NUREG/CR-7309, Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses

NUREG/CR-7309, Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses

ML25093A004
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Issue date:	04/30/2025
From:	Celik C, Dupont M, Karriem V, Lucas Kyriazidis, Lang A, Marshall W, Metwally W, Shaw A Office of Nuclear Regulatory Research, Oak Ridge
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ORNL/TM-2023/3243 NUREG/CR-7309
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NUREG/CR-7309 ORNL/TM-2023/3243 Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses Office of Nuclear Regulatory Research

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NUREG/CR-7309 ORNL/TM-2023/3243 Office of Nuclear Regulatory Research Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses Manuscript Completed: November 2023 Date Published: April 2025 Prepared by:

W. Metwally M. Dupont W. Marshall A. Lang V. Karriem C. Celik K. Fassino A. Shaw Oak Ridge National Laboratory Oak Ridge, TN 37831-6170 Lucas Kyriazidis, NRC Project Manager

iii ABSTRACT Interest is growing in the use of extended enrichment (between 5 and 8 wt% 235U) and higher burnup fuels (up to 80 GWd/MTU) in nuclear reactors. Therefore, safe storage and transportation of the resulting spent nuclear fuel (SNF) must be demonstrated. This report investigates the effects of extended enrichment and high-burnup fuels on validation of burnup credit (BUC) criticality safety analyses of SNF storage systems. The report presents results from a detailed similarity assessment study conducted to determine the appropriate criticality benchmark experiments for different application cases. In addition, the report presents the calculation of the bias and bias uncertainty, nuclear data-induced uncertainties, sensitivities, and BUC loading curves for selected application cases.

v TABLE OF CONTENTS ABSTRACT................................................................................................................................... iii LIST OF FIGURES....................................................................................................................... vii LIST OF TABLES.......................................................................................................................... xi EXECUTIVE

SUMMARY

............................................................................................................. xiii ACKNOWLEDGMENTS............................................................................................................... xv ABBREVIATIONS AND ACRONYMS....................................................................................... xvii 1 INTRODUCTION..................................................................................................................... 1-1 1.1 Background.................................................................................................................... 1-1 1.1.1 Burnup Credit..................................................................................................... 1-1 1.1.2 Previous Work.................................................................................................... 1-1 1.1.3 Developments and Needs.................................................................................. 1-1 1.2 Purpose and Outline....................................................................................................... 1-2 2 COMPUTATIONAL METHODS, DATA, AND MODELS...................................................... 2-1 2.1 Methods and Data.......................................................................................................... 2-1 2.1.1 Depletion............................................................................................................. 2-1 2.1.2 Criticality............................................................................................................. 2-2 2.1.3 Sensitivity, Uncertainty, and Validation.............................................................. 2-3 2.1.4 Nuclear Data....................................................................................................... 2-3 2.2 Models............................................................................................................................ 2-4 2.2.1 Fuel Assembly Depletion Model........................................................................ 2-4 2.2.2 GBC-32 Cask Storage Model............................................................................. 2-6 3 CRITICALITY BENCHMARK EXPERIMENTS...................................................................... 3-1 3.1 Selected Benchmarks.................................................................................................... 3-2 3.2 Similarity Assessment.................................................................................................... 3-4 3.2.1 Inventory Changes in the Application System................................................... 3-4 3.2.2 Summary and General Observations of Similarity Assessments...................... 3-7 3.2.3 Detailed Similarity Assessment Results.......................................................... 3-13 3.2.4 Influence of Nuclear Data Library and Covariance Data ck............................. 3-26 3.2.5 Similarity Assessment Conclusion................................................................... 3-36 4 BIAS AND BIAS UNCERTAINTY ANALYSES..................................................................... 4-1 4.1 Bias and Bias Uncertainty Estimates............................................................................. 4-1 4.2 Impacts on Bias and Bias Uncertainty Estimates........................................................ 4-13 4.2.1 Cross-Section and Covariance Data................................................................ 4-13 4.2.2 Extended Enrichment Fuel............................................................................... 4-25 4.2.3 High Burnup Fuel.............................................................................................. 4-29 4.3 Sensitivity...................................................................................................................... 4-30 4.3.1 Nuclear Data Libraries...................................................................................... 4-30 4.3.2 Decay Time...................................................................................................... 4-35

vi 5 CONCLUSIONS AND RECOMMENDATIONS..................................................................... 5-1 5.1 Continued Validity of NUREG/CR-7109 Conclusions................................................... 5-1 5.2 Additional Recommendations........................................................................................ 5-2 6 REFERENCES........................................................................................................................ 6-1 APPENDIX A SAMPLE INPUT FILES....................................................................................A-1 APPENDIX B BENCHMARK EXPERIMENTS........................................................................B-1 APPENDIX C keff RESULTS FOR GBC-32 CASES...............................................................C-1

vii LIST OF FIGURES Figure 2-1 Key Components Used in the SCALE Code System.......................................... 2-1 Figure 2-2 Depletion Sequence Used in the Analysis.......................................................... 2-2 Figure 2-3 A 2-D TRITON Model of the Westinghouse 17x17 OFA (1/4 Fuel Assembly)............................................................................................................ 2-5 Figure 2-4 Axial Burnup Profile............................................................................................. 2-5 Figure 2-5 Cutaway of the GBC-32 Cask and PWR Fuel Assemblies................................. 2-6 Figure 3-1 Node 17 235U, 239Pu, and 240Pu Number Densities for Increasing Assembly Average Burnup for 4 and 8 wt% 235U Fuels....................................................... 3-6 Figure 3-2 Node 17 235U and 239Pu Number Densities and 239Pu /(235U +239Pu) Ratio Compared to HTC 239Pu /(235U +239Pu) Ratio for Increasing Assembly Average Burnup and 4 wt% 235U Fuels................................................................ 3-6 Figure 3-3 Node 17 235U and 239Pu Number Densities and 239Pu /(235U +239Pu) Ratio Compared to HTC 239Pu /(235U +239Pu) Ratio for Increasing Assembly Average Burnup and 8 wt% 235U Fuel................................................................. 3-7 Figure 3-4 Number of Similar Benchmark Experiments Using the E7.1 Library................ 3-10 Figure 3-5 Number of Similar Benchmark Experiments Using the E8.0 Library................ 3-11 Figure 3-6 Comparison of the Number of Marginally Similar Benchmark Experiments (ck 0.8)............................................................................................................. 3-12 Figure 3-7 Comparison of the Number of Highly Similar Benchmark Experiments (ck0.9)............................................................................................................... 3-13 Figure 3-8 Similarity Coefficients of the 4 wt% 235U Enriched 10 GWd/MTU Burnup with E7.1 and Cov B (Application 1).................................................................. 3-15 Figure 3-9 Similarity Coefficients of the 4 wt% 235U Enriched 40 GWd/MTU Burnup with E7.1 and Cov B (Application 2).................................................................. 3-16 Figure 3-10 Similarity Coefficients of the 4 wt% 235U Enriched 80 GWd/MTU Burnup with E7.1 and Cov B (Application 3).................................................................. 3-17 Figure 3-11 Similarity Coefficients of the 8 wt% 235U Enriched 10 GWd/MTU Burnup with E7.1 and Cov B (Application 4).................................................................. 3-18 Figure 3-12 Similarity Coefficients of the 8 wt% 235U Enriched 40 GWd/MTU Burnup with E7.1 and Cov B (Application 5).................................................................. 3-19 Figure 3-13 Similarity Coefficients of the 8 wt% 235U Enriched 80 GWd/MTU Burnup with E7.1 and Cov B (Application 6).................................................................. 3-20 Figure 3-14 Similarity Coefficients of the 4 wt% 235U Enriched 10 GWd/MTU Burnup with E8.0 and Cov A (Application 7).................................................................. 3-21 Figure 3-15 Similarity Coefficients of the 4 wt% 235U Enriched 40 GWd/MTU Burnup with E8.0 and Cov A (Application 8).................................................................. 3-22

viii Figure 3-16 Similarity Coefficients of the 4 wt% 235U Enriched 80 GWd/MTU Burnup with E8.0 and Cov A (Application 9).................................................................. 3-23 Figure 3-17 Similarity Coefficients of the 8 wt% 235U Enriched 10 GWd/MTU Burnup with E8.0 and Cov A (Application 10)................................................................ 3-24 Figure 3-18 Similarity Coefficients of the 8 wt% 235U Enriched 40 GWd/MTU Burnup with E8.0 and Cov A (Application 11)................................................................ 3-25 Figure 3-19 Similarity Coefficients of the 8 wt% 235U Enriched 80 GWd/MTU Burnup with E8.0 and Cov A (Application 12)................................................................ 3-26 Figure 3-20 Uncertainty in 235U Reaction for Cov A and Cov B [19].................................. 3-27 Figure 3-21 Uncertainty in 239Pu Reaction for Cov A and Cov B [20]................................ 3-27 Figure 3-22 Uncertainty in 1H Elastic Scattering Reaction for Cov A and Cov B [20].......... 3-28 Figure 3-23 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 10 GWd/MTU Burnup with E7.1 and E8.0 (Applications 1 and 7)........................................... 3-31 Figure 3-24 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 10 GWd/MTU Burnup GBC-32 Application Model with E7.1 and E8.0 (Applications 4 and 10)............................................................................................................... 3-32 Figure 3-25 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 40 GWd/MTU Burnup with E7.1 and E8.0 (Applications 2 and 8)........................................... 3-33 Figure 3-26 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 40 GWd/MTU Burnup with E7.1 and E8.0 (Applications 5 and 11)......................................... 3-34 Figure 3-27 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 80 GWd/MTU Burnup with E7.1 and E8.0 (Applications 3 and 9)........................................... 3-35 Figure 3-28 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 80 GWd/MTU Burnup with E7.1 and E8.0 (Applications 6 and 12)......................................... 3-36 Figure 4-1 keff of Application Cases in Bias and Bias Uncertainty Analysis......................... 4-1 Figure 4-2 Significant ck Trending for 4 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library................................................................................... 4-3 Figure 4-3 Nonsignificant EALF Trending for 4 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library..................................................................... 4-4 Figure 4-4 Nonsignificant ck Trending for 4 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library................................................................................... 4-4 Figure 4-5 Nonsignificant EALF Trending for 4 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library..................................................................... 4-5 Figure 4-6 Significant ck Trending for 8 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library................................................................................... 4-5 Figure 4-7 Nonsignificant EALF Trending for 8 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library..................................................................... 4-6 Figure 4-8 Significant ck Trending for 8 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library................................................................................... 4-6

