitiai) for Takahama-3 samples from rod SF96 Sample ID SF96-1 SF96-2 SF96-3 SF96-4 SF96-5 Burnup I (GWd/MTU) 7.79 16.44 28.20 28.91 24.19 U-234 1.805E-04 1.522E-04 1.251E-04 1.250E-04 1.354E-04 U-235 1.944E-02 1.408E-02 8.638E-03 8.064E-03 9.937E-03 U-236 1.421E-03 2.411 E-03 3.244E-03 3.302E-03 3.013E-03 U-23 8 9.660E-0 1 9.580E-01 9.476E-01 9.475E-0 I 9.522E-0 I Pu-238 8.536E-06 4.172E-05 1.206E-04 1.248E-04 7.978E-05 Pu-239 3.781E-03 5.459E-03 6.001E-03 5.819E-03 5,.519E-03 Pu-240 6.764E-04 1.494E-03 2.303E-03 2.327E-03 1.964E-03 Pu-241 2.622E-04 8.684E-04 1.498E-03 1.480E-03 1,203E-03 Pu-242 2.440E-05 1.615E-04 5.103E-04 5.411 E-04 3.5 5 1E-04 Np-237 6.125E-05 1.323E-04 2.168E-04 2.252E-04 1.875E-04 Am-241 5.985E-06 1.735E-05 2.845E-05 3.094E-05 2.149E-05 Am-242m 1.218E-07 4.579E-07 6.413E-07 6.793E-07 5.647E-07 Am-243 1. 147E-06 1.728E-05 8.872E-05 9.598E-05 5.078E-05 Cm-242 8.502E-07 5.781E-06 1.628E-05 1.679E-05 1. 115E-05 Cm-244 9.560E-08 3.092E-06 2.862E-05 3.128E-05 1.280E-05 Nd-143 2.52 1E-04 4.778E-04 7.158E-04 7.184E-04 6.433E-04 Nd-144 1.536E-04 3.588E-04 7.292E-04 7.513E-04 5.927E-04 Nd-145 1.800E-04 3.575E-04 5.766E-04 5.880E-04 5.095E-04 Nd-146 1.536E-04 3.266E-04 5.795E-04 5.948E-04 4.910E-04 Nd- 148 8.770E-05 1.851E-04 3.201E-04 3.280E-04 2.733E-04 Nd-150 4.130E-05 8.972E-05 1.591E-04 1.628E-04 1.331E-04 Cs-134 8.609E-06 3.759E-05 1.002E-04 1.047E-04 7.146E-05 Cs-137 2.813E-04 5.983E-04 1.018E-03 1.053E-03 8.572E-04 Ce- 144 1.179E-04 2.250E-04 3.362E-04 3.453E-04 3.145E-04 Eu-154 2.309E-06 8.538E-06 1.973E-05 1.992E-05 1.423E-05 Ru- 106 2.830E-05 6.053E-05 1.402E-04 1.29 1E-04 1.344E-04 Sb-125 1.433E-06 2.829E-06 3.658E-06 4.645E-06 3.690E-06
'As reported in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
31
Table 3.18 Experimental results (g/g Uinitial) for Takahama-3 samples from rod SF97 Sample ID SF97-1 SF97-2 SF97-3 SF974 SF97-5 SF97-6 Burnup "
(GWd/MTU) 17.69 30.73 42.16 47.03 47.25 40.79 U-234 2.939E-04 2.348E-04 2.01 OE-04 1.872E-04 1.865E-04 2.057E-04 U-235 2.347E-02 1.571E-02 1.030E-02 8.179E-03 7.932E-03 1.016E-02 U-236 3.1 I15E-03 4.560E-03 5,3 12E-03 5.528E-03 5.532E-03 5.272E-03 U-238 9.493E-01 9.377E-01 9.282E-01 9.246E-0 I 9.247E-01 9.3 10E-01 Pu-238 2.370E-05 1.250E-04 2.581 E-04 3.199E-04 3.188E-04 2.175E-04 Pu-239 3.844E-03 5.928E-03 6.217E-03 6.037E-03 5.976E-03 5.677E-03 Pu-240 9.347E-04 1.871E-03 2.471E-03 2.668E-03 2.648E-03 2,326E-03 Pu-241 4.237E-04 1.235E-03 1.689E-03 1.770E-03 1.754E-03 1.494E-03 Pu-242 6.185E-05 3.152E-04 6.517E-04 8.246E-04 8.341E-04 5.977E-04 Np-237 1.521E-04 4.034E-04 5.845E-04 6.604E-04 6.701E-04 5.570E-04 Am-241 1.492E-05 4.017E-05 4.909E-05 5.31 IE-05 5.327E-05 4.297E-05 Am-242m 2.270E-07 8.838E-07 1.179E-06 1.233E-06 1.200E-06 9.756E-07 Am-243 4.448E-06 5.132E-05 1.410E-04 1.924E-04 1.935E-04 1.170E-04 Cm-242 2.134E-06 1.049E-05 1.839E-05 2.044E-05 1.903E-05 1.616E-05 Cm-243 2.483E-08 2.773E-07 6.921E-07 8.721E-07 8.670E-07 5.600E-07 Cm-244 4.981 E-07 1.384E-05 5,696E-05 8.81 OE-05 8.823E-05 4.221 E-05 Cm-245 1.087E-08 6.848E-07 3.735E-06 6.042E-06 5.915E-06 2.363E-06 Cm-246 3.866E- 10 4.222E-07 3.648E-07 7.440E-07 7.549E-07 2.481E-07 Cm-247 NA 4.043E-10 4.974E-09 1.098E-08 1.075E-08 3.139E-09 Nd-143 5.450E-04 8.307E-04 1.008E-03 1.048E-03 1.049E-03 9.736E-04 Nd- 144 4.66 1E-04 8.843E-04 1.33 IE-03 1.567E-03 1.599E-03 1.311 E-03 Nd-145 4.045E-04 6.480E-04 8.387E-04 9.118E-04 9.179E-04 8.247E-04 Nd- 146 3.502E-04 6.304E-04 8.929E-04 I .008E-03 1.014E-03 8.586E-04 Nd-148 1.945E-04 3.389E-04 4.662E-04 5.204E-04 5.226E-04 4.504E-04 Nd- 150 8.570E-05 1.582E-04 2.234E-04 2.516E-04 2.518E-04 2.130E-04 Cs-134 2.983E-05 1.030E-04 1.829E-04 2.139E-04 2.144E-04 1.632E-04 Cs-137 6.617E-04 1.15 1E-03 1.582E-03 1.749E-03 1.761E-03 1.53 1E-03 Ce-I144 2.026E-04 3.061E-04 3.720E-04 3.756E-04 3.750E-04 3.714E-04 Eu-154 5.253E-06 1.973E-05 3.293E-05 3.739E-05 3.707E-05 2.859E-05 Ru- 106 5.163 E-05 1.162E-04 1.829E-04 1.936E-04 1.162E-04 1.959E-04 Sb-125 2.462E-06 5.118E-06 4.966E-06 6.090E-06 7.507E-06 4.546E-06 Smi-147' 1.529E-04 2.050E-04 2.355E-04 2.468E-04 2.479E-04 2.37]E-04 Sm- 148 4,092E-05 1.194E-04 1.978E-04 2.338E-04 2.357E-04 1.809E-04 Sm-149 2.935E-06 3.976E-06 4.259E-06 3.943E-06 3.799E-06 3.843E-06 Sm-150 1.323E-04 2.499E-04 3.599E-04 4.074E-04 4.113E-04 3.409E-04 Sm- 151 9.324E-06 1.351E-05 1.503E-05 1.491E-05 1.465E-05 1.294E-05 Sm-152 6.526E-05 9.546E-05 1.191 E-04 1.298E-04 1.319E-04 1.207E-04 Sm-154 1.425E-05 2.977E-05 4.536E-05 5.252E-05 5.298E-05 4.231 E-05 a As reported in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
Measured data for samarium isotopes were reported at 3.96 years after discharge; at discharge time for all other isotopes.
32
4 ASSEMBLY AND IRRADIATION HISTORY DATA This section presents information on the fuel assembly geometry, irradiation history, and sample burnup necessary for developing a computational model to determine the isotopic composition of the samples under consideration.
4.1 TMI-1 SAMPLES The samples considered were selected from two different fuel assemblies, identified as NJ05YU and NJ070G, irradiated in the TMI-1 reactor. Details related to the geometry, material composition, and irradiation history were taken from Ref. 4. Both assemblies are a 15 x 15 design, with 208 fuel rods, 16 guide tubes, and one instrument tube, as illustrated in Figures 4.1 and 4.2.
The fuel assembly geometry and material information for the two assemblies are presented in Table 4.1.
Assembly NJ05YU was irradiated in the reactor for two consecutive cycles, cycle 9 and cycle 10. It contained 16 burnable poison rods (BPRs) with A120 3-B 4 C absorber, which were removed at the end of the cycle 9. All the fuel rods in this assembly had an initial fuel enrichment of 4.013 wt % 231U.
Assembly NJ070G was present in the reactor during cycle 10 only. It also contained 16 BPRs during this cycle.
235U. Four of its fuel rods had 2.0 wt % Gd 20 3 poison, and their initial fuel enrichment was 4.19 wt %
The other 204 regular fuel rods had an initial enrichment of 4.657 wt % 235U. Guide and instrument tube data were used as given elsewhere.18 The locations of the Gd 20 3 poison rods in the assembly were provided by AREVA.
Eleven of the 19 TMI-1 samples, those measured at ANL, were selected from a fuel rod identified as H6, located in assembly NJ05YU. The other eight TMI-1 samples, analyzed at GE-VNC, were selected from the rods identified as 01, 012, and 013, located in assembly NJ070G. The location of the measured fuel rods in the assembly is illustrated in Figures 4.1 and 4.2. Note that all three measured fuel rods from assembly NJ070G were located at the edge of the assembly; the rod identified as 01 was located at the corner of the assembly.
Two sets of burnup values were specified in Ref. 4 for each sample: cumulative bumup based on operational data provided at end of cycle (EOC) for cycles 9 and 10, and total measured burnup, determined based on isotopic measurements, corresponding to EOC-10. The specific average powers for cycle P9 and P10 used for the calculations in the current work were obtained as:
P9 = B9 Bmeas = B1 0 - B 9 Bmeas
-I0 (4-1)
At9 B10 Atl 0 B1 0 where B9 and B, 0 are the nominal bumup values at EOC-9 and EOC- 10, Bmeas is the sample measured bumup at EOC-10, and At9 and Atio is the cycle duration for cycles 9 and 10, respectively.
The effective full power days (EFPD) for cycle 9 and 10 are 639.4 days and 660.3 days, respectively.
The down time between cycles 9 and 10, not available in Ref. 4, was assumed to be 30 days. Bumrup and power data for each sample, as well as moderator density data are presented in Table 4.2. The variations with time of the soluble boron concentration in moderator and of the fuel temperature for assemblies NJ05YU and NJ070G are shown in Tables 4.3 and 4.4, respectively. Cooling time values corresponding to the measurement date for each sample are provided in Table 4.5.
33
Data available18 on the assemblies surrounding assembly NJ070G are illustrated in Figure 4.3. As the samples from this assembly are expected to be subjected to edge effects given their location at the periphery of the assembly, this information may be important for modeling purposes. The sensitivity of the calculated nuclide content to the inclusion of this type of geometry details in the computational model is discussed in detail in Appendix A. The measured fuel rods were located at the east edge of assembly NJ070G that neighbored an assembly from batch 12A with an initial fuel enrichment of 4 wt % 235U.
Assemblies in batch 12 were first irradiated in the core during cycle 10. Assemblies in batch 11 were present in the core since cycle 9; no data were available on the average burnup of these assemblies at BOC-10. It is not known whether assemblies surrounding assembly NJ070G have fuel rods containing gadolinia poison. Also unknown is the exact location of rods 01, 012, and 013 with respect to the assemblies located north and south of assembly NJ070G. However, given the symmetry, as seen in Figure 4.3, this detail is deemed to be of low importance for modeling purposes.
34
A B C D E F G H I J K L M N 0 Guide 1 _tube Fuel 2 rod Measured 3 fuel rod 4
5 6
7 8
9 10 11 12 13 14 15 4r Figure 4.1 Assembly layout for TMII-I samples--NJ05YU 35
A B C D E F G H I J K L M N 0 Guide 1 tube Fuel 2 rod Measured 3 fuel rod 4 Gd 2O_;rod 5
6 7
8 __
9 10 11 12 13 14 15 Figure 4.2 Assembly layout for TMI-1 samples-NJ070G 36
Table 4.1 Assembly design data for TMI-1 samples Parameter Data for assembly Data for assembly NJ05YU NJ070G Assembly and reactor dataa Reactor TMI-! TMI-1 Lattice geometry 15 x 15 15 x 15 Rod pitch (cm) 1.44272 1.44272 Number of fuel rods 208 208 Number of guide tubes 16 16 Number of instrument tubes 1 1 Assembly pitch (cm) 21.81098 21.81098 Fuel rod datae Fuel material type U0 2 U0 2 Fuel pellet density (g/cm 3) 10.196 10.217 Fuel pellet diameter (cm) 0.9362 0.9398 Fuel temperature (K) see Table 4.3 see Table 4.4 Enrichment (wt % 2351j) 4.013 4.657 Clad material Zircaloy-4 Zircaloy-4 Clad inner diameter (cm) 0.95758 0.95758 Clad outer diameter (cm) 1.0922 1.0922 Average clad temperature (K) 640 640 Number of rods with Gd 203 0 4 Gd 2 0O3 content (wt %) NA 2.0 Initial fuel composition (wt %)
234 235 u 0.040 0.045 (0.0) h 238 U 4.013 4.657 (4.019)
U 95.947 95.298 (95.981) h Moderator datae Moderator density (g/cm 3 ) see Table 4.2 see Table 4.2 Soluble boron in moderator (ppm) see Table 4.3 see Table 4.4 Burnable poison rod (BPR) dataa Absorber diameter (cm) 0.8636 0.8636 Clad inner diameter (cm) 0.9144 0.9144 Clad outer diameter (cm) 1.0922 1.0922 Absorber material A120 3-B4 C AI20 3 -B4C Absorber material density (g/cm 3) 3.7 3.7 B4C content (wt %) 1.7 2.1 Cladding material Zircaloy-4 Zircaloy-4 Guide/instrument tube data' Guide/instrument tube material Zircaloy-4 Zircaloy-4 Guide tube inner diameter (cm) 1.26492 1.26492 Guide tube outer diameter (cm) 1.3462 1.3462 Instrument tube inner diameter (cm) 1.12014 1.12014 Instrument tube outer diameter (cm) 1.25222 1.25222
'As provided in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisonsto SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 I, Rev. A (April 2002).
b Values in parentheses correspond to gadolinia-bearing fuel rods.
As provided in L. B. Wimmer, Summary Report of Commercial Reactor CriticalityDatafor Three Mile Island Unit 1, TDR-UDC-NU-000004 REV 01, Bechtel SAIC Company, LLC, Las Vegas, NV (August 2001).
37
Table 4.2 Burnup, power and moderator density data for TMI-1 samples Rod Burnup' Burnupa Measuredb Calculatedc Calculated Moderator Asebypower power dest Assembly ID Sample ID EOC-9 EOC-10 burnup cycle 9 cycle 10 density (GWd/MTU) (GWd/MTU) (GWd/MTU) (MW/MTU) (MW/MTU) (g/cm 3)
A2 28.338 51.861 50.6 43.242 34.759 0.7314 B2 28.444 52.089 50.1 42.787 34.442 0.7248 C1 28.132 51.545 50.2 42.849 34.533 0.6965 C3 28.230 51.696 51.3 43.813 35.266 0.7151 NJ05YU H6 D2 26.366 48.569 44.8 38.036 31.016 0.6787 AIB 24.767 45.687 44.8 37.983 31.067 0.7382 BIB 28.230 51.696 54.5 46.546 37.466 0.7151 B3J 28.338 51.861 53.0 45.293 36.407 0.7314 C2B 28.155 51.563 52.6 44.919 36.164 0.7057 DIA2 28.115 51.530 55.7 47.529 38.331 0.6934 DIA4 28.034 50.810 50.5 43.577 34.283 0.6875 O1 SI 27.498 25.8 39.073 0.7382 01 O1 S2 31.377 29.9 45.282 0.7057 00 O1 S3 30.848 26.7 40.436 0.6875 NJ070G 012 S4 25.592 23.7 35.893 0.7382 012 012S5 29.271 26.5 40.133 0.7057 012 S6 28.760 24.0 36.347 0.6875 O13 $7 25.331 22.8 34.530 0.7382 013 013 S8 29.020 26.3 39.830 0.7057 Based on operating history information.
b As provided in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisons to SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-0000 11, Rev. A (April 2002).
'See Eq. (4-1).