ix Figure 4-10 Significant EALF Trending with 4 wt% 235U 40 GWd/MTU with All Experiments using ENDF/B-VII.1 Nuclear Data Library................................... 4-13 Figure 4-11 Percent Contribution of Nuclear Data-Induced Uncertainty in keff from Major AC, Minor AC, and FP for 5 and 7 wt% 235U Cases................................ 4-22 Figure 4-12 Percent Contribution of Nuclear Data-Induced Uncertainty in keff from 235U, 238U, and 239Pu for 5 and 7 wt% 235U Cases............................................. 4-23 Figure 4-13 keff Results for AO Cases................................................................................... 4-25 Figure 4-14 keff Results for AFP Cases................................................................................. 4-26 Figure 4-15 keff Results for ALL Cases.................................................................................. 4-26 Figure 4-16 Difference in keff Results Between E7.1 and E8.0 for AFP Cases................... 4-27 Figure 4-17 Difference in Number Density of 239Pu in Zone 09 Between E7.1 and E8.0.................................................................................................................... 4-28 Figure 4-18 Difference in 239Pu Cross-Sections Between E7.1 and E8.0............................ 4-28 Figure 4-19 239Pu Fission Cross-Section Sensitivity Using E7.1 and E8.0.......................... 4-29 Figure 4-20 Comparison of Burnup Loading Curves Using ENDF/B-VII.1........................... 4-30 Figure 4-21 Comparison of Burnup Loading Curves Using ENDF/B-VIII.0.......................... 4-30 Figure 4-22 keff Difference Using Multigroup and Continuous Neutron Energy Libraries.... 4-31 Figure 4-23 Effect of Decay Time on keff and EALF.............................................................. 4-36 Figure 4-9 Nonsignificant EALF Trending for 8 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library..................................................................... 4-7

xi LIST OF TABLES Table 2-1 Nuclear Data Libraries and Covariance Data Used............................................ 2-4 Table 3-1 Characteristics of the Application Cases in Similarity Assessment.................... 3-2 Table 3-2 Source of Critical Experiments Used in Validation.............................................. 3-2 Table 3-3 Evaluations Used from the ICSBEP Handbook.................................................. 3-3 Table 3-4 Range of Key Parameters Used in Validation..................................................... 3-4 Table 3-5 Similarity Assessment Summary......................................................................... 3-8 Table 3-6 Summary of Presented Cases for Similarity Assessment................................ 3-14 Table 3-7 Nuclear Data Library and Covariance Data Combinations Used in the Study.................................................................................................................. 3-28 Table 3-8 Similarity Coefficients Between 4 wt% 235U Initial Enrichment and 10 GWd/MTU Burnup Case and 26 Critical Experiments Using Different Nuclear Data...................................................................................................... 3-29 Table 3-9 Main Contributors to the Similarity Coefficient of Four Application Cases....... 3-30 Table 4-1 Application Cases in Bias and Bias Uncertainty Analysis................................... 4-1 Table 4-2 Number of Applicable Critical Experiments in Bias and Bias Uncertainty Analysis............................................................................................ 4-2 Table 4-3 Bias and Bias Uncertainty Results for all Application Cases Obtained with the ck and EALF Trending Methods..................................................................... 4-9 Table 4-4 Bias and Bias Uncertainty Results for all Application Cases Obtained with the Nontrending Parametric and Nontrending Nonparametric Methods.......... 4-10 Table 4-5 Bias and Bias Uncertainty with and Without HTC Experiments........................ 4-11 Table 4-6 Effect of Critical Experiments Selection on Bias and Bias Uncertainty Using EALF Trending and ENDF/B-VII-1.......................................................... 4-12 Table 4-7 Nuclear Data-Induced Uncertainty (pcm) in keff for 4 wt% 235U Cases............ 4-14 Table 4-8 Nuclear Data-Induced Uncertainty (pcm) in keff for 5 wt% 235U Cases............ 4-16 Table 4-9 Nuclear Data-Induced Uncertainty (pcm) in keff for 6 wt% 235U Cases............ 4-17 Table 4-10 Nuclear Data-Induced Uncertainty (pcm) in keff for 7 wt% 235U Cases............ 4-18 Table 4-11 Nuclear Data-Induced Uncertainty (pcm) in keff for 8 wt% 235U Cases............ 4-19 Table 4-12 Percent Contribution in Nuclear Data-Induced Uncertainty for 4 wt% 235U Cases................................................................................................................. 4-20 Table 4-13 Percent Contribution in Nuclear Data-Induced Uncertainty for 5 wt% 235U Cases................................................................................................................. 4-20 Table 4-14 Percent Contribution in Nuclear Data-Induced Uncertainty for 6 wt% 235U Cases................................................................................................................. 4-21 Table 4-15 Percent Contribution in Nuclear Data-Induced Uncertainty for 7 wt% 235U Cases................................................................................................................. 4-21

xii Table 4-16 Percent Contribution in Nuclear Data-Induced Uncertainty for 8 wt% 235U Cases................................................................................................................. 4-22 Table 4-17 Worth and Uncertainty to Worth Ratios of Minor Actinides and Fission Products............................................................................................................. 4-24 Table 4-18 Nuclear Data-Induced Uncertainty in keff for 5 wt% 235U, 60 GWd/MTU (pcm).................................................................................................................. 4-32 Table 4-19 Nuclear Data-Induced Uncertainty in keff for 7 wt% 235U, 60 GWd/MTU (pcm).................................................................................................................. 4-34 Table 4-20 Effect of Decay Time on Bias and Bias Uncertainty......................................... 4-37 Table B-1 Benchmark Experiments Used in Similarity Assessments................................. B-1 Table C-1 Results with Initial Enrichment at 4 wt% 235U (ENDF/B-VII.1)............................ C-1 Table C-2 Results with Initial Enrichment at 4.5 wt% 235U (ENDF/B-VII.1)......................... C-2 Table C-3 Results with Initial Enrichment at 5.0 wt% 235U (ENDF/B-VII.1)......................... C-3 Table C-4 Results with Initial Enrichment at 5.5 wt% 235U (ENDF/B-VII.1)......................... C-4 Table C-5 Results with Initial Enrichment at 6.0 wt% 235U (ENDF/B-VII.1)......................... C-5 Table C-6 Results with Initial Enrichment at 6.5 wt% 235U (ENDF/B-VII.1)......................... C-6 Table C-7 Results with Initial Enrichment at 7.0 wt% 235U (ENDF/B-VII.1)......................... C-7 Table C-8 Results with Initial Enrichment at 7.5 wt% 235U (ENDF/B-VII.1)......................... C-8 Table C-9 Results with Initial Enrichment at 8.0 wt% 235U (ENDF/B-VII.1)......................... C-9 Table C-10 Results with Initial Enrichment at 4.0 wt% 235U (ENDF/B-VIII.0)...................... C-10 Table C-11 Results with Initial Enrichment at 4.5 wt% 235U (ENDF/B-VIII.0)...................... C-11 Table C-12 Results with Initial Enrichment at 5.0 wt% 235U (ENDF/B-VIII.0)...................... C-12 Table C-13 Results with Initial Enrichment at 5.5 wt% 235U (ENDF/B-VIII.0)...................... C-13 Table C-14 Results with Initial Enrichment at 6.0 wt% 235U (ENDF/B-VIII.0)...................... C-14 Table C-15 Results with Initial Enrichment at 6.5 wt% 235U (ENDF/B-VIII.0)...................... C-15 Table C-16 Results with Initial Enrichment at 7.0 wt% 235U (ENDF/B-VIII.0)...................... C-16 Table C-17 Results with Initial Enrichment at 7.5 wt% 235U (ENDF/B-VIII.0)...................... C-17 Table C-18 Results with Initial Enrichment at 8.0 wt% 235U (ENDF/B-VIII.0)...................... C-18

xiii EXECUTIVE

SUMMARY

The U.S. nuclear industry has seen increased interest in using extended enrichment and achieving higher burnup fuel in light-water reactors. One challenge that arises is the need to validate the criticality safety analysis associated with the fuel, especially because of the limited availability of critical benchmark experiments supporting validation studies that incorporate burnup credit (BUC).

Previous studies have demonstrated that, for up to 5 wt% 235U and 60 GWd/MTU burnup, there is a sufficient number of critical benchmark experiments through the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the French Haut Taux de Combustion (HTC) program that can be used in the validation studies.

The current study investigated the effects of extended enrichment (between 5 to 8 wt% 235U) and high-burnup (up to 80 GWd/MTU, assembly average) fuels and nuclear data libraries on the validation of BUC criticality safety analyses of spent nuclear fuel (SNF) storage systems. Both the Evaluated Nuclear Data File (ENDF)/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries were used in the study.