Table 4.3 Fuel temperature and concentration of soluble boron in moderator for TMI-1 samples from assembly NJ05YU Sample A2 B2 CI C3 D2 AIB BIB B3J C2B DIA2 DIA4 Cycle ID Boron"
- Time Temp.' (ppm)
(days) (K) 0.0 1670 74.2 1051.2 1085.4 1105.7 1098.3 1029.0 948.7 1098.3 1051.2 1106.1 1100.7 1091.2 1481 141.1 1040.9 1058.8 1069.1 1062.9 1025.2 957.5 1062.9 1040.9 1066.8 1068.4 1065.0 1342 214.0 1023.3 1030.3 1034.4 1029.4 1009.3 959.3 1029.4 1023.3 1031.2 1035.8 1035.8 1175 284.9 1002.0 1002.0 1003.2 998.2 995.7 953.5 998.2 1002.0 998.9 1006.2 1009.0 990 9 349.7 982.09 976.62 976.01 971.18 982.57 947.65 971.18 982.09 971.2 980.2 985.1 772 425.0 959.40 950.04 948.93 944.23 963.71 940.48 944.23 959.40 944.2 963.4 959.3 545 483.9 936.52 925.90 925.46 920.59 945.73 927.57 920.59 936.52 921.0 929*9 936.2 352 549.2 918.46 907.79 907.79 903.40 929.26 913.93 903.40 918.46 904.0 911.7 917.8 134 608.0 888.21 884.01 900.23 889.62 924.34 886.48 889.62 888.21 895.1 904.2 909.8 13 639.4 772.90 777.37 810.43 790.98 837.01 771.65 790.98 772.90 801.7 815.3 821.0 2 0.0 1800 68.0 835.54 861.01 871.01 871.32 825.07 787.87 871.32 835.54 874.84 865.43 843.34 1649 131.8 828.60 846.96 856.62 853.46 825.54 785.79 853.46 828.60 856.84 854.23 840.48 1521 209.0 824.52 935.65 844.68 838.76 829.98 786.23 838.76 824.52 841.84 845.32 840.87 1322 272.1 823.77 828.87 835.79 829.29 831.73 791.43 829.29 823.77 831.71 838.01 838.93 1140 10 347.4 822.13 823.12 828.46 822.09 832.12 796.54 822.09 822.13 823.98 831.46 836.07 918 416.4 818.71 816.71 821.65 815.18 831.46 799.84 815.18 818.71 816.96 825.09 832.48 718 486.4 813.82 809.93 815.29 808.54 829.93 801.23 808.54 813.82 810.51 818.96 828.43 506 556.3 807.62 802.59 808.43 801.59 827.37 800.98 801.59 807.62 803.76 812.15 823.34 298 626.1 801.96 796.93 802.65 796.15 823.76 799.18 796.15 801.96 798.34 806.21 817.73 103 660.3 799.90 795.18 799.87 794.37 819.26 797.96 794.37 799.90 796.26 803.01 813.51 1.8 As 4provided inJ. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisons to SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 1, Rev. A (April 2002).
Table 4.4 Fuel temperature and concentration of soluble boron in moderator for TMI-1 samples from assembly NJ070G Sample 01S1 01S2 OIS3 012S4 012S5 012S6 O13S7 O13S8 Cycle ID Borona
- Time Temp." (ppm)
(days) (K) 0.0 1800 68.0 960.29 1119.5 1.083.7 960.29 1119.5 1083.7 960.29 1119.51 1649 131.8 960.71 1084.8 1067.3 960.71 1084.8 1067.3 960.71 1084.79 1521 209.0 958.68 1043.2 1043.5 958.68 1043.2 1043.5 .958.68 1043.23 1322 272.1 954.18 1007.1 1016.4 954.18 1007.1 1016.4 954.18 1007.09 1140 10 347.4 946.12 978.57 991.65 946.12 978.57 991.65 946.12 978.57 918 416.4 937.15 951.57 967.21 937.15 951.57 967.21 937.15 951.57 718 486.4 926.04 929.82 945.98 926.04 929.82 945.98 926.04 929.82 506 556.3 914.37 912.15 928.04 914.37 912.15 928.04 914.37 912.15 298 626.1 904.09 896.84 912.12 904.09 896184 912.12 904.09 896.84 103 660.3 897.82 886.54 899.73 897.82 886.54 899.73 897.82 886.54 1.8 As provided in J. M. Scaglione. Three Mile Island Unit I RadiochemicalAssay Comparisons to SAS2H Calculations.Yucca Mountain Project Report.
CAL-UDC-NU-0000 11, Rev. A (April 2002).
0 Table 4.5 Cooling time at measurement date for TMI-1 samples Sample ID Cooling time a (days)
A2, B2, CI, C3, D2 1103 O1SI, O1S3, 012S4, 012S6 1298 O1S2, 012S5, 013S7, 013S8 1529 A1B, BIB, B3J, C2B, DIA2, DIA4 1711
'As provided in J. M. Scaglione, Three Mile Island Unit 1 RadiochemicalAssay Comparisons to SAS2I- Calculations,Yucca Mountain Project Report, CAL-UDC-NU-000011, Rev. A (April 2002).
NW N NE Batch 12C Batch 11C Batch 12E Enrich = 4.65% Enrich = 4% Enrich = 4.75%
W E Batch 11A Batch 12A Enrich = 3.63% Enrich = 4%
SW S SE Batch 12B Batch IIC Batch 12D Enrich = 4.65% Enrich = 4% Enrich = 4.75%
Figure 4.3 Assemblies surrounding assembly NJ070G 4.2 CALVERT CLIFFS SAMPLES The Calvert Cliffs measurements considered in this report were carried out on three samples from the fuel rod MKP-109 belonging to the CE 14 x 14 fuel assembly D047. The samples are identified as 87-8 1, 87-72, and 87-63. The rod was present in the reactor core for four consecutive cycles, from cycle 2 to cycle 5. The assembly had 176 fuel rods and five guide tubes, as illustrated in Figure 4.4. There were no burnable poison rods or gadolinia-bearing rods in the assembly during any of the irradiation cycles. The location of the rod from which the samples were selected is also shown in the figure.
The geometry data are presented in Table 4.6 and the burnup history data and soluble boron concentration in moderator are presented in Table 4.7. The fuel temperature, moderator temperature and density, and cooling times for each of the three samples are given in Table 4.8. All these data were taken from Refs. 11 and 15.
41
A B C D E F G H I J K L M N 1
S Guide tube Fuel rod 2
Measured 3 fuel rod 4
5 6
7 8
9 10 11 12 13 14 Figure 4.4 Assembly layout for Calvert Cliffs samples 42
Table 4.6 Assembly design data for Calvert Cliffs samples Parameter Data Assembly and reactor dataa Reactor Calvert Cliffs 1 Lattice geometry 14 x 14 Assembly design CE Rod pitch (cm) 1.4732 Number of fuel rods 176 Number of water rods 5 Assembly pitch (cm) 20.78 Fuel rod data' Fuel material type U0 2 Fuel density (g/cm 3) 10.045 Fuel pellet diameter (cm) 0.9563 Clad material Zircaloy-4 Fuel temperature (K) see Table 4.8 Clad inner diameter (cm) 0.9855 Clad outer diameter (cm) 1.1176 Average clad temperature b(K) 620 U isotopic 23 4 composition c (wt %)
23 5 U 0.027 23 6 U 3.038 238 U 0.014 U 96.921 Moderator dataa Moderator density (g/cm 3) see Table 4.8 Moderator temperature (K) see Table 4.8 Soluble boron content (ppm) see Table 4.7 Guide tube dataa Guide tube material Zircaloy-4 Inner radius (cm) 1.314 Outer radius (cm) 1.416 As provided in 0. W. Herman, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic CompositionAnalyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
b Assumed value.
' Initial values.
43
Table 4.7 Burnup history data for Calvert Cliffs samples Cycle Sample Burnup"b (GWd/MTU)
Cycle Start End Duration Down average
- date date (days) (days) boron a 87-81 87-72 87-63 (ppm) 2 3/22/77 1/22/78 306.0 71.0 330.8 5.28 7.56 9.52 3 4/3/78 4/20/79 381.7 81.3 469.4 12.69 17.78 21.93 4 7/10/79 10/18/80 466.0 85.0 503.7 20.63 28.42 34.14 5 1/11/81 4/17/82 461.1 1 492.1 27.35 37.12 44.34 "As provided in 0. W. Herman, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent FuelIsotopic Composition Analyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
Cumulative value.
Table 4.8 Moderator, fuel temperature, and cooling time data for Calvert Cliffs samples Parameter 87-81 87-72 87-63 Moderator temperature ' (K) 557 558 570 Moderator density a (g/cm 3) 0.7575 0.7569 0.7332 Fuel temperature "(K) 790 841 873 Cooling time b (days) 1870 I 4171* 1870 I 4656c 1870 I 4656c "As provided in 0. W. Herman, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
b At time of measurement; values correspond to PNNL and KRI measurement dates, respectively.
'Values obtained from PNNL private communication.
4.3 TAKAHAMA-3 SAMPLES Radiochemical analyses were performed at JAERI on 16 samples from three fuel rods identified as SF95, SF96, and SF97. 6"13,1 7 Rods SF95 and SF97 were standard fuel rods with 4.11 wt % 235U initial enrichment; whereas SF96 was a fuel rod with gadolinia poison that had a fuel initial enrichment of 2.6 wt % 2351j and a Gd 20 3 content of 6%. Rods SF95 and SF96 were from assembly NT3G23 and rod SF97 was from assembly NT3G24. Each of these two assemblies had a 17 x 17 configuration, with 264 fuel rods (14 of these containing gadolinial7) and 25 water-filled guide tubes. They resided in the reactor core for two (assembly NT3G23) or three (assembly NT3G24) consecutive cycles, starting from cycle 5.
The configuration of the assembly, including the location of the measured rods, is illustrated in Figure 4.5. Assembly parameters are listed in Table 4.9.
Bumup values, sample axial location along the fuel rod, moderator density and temperature, and cycle power for each sample are listed in Table 4.10. Operation history data and soluble boron concentration are presented in Tables 4.11 and 4.12, respectively. The moderator density was determined by interpolating on temperature vs. pressure data,' using the available moderator temperature data' 7 and a pressure value' 9 of 157 kg/cm 2. The cycle power for each sample was obtained by averaging the power data given in Ref. 17. The measured nuclide concentrations were reported at discharge time with the exception of those for samarium isotopes in samples from rod SF97 that were reported at 3.96 years after discharge.
44
J K L M N 0 P Guide 1 tube Fuel 2 _rod Measured 3 rod GdO 1 4 rod 5
6 7
8 9
10 11 12 13 14 15 16 17 Figure 4.5 Assembly layout for Takahama-3 samples 45
Table 4.9 Assembly design data for Takahama-3 samples Parameter Data Assembly and reactor dataa Reactor Takahama-3 Lattice geometry 17 x 17 Rod pitch (cm) 1.259 Number of fuel rods 264 Number of guide tubes 25 Assembly pitch (cm) 21.4 Fuel rod data' Fuel material type U0 2 Fuel pellet density (% TD) 95 Enrichment (wt % 235U) 4.11 (2.63)0 Fuel pellet diameter (cm) 0.805 Average fuel temperature (K) 900 Clad material Zircaloy-4 Clad inner diameter (cm) 0.822 Clad outer diameter (cm) 0.95 Average clad temperature (K) 600 Number of rods with Gd 203 14 Gd 20 3 content (wt %) 6.0 U isotopic composition b (Wt %)
234U 0.04 (0.02) 235 U 4.11 (2.63) 238 U 95.85 (97.25)
Moderator data' Moderator density (g/cm 3 ) see Table 4.10 Moderator temperature (K) see Table 4.10 Soluble boron (ppm) see Table 4.12 Guide tube data' Guide tube material Zircaloy-4 Inner radius (cm) 0.5715 Outer radius (cm) 0.6121
'As given in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/0 1), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
At beginning of life.
'Values in parentheses correspond to gadolinia-bearing fuel rods.
46
Table 4.10 Burnup, power, sample location, and moderator data for Takahama-3 samples Assembly Rod Burnup Powerb Power Powerb Sample' Moderator a Moderator Am Sample ID cycle 5 cycle 6 cycle 7 location temperature density ID (GWd/MTU) (MW/MTU) (MW/MTU) (MWIMTU) (cm) (K) (g/em 3)
SF95-1 14.30 19.21 17.17 20.1 593.04 0.6803 SF95-2 24.35 32.72 29.25 36.1 592.75 0.6810 SF95 SF95-3 35.52 47.59 42.54 88.1 589.37 0.6898 SF95-4 36.69 49.30 44.06 216.1 570.40 0.7324 NT323G SF95-5 30.40 40.85 36.51 356.1 554.19 0.7628 SF96-1 7.79 8.01 11.72 17.6 593.05 0.6803 SF96-2 16.44 16.90 24.71 33.6 592.82 0.6809 SF96 SF96-3 28.20 28.99 42.40 85.6 589.62 0.6892 SF96-4 28.91 29.71 43.46 213.6 570.82 0.7316 SF96-5 24.19 24.87 36.37 353.6 554.28 0.7627 SF97-I 17.69 14.76 15.74 13.97 16.3 593.05 0.6803 SF97-2 30.73 25.65 27.36 24.28 35.0 592.78 0.6810 SF97-3 42.16 35.19 37.53 33.31 62.7 591.48 0.6843 NT324G SF97 SF97-4 47.03 39.26 41.87 37.16 183.9 575.83 0.7211 SF97-5 47.25 39.44 42.06 37.33 292.6 559.14 0.7540 SF97-6 40.79 34.05 36.31 32.23 355.6 554.21 0.7628 aAs given in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01),
English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
b Cycle-averaged power calculated based on data provided in Y. Nakahara, Y. Suyamna, and T. Suzaki, TechnicalDevelopment on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
'Distance measured from top of fuel.
Table 4.11 Operation history data for Takahama-3 samples Cycle Start End Duration Down f0 date date (days) (days) 5 1990/01/26 1991/02/15 385 88 6 1991/02/15 1991/05/14 402 62 7 1991/05/14 1992/06/19 1 406 1 Table 4.12 Soluble boron concentration in moderator forTakahama-3 samples Cumulative Cycle # timea Boron content (days) (ppm) 0 1154 106 894 5 205 651 306 404 385 210 473 1132 592 864 6 704 613 817 358 875 228 937 1154 996 1001 1048 867 1100 732 7 1152 598 1204 463 1256 329 1308 195 1342 104 Measured from beginning of cycle 5.
48
5 COMPUTATIONAL MODELS 5.1 COMPUTATIONAL TOOLS The computational analysis of the measurements was carried out by using the 2-D depletion sequence TRITON in the SCALE computer code system.1 This sequence couples the 2-D arbitrary polygonal mesh, discrete ordinates transport code NEWT with the depletion and decay code ORIGEN-S in order to perform the burnup simulation. At each depletion step, the transport flux solution from NEWT is used to generate the cross sections for the ORIGEN-S calculation; the isotopic composition data resulting from ORIGEN-S is employed in the subsequent transport calculation to obtain cross sections for the next depletion step in an iterative manner throughout the irradiation history.
TRITON has the capability of individually simulating the depletion of multiple mixtures in a fuel assembly model. This is a very useful and powerful feature in a nuclide inventory analysis, as it allows a more appropriate representation of the local flux distribution and environmental effects on a specific measured fuel rod in the assembly. The flux normalization in a TRITON calculation can be performed using as a basis the power in a specified mixture, the total power corresponding to multiple mixtures, or the assembly power. The first of the above-mentioned options permits the burnup (power) in the measured sample, usually inferred from experimental measurements of burnup indicators (such as 1 Nd),
to be specified.
Individual TRITON models were developed for each of the 38 sample measurements discussed in the previous sections. In all cases, the calculations were carried out by normalizing the power to reproduce the measured concentration of 14'Nd in the sample within the experimental uncertainty. All TRITON calculations employed the SCALE 44-group cross-section library based on ENDF/B-V data and NITAWL as processor for the pin-cell cross-section treatment. Default values were used for the convergence parameters in the NEWT transport calculation. Selected TRITON input files are provided in Appendix B.
TRITON provides the user the option to control the number of nuclides in the depleted mixtures for which the cross sections used in the ORIGEN-S depletion calculation are updated at each depletion step based on the flux solution from the transport calculation with NEWT. The user should specify the control parameter "parm=(addnux=N)" on the first line of the TRITON input file, where N identifies the set of nuclides included in the transport calculation. The nuclides in the selected set that are not present in the initial fuel composition are set to trace concentrations (i.e., 10.20 atoms/b-cm). The calculation in the present report used the option "addnux=3" for which the set of nuclides considered in the transport calculation, and for which, therefore, the NEWT flux solution is used to update the cross sections for ORIGEN-S, contains 232 isotopes.
5.2 TMI-1 SAMPLES Assembly NJ05YU that hosted the fuel rod H6 (see Figure 4.1), from which 11 samples were selected, was irradiated in two consecutive cycles, cycle 9 and cycle 10. The BPRs present during cycle 9 were removed in cycle 10. Separate TRITON models were developed to accurately represent this change in the assembly geometry, as illustrated in Figure 5.1. Given the symmetry and the location of rod H6 in assembly NJ05YU, the models for the analysis of samples selected from this rod represent only half of the assembly geometry, with a reflective boundary condition on the left side of the configuration and white boundary conditions on the other three bounding surfaces. The geometry and material data were used as given in Table 4.1, and the power data as provided in Table 4.2. Six fuel mixtures were specified: one corresponding to the measured rod, four to the nearest neighbor fuel rods, and one to the rest of the fuel 49
rods in the assembly. At the end of the depletion simulation for cycle 9, the isotopic composition for each of these six fuel mixtures was extracted and used as input data in the model corresponding to cycle 10 (with no BPR present). A total of 232 nuclides, representing the main light elements, actinides, and fission products, were included to represent the fuel composition at the start of cycle 10. The variation of the soluble boron content in the coolant and of the temperature in fuel during irradiation, as given in Table 4.3, was modeled through the use of the TIMETABLE input block in TRITON; ten bumup steps per cycle were used.
As data became available on the assemblies surrounding the assembly NJ05YU, the TRITON geometry model was extended to include this information. However, as rod H6 is located toward the center of the assembly, it is not expected to be subject to significant edge effects due to the assembly surroundings.
These effects will be discussed in detail in Appendix A.
All three rods in assembly NJ070G, from which samples were selected for measurement, were edge rods located along one side of the assembly, with one of these rods placed at the corner of the assembly. The computational models used for the analysis of these samples include information on the assembly surroundings. As is shown in Appendix A, neglecting this type of detail could significantly affect the calculation of the nuclide content in the sample. The models for rods 012 and 013 are similar and include a quarter of assembly NJ070G and aquarter of the assembly surrounding it on the side on which the samples are located, as illustrated for rod 012 in Figure 5.2. As observed in this figure, in order to better approximate the local environment, given the close proximity of the measured rod to the assembly boundary, the nearest neighboring rods were represented by using different mixtures; one of these neighboring rods is located in a different assembly. In the case of the corner rod 01, the TRITON model included a quarter of assembly NJ070G and a quarter of each of the three surrounding assemblies that share the same corner point with assembly NJ070G. This model is illustrated in Figure 5.3. Note that in this model the average burnup at the beginning of cycle 10 for one of the adjacent assemblies (batch I IC in Figure 4.3) was not available; this assembly was assumed to be fresh fuel.