The current study concluded that there are a sufficient number of critical experiments to support the validation of BUC criticality safety analyses for up to 8 wt% 235U and 80 GWd/MTU. Using the applicable critical benchmark experiments, the bias and bias uncertainty for select application cases were calculated. The combined bias and bias uncertainty was found to be on the order of 1000 pcm. The study also showed that the unvalidated minor actinide and fission product nuclear data-induced uncertainty ranged from 0.7 to 1.3% of their worth. Examples of the resulting BUC loading curves for selected application cases are also presented.

xv ACKNOWLEDGMENTS This work was performed under contract with the U.S. Nuclear Regulatory Commission (NRC)

Office of Nuclear Regulatory Research. The authors thank L. Kyriazidis, the NRC Project Manager of the Office of Nuclear Regulatory Research (RES), and A. B. Barto of the Office of Nuclear Material Safety and Safeguards (NMSS) for their support and guidance. Many valuable review comments were received from NRC staff members of RES and NMSS. The authors also wish to thank Bret Brickner and Travis Greene for their reviews, and R. Roberts for assistance in editing, formatting, and preparing the final document.

xvii ABBREVIATIONS AND ACRONYMS 2-D two-dimensional AC actinides AFP actinides and 16 fission products ALL actinides and all fission products AC actinides AO actinides only ARP Automatic Rapid Processing BLO BNL-LANL-ORNL [uncertainty data]

BNL Brookhaven National Laboratory BUC burnup credit CE continuous-energy C/E calculated-to-expected ck correlation coefficient CSAS Criticality Safety Analysis Sequence CSAS5 Criticality Safety Analysis Sequence with KENO V.a transport code DICE Database for the International Handbook of Evaluated Criticality Safety Benchmark Experiments DOE U.S. Department of Energy EALF energy of average neutron lethargy causing fission ENDF Evaluated Nuclear Data File FP fission product GBC-32 generic burnup credit cask with 32 square storage cells HALEU high-assay low-enriched uranium HTC Haut Taux de Combustion ICSBEP International Criticality Safety Benchmark Evaluation Project IEU intermediate enriched uranium IRSN Institut de Radioprotection et de Sûreté Nucléaire IST IEU-SOL-THERM keff effective multiplication factor LANL Los Alamos National Laboratory LCE laboratory critical experiment LCT LEU-COMP-THERM LEU low-enriched uranium LMT LEU-MET-THERM LST LEU-SOL-THERM LT light elements MCT MIX-COMP-THERM MIX mixed uranium plutonium

xviii MG multigroup MOX mixed oxide MST MIX-SOL-THERM NEA Nuclear Energy Agency NEWT New ESC-based Weighting Transport [code]

NRC U.S. Nuclear Regulatory Commission

, nubar the average number of neutrons per fission OFA optimized fuel assembly ORIGAMI ORIGEN Assembly Isotopics ORIGEN Oak Ridge Isotope Generation ORNL Oak Ridge National Laboratory pcm per cent mille PWR pressurized-water reactor S/U sensitivity and uncertainty SDF sensitivity data file SNF spent nuclear fuel t-depl TRITON Depletion [sequence]

TRITON Transport Rigor Implemented with Time-dependent Operation for Neutronic

[depletion]

TSUNAMI Tools for Sensitivity and Uncertainty Analysis Methodology Implementation TSUNAMI-IP TSUNAMI Indices and Parameters VALID Verified, Archived Library of Inputs and Data WABA wet annular burnable absorber VADER Validation Analysis Data Evaluation Resource

1-1 1

INTRODUCTION

1.1 Background

1.1.1 Burnup Credit Criticality safety analyses are required to ensure the safe storage and transportation of spent nuclear fuel (SNF). Historically, it was common to assume a fresh fuel composition when performing these criticality safety analyses. With the increase in the amount of SNF and limitations on storage, allowance was given to account for the reduced reactivity of SNF compared to that of fresh fuel. The process of taking credit for this reduced reactivity in criticality safety analyses is commonly referred to as burnup credit (BUC). A discussion on the process for implementing the BUC methodology can be found in NUREG/CR-7109 [1]. One of the critical steps in implementing the BUC methodology is validation: determining the bias and bias uncertainty associated with the calculational method and nuclear data used.

1.1.2 Previous Work In 2012, NUREG/CR-7109 [1] was created to provide guidelines for the validation of keff calculations in BUC studies. The report presented results from a validation study for BUC criticality safety analyses of commercial SNF storage systems considering both the actinide-only (AO) and actinide and fission product nuclide sets. A total of 609 critical benchmark configurations were considered in the study, and application models included SNF from both pressurized water reactors (PWRs) and boiling water reactors (BWRs). The calculations were performed with the SCALE 6.1 code package and the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data library. The only available nuclear covariance data resource in SCALE 6.1 was the 44-group library first released with SCALE 6. The conclusions presented in NUREG/CR-7109 showed that for fuel up to 5 wt% 235U, Sufficient benchmark experiments were available to validate BUC criticality safety analysis confidently for the major actinides, and A conservative estimate for the bias associated with minor actinide and fission product (FP) nuclides of 1.5% of their worth may be used to account for the lack of adequate critical benchmarks containing these nuclides.

1.1.3 Developments and Needs The nuclear industry has seen increased interest in using extended enrichment (between 5 to 8 wt% 235U) fuel and higher burnup fuel. In addition to the release of the SCALE 6.3 code package

[2], the ENDF/B-VII.1 [3] and ENDF/B-VIII.0 [4] nuclear data libraries were also released subsequent to the publication of NUREG/CR-7109. To comply with the requirements for safe storage and transportation of SNF corresponding to those extended fuel enrichments and burnups and/or use of these new data, validation studies for BUC must be performed. Results from these studies are used to identify the appropriate critical benchmarks and to determine the effect of using these fuel types and/or new data on the bias and bias uncertainty.

1-2 1.2 Purpose and Outline There are two purposes for this report. The first purpose is to study the effects of extended enrichment and high burnup fuels on the validation of BUC criticality safety analyses of SNF storage systems. The second purpose is to study the effects of ENDF/B-VII.1 or ENDF/B-VIII.0 nuclear data on the validation. The study also investigates the applicability of the conclusions made in NUREG/CR-7109 [1] to these higher enrichment and higher burnup fuels and new nuclear data sets. An overview of the computational methods, nuclear data libraries, and models used in this report is presented in Section 2. Section 3 discusses the selected benchmark experiments in the validation studies and the results of the similarity assessments.

Section 4 presents the bias and bias uncertainty calculations for different application cases and the effect of using different nuclear data libraries and covariance data.

2-1 2 COMPUTATIONAL METHODS, DATA, AND MODELS 2.1 Methods and Data This section describes the various tools and the methodology used in the analyses. The SCALE 6.3 code package [2], along with ENDF/B-VII.1 [3] and ENDF/B-VIII.0 [4] 252-group neutron cross-section libraries and 56-group covariance data, were the main resources used in this study. Figure 2-1 shows the key components used in the SCALE code package and their sequence for this analysis.

CSAS:

Criticality Safety Analysis Sequence ORIGAMI:

Oak Ridge Isotope Generation (ORIGEN) Assembly Isotopics TRITON:

Transport Rigor Implemented with Time-dependent Operation for Neutronic [depletion]

TSUNAMI:

Tools for Sensitivity and Uncertainty Analysis Methodology Implementation TSUNAMI-IP:

TSUNAMI Indices and Parameters Figure 2-1 Key Components Used in the SCALE Code System 2.1.1 Depletion Figure 2-2 shows the TRITON depletion (t-depl) sequence that was used in SCALE for this analysis. The t-depl sequence and the Oak Ridge Isotope Generation (ORIGEN) program were used for lattice physics and depletion, fuel modeling, and generation of the fuel inventories for the various burnup, enrichment, and depletion conditions. The t-depl sequence uses the material and cross section processing module, XSProc, and the NEWT code for cross section processing and neutron transport calculation, respectively. XSProc accounts for temperature and resonance self-shielding to prepare the multigroup (MG) cross-section library, and NEWT obtains the two-dimensional (2-D) flux solution in the geometry of interest. ORIGEN then uses the NEWT flux solutions to calculate the time-dependent fuel compositions. TRITON controls nuclides for which high-fidelity self-shielding calculations are performed in fuel material during depletion using the ADDNUX parameter. For this work, the ADDNUX 3 (231 nuclides) was used to ensure a balance between fidelity and computational effort. Several studies were performed with an expanded set of nuclides using ADDNUX 4 (388 nuclides). A complete listing of the nuclides used with each ADDNUX value can be found in the SCALE user manual [2]. TRITON produces an ORIGEN library file (.f33) containing the transition matrices corresponding to different burnups. The ORIGEN libraries account for three separate sub-libraries: light elements (LT), actinides (AC), and fission products (FP). The Automatic Rapid Processing (ARP) code creates burnup and enrichment-dependent ORIGEN cross-section libraries by interpolating over Nuclear Data Depletion TRITON ORIGAMI Criticality CSAS Sensitivity and Uncertainty TSUNAMI-3D TSUNAMI-IP

2-2 reactor cross-section libraries generated in advance using reactor physics transport methods in TRITON. The ORIGEN Assembly Isotopics (ORIGAMI) program uses the ORIGEN and ARP libraries to compute the detailed nuclide inventory at a specified burnup, cooling time, and assembly power distribution for the light water reactor assembly. In performing these calculations, ORIGAMI used an 18-zone axial power distribution [5].

ARP:

Automatic Rapid Processing NEWT:

New ESC-based Weighting Transport [code]

ORIGEN:

Oak Ridge Isotope Generation Figure 2-2 Depletion Sequence Used in the Analysis The ORIGAMI sequence was used to generate standard composition blocks that can be used directly in Criticality Safety Analysis Sequence (CSAS) and Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI)-3D calculations. The composition block contains the axially variable nuclide concentrations. For each burnup, fuel enrichment, or decay time, a new composition data block is generated for the KENO or TSUNAMI input, thus producing a unique keff value with each change in condition.

The nuclides considered in this study were divided into three sets [5]:

1.

Actinides only (AO):

2.

Actinides and 16 FP (AFP):

3.

Actinides and all FP (ALL):

234U, 235U, 238U, 238Pu,239Pu,240Pu,241Pu,242Pu,241Am, and O The nuclides mentioned above plus the following actinides and fission products: 236U, 243Am, 237Np, 95Mo, 99Tc,101Ru,103Rh,109Ag,133Cs, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 143Nd, 145Nd, 151Eu, 153Eu, 155Gd The nuclides mentioned above plus those included in the TRITON depletion calculation with the ADDNUX=4 setting [2].

Additional fission products beyond this set are not expected to impact keff.

2.1.2 Criticality The SCALE Criticality Safety Analysis Sequence with KENO V.a transport (CSAS5) sequence was used to calculate keff values and their uncertainties for the different models in the criticality calculations. The CSAS5 sequence uses the Monte Carlo code KENO V.a to solve eigenvalue problems in 3-D models. To minimize the number of generations skipped for source convergence, the start option block was used to control the initial neutron distribution. Start type 7, which was used for all calculations, samples the initial distribution uniformly in the radial ORIGAMI TRITON NEWT ORIGEN ARP

2-3 direction and from a (1-cos z)2 distribution axially to start more particles at the ends of the fuel assembly. This approach generally aids source convergence for burned fuel. Twenty thousand particles per generation were simulated, with 100 generations skipped. The cases ran for a sufficient number of generations for the Monte Carlo uncertainty in keff to reduce to 10 per cent mille (pcm).

2.1.3 Sensitivity, Uncertainty, and Validation The SCALE Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI)-3D sequence was used to determine the sensitivity of the calculated keff values to each constituent nuclear data component used in the calculation. The sensitivity calculations were performed in multigroup (MG) mode and thus consisted of explicit forward and adjoint calculations. The sensitivities are reported as a function of the mixture, nuclide, and reaction and are stored in a sensitivity data file (SDF) for subsequent analysis. The calculated sensitivities can then be coupled with the energy-dependent covariance data to generate the nuclear data-induced uncertainty in keff. TSUNAMI-3D was used to generate SDFs for all evaluated critical benchmark experiments and application models. To verify appropriate parameters were selected for the TSUNAMI-3D sequence, direct perturbation eigenvalue calculations were performed. In direct perturbation calculations, atom densities of selected nuclides of high sensitivity are altered to test the accuracy of the sensitivity results.

The SCALE TSUNAMI Indices and Parameters (TSUNAMI-IP) program can also use sensitivity data from TSUNAMI-3D and nuclear covariance data to generate the data-induced uncertainty, and also determine the degree of similarity between two systems based on an integral index ck.

The uncertainty data can be reported by TSUNAMI-IP on a nuclide-reaction basis, or it can be integrated into a single uncertainty for the entire system. The individual nuclide-reaction uncertainties can be combined to determine the data-induced uncertainty contribution from a specific nuclide. Similarly, the contributions of each nuclide-reaction pair to the integral index ck can also be generated and reported by TSUNAMI-IP to support the analysis of similarity results.

The Validation Analysis Data Evaluation Resource (VADER) sequence in SCALE was used for the determination of bias and bias uncertainty values. The bias and bias uncertainty values for the different application cases are determined using trending, nontrending parametric, or nontrending nonparametric methods. These methods are described in Clarity et al. [6].

2.1.4 Nuclear Data The ENDF/B-VII.1 [3] and ENDF/B-VIII.0 [4] nuclear data libraries were used in this study. While most neutron transport calculations were performed with the 252-group MG libraries distributed with SCALE 6.3.1, some calculations were also performed with the continuous-energy (CE) data also distributed with SCALE. Compared to ENDF/B-VII.1, ENDF/B-VIII.0 has some important changes in nuclides like 1H, 3He, 9Be, 6Li, 16O, 56Fe, 235U, 238U, and 239Pu [4].

Three different nuclear covariance libraries were used in the analyses documented in this report. The oldest library is the 44-group SCALE covariance library first distributed with SCALE 6 [7], which was used for the work presented in NUREG/CR-7109 and is still distributed with SCALE 6.3.1. The SCALE 6.2 covariance library based primarily on ENDF/B-VII.1 data [3] was used in both the 56-and 252-group structures. The 56-group library was used predominantly because it is the default covariance library in SCALE 6.2 and 6.3.1. Finally, the ENDF/B-VIII.0 56-group covariance library distributed with SCALE 6.3.1 [24] was also used. It is important to note that the SCALE team curated covariance libraries released with SCALE 6 and SCALE 6.2

2-4 to include a set of covariance data that the SCALE developers believed were the best available data. The ENDF/B-VIII.0 library released with SCALE 6.3.1 no longer makes any attempt to improve on the ENDF data for any nuclides or reactions. The ENDF data is processed without replacement or modification. All four libraries contain a significant number of low-fidelity evaluations generated by a collaborative project among Brookhaven National Laboratory (BNL),

Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL)BLO

[8].

The covariance libraries do not contain complete covariance data for all reactions for all nuclides over the entire energy range being considered in SCALE calculations. In cases where covariance data is not available, TSUNAMI-IP draws on user-supplied default covariance data provided in the input. All calculations performed used covariance data fully correlated within the thermal, intermediate, and fast regions. The uncertainty values used are 5% in the thermal range, 15% in the intermediate range, and 40% in the fast range. TSUNAMI-IP flags any results that used these default values to patch the covariance data from the library. These covariance patches do not significantly impact any of the results documented here.

Table 2-1 lists all the nuclear data libraries and covariance data used in this study, as well as the corresponding symbols.

Table 2-1 Nuclear Data Libraries and Covariance Data Used 2.2 Models 2.2.1 Fuel Assembly Depletion Model The Westinghouse 17x17 optimized fuel assembly (OFA) design was used in this study. More details on the assembly specifications can be found in NUREG/CR-7109 [1]. Each assembly contains 24 wet annular burnable absorber (WABA) rodlets consisting of an annular absorber of alumina/boron carbide (Al2O3 / B4C) that is contained within two concentric Zircaloy-4 tubes during depletion [1].

The TRITON t-depl sequence was used to model the assembly and generate the fuel inventories to reflect various burnup, enrichment, and depletion conditions (i.e., burnable absorber history, soluble boron concentration, etc.). TRITON generates ORIGEN libraries that can be used in the ORIGAMI code to generate detailed nuclide inventories over the burnup Library Symbol Nuclear-Data ENDF/B-VIII.0 with 252 neutron energy groups E8.0 ENDF/B-VIII.0 with continuous neutron energy E8.0CE ENDF/B-VII.1 with 252 neutron energy groups E7.1 ENDF/B-VII.1 with continuous neutron energy E7.1CE Covariance Nuclear Data ENDF/B-VIII.0 56-group (added in SCALE 6.3)

Cov A SCALE 6.2 56-group based on ENDF/B-VII.1 Cov B SCALE 6.2 252-group based on ENDF/B-VII.1 Cov C

2-5 range for up to 90 GWd/MTU for each enrichment. A maximum node-average burnup of 90 GWd/MTU was necessary for assembly average burnups of 80 GWd/MTU. Because ORIGAMI is an interpolating tool, the enrichment range considered ranged from 3 to 9 wt% 235U to allow for the study of the 4 to 8 wt% 235U enrichment range of interest. The WABA rods were included throughout the depletion process to harden the neutron energy spectrum. The harder spectrum increases the production of Pu during depletion, resulting in a conservative representation of the SNF residual reactivity. Figure 2-3 shows a TRITON model of a 1/4 OFA assembly, and the axial burnup distribution is shown in Figure 2-4. WABA rods are represented in the figure with the off-colored square regions with concentric rings inserted into the guide tube locations, while fuel rods are shown in grey. The depletion calculations are unaffected by the nuclide set used in the criticality or sensitivity calculations. The desired set of nuclide number densities is extracted for use in the cask model after the desired burnup and cooling time is completed. APPENDIX A includes TRITON data, ARP data, and ORIGAMI sample files.

Figure 2-3 A 2-D TRITON Model of the Westinghouse 17x17 OFA (1/4 Fuel Assembly)

Figure 2-4 Axial Burnup Profile

2-6 2.2.2 GBC-32 Cask Storage Model The GBC-32 computational benchmark model developed for NUREG/CR-6747 was also used for this analysis. The GBC-32 includes square storage cells to accommodate PWR fuel assemblies. The cells have an inner dimension of 22 cm, and they are surrounded by stainless-steel walls and Boral panels measuring 365.76 cm tall and 19.05 cm wide. The panels are placed between storage cells and on the external faces of each cell. The 10B areal density in the panels is 0.0225 g 10B/cm2.

The cask was loaded with Westinghouse 17x17 OFA centered in the cask storage cells. The WABA rodlets present during depletion are not included in the assembly in the cask model.

Modeling of discharged BAs in the cask is generally neglected for conservatism. Neglecting the presence of the BA assemblies also allows greater flexibility in loading the storage cask. The cask and fuel are described in detail in NUREG/CR-6747 [5] and NUREG/CR-7109 [1]. Figure 2-5 shows a cutaway view of the KENO V.a cask and fuel model. A half-cask model is used in the transport calculations reported here to allow a finer mesh for the TSUNAMI-3D calculations.

The -Y face of the model has a mirror boundary condition applied to simulate the entire cask.

APPENDIX A includes CSAS5, TSUNAMI-3D, TSUNAMI-IP, and VADER sample files.

Figure 2-5 Cutaway of the GBC-32 Cask and PWR Fuel Assemblies

3 CRITICALITY BENCHMARK EXPERIMENTS The goal of computational validation is to determine the discrepancy between calculated and experimental results, or computational bias. In burnup credit (BUC) criticality safety analysis, for example, the computational bias is associated with the keff values predicted by a computational model of a criticality application system. The computational bias is derived by comparing laboratory critical experiment (LCE) results to calculational model results for those experiments.

The application model should be as similar as possible to the experiments to ensure the computational bias is applicable to the application model results. As discussed in Section 2.1.3, critical experiment similarity is judged in these analyses with an integral index ck called similarity coefficient [6].

In the case of BUC analysis, the selected LCEs should include fuel that most resembles spent fuel, with a significant amount of plutonium and fission products, for example. Other characteristics of interest are the fuel-to-moderator ratiousually lower than that of fresh fuel that is commonly used in critical experiments and the structural materials typically used for storage and transportation casks. It is difficult to find LCEs with those specific characteristics because the main objective of critical experiments is usually to validate fresh fuel. Therefore, the most similar LCEs used for BUC validation typically involve mixed oxide (MOX) or low-enriched-uranium (LEU) fuel types, depending on the burnup [1].