The power was adjusted by less than 1.5%, depending on the sample, in order to obtain a calculated la8Nd concentration, which is a direct measure of the integral number of fissions (burnup), in agreement with the experimental value.
50
Cycle 9 Cycle 10 U regular fuel pin U measured fuel pin H6 01 - F] neighbors of measured fuel pin BPR absorber U BPR clad Figure 5.1 TRITON assembly model for TMI-I samples in assembly NJ05YU 51
Emeasured fuel pin 0 12 U L--UH nearest neighbors of measured pin
- regular fuel~ins in assembly NJ070G nEgadolinia fuel pin E fuel pins in neighboring assembly moderator li BPR absorber Figure 5.2 TRITON assembly model for TMI-1 samples in rod 012 of assembly NJ070G 52
m measured fuel in O1 M I M U nearest neighbors of measured pin E gadolinia fuel pin
- BPR absorber
_ regular fuel pins in assembly NJ070G E U fuel pins in surrounding assemblies D moderator Figure 5.3 TRITON assembly model for TMI-1 samples in rod 01 of assembly NJ070G 53
5.3 CALVERT CLIFFS SAMPLES Half of the Calvert Cliffs D047 assembly was modeled for the analysis of the three measured fuel samples from fuel rod MKP-109, as illustrated in Figure 5.4. Geometry, material composition, temperature, coolant density, and soluble boron in moderator data as specified in Tables 4.6 to 4.8 were used.
The power (burnup) values given in Table 4.7 were adjusted by less than 1.6%, depending on the sample, such that the calculated 148Nd concentration was consistent with the corresponding measured value, within one standard deviation of the measurement.
- regular fuel in E measured fuel pinfE K nearest neighbors of measured fuel pin guide tube L moderator Figure 5.4 TRITON assembly model for Calvert Cliffs samples 54
5.4 TAKAHAMA-3 SAMPLES Fuel rod SF97, residing in assembly NT3G24, was simulated using a one-half assembly geometry model because the rod was located on a quarter-assembly symmetry axis. The models for fuel rods SF95 and SF96 from assembly NT3G23 used a one-quarter assembly model. The three models used for each rod are illustrated in Figures 5.5 to 5.7. In each model, the measured fuel rod, as well as the fuel rods adjacent to it, was individually depleted. The variation of the soluble boron in the moderator as given in Table 4.12, and the variation in moderator density and temperature provided in Table 4.10 were simulated through the use of the TIMETABLE input block in the TRITON input. Note that fuel rods SF95 and SF97 are located on the edge of the assembly and therefore possibly subjected to edge effects. However, as no information was available on the surrounding assemblies, these assemblies were not included in the model.
The cycle power data given in Table 4.10 was used for simulating the depletion for fuel samples SF95 and SF97, as these values yielded predicted 148 Nd concentrations that were in agreement with the measurements, within the experimental uncertainty. However, in the case of the samples from rod SF96, the simulation using the sample power (and burnup) in Table 4.10 yielded a calculated 14'Nd concentration that was with 4 to 10% less than the measured value, depending on the sample. This difference is much larger than the maximum 3% error in bumup specified in the JAERI report.' 7 The sample burnup determination by JAERI was made using the ASTM E 321-79 standard method that estimates the burnup (in GWd/MTU units) by multiplying the value of the burnup rate (%FIMA = Fission per Initial Metal Atom in percent value that is based on the measured 148Nd content) by a factor of 9.6 +/-L 0.3. 20 However, derivation of this factor is based on a recoverable energy per fission (MeV/fission) value obtained for a system that is near critical (i.e., the number of fissions is equal to the number of non-fission absorptions). While this assumption is valid for a large-scale reactor system, it may not apply on a local level. For the case of a gadolinia-bearing rod or other poison rod the absorption rate may significantly exceed the fission rate. The capture reactions in gadolinium contribute prompt capture gamma-ray energy to the system that is not accounted for in the ASTM method, but may be accounted for in modern depletion computer codes (such as ORIGEN-S). The applicability of simplified methods for bumup determination needs to be carefully considered, particularly when applied to nonstandard type fuel.
The cycle power values listed in Table 4.10 for rod SF96 were therefore increased to account for the discrepancy in bumup.
55
- regular fuel pin U measured fuel pin EU K neighbors of measured fuel pin
- gadolinia fuel pin E moderator Figure 5.5 TRITON assembly model for Takahama-3 SF95 samples 56
- regular fuel pin M measured fuel pin f lU - M neighbors of measured fuel pin M
gadolinia fuel pin L moderator Figure 5.6 TRITON assembly model for Takahama-3 SF96 samples 57
0 regular fuel pin U measured fuel pin U - E neighbors of measured fuel pin n gadolinia fuel pin E moderator Figure 5.7 TRITON assembly model for Takahama-3 SF97 samples 58
6 RESULTS 6.1 TMI-1 SAMPLES Results of the simulation analysis for the TMI-1 samples from assembly NJ070G that were measured at GE-VNC are illustrated in Figures 6.1-6.4; results for the samples from assembly NJ05YU that were measured at ANL are shown in Figures 6.5-6.8. Tables 6.1 and 6.2 list the calculated-to-experimental (C/E) ratio in percentage and the corresponding average, maximum, and minimum difference, for each of the measured nuclides. The results for samples in assembly NJ070G as shown in Table 6.1 correspond to the computational models illustrated in Figures 5.2 and 5.3. The sensitivity of the results to the level of details used in the computational model is discussed in Appendix A.
The calculations for the samples from assembly NJ070G show an average overestimation of the main actinides 2 35 U and 239Pu by 3.5% and 2.0%, respectively. In the case of the minor actinides, the calculated values are on average within 30% of the experimental values. Some of the largest errors in plutonium nuclides and minor actinides are seen for the samples from comer rod 01. A good agreement is observed for neodymium nuclides, with average overestimations of less than 2% for all measured isotopes except for 143Nd, for which it is 2.5%.
The results for the samples from assembly NJ05YU show a larger overestimation for 235U and 2 39pu, of 4.7% and 14.9% on average, respectively. The average overestimation in the case of the neodymium isotopes is 0.5%, 4.5% and 8.2% for 148Nd, 145Nd, and 143Nd, respectively. The average deviation for the minor actinides is less than 30%, but the spread of the values around the mean is quite large.
In addressing the significance of the comparison between the calculated and the experimental results one should take into consideration the magnitude of the measurement uncertainties. In the case of the samples from assembly NJ05YU measured at ANL, the reported experimental uncertainties are relatively large (see Table 3.1). The total measurement uncertainties (RSD) are between about 4% and 6% for uranium nuclides and in the 5% to 8% range for plutonium nuclides. As reported for the other measured fission products, relatively large experimental errors were seen for neodymium nuclides, in the 5% to 7% range.
The total measurement uncertainty for 148Nd, used to estimate the sample burnup, is very large. The large uncertainty in the burnup value used for calculations will consequently propagate into additional uncertainty in calculated nuclide concentrations and comparisons with measurements.
As previously mentioned, assemblies NJ070G and NJ05YU were removed from the core at the end of cycle 10 to investigate fuel failures that occurred during that cycle. As a result, the fuel condition may not be well known. As the large deviations between calculation and measurement observed for samples from assembly NJ05YU were not observed for samples from assembly NJ070G, the large differences may be related more to the measurement methods and accuracies than to the unknown fuel condition due to fuel failure.
59
Pu242 Pu241 Pu240 Pu239 Pu238 U236 M29.9GWdt 01 S2 M26.7 GWdtt 015S3 26.5 GWd/t 012S5 U235 M26.3 GWd/t 013S8 258 GWd/t 01S1 24.0 GWd/t 01 2S6 U234 237GWd/t012S4 M22,8GWd~t 01357
-25 -20 -15 -10 -5 0 5 10 15 20 25 C/E-1 (%)
Figure 6.1 TMI-I samples from assembly NJ070G--major actinides Cm245 Cm244 M Cm243 Cm242 Am243 q S29.9 GWd/t OlS2 Am242m M26.7 GWd/t 01S3 26.5 GWd/t 012S55 M26.3 GWd/t 0 13S8 Am241 /25.8 GWd/t 01S1 24.0 GWd/t 012S6 23.7 GWd/t 012S4 Np237 I 22.8 GWd/t 013S7 Ua 0 0I I 2I 3I 4I 5
-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 C/E -1 (%)
Figure 6.2 TMI-1 samples from assembly NJ070OG--minor actinides 60
U Nd150 Nd148 29.9 GWd/t 015S2 26.7 GWd/t 01S3 Nd146 F* 26.5 GWd/t 012S5 26.3 GWd/t 013S8 25.8 GWdit OlSi 24.0 GWd/t 012S6 23.7 GWd/t 012S4 22.8 GWd/t 013S7 Nd145 Nd143
-1 0 1 2 3 4 5 C/E -1 (%)
Figure 6.3 TMI-I samples from assembly NJ070G-fission products (Nd)
Gd155 Eu153 a Eu151 Sm152 Sm151 Sm150 F M 29.9 GWd/t OlS2 Sm149 M 26.7 GWd/t 01S3 CM 26.5 GWd/t 012S5 Sm147 26.3 GWd/t 013S8
- 25.8 GWdIt 011S Cs137 24.0 GWd/t 012S6 23.7 GWd/t 012S4 Cs134 22.8 GWd/t 013S7
-50 30 10 0 10 .... 3 .... 40 C/E -1 (%)
Figure 6.4 TMI-1 samples from assembly NJ070G-fission products (Cs, Sm, Eu, Gd) 61
Pu242 557 GW t M 545 GWdIt D1A2 B1B B3J zr:7 Pu241 530 GWd/t M 52.6 GWd/t C2B 51.3 GWd/t C3 50.6 GWd/t A2 Pu240 M 5.5 GWd/t D1A4 50.2 GWd/t C1 50.1 GWdIt B2 Pu239 44.8 GWd/t 02 448 GWdt A18 Pu238 U236 U235 U234 2I I .... 40..
I I50..
-10 0 10 20 30 40 50
-50 -40 -30 -20 C/E -1 (%)
Figure 6.5 TMI-1 samples from assembly NJ05YU-major actinides Am243
ý=.A Am242m M 55.7 GWd/t D1A2 S54.5 GWdtt B1B Am241 M 53.0 GWd/t B3J M 52.6 GWd/t C2B M 51.3 GWdit C3 "i
M50.6 GWd/t A2 M .5 GWd/t D1A4 50.2 GWd/t Cl M50.1 GWd/t B2 Np237 44.8 GWd/t 02 S44.8 GWd/t AlB I
+ 4 6 8I0 . I -1 - I I -141 2- - I - I " "- I 0 I
-140-120-100 -80 40 -20 0 20 40 60 80 100 120 140 C/E-1 (%)
Figure 6.6 TMI-1 samples from assembly NJ05YU-minor actinides 62
Ag109 Rh103 557 GWdt D1A2 RulO 54.5 GWd B1B M 53.0 GWd/t B3J 52.6 GWd/t C2B M 51.3 GWdA C3 50.6 GWdIt A2 Tc99 M 50.5GWd/t D1A4 S50.2 GWdA C1 5.1 GWd/t B2 M 44.8 GWd/t D2 M44.8 GWCt A1B Mo95
-40 -20 0 20 40 60 80 100 C/E -1 (%)
Figure 6.7 TMI-1 samples from assembly NJ05YU-fission products (metallics)
Gd155 Eu155 Eu153 Eu151 Sm152 Sm151 Sm150 M155.7 GWd/tD01A2 Sm149 54.5 GWd/t B1B M53.0 GWd/t B3J Sm147 52.6 GWd/t C2B M 51.3 GWd/t C3 Cs137 M 50.6 GWd/t M150.5 A2 GWd/t D1A4 Nd148 M 50.2 GWd/t Cl M 50.1 GWd/t B2 Nd145 44.8 GWd/t D2 M 44.8 GWd/t A1B Nd143
-80 -60 -40 -20 0 20 40 60 80 C/E -1 (%)
Figure 6.8 TMI-1 samples from assembly NJ05YU-fission products (Nd, Cs, Sm, Eu, Gd) 63
Table 6.1 C/E-1 (%) for TMI-1 samples from assembly NJ070G Sample ID O!3S7 012S4 I 012S6 I01S1 013S8 012S5 I OIS3 I O1S2 Burnup a (GWd/MTU) 22.8 23.7 1 24.0 1 25.8 26.3 26.5 1 26.7 29.9 Avg Max Min U-234 -2.4 -0.1 0.5 -0.4 0.6 2.3 0.9 0.5 0.2 2.3 -2.4 U-235 3.8 3.6 1.3 3.7 4.0 3.9 2.4 5.3 3.5 5.3 1.3 U-236 -3.0 -3.9 -4.0 -3.5 -3.0 -3.4 -2.9 -2.6 -3.3 -2.6 -4.0 Pu-238 -15.2 -15.5 -22.9 -12.7 -18.6 -18.5 -18.9 -11.7 -16.8 -11.7 -22.9 Pu-239 2.2 2.3 -4.1 3.0 2.0 0.0 0.4 9.8 2.0 9.8 -4.1 Pu-240 -0.9 -0.8 -4.1 -0.8 -2.4 -3.6 -4.8 -1.3 -2.3 -0.8 -4.8 Pu-241 -6.8 -6.1 -11.8 -5.4 -8.3 -9.6 -9.6 -2.9 -7.6 -2.9 -11.8 Pu-242 -10.7 -9.8 -11.8 -8.5 -11.8 -12.2 -13.3 -11.8 -11.2 -8.5 -13.3 Np-237 -2.7 -7.6 -7.8 1.7 -3.8 -3.4 -4.7 0.5 -3.5 1.7 -7.8 Am-241 -4.6 -9.0 9.7 33.2 -6.3 -8.3 0.7 12.2 3.5 33.2 -9.0 Am-242m 4.4 -4.9 2.2 35.2 -10.2 -13.1 3.7 15.7 4.1 35.2 -13.1 Am-243 1.1 0.9 20.9 50.1 -1.7 -3.6 13.0 19.1 12.5 50.1 -3.6 Cm-242 -24.2 -28.9 -21.4 -10.7 -37.5 -34.2 -30.3 -40.0 -28.4 -10.7 -40.0 Cm-243 -29.1 -29.2 -22.5 6.8 -31.4 -34.2 -25.3 -13.5 -22.3 6.8 -34.2 Cm-244 -15.9 -16.7 -6.2 30.1 -19.4 -21.6 -7.0 2.2 -6.8 30.1 -21.6 Cm-245 -42.2 -42.7 -42.1 -8.1 -46.4 -47.9 -37.9 -23.3 -36.3 -8.1 -47.9 Nd-143 2.2 2.0 2.1 2.7 2.6 2.5 2.8 3.4 2.5 3.4 2.0 Nd-145 1.4 1.2 2.1 1.6 1.7 1.8 2.2 1.6 1,7 2.2 1.2 Nd-146 0.4 0.4 0.3 0.7 0.3 0.3 0.9 0.9 0.5 0.9 0.3 Nd-148 0.0 0.0 0.0 -0.1 0.0 0.0 0.0 0.0 0.0 0.0 -0.1 Nd-150 -0.1 0.6 -0.2 0.0 0.2 0.1 -0.3 0.3 0.1 0.6 -0.3 Cs-134 -23.1 -22.4 -23.6 -20.1 -23.4 -22.5 -21.1 -18.2 -21.8 -18.2 -23.6 Cs-137 -5.9 -4.5 -3.6 -3.4 -6.4 -4.9 -4.1 -7.2 -5.0 -3.4 -7.2 Sm-147 -1.8 -4.9 -3.7 -5.1 -2.4 -3.0 -5.2 -4.7 -3.9 -1.8 -5.2 Sm-149 13.4 11.6 8.5 14.6 19.0 18.6 13.6 28.2 15.9 28.2 8.5 Sm-150 1.7 1.0 1.4 2.0 1.6 1.0 1.4 2.4 1.6 2.4 1.0 Sm-151 28.8 27.1 18.0 30.1 24.8 24.9 24.8 32.8 26.4 32.8 18.0 Sm-152 13.8 13.4 17.9 14.7 16.9 16.6 15.4 14.2 15.4 17.9 13.4 Eu-151 33.3 19.8 11.6 24.3 29.1 28.4 20.5 39.8 25.9 39.8 11.6 Eu-153 -7.7 -8.8 -8.4 -6.1 -8.5 -8.1 -6.5 -5.1 -7.4 -5. -8.8 Gd-155 -38.7 -42.4 -47.1 -46.2 -41.9 -40.9 -48.2 -36.7 -42.8 -36.7 -48.2 a As reported in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisonsto SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-0000 11, Rev. A (April 2002).
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Table 6.2 C/E-1 (%) for TMI-1 samples from assembly NJ05YU Sample ID IA1B D2 B2 C1 DIA4I A2 C3 IC2B [ B3J I BIB ID1A2 Burnup a (GWd/MTU) 44.8 1 44.8 1 50.1 1 50.2 50.5 50.6 51.3 1 52.6 1 53.0 1 54.5 1 55.7 Avg Max Min U-234 5.6 10.0 5.6 -0.8 1.0 2.4 4.8 7.6 5.7 1.2 -3.4 3.6 10.0 -3.4 U-235 0.9 24.7 14.5 11.0 -2.5 9.0 9.2 5.2 1.7 -6.5 16.1 4.7 24.7 -16.1 U-236 4.9 0.5 2.5 1.2 3.3 1.0 4.5 8.0 2.8 4.5 3.6 3.4 8.0 0.5 Pu-238 -34.6 -12.3 7.4 5.8 -6.1 -3.5 41.9 -18.8 -7.8 -8.8 9.7 -2.5 41.9 -34.6 Pu-239 14.6 17.2 12.3 14.7 16.5 9.9 9.1 22.3 15.1 17.7 14.0 14.9 22.3 9.1 Pu-240 19.0 7.0 11.1 11.4 18.1 9.2 8.5 23.9 18.0 22.1 21.1 15.4 23.9 7.0 Pu-241 2.0 5.3 5.5 6.6 -0.6 7.3 6.9 7.0 1.1 5.0 .1.2 4.3 7.3 -0.6 Pu-242 8.6 -7.8 0.6 1.9 -1.8 1.7 4.0 7.8 -7.4 12.4 15.0 3.2 15.0 -7.8 Np-237 2.5 -4.3 1.7 2.0 6.2 1.8 6.1 9.0 4.8 9.6 12.5 4.7 12.5 -4.3 Am-241 8.2 -14.8 -13.6 -18.7 -18.3 -3.0 -0.5 -15.9 -18.4 48.4 32.4 -1.3 48.4 -18.7 Am-242m -86.3 -84.2 -85.4 -84.4 75.8 -85.7 -85.1 -17.1 4.7 31.0 133.0 -25.8 133.0 -86.3 Am-243 46.6 -1.7 -1.8 3.7 40.7 1.0 8.6 46.6 37.1 53.2 61.4 26.9 61.4 -1.8 Nd-143 4.6 14.9 8.8 12.0 2.2 14.3 15.6 7.9 4.2 3.5 2.7 8.2 15.6 2.2 Nd-145 3.4 5.8 5.3 6.2 -0.4 9.5 8.1 4.8 1.6 2.8 2.2 4. 5 9.5 -0.4 Nd-148 0.8 0.9 0.4 0.5 0.5 0.3 0.3 0.3 0.2 0.3 -0.1 0.4 0.9 -0.1
-3.6 Cs-137 -13.7 -6.7 -3.7 -6.9 -1.3 -3.8 1.3 -1.3 0.1 17.0 -2.1 17.0 -13.7 Sm-147 11.0 17.8 19.2 17.7 7.2 12.8 21.7 C 11.4 3.5 0.6 1.2 11.3 21.7 0.6 Sm-149 16.9 27.0 18.5 25.8 12.3 0.9 36.3 20.6 23.2 18.9 9.5 19.1 36.3 0.9 Sm-150 8.8 12.1 17.1 15.1 7.6 18.7 24.6 10.7 3.0 3.0 8.7 11.8 24.6 3.0 Sm-151 35.8 54.4 41.7 59.4 41.5 50.1 55.0 48.5 28.6 31.6 33.0 43.6 59.4 28.6 Sm-152 34.5 33.8 37.9 40.2 32.7 36.3 44.3 41.8 31.4 32.5 34.7 36.4 44.3 31.4 Eu- 151 1.5 -30.8 -40.8 -28.2 13.7 -47.3 -43.3 6.3 -4.0 31.0. 17.9 -11.3 31.0 -47.3 Eu-153 7.6 1.7 8.2 8.6 4.8 7.1 15.8 11.1 5.3 7.5 8.0 7.8 15.8 1.7 Eu-155 -61.5 -59.2 -55.5 -58.8 -63.2 -53.9 -52.5 -50.7 -52.2 -66.7 -45.9 -56.4 -45.9 -66.7
Table 6.2 C/E-1 (%) for TMI-1 samples from assembly NJ05YU (continued)
Sample D [ AIB D2 I B2 CI DIA4 A2 C3 [ C2B I B3J BIB DIA2 Burnup a (GWd/MTU) 33.8 1 44.8 1 50.1 1 50.2 1 50.5 1 50.6 1 51.3 1 52.6 1 53.0 1 54.5 1 55.7 Avg Max Min Gd-155 -52.0 -47.9 -48.4 -46.3 -65.9 -34.5 -47.5 -46.8 -51.8 -47.7 -46.9 -48.7 -34.5 -65.9 Mo-95 -3.9 8.2 -3.2 -0.8 0.5 -1.6 10.4 3.3 1.5 1.4 6.3 2.0 10.4 -3.9 Tc-99 -26.0 7.4 5.3 6.3 -3.2 7.1 13.1 -12.1 -3.5 -6.7 9.2 -0.3 13.1 -26.0 Ru-101 -6.5 9.9 -3.6 -0.4 6.1 1.2 15.6 3.5 4.3 5.7 13.0 4.4 15.6 -6.5 Rh-103 2.6 19.5 5.5 7.8 11.1 7.7 23.2 11.8 10.8 12.1 15.6 11.6 23.2 2.6 Ag-109 100.7 123.4 127.2 125.3 44.0 103.2 34.2 96.5 65.6 205.4 200.1 111.4 205.4 34.2 a As reported in J. M. Scaglione, Three Mile Island Unit I RadiochemicalAssay Comparisonsto SAS2H Calculations,Yucca Mountain Project Report, CAL-UDC-NU-00001 1, Rev. A (April 2002).
ON~
6.2 CALVERT CLIFFS SAMPLES TRITON depletion simulations were carried out for each sample by slightly adjusting, by about 1%, the power (burnup) data, as given in Table 4.7, in order to obtain a calculated 148Nd concentration in agreement with the measured value. Calculated results are illustrated in Figures 6.9-6.11 and listed in Table 6.3. The measured results that were used for comparison to calculation are those provided in Table 3.14.
Figure 6.9 shows good agreement between calculation and measurement for actinides. Computed concentrations for all uranium and plutonium nuclides, except for 2-*Pu, are within 6% of the measured values. The 238 Pu and 241 Am nuclides, both important contributors for decay heat applications, are each underestimated by about 8% on average. As observed in Figure 6.10, all cesium isotopes (except 314Cs) are predicted within about 6% of the experimental values; 1 3Cs and 137 Cs are predicted to within 1.9%
and 0.7%, respectively, on average. The nuclide 134Cs is underestimated by about 14% on average; this underprediction is consistent with results of previous analyses with SCALE. 15 Very good predictions were obtained for neodymium: all neodymium isotopes except for 151Nd were estimated on average within about 1% of the experimental values. The comparison for other measured fission products is illustrated in Figure 6.11. As seen, 9°Sr and 99Tc, important in decay heat and bumup credit applications, respectively, are well predicted, being overestimated by 2% and 9% on average.
Am241 Np237 Pu242 Pu241 Pu241 Pu239 Pu238 U238 U236 U235 44 GWd/t 87-63 M37 GWd/t 87-72 U234 l27 GWd/t 8781
-20 -15 -10 -5 0 5 10 15 20 C/E -1 (%)
Figure 6.9 Calvert Cliffs samples-actinides 67
Cs137 Cs135 U
Cs134 Cs133 Ndl50 Nd146 Nd148 Nd145 I
Nd144 44 GWd/t 87-63 37 GWd/t 87-72 1 Nd143 27 GWd/t 87-81 F-
-25 -20 -15 -10 -5 0 5 10 C/E -1 (%)
Figure 6.10 Calvert Cliffs samples-fission products (Nd, Cs)
Tc99 Sr9O Gd155 Eu155 Eu154 Sm154 Sm152 Sm151 Sm150 Sm149 S 44 GWd/t 87-63 Sm148 M 37 GWd/t 87-72 Sm147 27 GWd/t 87-81
-50 30 -20 -10 0 10 20 30 40 50 C/E -1 (%)
Figure 6.11 Calvert Cliffs samples-fission products (Sm, Eu, Gd, Sr, Tc) 68
Table 6.3 C/E-1 (%) for Calvert Cliffs samples Sample ID 87-81 87-72 1 87-63 Burnup (GWd/MTU) 27.35 37.12 44.34 Avg b Max b Min b U-234 -1.4 -2.7 2.2 -0.6 2.2 -2.7 U-235 -1.5 -2.4 -1.1 -1.7 -1.1 -2.4 U-236 2.1 2.4 1.8 2.1 2.4 1.8 U-238 -0.7 -0.5 -0.2 -0.4 -0.2 -0.7 Pu-238 -9.6 -7.3 -6.6 -7.8 -6.6 -9.6 Pu-239 2.5 3.5 6.2 4.1 6.2 2.5 Pu-240 0.0 0.1 0.7 0.3 0.7 0.0 Pu-241 -2.5 -2.2 -0.2 -1.6 -0.2 -2.5 Pu-242 -0.9 -0.3 -2.1 -1.1 -0.3 -2.1 Np-237 6.4 15.5 6.9 9.6 15.5 6.4 Am-241 -4.6 -9.8 -8.1 -7.5 -4.6 -9.8 Cs-133 0.7 1.9 3.1 1.9 3.1 0.7 Cs-134 -4.8 -14.8 -20.8 -13.5 -4.8 -20.8 Cs-135 6.2 5.0 4.6 5.3 6.2 4.6 Cs-137 -0.8 -0.4 -1.0 -0.7 -0.4 -1.0 Nd-143 0.5 0.9 2.2 1.2 2.2 0.5 Nd-144 -0.8 -1.0 -1.2 -1.0 -0.8 -1.2 Nd-145 -0.6 -0.9 -0.9 -0.8 -0.6 -0.9 Nd-146 0.7 0.8 0.9 0.8 0.9 0.7 Nd-148 0.0 0.0 0.0 0.0 0.0 0.0 Nd-150 2.2 3.2 4.1 3.2 4.1 2.2 Sm-147 3.8 0.3 -8.6 -1.5 3.8 -8.6 Sm-148 -1.1 -1.3 -8.3 -3.5 1.1 -8.3 Sm-149 -26.3 28.8 -42.4 -13.3 28.8 -42.4 Sm-150 6.4 8.2 4.8 6.5 8.2 4.8 Sm-151 42.7 30.9 30.3 34.6 42.7 30.3 Sm-152 23.61 30.8 24.3 26.2 30.8 23.6 Sm-154 -11.4 8.3 -6.3 -3.1 8.3 -11.4 Eu-151 -43.4 23.5 -57.4 -20.8 23.5 -57.4 Eu-152 -69.0 -48.4 -93.8 -70.4 -48.4 -93.8 Eu-153 3.2 3.0 3.4 3.2 3.6 3.0 Eu-154 -4.2 2.5 9.3 2.5 9.3 -4.2 Eu-155 -31.7 -29.9 -30.1 -30.6 -29.9 -31.7 Gd-154 -20.7 32.3 32.7 14.8 32.7 -20.7 Gd-155 -48.7 -26.4 -28.8 -34.6 -26.4 -48.7 Gd-156 -24.9 37.2 64.6 25.7 64.6 -24.9 Gd-158 -19.1 -99.4 -99.4 -72.6 -18.7 -99.4 Gd- 160 -48.6 Tc-99 5.2 6.8 14.1 8.7 14.1 5.2 Sr-90 3.2 1.1 2.4 2.2 3.2 1.1 AS provideG in u. W. Herman, S. M. Bowman, M. C. tBraoy, and C. V. ParKs, Valitdalon oj the ,bALE System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
b Shown only for isotopes measured in all three samples.
69
6.3 TAKAHAMA-3 SAMPLES The results of the simfulations are presented in Table 6.4 as calculated-to-experimental concentration ratios in percent and illustrated in Figures 6.12-6.19. The uranium isotopes are well predicted except for
- 234U, for which there is a large spread of the errors (see Figure 6.12); this could be attributed to possible uncertainties in the initial 234U concentration in the fuel. Predictions for 235U and 2 36 U are on average within 2% and 1%, respectively, of the experimental values.
As seen from Figures 6.13 to 6.15, there is a systematically large overprediction of plutonium and some higher actinides (americium, curium) in the case of samples (SF97-1, SF96-1, and SF95-1) located near the end of the active fuel length as compared to samples not subjected to possible rod end effects. The three above-mentioned samples were cut from axial locations at 16.3 cm, 17.6 cm, and 20.1 cm, respectively, from the top of the rod, corresponding to about 4 mm, 17 mm, and 41 mm distance from the end of the active fuel length. Large deviations have also been observed in previous analyses of these samples with the HELIOS code. 6 The effect is most pronounced for samples SF97-1 and SF96-1, located at a shorter distance from the end of the active fuel region than SF95-1. For example, the overestimation of 239 Pu is 32%, 22%, and 13% for samples SF97-1, SF96-1, and SF95-1, respectively. These values are very large as compared to the average overestimation corresponding to the other 14 samples, which is about 4%. All these three samples are located in a region of the fuel characterized by high leakage and large flux gradients. Although results for these samples are shown here, they would likely be excluded from code validation studies and uncertainty evaluation analyses that are usually carried out using a consistent set of experimental data typical of the average fuel behavior. However, analyses of these samples are valuable as they provide useful information on fuel characteristics for fuel regions in the proximity of the lower burnup assembly ends, regions of importance in burnup credit applications. A more appropriate representation of these samples would require a three-dimensional model; this is not possible with TRITON/NEWT, which is limited to 2-D geometry.
The results for the fission product group consisting of neodymium, cesium, cerium, and samarium isotopes as well as two metallic ruthenium and antimony nuclides of importance to bumup credit are illustrated in Figures 6.16 to 6.19. With the exception of 142Nd, which was measured only in the SF95 samples, the other neodymium nuclides (see Figure 6.16) are well predicted, with an average overestimation about I% for 14 5,146 ,'14 8"'5 Nd nuclides and an average underestimation of about 1% and 3%
for 143Nd and '44Nd, respectively. Most of samarium nuclides (see Figure 6.19) are well predicted, within 5% on average, except for .5.Sm and 151 m, for which there is a systematic overestimation in the 30%
range. Cesium isotopes (see Figure 6.17) are underestimated by about 10% and 3% in the case of 134Cs and 137Cs, respectively. The nuclides 144Ce and 154Eu are well predicted, within 1% and 4% of the measurement, on average. Note that the average deviations shown in Table 6.4 include the results corresponding to the gadolinia fuel rod SF96.