The similarity coefficient, ck, ranges between -1 and 1 and is based on the related nuclear data uncertainties in the calculated keff values between an LCE and an application. A value close to 1 indicates that the two models have a strong positive correlation in terms of nuclear data uncertainty, and the LCE is applicable for estimating the computational bias for the application model. The current guidance is to use LCEs with correlation coefficients ck 0.9, which are considered highly similar. When available, experiments with 0.8 ck < 0.9 are also used because they are considered marginally similar [10]. Experiments with ck < 0.8 are not recommended to be used in validation and computational bias determination studies [11]. The ck values are calculated with TSUNAMI-IP in SCALE by combining the TSUNAMI-3D sensitivity data and the nuclear data uncertainty from covariance matrices. Compared to the analysis presented in NUREG/CR-7109 [1], the study described herein analyzes the similarity implications of using spent fuel in a GBC-32 cask with initial enrichment ranging from 4 wt% 235U up to 8 wt% 235U and burnup ranging from 10 GWd/MTU up to 80 GWd/MTU, thus corresponding to the needs of some of the slightly higher enrichments and burnups currently planned for incorporation in light-water commercial power reactors [12]. These studies also consider nuclear data from ENDF/B-VII.1 and ENDF/B-VIII.0, which have been released since the publication of NUREG/CR-7109. Table 3-1 summarizes the cases considered in the similarity assessment for the GBC-32 spent fuel storage cask application. For comparison, the SCALE 6.1 44-group covariance data were used with the ENDF/B-VII.0 nuclear data library in NUREG/CR-7109.

3-1

3-2 Table 3-1 Characteristics of the Application Cases in Similarity Assessment Enrichment (wt% 235U) 4, 5, 6, 7, 8 Burnup (GWd/MTU) 10, 40, 60, 70, 75, 80 Nuclear data library ENDF/B-VII.1 (E7.1) with SCALE 6.2 covariance library (Cov B)

ENDF/B-VIII.0 (E8.0) with ENDF/B-VIII.0 covariance library (Cov A)

Nuclide set Actinides and 16 fission products (AFP) 3.1 Selected Benchmarks Sensitivity data files generated with TSUNAMI-3D for 2,104 critical experiments are used as candidate experiments for the validation set with sources shown in Table 3-2. The experiments were selected based on having a thermal fission neutron energy, fissile material of LEU, intermediate enriched uranium (IEU), or a mixture of uranium and plutonium (MIX), and available SCALE-compatible sensitivity data. Other critical experiments exist but could not be rigorously assessed in TSUNAMI-IP, or they were considered unlikely to be relevant given differences in fissile material and/or neutron energy spectrum. The primary source of critical experiments for this effort is the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook [13]. Of the more than 5,000 configurations evaluated in the ICSBEP Handbook, 1,949 critical experiments were considered in this work. Among these experiments, 218 are included in the ORNL Verified, Archived Library of Inputs and Data (VALID) [14], 1,323 are distributed by the 2021 Nuclear Energy Agency (NEA) database for ICSBEP (DICE) [15],

and 408 are ICSBEP benchmarks modeled internally at ORNL that are in the process of being added to the VALID database. Among the 1,323 NEA-generated experiments, 1,097 come from LEU experiments, four from IEU experiments, and 222 from MIX experiments. The remaining 155 experiments are from the French Haut Taux de Combustion (HTC) [16-19] program in France. Note that 156 experiments were actually evaluated in the HTC program, but only 155 were used in this study because the MIX-HTC2B-013 similarity coefficient appears abnormally low for all cases studied. This issue remains unexplained, so it was decided to discard MIX-HTC2B-013 from the validation set.

Table 3-2 Source of Critical Experiments Used in Validation Source Sub-category per source Number of critical experiments ICSBEP VALID database 218 NEA LEU 1,097 NEA IEU 4

NEA MIX 222 Benchmarks In the process of being added to the VALID database 408 HTC 155 Total 2,104

3-3 Table 3-3 details the 1,949 critical configurations used from the ICSBEP Handbook. As seen in the table, many LEU experiments are used. The LEU experiments all use thermal neutron energies and include different fuel forms such as compounds, solutions, metallic, and miscellaneous fuel (e.g., LEU-MISC-THERM-001, an experiment using 5% enriched fuel rods in 6% enriched uranyl nitrate solutions with water reflection). The LEU experiments were expected to be marginally similar for low burnup and low initial 235U enrichment. However, the lack of plutonium in the systems would make them unlikely candidates for validation of high burnup fuel. The MIX experiments are more likely to be applicable to BUC validation because those systems include uranium and plutonium in different ratios, which is similar to a spent fuel application. The MIX experiments include natural or depleted uranium, and a few have plutonium isotopics similar to spent fuel in compound and solution systems. The few IEU experiments could apply to the low-burnup extended enrichment application model cases.

Table 3-3 Evaluations Used from the ICSBEP Handbook The fuel rods used in the HTC experiments were specifically designed to be similar to PWR spent fuel up to 37.5 GWd/MTU with an initial enrichment of 4.5 wt% 235U, thus including significant plutonium amounts and 241Am from the decay of 241Pu. The HTC fuel rods have a uniform axial burnup profile that differs from the real spent fuel burnup profile shown in Figure 2-4. Some configurations also include structural materials such as stainless steel, similar to Experiment type Number of experiments Number of evaluations List of evaluations IEU-COMP-THERM 103 5

001, 002, 011, 015, 016 IEU-SOL-THERM 63 5

001, 002, 003, 004, 005 LEU-COMP-THERM 1,250 83 001, 002, 003, 004, 005, 006, 008, 009, 010, 011, 012, 013, 014, 015, 016, 017, 018, 020, 021, 022, 023, 024, 025, 026, 027, 028, 029, 030, 031, 032, 033, 034, 035, 036, 037, 038, 039, 040, 042, 043, 044, 045, 046, 047, 048, 049, 050, 051, 052, 053, 054, 055, 057, 058, 061, 062, 065, 066, 068, 069, 070, 071, 072, 073, 074, 075, 076, 077, 078, 079, 080, 082, 083, 084, 085, 089, 090, 091, 092, 094, 096, 097, 101 LEU-MET-THERM 79 6

001, 002, 004, 006, 007, 015 LEU-MISC-THERM 48 5

001, 002, 003, 005, 006 LEU-SOL-THERM 115 22 001, 002, 003, 004, 005, 006, 007, 008, 009, 011, 012, 013, 016, 017, 018, 019, 020, 021, 022, 023, 024, 025 MIX-COMP-THERM 249 11 001, 002, 004, 006, 007, 008, 012, 013, 014, 016, 017 MIX-SOL-THERM 42 5

001, 002, 003, 007, 010

3-4 those in the GBC-32 application models used in this study. Some cases contain materials not present in the application model, such as lead or soluble poisons, including boron or gadolinium.

These 155 experiments are most likely to be highly similar to the application models at moderate and high burnups. An evaluation of the HTC experiment data is documented in NUREG/CR-6979 [16].

The energy of average neutron lethargy causing fission (EALF) was used in the validation process, along with uranium enrichment, plutonium ratio, boron concentration, and gadolinium concentration. The ranges of key parameters in the 2,104 critical experiments used are summarized in Table 3-4. A complete list of the 2,104 experiments used is provided in APPENDIX B. From the 2,104 selected experiments, only the LCEs with ck correlation coefficients greater than or equal to 0.8 are recommended to be used to estimate the computational bias for each case, as described in Section.

Table 3-4 Range of Key Parameters Used in Validation 3.2 Similarity Assessment 3.2.1 Inventory Changes in the Application System Before the results of the similarity assessments are presented, it is important to explain how the fissile inventory changes with respect to enrichment and burnup in the application system.

Results from previous studies [20, 21] show that major actinide sensitivity dominates similarity coefficient determinations between experiments and applications. As such, finding similar quantities of uranium, plutonium, and ratios of uranium and plutonium in two systems can lead to a higher similarity coefficient, ck, assuming that neutron energy spectra are similar. The observations presented in this section aim to predict the similarity changes with increasing burnup and initial enrichment by analyzing the 235U and 239Pu number densities changes.

Because there are many other competing factors, the analysis given here should serve as a qualitative indication of the expected similarity trends.

During depletion, the quantity of uranium in the fuel decreases while the quantity of plutonium increases. Figure 3-1 shows the number densities for 235U, 239Pu, and 240Pu for initial fuel enrichments of 4 and 8 wt% 235U in node 17 (a region located 325 to 345 cm from the bottom of the fuel rod) as a function of assembly average burnup. Node 17 is chosen for this study because it is representative of the positions with the highest fission density and reactivity in the spent fuel. Because node 17 burnup corresponds to 73.8 % of the assembly average burnup, as shown in Figure 2-4, the number densities given correspond to a node burnup that is 0.738 Parameter Minimum Maximum EALF (eV) 0.011 7.56 Uranium enrichment (wt% 235U) 0.15 36.96 Pu/(Pu+U) (ratio) 0 0.29 Soluble boron concentration (g/L) 0 95.74 Soluble gadolinium concentration (g/L) 0 1.48

3-5 times the assembly average burnup. For example, a burnup of 40 GWd/MTU assembly average corresponds to 40 x 0.738 = 29.5 GWd/MTU node 17 burnup. The figure shows that the 235U continually decreases, whereas the 240Pu continually increases over the burnup range at a similar rate between the 4 and 8 wt% cases. The 235U number densities at the highest burnup are very different in both cases because of the different initial inventory. Therefore, it can be conjectured that the similarity coefficients of those two cases would be significantly different when compared to the same experiments. The 240Pu number densities are similar, but this nuclide does not contribute significantly to ck. The number density for the other major nuclide for ck contribution, 239Pu, steeply increases in the 4 wt% case until reaching a plateau at approximately 40-50 GWd/MTU burnup. In the 8 wt% case, the 239Pu number density also steeply increases until approximately 40-50 GWd/MTU and then continues to slowly increase.