70
Table 6.4 C/E-1 (%) for Takahama-3 samples Sample ID a
SF95-1 SF95-2 SF95-3 SF95-4 SF95-5 I SF97-1I SF97-2 SF97-3 [ SF97-4 SF97-5 SF97-6 I SF96-1 I SF96-2 SF96-3 SF96-4 SF96-5 Burnup (GWd/MTIY 14.3 24.4 35.4 36.7 30.4 17.69 30.7 42.16 47.03 1 47.25 40.79 8.55 I17.38 I29.58 30.35 25.35 1 8.55 173 1 295 30.3 25.-35_
Avg Max Min U-234 9.4 -1.4 26.4 25.1 -8.2 6.4 9.6 6.8 -6.6 7.2 8.4 -7.6 -5.9 -7.6 -7.9 -6.9 3.8 26.4 -8.2 U-235 0.9 2.7 3.1 3.9 2.1 3.3 0.7 1.4 1.9 0.6 2.0 1.9 1.9 33 1.5 1.9 2.1 3.9 0.6 U-236 0.6 -1.7 -0.1 -0.3 -1.0 1.2 -0.2 -0.7 -2.0
-0.3 -0.4 -0.3 -1.6 -0.4 -0.8 -0.4 -0.5 1.2 -2.0 Pu-238 8.9 -3.7 4.9 4.4 2.0 31.7 -5.5 -7.6 20.4
-10.4 -12.6 -7.0 -1.4 -1.2 -6.8 1.1 1.1 31.7 -12.6 Pu-239 12.8 8.0 8.7 7.9 8.3 31.7 4.0 4.1 4.9 3.3 0.8 22.0 3.8 1.7 0.5 2.2 7.8 31.7 0.5 Pu-240 7.0 3.4 5.9 6.2 6.2 14.9 7.0 7.6 6.3 5.8 7.3 13.2 5.5 7.4 7.3 9.0 7.5 14.9 3.4 Pu-241 11.4 0.6 1.0 1.6 2.3 30.4 -2.6 -2.5 -3.0 -5.4 -1.4 26.4 1.9 1.2 -2.5 2.3 3.9 30.4 -5.4 Pu-242 13.6 0.0 -0.8 -0.1 3.1 23.3 1.0 -1.1 -2.4 -2.9 -0.8 21.9 5.6 3.9 0.7 7.9 4.6 23.3 -2.9 Np-237 24.5 1.8 3.8 1.2 -1.9 -0.6 36.5 45.1 65.9 58.5 52.6 26.1 65.9 -1.9 Am-241 -6.5 23.4 23.9 70.3 18.7 74.5 29.2 27.9 15.4 10.4 31.9 67.6 45.0 29.1 9.7 47.2 32.4 74.5 -6.5 Am-242m 22.5 17.5 17.3 19.7 22.5 111.4 25.8 18.0 6.7 1.7 18.9 31.0 9.0 23.5 2.6 10.8 22.4 111.4 1.7 Am-243 35.8 22.2 21.7 23.8 23.8 77.0 15.3 13.6 10.9 9.0 15.4 65.4 25.8 24.2 16.5 29.5 26.9 77.0 9.0 Crn-242 -16.0 -30.5 -37.5 -45.5 -18.1 19.6 -4.4 2.2 0.3 1.9 6.9 12.6 -18.6 -17.8 -22.8 -14.4 -11.4 19.6 -45.5 Cm-243 -6.3 -19.2 -10.5 -13.5 -20.9 29.9 -18.5 -17.4 -17.9 -20.6 -17.8 -12.1 29.9 -20.9 Cm-244 30.3 -0.3 8.1 5.8 13.0 86.6 -2.3 -4.0 -6.8 -9.3 -1.1 51.8 7.3 6.7 -0.9 13.1 12.4 86.6 -9.3 Cm-245 13.9 -21.5 -13.9 -19.7 -14.1 79.9 -28.4 -30.2 -33.1 -36.3 -26.2 -11.8 79.9 -36.3 Cmn-246 -45.1 -66.1 -21.2 -23.8 25.0 48.2 -93.2 -34.0 -38.2 -40.7 -33.5 -29.3 48.2 -93.2 Crn-247 -31.5 -32.1 -35.9 -38.4 -36.0 -34.8 -31.5 -38.4 Nd-142 -19.0 -7.3 -15.3 -13.1 -5.4 -12.0 -5.4 -19.0 Nd-143 -2.0 -2.0 -1.9 -1.0 -1.4 0.4 0.2 0.2 0.7 -0.2 0.3 -4.2 -2.6 -1.9 -1.8 -3.0 -1.3 0.7 -4.2 Nd-144 0.0 -1.6 -1.0 -4.7 -1.2 3.3 1.4 -2.1 -2.3 -6.8 -6.0 -10.3
-3.9 -4.3 -8.0 -7.1 -3.4 3.3 -10.3 Nd-145 0.3 -0.3 -0.5 -0.1 0.2 0.4 1.3 0.9 1.1 1.1 0.8 -1.2 0.3 0.4 1.0 0.0 0.4 1.3 -1.2 Nd-146 2.8 1.8 1.7 1.6 2.3 2.4 1.5 0.8 0.7 0.6 1.0 -1.0 0.1 0.3 0.5 0.1 1.1 2.8 -1.0 Nd-148 1.0 -0.3 -0.5 -0.7 -0.1 1.7 0.9 0.2 0.0 0.0 0.4 -0.2 0.0 -0.1 -0.1 0.0 0.1 1.7 -0.7 Nd-150 -0.2 0.4 -0.4 -0.3 1.1 4.3 1.9 1.6 1.3 1.4 2.1 0.7 0.7 1.3 1.0 1.7 1.2 4.3 -0.4 Cs-134 -8.6 -13.9 -12.5 -12.7 -12.8 -5.4 -19.7 -17.6 -14.9 -15.4 -16.2 1.0 -9.1 1.1 -19.7 1.1 0.9 1.0 0.9 Cs-137 -1.6 1 -3.0 1 -2.3 -2.7 -1.6 -2.1 -2.3 -2.6 -1.8 1 -2.0 -2.6 5. -8.0 I -4.8 -6.0 -3.9 -2.6 5.2 -8.0 I I I I
Table 6.4 C/E-1 M) for Takahama-3 samDles (continued)
Sample ID SF95-i IF95-2 SF95-3 SF95-4 SF95- S SF97-1 S9F972 97-3 SF97-4 SF97-5 F97-6 SF96-1 SF96-2 I SF96-3 SF96-4 SF96-5 Burnup' (GWdI MTUI 14.3 24.4 35.4 36.7 30.4 17.69 30.7 42.16 47.03 47.25 40.79 8.55 17.38 29.58 30.35 25.35 Avg Max Min Ce-144 -2.1 -2.0 -5.0 4.2 -1.7 -17.7 -10.2 -2.7 6.0 6.7 -4.8 -0.3 5.1 15.5 15.4 7.9 09 15.5 -17.7 Eu-154 7.3 0.8 5.1 1.8 3.2 21.4 -2.7 1.3 1.9 0.3 2.4 5.0 -2.3 7.9 5.3 7.6 4.1 21.4 -2.7 Ru-106 2.8 20.5 29.2 30.4 12.5 -6.5 -4.7 -2.2 8.3 80.4 -15.0 27.5 53.0 36.3 50.5 12.7 21.0 80.4 -15.0 Sb- 125 94.9 86.3 126.3 175.9 112.8 31.4 22.5 84.4 70.6 38.6 91.3 30.5 52.8 118.7 75.2 79.6 80.7 175.9 22.5 Sm-147 -2.3 0.6 -1.7 -2.9 -2.3 -0.9 -1.6 0.6 -2.9 Sm-148 17.8 -4.2 -9.6 -12.7 -13.6 -8.1 -5.1 17.8 -13.6 Sm-149 3.2 -8.5 -2.5 6.1 6.3 -3.0 0.3 6.3 -8.5 Sm-150 5.4 4.9 5.1 5.0 4.4 6.4 5.2 6.4 4.4 Sm-151 51.4 30.0 34.3 33.3 28.0 33.9 35.2 51.4 28.0 Sm-152 7.2 23.1 29.1 30.8 30.3 26.3 24.5 30.8 7.2 Sm-154 3.5 0.2 0.3 -0.2 -1.1 1.0 0.6 3.5 -1.1
'aAS stated in Y. Nakahara, Y. Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent LWR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/01), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002). -
U236 4,
I L
SF96-5 SF96-4 U235 SF96-3 SF96-2 SF96-1 SF97-6 SF97-5 SF97-4 SF97-3 SF97-2 SF97-1 U234 SF95-5 SF95-4 a w SF95-3 SF95-2 SF95-1
-10 -5 0 5 10 15 20 25 30 (C/E-1) (%)
Figure 6.12 Takahama-3 samples-uranium nuclides Lý-
Pu242 bm Pu241 l m 1 SF96-5 1 SF96-4 Pu240 M SF96-3 M SF96-2 SF96-1
-V97-6 SF97-5 Pu239 - SF97-4 l SF97-3 SF97-2 sF97-I U I SF95-5 Pu238 -MF SF95-4 SF95-3
~1~~~~ SF95-2 SF95-1 m
-20 -10 0 10 20 30 40 (C/E-i1) (0/)
Figure 6.13 Takahama-3 samples-plutonium nuclides 73
Am243
-V Am242m M SF96-5 M SF96-4 M SF96-3 n SF96-2 NSF96-i M SF97-6 Am241 M SF97-5 n SF97-4 M SF97-3 M SF97-2 M SF97-1 M SF95-5 I SF95-4 Np237 m SF95-3 m SF95-2 M SF95-1
-20 0 20 40 60 80 100 120 (CiE-1) (%)
Figure 6.14 Takahama-3 samples--minor actinides (Np, Am)
Cm247 Cm246 Cm245 M. ____________ miSF96-5 SF96-4 SF96-3 SF96-2 I SF96-1 Cm244 m SF97-6 I SF97-5
. SF97-4 SF97-3 Cm243 SF97-2 I SF97-1
.SF95-5 M SF95-4 M SF95-3 Cm242 M SF95-2 M SF95-1
-100 -80 40 -20 0 20 40 60 80 100 (C/E-1) (%)
Figure 6.15 Takahama-3 samples-minor actinides (Cm) 74
Nd150 d
Nd148 Nd146 SF96-5 SF96-4 Nd145 L SF96-3 SF96-2 SF96-1 SF97-6 Nd144 SF97-5 SF97-4 SF97-3 SF97-2 Nd143 SF97-1 SF95-5 SF95-4 Nd142 MSF95-3 MSF95-2 N
o ______________ SF95-1 2 5 I
-20 -15 -10 -5 0 5 (C/E-1) (%)
Figure 6.16 Takahama-3 samples-fission products (Nd)
Eu154 EMPNIIIIIIIIIIIIIIIII Ce144 M SF96-5 M SF96-4 SF96-3 I SF96-1 SF97-6 SF97-5 Cs137 SF97-4 SF97-3 SF97-2 IMSM-1 i M SF95-5 M SF95-4 M SF95-3 Cs134 M SF95-2 M SF95-1 i v . . . . .. . . . .
20 10 -5 0 5 10 15 20 25 (C/E-1) (%)
Figure 6.17 Takahama-3 samples-fission products (Cs, Ce, Eu) 75
T Sb125
- SF96-5 iSF96-4 SF96-3 M SF96-2 MSF96-i M SF97-6 M SF97-5 M SF97A4 i SF97-3 Ru 106 M SF97-2 M SF97-1 i SF95-5 M SF96-4 M SF95-3 M SF95-2 M SF95-I
.... ...... i ...
i....i i'...i.... i.... i.... 1
-20 0 20 40 60 80 100 120 140 160 180 200 (C/E-1) (%)
Figure 6.18 Takahama-3 samples-fission products (metallics)
Sm154 Sm152 Sm151 Sm 150 Sm149 SF97-6 Sm148 M SF97-5 SF97-4
.. SF97-3 Sm147 SF97-2 SF97-1
'... 'I' "'I ........
' I ' 'I ' I
-20 -10 0 10 20 30 40 50 60 Figure 6.19 Takahama-3 samples-fission products (Sm) 76
7
SUMMARY
The purpose of the work described in this report was to evaluate available isotopic measurements involving high burnup fuel and to analyze the data using the ORNL SCALE computer code system. This information is needed to assess and quantify the uncertainties associated with the high burnup fuel characteristics of importance for spent fuel storage applications involving decay heat, radiation sources, criticality with bumup credit, and for reactor safety studies. Previously available experimental data for low- and medium-range burnup fuel was also considered so that the set of data used for uncertainty evaluations would cover a large bumup range, which would allow possible trends with high burnup to be evaluated.
The measurements analyzed in this report include 38 spent fuel samples from fuel irradiated in three PWRs operated in the United States and Japan. The samples cover a large burnup range, from 14 to 56 GWd/MTU, and an initial fuel enrichment domain from 2.6 to 4.7 wt % 235U. Twenty-two of the 38 samples considered are of domestic origin (TMI-1 and Calvert Cliffs-1 reactors) and 16 are from experiments carried out in Japan (Takahama-3 reactor). Information is presented on the fuel assembly geometry, irradiation history, and sample burnup. This information is necessary for developing a computational model to simulate the irradiation and decay of the samples under consideration. The data are presented in sufficient detail to allow an independent analysis to be performed.
The analysis of the measurements in this report was carried out by employing the two-dimensional depletion sequence of the TRITON module in the SCALE computer code system. Individual TRITON models were developed for each of the samples considered, including as many geometry and irradiation history details as available. The results of the simulations reported here were obtained using the fuel sample burnup that reproduced, within the experimental uncertainty margins, the measured concentration of the fission product burnup indicator 148Nd.
Some of the key modeling issues in isotopic assay data analysis are discussed in Appendix A in relation to the analysis of TMI-1 samples. The effect on predicted nuclide concentrations of modeling details, such as information on nearest assemblies (enrichment, burnup) or poison rod location was assessed and shown to be significant for samples selected from edge rods.
77
8 REFERENCES
- 1. SCALE: A ModularCode System for PerformingStandardized ComputerAnalyses for Licensing Evaluation,ORNL/TM-2005/39, Version 5.1, Vols. 1-111, Oak Ridge National Laboratory, Oak Ridge, Tennessee, November 2006. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-732.
- 2. S. F. Wolf, D. L. Bowers, and J. C. Cunnane, Analysis ofSpent Nuclear Fuel Samples from Three Mile Islandand Quad Cities Reactors. Final Report, Argonne National Laboratory, Argonne, Illinois (November 2000).
- 3. R. D. Reager and R. B. Adamson, TRW Yucca Mountain Project, Test Report, Phase 2, Ref. TRW Purchase Order No. A09112CC8A, GE Nuclear Energy.
- 4. J. M. Scaglione, Three Mile Island Unit 1 RadiochemicalAssay Comparisons to SAS2H Calculations, Yucca Mountain Project Report, CAL-UDC-NU-00001 1, Rev. A (April 2002).
- 5. I. C. Gauld, Strategiesfor Application of Isotopic Uncertaintiesin Burnup Credit, NUREG/CR-6811 (ORNL/TM-2001/257), Oak Ridge National Laboratory, Oak Ridge, Tennessee (June 2003).
- 6. C. E. Sanders and I. C. Gauld, Isotopic Analysis of High-BurnupPWR Spent Fuel Samples from the Takahama-3 Reactor, NUREG/CR-6798 (ORNL/TM-2001/259), Oak Ridge National Laboratory, Oak Ridge, Tennessee (January 2003).
- 7. N. Thiollay, J. P Chauvin, B. Roque, A. Santamarina, J. Pavageau, J. P. Hudelot, and H. Toubon, "Burnup Credit for Fission Product Nuclide in PWR (U0 2) Spent Fuels," International Conference on Nuclear Criticality Safety (ICNC 99), Versailles, France (1999).
- 8. www.epri.com, abstract for "TMI-1 Cycle 10 Fuel Rod Failures," TR-108784 (October 1998).
- 9. R. J Guenther et.al., CharacterizationofLWR Spent Fuel MCC-Approved Testing MaterialATM-104, PNL-5109-104 (1991).
- 10. M. C. Brady-Raap and R. J. Talbert, Compilation of RadiochemicalAnalyses of Spent NuclearFuel Samples, PNNL-13677, Pacific Northwest National Laboratory, Richland, Washington (September 2001).
- 11. S. R. Bierman and R. J. Talbert, Benchmark Datafor ValidatingIrradiatedFuel Compositions Used in CriticalityCalculations,PNL-10045, Pacific Northwest Laboratory (October 1994).
- 12. S. R. Bierman, "Spent Reactor Fuel Benchmark Composition Data for Code Validation," Proceedings of InternationalConference on Nuclear CriticalitySafety, Oxford, United Kingdom (September 1991).
- 13. http://www.nea.fr/html/science/wpncs/sfcompo/, SFCOMPO-Spent Fuel Isotopic Composition Database, operated by the NEA Nuclear Science Division under the supervision of the Working Party on Nuclear Criticality Safety.
- 14. ARIANE InternationalProgramme-FinalReport, ORNL/SUB/97-XSV750- 1, Oak Ridge National Laboratory, Oak Ridge, Tennessee (May 2003).
- 15. 0. W. Herman, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System for PWR Spent FuelIsotopic CompositionAnalyses, ORNL/TM-12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee (March 1995).
- 16. A. A. Rimski-Korsakov, A. V. Stepanov, and A. D. Kirikov, RadiochemicalAnalysis of Spent Reactor Fuel Sample-Reportof Results, V. G. Khlopin Institute, St. Petersburg, Russia, communication to PNNL (1993).
- 17. Y. Nakahara, K Suyama, and T. Suzaki, Technical Development on Burnup Creditfor Spent L WR Fuels, JAERI-Tech 2000-071 (ORNL/TR-2001/0 1), English Translation, Oak Ridge National Laboratory, Oak Ridge, Tennessee (2002).
- 18. L. B. Wimmer, Summary Report of CommercialReactor Criticality Datafor Three Mile Island Unit 1, TDR-UDC-NU-000004 REV 01, Bechtel SAIC Company, LLC, Las Vegas, Nevada (August 2001).
- 19. 2004 World Nuclear Industry Handbook,edited by Nuclear Engineering International (2005).
- 20. "Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Neodymium-148 Method)," American Society for Testing and Materials, ANSI/ASTM E 321-79.
79
APPENDIX A EFFECT OF MODELING DETAILS ON PREDICTED NUCLIDES FOR TMI-1 SAMPLES
)
A.I Assembly NJ05YU The transport calculation for a fuel assembly in TRITON usually assumes a reflective boundary condition, given the unavailability in most cases of detailed information on the assembly surroundings.
This is expected to be a reasonable approximation for assemblies located in a generally uniform core and to not significantly influence the depletion of fuel rods located far from the assembly edge. As the H6 fuel rod, from which the samples were selected, was located toward the center of the assembly, and therefore far from the assembly edge, a more accurate representation of the assembly environment (i.e.,
more rigorous boundary condition) is expected to have less influence on the flux in the samples under consideration. However, the magnitude of the effect would depend on the characteristics (burnup, fuel enrichment, etc.) of the assemblies surrounding the NJ05YU assembly. The effect of using a more rigorous boundary condition in the depletion simulation on the predicted nuclide concentrations for this type of assembly is analyzed in this Appendix.
Data were available from Ref. 18 on the assemblies surrounding assembly NJ05YU during cycles 9 and 10, as illustrated in Figures A. 1 and A.2, respectively. During cycle 9, all the first-order neighboring assemblies (at N, S, E and W locations) were from batch IOB with an initial fuel enrichment of 3.63%;
these assemblies were irradiated since cycle 8 and did not contain BPRs or gadolinia fuel rods. The assembly located at the NW position was also from batch 10B; whereas, the assembly at the NE position was from batch 11 B and had BPRs with a load of 1.1 wt % B 4C. Assemblies at SW and SE were from batch I IC and had BPRs with 2.1 wt % B 4 C. The burnup of the assemblies from batch IOB at the.
beginning of cycle 9 was not known. As shown in Figure A.2, the arrangement of the assemblies surrounding NJ05YU is symmetric. During cycle 10, the first-order neighbors of NJ05YU at N, S, E, and W were fresh fuel assemblies with 4.65 wt % 235U initial enrichment and containing BPRs with 2.1% B 4C load. The assemblies located at NW, NE, SW, and SE were all from batch 11, with 4.0 wt % 235U initial enrichment and no BPRs present.
The first computational model (called "model #1") represents a half assembly, as illustrated in Figure 5.1.
The second TRITON model (called "model #2") represents half of the test assembly NJ05YU and surrounding assemblies, taking advantage of the configuration symmetry as shown in Figures A. 1 and A.2. The TRITON model in this case is illustrated in Figure A.3.
A-I
NW N NE Batch I OB Batch IOB Batch 1IB Enrich = 3.63% Enrich = 3.63% Enrich = 4%
BP load= 1.1%
w E Batch 10B Batch 10B Enrich = 3.63% Enrich = 3.63%
SW S SE Batch 1IC Batch IOB Batch 11C Enrich = 4% Enrich = 3.63% Enrich = 4%
BP load= 2.1% BP load= 2.1%
Figure A.1 Assemblies surrounding NJ05YU during cycle 9 NW N NE Batch 11C Batch 12B Batch 11A Enrich = 4% Enrich = 4.65% Enrich = 4%
BP load = 2.1%
W E Batch 12B Batch 12B Enrich = 4.65% Enrich = 4.65%
BP load= 2.1% BP load= 2.1%
SW S SE Batch 11A Batch 12B Batch 11C Enrich = 4% Enrich = 4.65% Enrich = 4%
BP load= 2.1%
Figure A.2 Assemblies surrounding NJ05YU during cycle 10 A-2
- regular fuel pin in test assembly U regular fuel pin in test assembly Etest fuel pin
- test fuel pin
- - EK neighbors of test pin U FEE- [ neighbors of test pin
- pins in N and S assembly r pins in N and S assembly pins in E assembly pins in E assembly F pins in SE assembly F pins in SE assembly 0 pins in NE assembly
- pins in NE assembly
- BPR absorber in test assembly
- BPR absorber in N, S, E assembly
- BPR absorber in NE assembly
- BPR absorber in SE assembly (a) cycle 9 (b) cycle 10 Figure A.3 TRITON model #2 for TMI-1 samples in assembly NJ05YU A-3
The two modeling approaches described above were used to assess the effect of modeling the assembly surroundings on the calculated nuclide inventory. All calculations reported here used the sample burmup as8 provided in Ref. 4, with no power (burnup) adjustment done to match the measured concentration of 14Nd because the calculated value for this nuclide was within one standard deviation reported for the measurement. A comparison of the C/E average and standard deviation over all II samples obtained with each of the two computational models is presented in Table A. 1 and illustrated in Figure A.4 for uranium and plutonium nuclides. Results for all samples and both modeling approaches are shown in Tables A.2 and A.3.
For the considered samples, the effect of modeling the assemblies surrounding the assembly NJ05YU has small, but not a major impact. In the case of uranium isotopes, for example, the average overprediction decreases from 4.7% to 3.9% for 231U, and remains practically unchanged for 234 U and 2 36 U when going from model #1 to model #2. For plutonium nuclides, the change is less than 0.5% for all measured nuclides except for 2 39 pu for which there is a 1.3% decrease in the average overestimation, from 14.9% to 13.6%. The change in fission products average C/E does not exceed 1.4%. Note though that part of the differences observed may be due to the modeling of the assemblies from batch 1OB during cycle 9 as fresh fuel, as their burnup was not known.
A-4
Table A.1 Effect of modeling assumptions on C/E-1 (%)
for samples from assembly NJ05YU Model #1' Model H2b Nuclide ID Avg a Max Min Avg a Max Min U-234 3.6 3.9 10.0 -3.4 3.5 3.9 10.0 -3.5 U-235 4.7 11.0 24.7 -16.1 3.9 11.0 24.0 -16.9 U-236 3.4 2.2 8.0 0.5 3.4 2.2 8.2 0.6 Pu-238 -2.5 19.4 41.9 -34.6 -2.6 19.3 41.6 -34.7 Pu-239 14.9 3.7 22.3 9.1 13.6 3.7 20.9 7.9 Pu-240 15.4 6.1 23.9 7.0 15.0 6.0 23.5 6.8 Pu-241 4.3 2.8 7.3 -0.6 3.9 2.8 6.9 -1.0 Pu242 3.2 7.4 15.0 -7.8 3.7 7.4 15.7 -7.4 Nd- 143 8.2 5.2 15.6 2.2 7.9 5.2 15.2 1.9 Nd-145 4.5 2.9 9.5 -0.4 4.4 2.9 9.4 -0.5 Nd-148 0.4 0.3 0.9 -0.1 0.3 0.3 0.8 -0.1 Cs-137 -2.1 7.5 17.0 -13.7 -2.1 7.5 16.9 -13.7 Eu-151 -11.3 27.3 31.0 -47.3 -13.2 27.0 29.3 -47.8 Eu-153 7.8 3.6 15.8 1.7 7.7 3.6 15.8 1.6 Eu-155 -56.4 6.1 -45.9 -66.7 -56.5 6.1 -46.1 -66.8 Sm-147 11.3 7.4 21.7 0.6 11.1 7.5 21.5 0.2 Sm-149 19.1 9.5 36.3 0.9 18.2 9.4 35.3 0.2 Sm-150 11.8 6.6 24.6 3.0 11.7 6.7 24.5 2.8 Sm-151 43.6 10.6 59.4 28.6 42.2 10.5 57.9 27.3 Sm- 152 36.4 4.2 44.3 31.4 36.2 4.2 44.2 31.3 Gd-155 -48.7 7.3 -34.5 -65.9 -48.9 7.3 -34.7 -66.0 Am-241 -1.3 22.6 48.4 -18.7 -1.8 22.4 47.3 -19.1 Am-242m -25.8 78.1 133.0 -86.3 -26.6 77.1 130.3 -86.4 Am-243 26.9 24.8 61.4 -1.8 27.5 24.9 62.3 -1.3 Np-237 4.7 4.7 12.5 -4.3 4.0 4.6 11.7 -4.9 Mo-95 2.0 4.6 10.4 -3.9 2.0 4.6 10.4 -3.9 Tc-99 -0.3 11.5 13.1 -26.0 -0.4 11.5 13.0 -26.1 Ru-101 4.4 6.7 15.6 -6.5 4.4 6.7 15.5 -6.5 Rh-103 11.6 6.0 23.2 2.6 11.3 6.0 22.8 2.4 Ag-109 111.4 55.1 205.4 34.2 111.4 55.0 204.9 34.2 As illustrated in Figure 5.1.
bAs illustrated in Figure A.3.
A-5
20-model #1 model #2 1 15-U.I 5-0
-5I.Ihn U-234 U-235 U-236 Pu-238 Pu-239 Pu-240 Pu-241 Pu242 Figure A.4 Effect of modeling assumptions on U and Pu-assembly NJ05YU A-6
Table A.2 C/E-1 (%) for samples in assembly NJ05YU - computational model #1 Sample ID I A1B I D2 I B2 CI I DIA4 I A2 C3 I C2B I B3J I RiB I DIA2 Burnup (GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 Avg Max Min U-234 5.6 10.0 5.6 -0.8 1.0 2.4 4.8 7.6 5.7 1.2 -3.4 3.6 10.0 -3.4 3.9 U-235 0.9 24.7 14.5 11.0 -2.5 9.0 9.2 5.2 1.7 -6.5 -16.1 4.7 24.7 -16.1 11.0 U-236 4.9 0.5 2.5 1.2 3.3 1.0 4.5 8.0 2.8 4.5 3.6 3.4 8.0 0.5 2.2 Pu-238 -34.6 -12.3 7.4 5.8 -6.1 -3.5 41.9 -18.8 -7.8 -8.8 9.7 -2.5 41.9 -34.6 19.4 Pu-239 14.6 17.2 12.3 .14.7 16.5 9.9 9.1 22.3 15.1 17.7 14.0 14.9 22.3 9.1 3.7 Pu-240 19.0 7.0 11.1 11.4 18.1 9.2 8.5 23.9 18.0 22.1 21.1 15.4 23.9 7.0 6.1 PU-241 2.0 5.3 5.5 6.6 -0.6 7.3 6.9 7.0 1.1 5.0 1.2 4.3 7.3 -0.6 2.8 Pu242 8.6 -7.8 0.6 1.9 -1.8 1.7 4.0 7.8 -7.4 12.4 15.0 3.2 15.0 -7.8 7.4 Nd-143 4.6 14.9 8.8 12.0 2.2 14.3 15.6 7.9 4.2 3.5 2.7 8.2 15.6 2.2 5.2 Nd-145 3.4 5.8 5.3 6.2 -0.4 9.5 8.1 4.8 1.6 2.8 2.2 4.5 9.5 -0.4 2.9.
Nd-148 0.8 0.9 0.4 0.5 0.5 0.3 0.3 0.3 0.2 0.3 -0.1 0.4 0.9 -0.1 0.3 Cs-137 -13.7 -6.7 -3.7 -6.9 -1.3 -3.8 1.3 -3.6 -1.3 0.1 17.0 -2.1 17.0 -13.7 7.5 Eu-151 1.5 -30.8 -40.8 -28.2 13.7 -47.3 -43.3 6.3 -4.0 31.0 17.9 -11.3 31.0 -47.3 27.7 Eu-153 7.6 1.7 8.2 8.6 4.8 7.1 15.8 11.1 5.3 7.5 8.0 7.8 15.8 1.7 3.6 Eu-155 -61.5 -59.2 -55.5 -58.8 -63.2 -53.9 -52.5 -50.7 -52.2 -66.7 -45.9 -56.4 -45.9 -66.7 6.1 Sm-147 11.0 17.8 19.2 17.7 7.2 12.8 21.7 11.4 3.5 0.6 1.2 11.3 21.7 0.6 7.4 Sm-149 16.9 27.0 18.5 25.8 12.3 0.9 36.3 20.6 23.2 18.9 9.5 19.1 36.3 0.9 9.5 Sm-150 8.8 12.1 17.1 15.1 7.6 18.7 24.6 10.7 3.0 3.0 8.7 11.8 24.6 3.0 6.6 Sm-151 35.8 54.4 41.7 59.4 41.5 50.1 55.0 48.5 28.6 31.6 33.0 43.6 59.4 28.6 10.6 Sm-152 34.5 33.8 37.9 40.2 32.7 36.3 44.3 41.8 31.4 32.5 34.7 36.4 44.3 31.4 4.2 Gd-155 -52.0 -47.9 -48.4 -46.3 -65.9 -34.5 -47.5 -46.8 -51.8 -47.7 -46.9 -48.7 -34.5 -65.9 7.3 Am-241 8.2 -14.8 -13.6 -18.7 -18.3 -3.0 -0.5 -15.9 -18.4 48.4 32.4 -1.3 48.4 -18.7 22.6 Am-242m -86.3 -84.2 -85.4 -84.4 75.8 -85.7 -85.1 -17.1 4.7 31.0 133.0 -25.8 133.0 -86.3 78.1 Arn-243 46.6 -1.7 -1.8 3.7 40.7 1.0 8.6 46.6 37.1 53.2 61.4 26.9 61.4 -1.8 24.8 Np-237 2.5 -4.3 1.7 2.0 6.2 1.8 6.1 9.0 4.8 9.6 12.5 4.7 12.5 -4.3 4.7 Mo-95 -3.9 8.2 -3.2 -0.8 0.5 -1.6 10.4 3.3 1.5 1.4 6.3 2.0 10.4 -3.9 4.6 Tc-99 -26.0 7.4 5.3 6.3 -3.2 7.1 13.1 -12.1 -3.5 -6.7 9.2 -0.3 13.1 -26.0 11.5 Ru-101 -6.5 9.9 -3.6 -0.4 6.1 1.2 15.6 3.5 4.3 5.7 13.0 4.4 15.6 -6.5 6.7 Rh-103 2.6 19.5 5.5 7.8 11.1 7.7 23.2 11.8 10.8 12.1 15.6 11.6 23.2 2.6 6.0 Ag-109 100.7 123.4 127.2 125.3 44.0 103.2 34.2 96.5 65.6 205.4 200.1 111.4 205.4 34.2 55.1
Table A.3 C/E-1 (%) for samples in assembly NJ05YU - computational model #2 Sample ID A1B D2 B2 C1I DIA4I A2 C3 C2B B3J I BIB IDIA2 Burnup (GWd/MTU) 44.8 1 44.8 1 50.1 1 50.2 50.5 1 50.6 51.3 1 52.6 53.0 1 54.5 55.7 Avg Max Min a U-234 5.5 10.0 5.5 -0.9 0.9 2.3 4.7 7.5 5.6 0.9 -3.5 3.5 10.0 -3.5 3:9 U-235 0.4 24.0 13.7 10.2 -3.3 8.3 8.4 4.4 0.8 -7.5 -16.9 3.9 24.0 -16.9 11.0 U-236 5.0 0.6 2.6 1.3 3.4 1.1 4.6 8.2 2.9 4.4 3.7 3.4 8.2 0.6 2.2 Pu-238 -34.7 -12.5 7.2 5.6 -6.2 -3.7 41.6 -18.9 -7.9 -9.1 9.5 -2.6 41.6 -34.7 1.9.3 Pu-239 13.6 16.2 11.0 13.5 15.3 8.7 7.9 20.9 13.8 16.1 12.6 13.6 20.9 7.9 3.7 Pu-240 18.7 6.8 10.7 11.1 17.7 8.9 8.1 23.5 17.5 21.4 20.6 15.0 23.5 6.8 6.0 Pu-241 1.7 5.1 5.1 6.3 -1.0 6.9 6.4 6.6 0.6 4.4 0.7 3.9 6.9 -1.0 2.8 Pu242 9.1 -7.4 1.1 2.5 -1.2 2.2 4.5 8.4 -6.9 12.8 15.7 3.7 15.7 -7.4 7.4 Nd-143 4.5 14.6 8.4 11.8 1.9 13.9 15.2 7.5 3.8 3.0 2.4 7.9 15.2 1.9 5.2 Nd-145 3.3 5.7 5.2 6.1 -0.5 9.4 8.0 4.7 1.5 2.5 2.1 4.4 9.4 -0.5 2.9 Nd-148 0.7 0.8 0.4 0.5 0.4 0.2 0.3 0.3 0.1 0.0 -0.1 0.3 0.8 -0.1 0.3 Cs-137 -13.7 -6.7 -3.8 -7.0 -1.4 -3.8 1.3 -3.6 -1.4 -0.1 16.9 -2.1 16.9 -13.7 7.5 Eu-151 0.6 -31.4 -41.4 -29.0 12.6 -47.8 -43.9 5.2 -5.0 29.3 16.6 -12.2 29.3 -47.8 27.3 00 Eu-153 7.5 1.6 8.1 8.6 4.8 7.1 15.8 11.1 5.3 7.2 7.9 7.7 15.8 1.6 3.6 Eu-155 -61.6 -59.3 -55.6 -59.0 -63.3 -54.0 -52.6 -50.9 -52.4 -66.8 -46.1 -56.5 -46.1 -66.8 6.1 Sm-147 10.9 17.7 19.0 17.5 7.0 12.7 21.5 11.2 3.2 0.2 0.9 11.1 21.5 0.2 7.5 Sm-149 16.1 26.1 17.7 25.0 11.5 0.2 35.3 19.7 22.3 17.8 8.7 18.2 35.3 0.2 9.4 Sm-150 8.7 12.1 17.1 15.1 7.5 18.6 24.5 10.7 3.0 2.8 8.6 11.7 24.5 2.8 6.7 Sm-151 34.6 53.1 40.3 57.9 40.1 48.7 53.4 47.1 27.3 30.0 31.6 42.2 57.9 27.3 10.5 Sm-152 34.4 33.7 37.7 40.0 32.7 36.2 44.2 41.7 31.3 32.1 34.6 36.2 44.2 31.3 4.2 Gd-155 -52.2 -48.1 -48.6 -46.5 -66.0 -34.7 -47.7 -47.0 -51.9 -48.0 -47.0 -48.9 -34.7 -66.0 7.3 Am-241 7.8 -15.1 -14.0 -19.1 -18.6 -3.4 -1.0 -16.3 -18.9 47.3 31.7 -1.8 47.3 -19.1 22.4 Arn-242m -86.4 -84.3 -85.6 -84.6 74.1 -85.8 -85.3 -18.0 3.5 29.3 130.3 -26.6 130.3 -86.4 77.1 Am-243 47.2 -1.3 -1.3 4.2 41.4 1.5 9.2 47.4 37.9 53.8 62.3 27.5 62.3 -1.3 24.9 Np-237 1.9 -4.9 1.0 1.3 5.6 1.2 5.4 8.2 4.0 8.7 11.7 4.0 11.7 -4.9 4.6 Mo-95 -3.9 8.2 -3.2 -0.7 0:5 -1.5 10.4 3.2 1.5 1.2 6.3 2.0 10.4 -3.9 4.6 Tc-99 -26.1 7.4 5.3 6.2 -3.3 7.1 13.0 -12.1 -3.6 -7.0 9.1 -0.4 13.0 -26.1 11.5 Ru-101 -6.5 10.0 -3.6 -0.4 6.0 1.2 15.5 3.4 4.2 5.5 12.9 4.4 15.5 -6.5 6.7 Rh-103 2.4 19.2 5.2 7.5 10.7 7.4 22.8 11.5 10.5 11.5 15.2 11.3 22.8 2.4 6.0 Ag-109 100.7 123.5 127.2 125.5 44.0 103.3 34.2 96.6 65.6 204.9 200.0 111.4 204.9 34.2 55.0
A.2 Assembly NJ070G The initial model for assembly NJ070G, built based on the information available at the time, did not include any details on the assembly surroundings. It also assumed the location for the gadolinia fuel rods, as shown in Figure A.5, and assumed that the assembly pitch was 15 times larger than the rod pitch (i.e., no extra water between adjacent assemblies). The initial TRITON model (called here model #1) is shown in Figure A.6 for samples in comer rod 01.
A B C D E F G H I J K L M N 0 Guide 1 tube Fuel 2 rod Measured 3 fuel rod Gd 201 4 fuel rod 5
6 7
8 9
10 11 12 13 14 15 Figure A.5 Initial layout-assembly NJ070G A-9
- regular fuel pin U test fuel pin U ] U neighbors of test pin U gadolinia fuel pin a clad L moderator 7" BPR absorber U BPR clad Figure A.6 Initial TRITON model for samples in corner rod 01 of assembly NJ070G A-10
As additional information (Ref. 18) on the assemblies surrounding the NJ070G assembly, or data related to assembly geometry became available, new computational models were developed to include this information and therefore assess the effect on the calculated isotopic inventory. All of the three rods measured were located at the edge of the assembly; therefore, the boundary condition is expected to influence the flux spectrum in the samples under consideration.
Important information on the location of the gadolinia fuel rods was obtained from AREVA.
These rods were actually located at B2, B14, N2, and N14 with respect to the layout illustrated in Figure A.5, therefore being close to the rods from which samples were selected. Assembly pitch also became available (Ref. 18); the initial model considered the assembly pitch to be the product of the rod pitch and the number of fuel rods in a row of the assembly lattice. Consideration of the actual assembly pitch value is expected to slightly increase the moderation for the measured rod near the edge of the assembly. Another change was for the dimensions of the guide and instrument tubes - the actual dimensions were smaller than the initially assumed values. This led to an increase in the moderator volume in the assembly by 0.7%. Data on the assemblies surrounding assembly NJ070G are illustrated in Figure 4.3. All the available data was included to build a more detailed TRITON model (called model #2) as discussed in Section 5.1 and illustrated in Figures 5.2 and 5.3.
Both modeling approaches described above were used in simulations to assess the effect of including more detailed information on the assembly configuration (i.e. assembly pitch, gadolinia rod location), as well as more accurately modeling the assembly surroundings, on the calculated nuclide inventory. It was found that the use of a more detailed model has a significant effect on the calculated concentration for some of the main actinides, as illustrated in Figure A.7 for 2 35U and 239 pu. The data shown in the figure correspond to a calculation that used the sample burnup as provided in Ref. 4. The calculated concentration of 4a' Nd, in this case, was within the experimental uncertainty of 1.5% for most of the samples considered. The more accurate model led to a decrease of the 235U and 239 Pu average overestimation from 4.7% and 10.4%, respectively, to 3.3% and 1.8%. The comparison for all other nuclides is presented in Table A.4.
A-1I
2 35 u
31 30 29 28 I- 27 26 Q.9 25 E
- 3 24 I model 2 M model 1 23 22 I. .