Note that at burnups below approximately 20 GWd/MTU, the 239Pu number density is higher for the 4 wt% case than for the 8 wt% case. This is caused by the lower 235U capture and higher 238U capture at lower initial enrichment compared to higher ones, thus increasing the 239Pu production at lower initial enrichment and low burnup.

Figure 3-1 shows that despite large differences in 235U number densities between the two bounding cases of 4 and 8 wt% initial enrichments over the burnup range, the 239Pu number densities remain relatively similar in both cases between 40 and 80 GWd/MTU. This asymptotic behavior at 4 wt% and a slower rate of change at 8 wt% at high burnup indicate that the similarity to experiments should change much less between 40 and 80 GWd/MTU than between 10 and 40 GWd/MTU.

Figure 3-2 shows the number densities of 235U, 239Pu, and the 239Pu/(235U +239Pu) ratio for an initial enrichment of 4 wt% 235U as a function of burnup compared to the 239Pu/(235U +239Pu) ratio of the HTC experiments. As previously detailed, the HTC experiments include 155 configurations designed to be similar to PWR fuel irradiated to 37.5 GWd/MTU with an initial enrichment of 4.5 wt% 235U. The 239Pu/(235U +239Pu) ratio for the HTC experiments is 0.29 and is constant, so it is shown as a straight line. The figure illustrates that the 239Pu/(235U +239Pu) ratio of the burned fuel application gradually increases with burnup as the 235U number density decreases and the 239Pu number density increases. Between 10 and 40 GWd/MTU, the 239Pu/(235U +239Pu) ratio of the burned fuel application is lower than that of the HTC experiments.

At approximately 40 GWd/MTU, the ratios of the experiment and application systems are similar, whereas above 40 GWd/MTU, the 239Pu/(235U +239Pu) ratio of the burned fuel in the application system is higher than that of the HTC experiments. Between 10 and approximately 40 GWd/MTU, the behavior becomes increasingly similar to that seen in the HTC experiments, so their similarity, as assessed by ck, is also expected to increase. Above 40 GWd/MTU, the application should become less similar to the HTC experiments, so their similarity is expected to decrease. According to this simplistic analysis, at approximately 42 GWd/MTU assembly average burnup, the similarity coefficient of the application system and the experiment is expected to be the highest, matching the HTC fuel characteristics of 4.5 wt% 235U initial enrichment and 37.5 GWd/MTU. The results presented in Section 3.2.2 further assert these predictions. However, the exact burnup at which the maximum similarity occurs was not estimated.

3-6 Figure 3-1 Node 17 235U, 239Pu, and 240Pu Number Densities for Increasing AssemblyAverage Burnup for 4 and 8 wt% 235U Fuels Figure 3-2 Node 17 235U and 239Pu Number Densities and 239Pu /(235U +239Pu) Ratio Compared to HTC 239Pu /(235U +239Pu) Ratio for Increasing Assembly Average Burnup and 4 wt% 235U Fuels Similarly, the number densities of 235U, 239Pu, and 239Pu /(235U +239Pu) ratio are compared to the 239Pu/(235U +239Pu) ratio of the HTC experiments for an 8 wt% 235U initial enrichment as a function of burnup.

3-7 239Pu /(235U +239Pu) ratio Figure 3-3 shows that the 239Pu/(235U +239Pu) ratio of the burned fuel application still gradually increases with burnup because the 235U number density continually decreases and the 239Pu in the fuel increases. Over the entire burnup range, the 239Pu /(235U +239Pu) ratio of the burned fuel application is lower than that for the HTC 239Pu /(235U +239Pu) ratio. Based on this result, it can be conjectured that the application is becoming increasingly similar to the HTC experiments with burnup, so the ck value is also expected to increase for all increasing burnup values. Because the 239Pu/(235U +239Pu) ratio from the fuel is significantly below the HTC ratio at low burnup, a low ck is expected. The ck value should increase with burnup and should reach a maximum value at 80 GWd/MTU. The results presented in Section 3.2.2 and beyond verify these expectations.

Figure 3-3 Node 17 235U and 239Pu Number Densities and 239Pu /(235U +239Pu) Ratio Compared to HTC 239Pu /(235U +239Pu) Ratio for Increasing Assembly Average Burnup and 8 wt% 235U Fuel The analysis described in this section was performed only for 4 and 8 wt% 235U enrichments. It is expected that the logic applied to intermediate cases (i.e., 5, 6, and 7 wt% 235U) would be consistent.

3.2.2 Summary and General Observations of Similarity Assessments This section discusses the results of the similarity study performed with the SCALE sensitivity and uncertainty (S/U) toolsTSUNAMI-3D and TSUNAMI-IPfor the 60 application cases defined in Table 3-1 and the 2,104 critical experiments detailed in Table 3-2 and Table 3-3. The number of similar experiments for each application case and ck range is summarized in Table 3-5.

3-8 The results are tabulated for (1) the ENDF/B-VII.1 nuclear data library, with 252 neutron energy groups with the SCALE 6.2 56-group covariance library based on ENDF/B-VII.1, (E7.1 with Cov B), and (2) the ENDF/B-VIII.0 nuclear data library with 252 neutron energy groups, with the ENDF/B-VIII.0 56-group covariance library that was added to SCALE 6.3 (E8.0 with Cov A).

Table 3-5 Similarity Assessment Summary 235U (wt%)

Burnup (GWd/MTU)

Number of critical experiments meeting ck criteria E7.1 with Cov B E8.0 with Cov A ck < 0.8 ck 0.8 0.8 ck < 0.9 ck 0.9 ck < 0.8 ck 0.8 0.8 ck < 0.9 ck 0.9 4

10 1,870 234 211 23 1,177 928 911 17 40 1,952 152 66 86 1,831 273 138 135 60 1,893 211 135 76 1,826 278 150 128 70 1,888 216 159 57 1,820 284 170 114 75 1,889 216 209 7

1,820 284 176 108 80 1,896 208 205 3

1,819 285 189 96 5

10 1,868 236 190 46 1,019 1,085 1,085 0

40 1,922 182 113 69 1,862 242 117 125 60 1,925 179 83 96 1,831 273 139 134 70 1,900 204 109 95 1,829 275 148 127 75 1,897 207 124 83 1,828 276 149 127 80 1,894 210 140 70 1,827 277 153 124 6

10 1,836 268 222 46 865 1,239 1,233 6

40 1,825 279 273 6

1,925 179 72 107 60 1,953 151 68 83 1,860 244 118 126 70 1,952 152 57 95 1,833 271 145 126 75 1,933 171 76 95 1,833 271 145 126 80 1,925 179 84 95 1,832 272 147 125 7

10 1,818 286 240 46 906 1,198 1,004 194 40 1,759 345 345 0

1,954 150 61 89 60 1,910 194 128 66 1,867 237 128 109 70 1,956 148 78 70 1,861 243 131 112 75 1,956 148 66 82 1,860 244 126 118 80 1,954 150 64 86 1,860 244 126 118 8

10 1,825 280 234 46 899 1,205 875 330 40 1,728 376 376 0

1,961 143 112 31 60 1,821 283 279 4

1,949 155 61 94 70 1,915 189 159 30 1,894 210 104 106 75 1,957 147 84 63 1,868 236 130 106 80 1,957 147 80 67 1,865 239 132 107

3-9 Several observations are made based on the information presented in Table 3-5:

The minimum number of at least marginally similar experiments (ck 0.8) is 147 for ENDF/B-VII.1 with the SCALE 6.2 covariance library and 143 for ENDF/B-VIII.0 nuclear data and covariance library, thus indicating that enough experiments exist to perform validation.

The number of highly similar experiments (ck 0.9) varies significantly, depending on the combination of enrichment, burnup, and nuclear data included, with a minimum of 0 for several cases and a maximum of 330. This shows that despite a good pool of marginally similar experiments, improvements in the available experiments are possible, as detailed below.

Among the 2,104 critical experiments, at least 1,728 are not applicable to BUC analysis studies with the ENDF/B-VII.1 nuclear data, and 865 are not applicable with ENDF/B-VIII.0 nuclear data. This demonstrates that many of the ICSBEP Handbook experiments are not applicable for PWR BUC validation.

The number of similar experiments varies significantly between the two sets of nuclear data used, especially at 10 GWd/MTU at all initial enrichments, where there are hundreds more applicable experiments when using ENDF/B-VIII.0 nuclear data compared to ENDF/B-VII.1.

To better analyze the number of similar experiments with the initial enrichment, burnup, and nuclear data library used, the data from Table 3-5 are plotted below in Figure 3-4 through Figure 3-7. Figure 3-4 shows the number of at least marginally similar (ck 0.8) and highly similar (ck 0.9) experiments for all the enrichment/burnup combinations using the ENDF/B-VII.1 nuclear data. The following observations can be made:

The number of marginally similar experiments within the same enrichment does not vary significantly at higher burnups (70-80 GWd/MTU).

The number of marginally similar experiments increases with enrichment at low and medium burnups (10-40 GWd/MTU).

At low initial enrichments of 4-5 wt% 235U, the number of highly similar experiments is greatest at medium burnup and lowest for the 4 wt% 235U case at low and high burnups.

At medium initial enrichments of 6 wt% 235U and at high initial enrichments ranging from 7 to 8 wt% 235U, the number of highly similar experiments is greatest at low and high burnup and is lowest at medium burnup (40 GWd/MTU).

3-10 Figure 3-4 Number of Similar Benchmark Experiments Using the E7.1 Library Figure 3-5 shows the number of at least marginally similar (ck 0.8) and highly similar (ck 0.9) experiments for all the enrichment/burnup combinations using the ENDF/B-VIII.0 nuclear data.

The following observations can be made:

The number of marginally and highly similar experiments within the same enrichment does not vary significantly at higher burnups of 70-80 GWd/MTU.

The number of marginally similar experiments is greatest for all enrichments at low burnup (10 GWd/MTU).

At low and medium initial enrichments ranging from 4 to 6 wt% 235U, the number of highly similar experiments is greatest at medium burnup and is lowest at low burnups (10 GWd/MTU).

At high initial enrichments of 7-8 wt% 235U, the number of highly similar experiments is greatest at low burnup and is lower at medium burnup (40 GWd/MTU).

3-11 Figure 3-5 Number of Similar Benchmark Experiments Using the E8.0 Library Figure 3-6 shows the number of experiments that are at least marginally similar (ck 0.8) for all enrichment/burnup combinations for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data. The following observations can be made:

At low initial enrichments ranging between 4 and 5 wt% 235U, the number of marginally similar experiments is always higher when using the ENDF/B-VIII.0 nuclear dataset than the ENDF/B-VII.1 nuclear dataset.

At medium and high initial enrichments of 6-8 wt% 235U, the number of marginally similar experiments is higher at medium burnup (40 GWd/MTU) when using the ENDF/B-VII.1 nuclear dataset than the ENDF/B-VIII.0 nuclear dataset.

The number of marginally similar experiments is within a factor of approximately 2 between both nuclear datasets except for the 10 GWd/MTU datapoint at all enrichments, for which the ENDF/B-VIII.0 nuclear dataset shows about three times more marginally similar experiments than for the ENDF/B-VII.1 nuclear dataset.

3-12 Figure 3-6 Comparison of the Number of Marginally Similar Benchmark Experiments (ck 0.8)

Figure 3-7 shows the number of highly similar (ck 0.9) experiments for all the enrichment/burnup combinations for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data. The following observations can be made:

The number of highly similar experiments is larger when using the ENDF/B-VIII.0 nuclear data than with ENDF/B-VII.1 nuclear data in 27 of the 30 enrichment-burnup pairs cases.

In those 27 cases, the difference between the number of highly similar experiments and the sets of nuclear data is usually very large, as shown in the case with an initial enrichment of 4 wt% 235U and a burnup of 80 GWd/MTU. The number of highly similar experiments increases from 3 to 96 when moving from ENDF/B-VII.1 to ENDF/B-VIII.0 data.

The three cases in which the ENDF/B-VII.1 nuclear data have a larger number of similar experiments than the ENDF/B-VIII.0 nuclear dataset are at 10 GWd/MTU for 4, 5, and 6 wt% 235U initial enrichments.

3-13 Figure 3-7 Comparison of the Number of Highly Similar Benchmark Experiments The results shown in Table 3-5 indicate that enough experiments are available for validation of the initial enrichment in the range of 4 to 8 wt% 235U over the burnup range of 10 to 80 GWd/MTU and for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data.

3.2.3 Detailed Similarity Assessment Results This section presents a more in-depth analysis regarding which critical experiments provide the highest and lowest similarity coefficients in a limited number of cases. Because the number of similar experiments per burnup and enrichment follows a pattern between the boundary cases, only a few specific cases were studied at this level of detail: these are 10, 40, 80 GWd/MTU burnups with an initial enrichment of 4 and 8 wt% 235U. Results are presented for the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data. The plots of the resulting ck values for the 2,104 experiments for each of the 12 cases are shown in the following subsections. The number of experiments with ck above 0.8 and 0.9 is also shown in each figure. A summary of the presented cases is provided below in Table 3-6.

(ck 0.9)

3-14 Table 3-6 Summary of Presented Cases for Similarity Assessment Application Burnup (GWd/MTU)

Initial enrichment (wt% 235U)

Nuclear data library Covariance library Relevant section 1

10 4

E7.1 Cov B 3.2.3.1 2

40 3

80 4

10 8

E7.1 Cov B 3.2.3.2 5

40 6

80 7

10 4

E8.0 Cov A 3.2.3.3 8

40 9

80 10 10 8

E8.0 Cov A 3.2.3.4 11 40 12 80 3.2.3.1 Applications 1-3: 4 wt% initial 235U Enrichment and ENDF/B-VII.1 Nuclear Data Library and SCALE 6.2 Covariance Library The similarity results for the first application shown in Figure 3-8 correspond to an initial 235U enrichment of 4 wt% and a burnup of 10 GWd/MTU. This application has 234 experiments that are at least marginally similar, of which 23 are highly similar. The results from experiments with the highest similarity come from ICSBEP categories LEU-COMP-THERM and LEU-MET-THERM and from HTC cases. Because this is a low burnup, the application still contains uranium and a low amount of plutonium, so both the LEU experiment categories and the HTC experiments are similar. The number of highly similar experiments is low because only a few experiments correspond to an LEU/low burnup application case. The most similar experiment is LEU-COMP-THERM-008, with ck values between 0.90 and 0.92.

3-15 Figure 3-8 Similarity Coefficients of the 4 wt% 235U Enriched 10 GWd/MTU Burnup with E7.1 and Cov B (Application 1)

The similarity results for the second application shown in Figure 3-9 correspond to an application with an initial 235U enrichment of 4 wt% and a burnup of 40 GWd/MTU. This application has 152 experiments that are at least marginally similar, of which 86 are highly similar from the ICSBEP category MIX-COMP-THERM and HTC cases. The only marginally similar experiment from the ICSBEP Handbook is MIX-COMP-THERM-008-001; the remainder are from the HTC experiments. The HTC experiments were designed to correspond to a 4 wt%

235U enriched fuel with a burnup of 37.5 GWd/MTU, so the similarity with this application was expected to be very high. Because this is a medium burnup, the application contains less 235U, and the amount of plutonium is greater than in the 10 GWd/MTU case previously shown; therefore, the similarity of the LEU experiments decreased, and the similarity of any mixed uranium/plutonium experiments increased. The most similar experiment is HTC2B-009, which has a ck value of 0.96.

3-16 Figure 3-9 Similarity Coefficients of the 4 wt% 235U Enriched 40 GWd/MTU Burnup with E7.1 and Cov B (Application 2)

The similarity results for the third application shown in Figure 3-10 correspond to an application with an initial 235U enrichment of 4 wt% and a burnup of 80 GWd/MTU. This application has 208 experiments that are at least marginally similar of which three are highly similar, which are from the ICSBEP MIX-COMP-THERM category and HTC data. Because this is a high burnup, the application contains less 235U and more plutonium than the 40 GWd/MTU case previously shown. Therefore, the similarity of the LEU experiments and most of the HTC experiments decreased, and the similarity of all the MIX-COMP-THERM experiments increased, especially MIX-COMP-THERM-007 and MIX-COMP-THERM-008. As expected from the analysis of the 239Pu/(235U +239Pu) ratios in Section 3.2.1, the similarity of the application to the HTC experiments is increasing from 10 to 40 GWd/MTU and decreasing from 40 to 80 GWd/MTU.

The most similar experiment is from the HTC series, with a ck value of 0.92 for HTC2B-009.

Note that many of the HTC cases are similar, even at this high burnup, indicating that for the HTC cases, data will likely remain applicable, even in the high burnup ranges being considered in the industry today.

3-17 Figure 3-10 Similarity Coefficients of the 4 wt% 235U Enriched 80 GWd/MTU Burnup with E7.1 and Cov B (Application 3) 3.2.3.2 Applications 4-6: 8 wt% initial 235U enrichment and ENDF/B-VII.1 nuclear data library and SCALE 6.2 covariance library The similarity results for the fourth application shown in Figure 3-11 corresponds to an application with an initial 235U enrichment of 8 wt% and a burnup of 10 GWd/MTU. For this application, there are 279 experiments that are at least marginally similar, of which 46 are highly similar from ICSBEP categories LEU-COMP-THERM and LEU-MET-THERM. Because this is a high initial enrichment and low burnup, the application still contains a significant amount of 235U and a very low amount of plutonium. Hence, the LEU experiments are more similar than in the 4 wt% 10 GWd/MTU case. The HTC experiments are not applicable at this low burnup and relatively high enrichment. The most similar experiment is LEU-COMP-THERM-014-005, which has a ck value of 0.94.

3-18 Figure 3-11 Similarity Coefficients of the 8 wt% 235U Enriched 10 GWd/MTU Burnup with E7.1 and Cov B (Application 4)

The similarity results for the fifth application shown in Figure 3-12 correspond to an application with an initial 235U enrichment of 8 wt% and a burnup of 40 GWd/MTU. In this case, there are 376 experiments that are at least marginally similar, of which 0 are highly similar from the LEU-COMP-THERM and LEU-MET-THERM ICSBEP categories and from the HTC experiments. The HTC experiments were designed to correspond to a 4 wt% enriched and 40 GWd/MTU burnup fuel, so the HTC cases were expected to be marginally similar to this application, although the ck values are lower than for the 4 wt% initial enrichment cases presented in the previous section. At this medium burnup, the application contains less 235U and more plutonium compared to the 10 GWd/MTU case previously shown. The similarity of the LEU experiments decreased, and the similarity of any mixed uranium/plutonium experiments, including the HTC cases, increased. There are no highly similar experiments because this case contains an actinide mixture with too much plutonium for the applicability of LEU experiments and too much 235U for the applicability of the HTC data or other MOX experiments. The most similar experiment is LEU-COMP-THERM-014-005, with a ck value of 0.88.