0 1 2 3 4 5 6 7 8 9 CIE -1 (%)
31 30 29 28 27 CL 26 C
- 3 25 24 I model 2 model 1 23 22 r
-5 0 5 10 15 20 25 C/E -1 (%)
23 5 239 Figure A.7 Effect of modeling assumptions on U and Pu-assembly NJ070G A-12
Table A.4 C/E -1 (%) for samples in assembly NJ070G - computational models # 1 and #2 SampleID O13S7 012S4 012S6 O1SI O13S8 012S5 O1S3 01S2 Avg Burnup (GWd/MTU) 22.8 23.7 24.0 25.8 26.3 26.5 26.7 29.9 Nuclide ID modla':mi0 j modl imiod2j modl mod2 modl m-d2 modl I rod2 - modl ni6d2,* modl modt2t modl ifnod2 modl Imod2.
U-234 -2.6 -1.9 -1.1 -0.3 -0.2 0.6 -1.8 -0.7 -0.5 0.3 1.0 1.9 0.1 1.2 -0.6 0.6 -0.6 0.2 U-235 6.3 4,9 4.4 3.0 3.0 1.5 5.2 2.7 5.0 3.3 4.8 3.0 5.2 2.5 8.3 5.0 4.7 3.3 U-236 -3.8 -4.1 -3.0 -3.4. -3.8 -4.2 -2.4 -2.7. -2.1 -2.3 -2.3 -2:-7 -2.9 -3.3 -2.3 -2.6 -2.5 -3.1:i Pu-238 -11.1 ,-18A"&i -6.0 -13.7:. -16.6 -23* 1.0 "-10.6 -9.5 -i16.7, -9.2 :-16.3 -10.2 <-20.3 -1.2 -12.0 -7.0 -16.4 Pu-239 9.5 <1.6 10.6 ..2.6' 3.5 -4.3 15.1 - 2.8 10.9 *2A4. 8.6 0.'4i 11.9 -0.4 23.0 9.1 10.4 1.8-Pu-240 -0.8 ,-2.9 2.6 0.3 -2.2 -4.4 3.6 .0.5. 0.7 -1.3 0.0 -2.3 -2.5 -5.4 1.8 . -1.2 0.4 -2.1 Pu-241 -2.5 .-9.0 2.1 -5.0 -5.8 -12.2 5.9 ' -44* -0.4 -7.0 -1.5 -8.2 -1.5 '-10.9 6.8 -3.5 0.3 -7.5 Pu242 -12.5 .,14.5. -5.2 -7.8 -10.3 -12.5 -2.9 -'<5.9, -8.1 - -9.5 -7.8 J'-9.'7.* -12.5 '"-14.5 -10.1 -11.8 -7.7 -10.8.
Nd-143 1.1 -0.*9. 2.9 .' 27* 2.0 :1. '8: 4.1 K '3'7' 3.6 3',-3.4 3.6 3.4 2.8 2.4 4.2 3.5 1!" 2.7 2..-7*
Nd-145 -0.5 ,-0.1, 1.5 2.0 1.4 .1.9' 2.0 - '2.7 2.2 '2.6 2.3 :18 1.2 1.9 1.1 1.8 1.2 2.0?ý Nd-146 - .5,d14-1.2
-1.0 4.....16 1.5 1.0 ;*:!l0.:8i1.3 0.3
-0.2 -.0.0
- }:}f0.3} 2.3 1.4 2.0 Li*:::::2: 1.6 1.47 1.1 .', 1.!!i~*:0:, 1.7 1.2 1.5
- l*l!
1.t:i -0.2 0.4 0.8 i::-0ý4j* 1.5 ?:i.0i2}
0.4 1.1 1.0 0.4 0.8 6 2*:
0,:
Nd-150 -1.1 -1.8 2.3 :1.6' 0.2 i -05,O. 2.4 ' 1.3 2.0 ,12.2 2.1 1.31 0.3 -0.8 1.5 0.4 1.1 0.4 Cs-134 -22.5 '--25.5: -18.2 -21.2 -21.2 2-24.1' -14.2 -18.6 -19.1 --22.0: -18.0 -20,91 -18.1 -22.2! -14.4 -18.4, -16.2 -21.6 Cs-137 -7.4 -4 7 -3.7 2-3. -38 -3.8 -2.2 <-2.'2 -5.5 -5.5 -3.8 -- 3-8 -4.4 ~-4.5 -7.0 -7.0 -4.2 -4.8' Eu-151 417 326 28.4 2' 19.5 11:5 38.2 .24.2 .- 39.4 29.5 38.5 28.9 33.6 19.6 56.1 39.00 32.8 25.7 Eu-153 -8.6 -9.8 -6.5 '--7.7' -7.7 -8.8' -2.9 ' -4.6 -6.1 -7.2' -5.5 - -6,7' -5.7 -7.2 . -3.6 -5.0' -5.2 -77.1, Sm-147 -4.1 -2.8 -5.8 j'2-4.5. -5.2 -3.8' -6.2 4..3 -3.2 -1.8 -3.7 ' -2.3 -7.1 '--5.2 -6.4 -4.4 -4.6 -- 3.6.
Sm-149 20.1 1*'42.5, 19.4 i.12A1l 16.0 j '84"; 26.1 ,11*4-7 28.2 19.6 27.5 19.3 24.4 . 12.5 40.6 27.-4 22.5 15.8 Sm-150 0.3 .02 2.5 2.i20:' 1.5 1.1 4.2 3.4 3.2 2.7 12.1 11.6' 1.4 :0.8 3.2 2.5 3.2 3.0.
Sm-151 36.7 28..6 36.2 27.5 26.1 17.8 44.4 301. 34.7 25.3 34.7 25.5' 38.1 23.9 48.1 32.1 33.2 26.31 Sm-152 11.1 120'* 13.4 . -14.3, 16.6 ', 17.6 14.7 "16.2 16.9 180* 16.8 17.8., 13.6 . 15.1 13.1 14.6 12.9 15.7%
Gd-155 -39.1 -440.2 -40.5 -41.6; -46.2 -47.3' -43.7 7445A1 ' -39.8 -41.0 -38.7 -39.8' -47.2 -48.6 -34.9 -36.6 -36.7 -42.5.
Cm-242 -22.0 * ] -22.2 :*27. -16.6 -*22:0! 0.22 -18.7 -31.9 "-36.0" -27.7 -32.5. -25.3 -231.6 -34.7 -20.0 ,28.2i240.3 Cm-243 -25.1 ".33.l -17.7 '-26.9 -13.7 -23.3 30.0 ' 10.3 -21.0 i-29.0 -23.4 -31.6 -14.9 -27.3 0.0 -14.1 -9.5 -21.9 Cm-244 -11.7 ,'-22.3' -0.7 -13.2 5.2 -7.5 63.7 '36.2' -5.1 5 -6.7 '-174* 6.5 -10.3 19.0 1.3, 7.8 -6.1 Cm-245 -36.4 -47.5 -26.6 -*39.7:' -31.1 -43.1' 27.3 }. -3.3 -32.2 -43.2 -33.2 -44.5 -23.1 -40. 8 -2.6 -24.3: -17.6 -35.8:.
Am-241 -0.2 'ij-6*9 -1.0 -7:8'i 17.2 '9.2' 49.1 .K'34.5 1.8 -5.0' -0.1 -:6.9 9.9 -0.7. 23.5 11.5 11.1 3.5 Am-242m 12.5 1.6. 7.4 i-3.6 13.1 1:.7 59.1 '36.2 1 1.1 . -8.9 -1.6 -11.7 18.8 1.8 34.3 14.5 16.1 4.0.
Am-243 3.3 :2i44:'82s 13.6 ' '4.1 29.7 119.6, 74.7 55.5 9.3 1.8- 8.2 0.3 22.5 10.2 30.8 18.6 21.4 13.2' Np-237 0.8 :-j<5M0; -1.0 -6.5 -2.8 -8.:17; 11.9 1 ' 3.2 3.1 1 -2.5 3.6 -1.9 2.1 -5.6 8.4 0.4 2.9 -3.2 Assembly surroundings modeled with a reflective boundary condition.
Assemblies surrounding the test assembly close to the boundary the samples were selected are explicitly modeled.
In order to estimate the relative importance of various assumptions used in the computational model on the calculated nuclide concentrations, separate calculations were carried out for sample 013S7 by changing some of the model parameters one at a time. The following modeling parameters were considered:
- boundary condition on the assembly sides (modeling of surrounding assemblies);
- location of gadolinia fuel rods in the assembly (in some cases no precise information on position is available and therefore it must be assumed); and
- assembly pitch.
The model illustrated in Figure 5.3 was considered as a reference. Three models were developed starting from this reference model, as follows:
- 1. The assemblies neighboring the assembly NJ070G were not explicitly modeled; a reflective boundary condition was employed on the sides of NJ070G. The TRITON geometry in this case represented a quarter of an assembly.
- 2. The location of the gadolinia rod in the assembly was assumed, as shown in Figure A.5, farther from its actual location shown in Figure 5.2.
- 3. The value used for the assembly pitch was assumed as rod pitch times 15 instead of the actual value.
Given the proximity of the measured rod 013 to the assembly boundary and to the gadolinia rod location, each of the three above-mentioned changes are expected to influence the flux spectrum in the measured fuel rod and therefore the calculated nuclide concentrations for samples selected from that rod.
The results of the analysis are shown in Table A.5. One parameter only was changed at a time; the other remained as in the model shown in Figure 5.2. The effect of the assumptions is illustrated 48 in Figure A.8 for three nuclides: 235U, 239 Pu and 148Nd. As expected, the content of the 1 Nd'nuclide, which is a burnup indicator, does not change appreciably when changing the 23 9 model. The effect is, however, significant in case of pu: the C/E ratio increases by about 2.5%
as compared to the reference model when either the gadolinia rod location or the surrounding assemblies are not exactly represented; when the assembly pitch is assumed, the C/E increases to 9.7%, as compared to the reference model for which it is 1.6%. The C/E change for 235U is not as dramatic as for 239pu. It increases by about one half of a percent when the gadolinia rod location or the surrounding assemblies' effect is assumed and about 1.5% when the assembly pitch is assumed.
A-14
Table A.5 Effect of modeling assumptions on C/E-1 (%) for sample 013S7 from assembly NJ070G Nuclide Reference Surrounding assemblies Assumed location Assumed value ID model not modeled explicitly for gadolinia rod for assembly pitch U-234 -1.9 -2.3 -2.1 -2.7 U-235 4.9 5.5 5.4 6.5 U-236 -4.1 -4.0 -4.1 -3.8 Pu-238 -18.4 -15.6 -16.2 -10.7 Pu-239 1.6 4.3 4.0 9.7 Pu-240 -2.9 -1.7 -2.3 -0.2 Pu-241 -9.0 -6.4 -7.0 -2.0 Pu242 -14.5 -13.1 -14.0 -11.7 Nd-143 0.9 0.9 1.0 1.0 Nd- 145 -0.1 -0.2 -0.2 -0.6 Nd-146 -1.2 -1.2 -1.2 -1.1 Nd-148 -1.6 -1.6 -1.6 -1.5 Nd-150 -1.8 -1.5 -1.6 -1.0 Cs-134 -25.5 -24.4 -24.4 -22.6 Cs-137 -7.4 -7.5 -7.4 -7.4 Eu-151 32.6 35.7 35.4 41.9 Eu-153 -9.8 -9.2 -9.3 -8.5 Sm-147 -2.8 -3.3 -3.3 -4.1 Sm-149 12.5 15.0 15.0 20.1 Sm-150 -0.2 0.0 0.0 0.2 Sm-151 28.0 30.9 30.6 36.8 Sm-152 12.0 11.7 11.6 11.2 Gd-155 -40.2 -39.8 -39.8 -39.0 Cm-242 -27.2 -24.7 -25.8 -21.1 Cm-243 -33.1 -29.7 -31.0 -24.0 Cm-244 -22.3 -18.1 -19.5 -10.5 Cm-245 -47.5 -43.1 -44.3 -35.5 Am-241 -6.9 -4.1 -4.8 0.4 Am-242m 1.6 6.3 4.8 13.9 Am-243 -4.8 -1.3 -2.7 4.5 Np-237 -5.0 -2.8 -3.1 1.0 A-15
Nd-148 Pu-239 assembly pitch U-235 gadolinia rod location O surrounding assemblies reference model 3 1 0 1 2 3 4 5 6 7 8 9 10 11 12 C/E -1 (%)
23 5 239 Figure A.8 Effect of modeling assumptions on U, Pu, and 148Nd for sample 01S7 A-16
APPENDIX B SELECTED TRITON INPUT FILES
B.1 Sample 012S4 from Rod 012 in TMI-1 Assembly NJ070G
=t-depl parm=(nitawl,addnux=3)
TMI-1 Assembly NJ070G, rod 012, sample 012S4 44groupndf5 read alias
$fuell 10 11 12 13 14 end
$fuel2 15 end
$fuel3 16 17 end
$cladl 20 21 22 23 24 end
$clad2 25 end
$clad3 26 27 end
$modl 30 31 32 33 34 end
$mod2 35 end
$mod3 36 37 end
$gapl 40 41 42 43 44 end
$gap2 45 end
$gap3 46 47 end read comp
,fuel uo2 $fuell den=10.217 1 960.29 92234 0.045 92235 4.657 92238 95.298 end uo2 $fuel2 den=10.217 0.98 960.29 92234 0.037 92235 4.190 92236 0.019 92238 95.754 end arbmgd 10.217 2 0 1 0 64000 2 8016 3 $fuel2 0.02 960.29 end uo2 $fuel3 den=10.412 1 960.29 92234 0.040 92235 4.013 92238 95.947 end
'clad zirc4 $cladl 1 640 end zirc4 $clad2 1 640 end zirc4 $clad3 1 640 end
'moderator h2o $modl den=0.7382 1 582 end arbm-bormod 0.7382 1 1 0 0 5000 100 $modl 1800.0-6 582 end h2o $mod2 den=0.7382 1 582 end arbm-bormod 0.7382 1 1 0 0 5000 100 $mod2 1800.0-6 582 end h2o $mod3 den=0.7382 1 582 end arbm-bormod 0.7382 1 1 0 0 5000 100 $mod3 1800.0-6 582 end
'gap n $gapl den=0.00125 1 640 end n $gap2 den=0.00125 1 640 end n $gap3 den=0.00125 1 640 end I BPR A1203-B4C Al 50 0 3.817e-2 582 end 0-16 50 0 5.726e-2 582 end C 50 0 7.547e-4 582 end B-10 50 0 6.015e-4 582 end B-11 50 0 2.421e-3 582 end B-I
I BPR clad zirc4 51 1 582 end 1
end comp I
read celldata latticecell squarepitch pitch=1.44272 $modl fuelr=0.4699 $fuell cladr=0.5461 $cladl gapr=0.47879 $gapl end latticecell squarepitch pitch=l.44272 $mod2 fuelr=0.4699 $fuel2 cladr=0.5461 $clad2 gapr=0.47879 $gap2 end latticecell squarepitch pitch=1.44272 $mod3 fuelr=0.4699 $fuel3 cladr=0.5461 $clad3 gapr=0.47879 $gap3 end end celidata read depletion 10 -11 12 13 14 15 16 17 50 end depletion read timetable I change B in moderator densmult $modl 2 5010 5011 0.0 1.0000 68.0 0.9161 131.8 0.8450 209.0 0.7344 272.1 0.6333 347.4 0.5100 416.4 0.3989 486.4 0.2811 556.3 0.1656 626.1 0.0572 660.3 0.0010 end densmult $mod2 2 5010 5011 0.0 1.0000 68.0 0.9161 131.8 0.8450 209.0 0.7344 272.1 0.6333 347.4 0.5100 416.4 0.3989 486.4 0.2811 556.3 0.1656 626.1 0.0572 660.3 0.0010 end densmult $mod3 2 5010 5011 0.0 1.0000 68.0 0.9161 131.8 0.8450 209.0 0.7344 272.1 0.6333 347.4 0.5100 416.4 0.3989 486.4 0.2811 556.3 0.1656 B-2
626.1 0.0572 660.3 0.0010 end
. change temperature in fuel temperature $fuell 0.0 960.29 68.0 960.29 131.8 960.71 209.0 958.68 272.1 954.18 347.4 946.12 416.4 937.15 486.4 926.04 556.3 914.37 626.1 904.09 660.3 897.82 end temperature $fuel2 0.0 960.29 68.0 960.29 131.8 960.71 209.0 958.68 272.1 954.18 347.4 946.12 416.4 937.15 486.4 926.04 556.3 914.37 626.1 904.09 660.3 897.82 end temperature $fuel3 0.0 960.29 68.0 960.29 131.8 960.71 209.0 958.68 272.1 954.18 347.4 946.12 416.4 937.15 486.4 926.04 556.3 914.37 626.1 904.09 660.3 897.82 end end timetable I
read burndata power= 35.893 burn= 68.0 down=0 end power= 35.893 burn= 63.8 down=0 end power= 35.893 burn= 77.2 down=0 end power= 35.893 burn= 63.1 down= 0 end power= 35.893 burn= 75.3 down=0 end power= 35.893 burn= 69.0 down= 0 end power= 35.893 burn= 70.0 down=0 end power= 35.893 burn= 69.9 down=0 end power= 35.893 burn= 69.8 down=0 end power= 35.893 burn= 34.2 down=1298 end end burndata I
read opus units=grams symnuc=u-234 u-235 u-236 u-238 B-3
pu-238 pu-239 pu-240 pu-241 pu-242 nd-143 nd-145 nd-146 nd-148 nd-150 cs-134 cs-137 eu-151 eu-153 sm-147 sm-149 sm-150 sm-151 sm-152 gd-155 cm-242 cm-243 cm-244 cm-245 am-241 am-242m am-243 np-237 end matl= 10 11 12 13 14 15 end end opus read model TMI-I Assy NJ070G rod 012 sample 012S4 read parm run=yes drawit=yes fillmix=30 epsinner=-le-4 cmfd=yes xycmfd=4 echo=yes end parm I
read materials 10 2 ! fuel pin end 11 2 ! test pin ! end 12 2 ! N neighbor ! end 13 2 I W neighbor I end 14 2 1 E neighbor I end 15 2 I Gd pin ! end 16 2 I S neighbor ! end 17 2 I fuel neighbor assy end 20 2 clad ! end 30 2 I moderator! end 40 0 gap end 50 2 I BPR abs I end 51 2 1 BPR clad I end end materials read geom unit 1 com='fuel pin cell' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0 .72136 media 10 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 2 com='test pin 1 cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0 .72136 media 11 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 3 B-4
com='N neighbor test pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0.72136 media 12 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 4 com='E neighbor test pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0.72136 media 13 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 5 com='W neighbor test pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0.72136 media 14 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 6 com='Gd fuel pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0.72136 media 15 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 7 com='BPR' cylinder 1 0.4572 cylinder 2 0.5461 cylinder 3 0.63246 cylinder 4 0.6731 cuboid 5 4p0.72136 media 50 1 1 media 51 1 2 -1 media 30 1 3 -2 media 20 1 4 -3 media 30 1 5 -4 boundary 5 4 4 unit 8 B-5
com='left half fuel pin cell' cylinder 1 0.4699 chord -x=O cylinder 2 0.47879 chord -x=O cylinder 3 0.5461 chord -x=O cuboid 4 0.0 -0.72136 2p0.72136 media 10 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 2 4 unit 9 com='bottom half fuel pin cell' cylinder 1 0.4699 chord -y=O cylinder 2 0.47879 chord -y=O cylinder 3 0.5461 chord -y=0 cuboid 4 2p0. 7 2136 0.0 -0.72136 media 10 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 2 unit 10 com='quarter left-bottom instrument tube' cylinder 1 0.56007 chord -x=0 chord -y=0 cylinder 2 0.62611 chord -x=0 chord -y=0 cuboid 3 0.0 -0.72136 0.0 -0.72136 media 30 1 1 media 20 1 2 -1 media 30 1 3 -2 boundary 3 2 2 unit 11 com='S neighbor test pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4 p0. 7 2 136 media 16 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4 unit 12 com='neighbor assy pin' cylinder 1 0.4699 cylinder 2 0.47879 cylinder 3 0.5461 cuboid 4 4p0.72136 media 17 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 4 4.
unit 13 com='left half fuel pin cell neigh assy' cylinder 1 0.4699 chord -x=0 cylinder 2 0.47879 chord -x=0 cylinder 3 0.5461 chord -x=0 B-6
cuboid 4 0.0 -0.72136 2p0.72136 media 17 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1 4 -3 boundary 4 2 4 unit 14 com='top half fuel pin cell, cylinder 1 0.4699 chord +y=0 cylinder 2 0.47879 chord +y=0 cylinder 3 0.5461 chord +y=0 cuboid 4 2 p0. 7 2 136 0.72136 0.0 media 17 1 1 media 40 1 2 -1 media 20 1 3 -2 media 30 1.4 -3 boundary 4 4 2 unit 15 com='quarter left-top instrument tube' cylinder 1 0.56007 chord -x=0 chord +y=0 cylinder 2 0.62611 chord -x=0 chord +y=0 cuboid 3 0.0 -0.72136 0.72136 0.0 media 30 1 1 media 20 1 2 -1 media 30 1 3 -2 boundary 3 2 2 unit 16 com='guide tube' cylinder 3 0.63246 cylinder 4 0.6731 cuboid 5 4 p0. 7 2 1 3 6 media 30 1 3 media 20 1 4 -3 media 30 1 5 -4 boundary 5 4 4 unit 17 com='1/4 of top assembly' cuboid 10 10.90549 0.0 10.90549 0.0 array 1 10 place 1 1 0.80645 0.80645 media 30 1 10 boundary 10 15 30 unit 18 com='1/4 of bottom assembly' cuboid 10 10.90549 0.0 10.90549 0.0 array 2 10 place 1 1 0.80645 0.0 media 30 1 10 boundary 10 15 30 global unit 20 cuboid 10 10.90549 0.0 21.81098 0.0 array 3 10 place 1 1 0.0 0.0 media 30 1 10 boundary 10 15 30 end geom read array ara=1 nux=8 nuy=8 typ=cuboidal fill 1 1 4 2 5 1 1 8 B-7
1 6 1 3 1 1 1 8 1 1 1 1 1 7 1 8 1 1 1 7 1 1 1 8 1 1 1 1 1 1 1 8 1 1 7 1 1 7 1 8 1 1 1 1 1 1 1 8 9 9 9 9 9 9 91i0 end f ill ara=2 nux=8 nuy=8 typ=cuboidal fill 14 14 14 14 14 14 14 15 12 12 12 12 12 12 12 13 12 12 16 12 12 16 12 13 12 12 12 12 12 12 12 13 12 12 12 16 12 12 12 13 12 12 12 12 12 16 12 13 12 12 12 12 12 12 12 13 12 12 12 1l 12 12 12 13 end fill ara=3 nux=l nuy=2 typ=cuboidal fill 18 17 end fill end array read bounds
-x=white +x=ref -y=ref +y=ref end bounds end model end
=shell cp ft712f01 $RTNDIR/112S4.den end B-8
B.2 Sample 87-72 from Rod MKP-109 in Calvert Cliffs Assembly D047
=t-depl parm=(nitawl,addnux=3)
Calvert Cliffs Assembly D047 Rod MKP109 Sample 87-72 44groupndf read alias
$fuell 10 11 12 13 14 15 end
$cladl 20 21 22 23 24 25 end
$modl 30 31 32 33 34 35 end
$gapl 40 41 42 43 44 45 end end alias read comp uo2 $fuell den=10.045 1 841 92234 0.027 92235 3.038 92236 0.014 92238 96.921 end zirc4 $cladl 1 620 end h2o $modl den=0.7569 1 558 end arbmb 0.7569 1 1 0 0 5000 100 $modl 330.8e-06 558 end n $gapl den=0.00125 1 620 end guide tube zirc4 5 1 558 end end comp read celldata latticecell squarepitch pitch=l.4732 $modl fueld=0.9563 $fuell gapd=0.9855 $gapl cladd=l.1176 $cladl end end celldata 11 read timetable density $modl 2 5010 5011 0.00 1.000 377.00 1.000 377.01 1.419 840.00 1.419 840.01 1.523 1391.001.523 1391.011.488 1852.101.488 end end timetable read depletion 10 -11 12 13 14 15 end end depletion read burndata power=24.53 burn=306 nlib=3 down=71 end power=26.55 burn=381.7 nlib=3 down=81.3 end power=22.66 burn=466 nlib=3 down=85 end power=18.72 burn=461.1 nlib=3 down=1870 end end burndata read opus B-9
units=grams symnuc= u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 np-237 am-241 cm-243 cm-244 cs-133 cs-134 cs-135 cs-137 eu-154 nd-143 nd-144 nd-145 nd-146 nd-148 nd-150 pm-147 sm-147 sm-148 sm-149 sm-150 sm-151 sm-152 sm-154 eu-151 eu-153 eu-154 eu-155 gd-154 gd-155 end matl= 0 10 11 12 13 14 15 end end opus I
read model Calvert Cliffs Rod MKPI09 Sample mkpl09-2 read parm run=yes drawit=no fillmix=30 echo=yes cmfd=yes xycmfd=4 epsinner=-le-4 end parm read materials 10 1 regular pin end 11 1 test pin end 12 1 !N test pin end 13 1 ! S test pin end 14 1 l E test pin end 15 1 ! W test pin end 20 1 1 clad I end 30 2 I moderator end 40 0 I gap I end 5 1 !guide tube I end end materials read geom unit 1 com=Iregular fuel pin, cylinder 10 .47815 cylinder 20 .49275 cylinder 30 .5588 cuboid 40 4 p0.736 6 media 10 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 2 com='test fuel pint cylinder 10 .47815 cylinder 20 .49275 cylinder 30 .5588 cuboid 40 4p0.7366 media 11 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 3 com='N test fuel pin' cylinder 10 .47815 cylinder 20 .49275 cylinder 30 .5588 cuboid 40 4p0.7366 B-10
media 12 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 4 com='S test fuel pi cylinder 10 .47811 cylinder 20 .49271 cylinder 30 .5588 cuboid 40 4 p07.
media 13 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 5 com='E test fuel pi cylinder 10 .4781E cylinder 20 .49271 cylinder 30 .5588 cuboid 40 4p0.',
media 14 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 6 com='W test fuel pi cylinder 10 .4781E cylinder 20 .4927E cylinder 30 .5588 cuboid 40 4pO.2 media 15 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 71 com='guide tube - I-/4 NE' cylinder 1 1.314 chord +x=0 chord +y=0 cylinder 2 1.416 chord +x=0 chord +y=0 cuboid 3 1.473 0 1.473 0 media 30 1 1 media 5 1 2 -1 media 30 1 3 -2 boundary 3 4 4 unit 72 com=,guide tube - I7/4 SE' cylinder 1 1.314 origin x=0 y=1.473 chord +x=0 chord -y=1.473 cylinder 2 1.416 origin x=0 y=1.473 chord +x=0 chord -y=1.473 cuboid 3 1.473 0 1.473 0 media 30 1 1 media 5 1 2 -1 media 30 1 3 -2 boundary 3 4 4 unit 73 B-I I
com='guide tube - 1/4 SW' cylinder 1 1.314 origin x=1.473 y=1.473 chord -x=1.473 chord -y=1.473 cylinder 2 1.416 origin x=1.473 y=1.473 chord -x=1.473 chord -y=1.473 cuboid 3 1.473 0 1.473 0 media 30 1 1 media 5 1 2 -1 media 30 1 3 -2 boundary 3 4 4 unit 74 com='guide tube - 1/4 NW' cylinder 1 1.314 origin x=1.473 y=0 chord -x=1.473 chord +y=O cylinder 2 1.416 origin x=1.473 y=0 chord -x=1.473 chord +y=0 cuboid 3 1.473 0 1.473 0 media 30 1 1 media 5 1 2 -1 media 30 1 3 -2 boundary 3 4 4 global unit 10 cuboid 10 20.78 0.0 10.39 0.0 array 1 10 place 1 1 0.8142 0.7366 media 30 1 10 boundary 10 28 14 end geom read array ara=l nux=14 nuy=7 fill 1 1 1 1 1 1 74 71 1 1 1 1 1 1 1 1 73 72 1 1 4 1 1 1 73 72 1 1 1 1 74 71 1 6 2 5 1 1 74 71 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 end fill end array read bounds all=refl end bounds end data end
=shell cp ft7lfOO $RTNDIR/87-72.den end B-12
B.3 Sample SF97-3 from Rod SF97 in Takahama-3 Assembly NT3G23
=t-depl parm=(nitawl,addnux=3)
Takahama-3 Rod SF97 Sample SF97-3 44groupndf read alias
$fuell 10 11 12 13 14 15 16 end
$cladl 20 21 22 23 24 25 26 end
$modl 30 31 32 33 34 35 36 end
$gapl 40 41 42 43 44 45 46 end
$fuel2 17 end
$clad2 27 end
$mod2 37 end
$gap2 47 end end alias read comp uo2 $fuell den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end zirc4 $cladl 1 600 end h2o $modl den=0.6843 1 591.48 end arbmb 0.6843 1 1 0 0 5000 100 $modl 1154e-06 591.48 end n $gapl den=0.00125 1 600 end uo2 $fuel2 den=10.412 0.94 900 92234 0.02 92235 2.63 92238 97.35 end arbmgd 10.412 2 0 1 0 64000 2 8016 3 $fuel2 0.06 900 end zirc4 $clad2 1 600 end h2o Smod2 den=0.6843 1 591.48 end arbmb 0.6843 1 1 0 0 5000 100 $mod2 1154e-06 591.48 end n Sgap2 den=0.00125 1 600 end end comp I
read celldata latticecell squarepitch pitch=1.259 $modl fueld=0.805 $fuell gapd=0.822 $gapl cladd=0.950 $cladl end latticecell squarepitch pitch=l.259 $mod2 fueld=0.805 $fuel2 gapd=0.822 $gap2 cladd=0.950 $clad2 end end celldata read depletion 10 -11 12 13 14 15 16 17 end end depletion I
read burndata power=35-.162 burn=385 nlib=3 down=88 end power=37.498 burn=402 nlib=3 down=62 end power=33.282 burn=406 nlib=3 down=1446 end end burndata read opus B-13
units=grams symnuc= u-234 u-235 u-236 u-238 np-237 pu-238 pu-239 pu-240 pu-241 pu-242 am-241 am-242m am-243 cm-242 cm-243 cm-244 cm-245 cm-246 cm-247 nd-143 nd-144 nd-145 nd-146 nd-148 nd-150 cs-137 cs-134 eu-154 ce-144 sb-125 ru-106 sm-147 sm-148 sm-149 sm-150 sm-151 sm-152 sm-154 end matl=0 10 11 12 13 14 15 16 17 end end opus read timetable density $modl 2 5010 5011 0 1.000 106 0.775 205 0.564 306 0.350 385 0.182 473 0.981 592 0.749 704 0.531 817 0.310 875 0.198 937 1.000 996 0.867 1048 0.751 1100 0.634 1152 0.518 1204 0.401 1256 0.285 1308 0.169 1342 0.090 end density $mod2 2 5010 5011 0 1.000 106 0.775 205 0.564 306 0.350 385 0.182 473 0.981 592 0.749 704 0.531 817 0.310 875 0.198 937 1.000 996 0.867 1048 0.751 1100 0.634 1152 0.518 1204 0.401 1256 0.285 1308 0.169 1342 0.090 end end timetable read model Takahama-3 Rod SF97 Sample SF97-3 read parm run=yes drawit=no fillmix=30 echo=yes B- 14
cmfd=yes xycmfd=4 epsinner=-le-4 end parm read materials 10 1 regular pin end 11 1 1 test pin I end 12 1 N test pin I end 13 1 I NE test pin I end 14 1 E test pin end e
15 1 !W test pin I end 16 1 1 SW test pin I end 17 1 I gadolinia pin I end 20 1 I clad I end 30 2 moderator I end 40 0 gap end end materials read geom unit 1 com='regular fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 10 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 2 com='test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4 p0.6 2 95 media 11 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 3 com='N test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 12 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 4 com='NE test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 13 1 10 media 40 1 20 -10 B-15
media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 5 com='E test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 14 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 6 com='Gd2o3 fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 17 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 7 com='guide tube' cylinder 10 .5715 cylinder 20 .6121 cuboid 40 4pO.6295 media 30 1 10 media 20 1 20 -10 media 30 1 40 -20 boundary 40 4 4 unit 8 com='W test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 15 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 9 com='SW test fuel pin' cylinder 10 .4025 cylinder 20 .411 cylinder 30 .475 cuboid 40 4p0.6295 media 16 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 4 unit 12 B-16
com='bottom half of regular fuel pin' cylinder 10 .4025 chord -y=0 cylinder 20 .411 chord -y=0 cylinder 30 .475 chord -y=0 cuboid 40 2 p0.6 2 95 0.0 -0.6295 media 10 1 10 media 40 1 20 -10 media 20 1 30 -20 media 30 1 40 -30 boundary 40 4 2 unit 72 com='bottom half of guide tube' cylinder 10 .5715 chord -y=0 cylinder 20 .6121 chord -y=0 cuboid 40 2p0.6295 0.0 -0.6295 media 30 1 10 media 20 1 20 -10 media 30 1 40 -20 boundary 40 4 2 global unit 10 cuboid 10 21.403 0.0 10.7015 0.0 array 1 10 place 1 1 0.6295 0.6295 media 30 1 10 boundary 10 34 34 end geom I
read array ara=l nux=17 nuy=9 fill 1 1 1 5 2 8 1 1 1 1 1 1 1 7 6 1 11 41 3 91 1 1
7 7 7 6 1 1 1 1 1 1 1 1 6 1 1 1 1 17 111 1 1 1 1 1 6 1 1 1 1 1 1 1 1 1 111 1 16 11 1 1 1 1 7 1 1 7 1 1 7 1 1 7 11 7 1 1 1 1 6 1 1 1 6 1 1 1 1 1 1 1 1 1 12 12 72 12 12 72 12 12 72 12 12 72 12 12 72 12 12 end fill end array read bounds all=refl end bounds end data end
=shell cp ft71f001 end $RTNDIR/sf97-3.den B-17
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (9-2004) (Assigned by NRC, Add Vol., Supp., Rev.,
NRCMD 3.7 and Addendum Numbers, if any.)
BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse) NUREG/CR-6968 (ORNLIM-2008/071)
- 2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation-Calvert MONTH YEAR Cliffs, Takahama, and Three Mile Island Reactors February 2010
- 5. AUTHOR(S) 6. TYPE OF REPORT G. Ilas, I. C. Gauld, F. C. Difilippo, and M. B. Emmett Technical
- 7. PERIOD COVERED (inclusive Dates)
- 8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address;if contractor, provide name and mailing address.)
Oak Ridge National Laboratory Managed by UT-Battelle, LLC Oak Ridge, TN 37831-6170
- 9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above'; if contractor,provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address.)
Division of Systems Analysis Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
- 10. SUPPLEMENTARY NOTES M. Aissa, NRC Project Manager
- 11. ABSTRACT (200 words orless)
This report is part of a report series designed to document benchmark-quality radiochemical isotopic assay data against which computer code accuracy can be quantified to establish the uncertainty and bias associated with the code predictions. The experimental data included in the report series were acquired from domestic and international programs and include spent fuel samples that cover a large burnup range. The measurements analyzed in the current report, for which experimental data is publicly available, include 38 spent fuel samples selected from fuel rods with a 2.6 to 4.7 wt % U-235 initial enrichment that were irradiated in three pressurized water reactors operated in the United States and Japan and achieved burnup values from 14 to 56 GWd/MTU. The analysis of the measurements was performed by employing the two-dimensional depletion sequence TRITON in the SCALE code system.
- 12. KEY WORDS/DESCRIPTORS (List words or phrases that wiltassist researchers in locatingthe report.) 13. AVAILABILITY STATEMENT SCALE, TRITON, ORIGEN-S, spent nuclear fuel, isotopic validation, burnup credit, radiochemical unlimited analysis, experimental programs, Calvert Cliffs, Takahama, TMI-1 14. SECURITY CLASSIFICATION (This Page) unclassified (This Report) unclassified
- 15. NUMBER OF PAGES
- 16. PRICE NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER
meyciS paparn rdPrinted FederalRecycling Program
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, DC 20555-0001 OFFICIAL BUSINESS