3-19 Figure 3-12 Similarity Coefficients of the 8 wt% 235U Enriched 40 GWd/MTU Burnup with E7.1 and Cov B (Application 5)

The similarity results for the sixth application shown in Figure 3-13 correspond to an application with an initial 235U enrichment of 8 wt% and a burnup of 80 GWd/MTU. In this application, there are 147 experiments that are at least marginally similar of which 67 are highly similar, all from the HTC experiments. At this high burnup, the application contains less 235U and more plutonium compared to the 40 GWd/MTU case previously shown. The similarity of the LEU experiments decreased, and the similarity of the HTC and MIX-COMP-THERM (MCT) experiments increased. This case is similar to the 4 wt% 235U enriched and 40 GWd/MTU burnup case, with enough uranium burned and plutonium produced to match the characteristics of the HTC experiments. As expected from the analysis of the 239Pu/(235U +239Pu) ratios described in Section, the similarity of the 80 GWd/MTU application to the HTC experiments is increased compared to the 40 GWd/MTU application. Another interesting behavior is that the MIX-COMP-THERM experiment ck values are unexpectedly higher for cases at 4 wt% 235U initial enrichment than for cases at 8 wt% initial enrichment. This can be explained by the relatively high 239Pu/(235U +239Pu) ratio in most of the MIX-COMP-THERM experiments. The 4 wt% 235U enriched cases compare more favorably to the MIX-COMP-THERM cases in the ICSBEP Handbook than the 8 wt% enrichment cases because the ICSBEP MIX-COMP-THERM experiments contain natural or depleted uranium. The amount of 235U remaining from the higher initial enrichment causes a more significant actinide mismatch and correspondingly lower ck values for the 8 wt% cases. The most similar experiments for this application are from the HTC2B series, with ck values between 0.91 and 0.93. As with the 4 wt% initial enrichment cases, the HTC experiments remain applicable for validation at these high burnups.

3-20 Figure 3-13 Similarity Coefficients of the 8 wt% 235U Enriched 80 GWd/MTU Burnup with E7.1 and Cov B (Application 6) 3.2.3.3 Applications 7-9: 4 wt% Initial 235U Enrichment and ENDF/B-VIII.0 Nuclear Data Library and Covariance Library A similar analysis was performed for the six cases using the ENDF/B-VIII.0 nuclear data.

Overall, the pattern of results for each initial enrichment and burnup is similar to the ENDF/B-VII.1 nuclear data, but there is a large increase in the number of similar experiments. The 235U uncertainties are larger in the ENDF/B-VIII.0 covariance data, so the assessed similarity of LEU experiments is generally higher with the newer covariance data. The larger uncertainty places more weight on the 235U similarity, generally increasing the ck values for LEU experiments.

Because the LEU category of experiments is the largest pool of experiments, the number of similar experiments identified for each application tends to increase. The ENDF/B-VIII.0 covariance data also include a significant increase in the 239Pu(n,) reaction uncertainty. In some cases, this will result in an increase of the overall importance of 239Pu, depending on the interplay of the various reaction sensitivity profiles. Changes in the covariance data for 239Pu fission and 238U(n,) also influence the similarity assessments performed with ENDF/B-VIII.0 covariance data [20, 21].

Figure 3-14 provides the similarity results corresponding to the seventh application, with an initial 235U enrichment of 4 wt% and a burnup of 10 GWd/MTU. For this application, there are 927 experiments that are at least marginally similar of which 17 are highly similar from the ICSBEP categories LEU-COMP-THERM, LEU-MET-THERM, LEU-MISC-THERM, and from the HTC cases.

3-21 At this low burnup, the application contains a significant amount of 235U and only a low amount of plutonium, so both the LEU experiments and the HTC experiments are applicable. There is a low number of highly similar experiments because only a few experiments correspond to a low-enriched, low burnup application case. The most similar experiments are HTC4FE and HTC4PB, with ck values between 0.90 and 0.91.

Figure 3-14 Similarity Coefficients of the 4 wt% 235U Enriched 10 GWd/MTU Burnup with E8.0 and Cov A (Application 7)

The similarity results shown in Figure 3-15 correspond to the eighth application, with an initial 235U enrichment of 4 wt%, and a burnup of 40 GWd/MTU. For this application, there are 273 experiments that are at least marginally similar, of which 135 are highly similar from the ICSBEP category MIX-COMP-THERM and from the HTC configurations. The HTC experiments were designed to correspond to a 4 wt% 235U enriched fuel with a 40 GWd/MTU burnup, so the similarity with this application was expected to be very high. At this medium burnup, the application contains less 235U and more plutonium compared to the 10 GWd/MTU case previously shown. The similarity of the LEU experiments decreased, and the similarity of the mixed uranium/plutonium experimentsboth MIX-COMP-THERM and HTCincreased. The most similar experiment is HTC2B-007, with a ck value of 0.98.

3-22 Figure 3-15 Similarity Coefficients of the 4 wt% 235U Enriched 40 GWd/MTU Burnup with E8.0 and Cov A (Application 8)

The similarity results shown in Figure 3-16 correspond to the ninth application, with an initial 235U enrichment of 4 wt% and a burnup of 80 GWd/MTU. For this application, there are 285 experiments that are at least marginally similar, of which 96 are highly similar from ICSBEP categories MIX-COMP-THERM and MIX-SOL-THERM, as well as some HTC configurations. At this high burnup, the application contains less 235U and more plutonium compared to the 40 GWd/MTU case previously shown. The similarity of the LEU experiments drops again, and most of the HTC experiments decrease. The similarity of all the mixed uranium/plutonium experiments increased. For the 4 wt% 235U initial enrichment, 80 GWd/MTU is an extremely high burnup, so the 235U content is dropping into the range of the natural and depleted uranium used in the ICSBEP Handbook MIX experiments. The most similar experiment, however, is still an HTC, with a ck value of 0.95 for HTC2B-008. As expected from the analysis of the 239Pu/(235U

+239Pu) ratios presented in Section 3.2.1, the similarity of the application to the HTC experiments increases from 10 to 40 GWd/MTU, and it decreases somewhat from 40 to 80 GWd/MTU.

3-23 Figure 3-16 Similarity Coefficients of the 4 wt% 235U Enriched 80 GWd/MTU Burnup with E8.0 and Cov A (Application 9) 3.2.3.4 Applications 10-12: 8 wt% Initial 235U Enrichment and ENDF/B-VIII.0 Nuclear Data Library and Covariance Library The results shown in Figure 3-17 correspond to the tenth application, with an initial 235U enrichment of 8 wt% and a burnup of 10 GWd/MTU. For this application, there are 1,205 experiments that are at least marginally similar of which 330 are highly similar from the ICSBEP categories IEU-SOL-THERM, LEU-COMP-THERM, LEU-MET-THERM, LEU-MISC-THERM, and LEU-SOL-THERM. At this relatively high initial enrichment and low burnup, the application still contains a significant amount of 235U and a very low amount of plutonium. The LEU experiments are more applicable than in the 4 wt% 235U and 10 GWd/MTU cases, and the HTC experiments are not applicable. The most similar experiment is LEU-COMP-THERM-015-112, which has a ck value of 0.93.

3-24 Figure 3-17 Similarity Coefficients of the 8 wt% 235U Enriched 10 GWd/MTU Burnup with E8.0 and Cov A (Application 10)

The similarity results shown in Figure 3-18 correspond to the eleventh application with an initial 235U enrichment of 8 wt% and a burnup of 40 GWd/MTU. For this application, there are 143 experiments that are at least marginally similar, of which 31 are highly similar, all HTC experiments. At this medium burnup, the application contains less 235U and more plutonium compared to the 10 GWd/MTU case previously shown. The similarity of the LEU experiments decreased, and the similarity of all the mixed uranium/plutonium experiments, including the HTCs, increased. At this point, most of the HTC cases are either marginally similar or highly similar, although they are still slightly less applicable than in the 4 wt% 235U initial enrichment case. The most similar experiments are HTC2B-001 through HTC2B-008, with ck values of 0.92.

3-25 Figure 3-18 Similarity Coefficients of the 8 wt% 235U Enriched 40 GWd/MTU Burnup with E8.0 and Cov A (Application 11)

The similarity results shown in Figure 3-19 correspond to the twelfth application, with an initial 235U enrichment of 8 wt% and a burnup of 80 GWd/MTU. For this application, there are 239 experiments that are at least marginally similar, of which 107 are highly similar from the ICSBEP MIX-COMP-THERM category and the HTC experiments. At this high burnup, the application contains less 235U and more plutonium compared to the 40 GWd/MTU case previously shown.

The similarity of the LEU experiments decreased further, and the similarity of all the mixed uranium/plutonium experiments, including the HTCs, increased further. This case is similar to the 4 wt% 235U enriched and 40 GWd/MTU burnup case discussed in Section 3.2.1 (application 8), with enough uranium depletion and plutonium generation to match the characteristics of the HTC experiments. As expected from the analysis of the 239Pu/(235U +239Pu) ratios described in Section 3.2.1, the similarity of the application to the HTC experiments increases over the whole burnup range considered here. This result is consistent with both nuclear data libraries. The most similar experiments are also HTC2B-001 through HTC2B-008, with ck values between 0.96 and 0.97.

3-26 Figure 3-19 Similarity Coefficients of the 8 wt% 235U Enriched 80 GWd/MTU Burnup with E8.0 and Cov A (Application 12) 3.2.4 Influence of Nuclear Data Library and Covariance Data ck 3.2.4.1 Covariance Data Library Analysis The previous sections show that the similarity study results can drastically change according to the different nuclear data and covariance libraries used. These differences are highlighted and explained in this section. The covariance data updates mostly cause the differences between ENDF/B-VII.1 and ENDF/B-VIII.0 similarity assessments. As noted in previous work [20], the update of the covariance data between Cov B (SCALE 6.2 based on ENDF/B-VII.1) and Cov A (ENDF/B-VIII.0 in SCALE 6.3) can have a significant impact on the similarity assessments because of increases in the uncertainty of very important reactions for BUC studies, such as the uncertainty increase in (also written as nubar), the average neutron emission per fission, for 235U and 239Pu, as shown in Figure 3-20 and Figure 3-21, respectively. Because of these changes, the uncertainty in nuclear data shared by application models and critical experiments involving 235U and 239Pu fission is higher, and ck tends to increase, thus providing more similar experiments when using Cov A compared to Cov B. A similar effect is observed with the increase of the 1H elastic scattering uncertainty in Cov A compared to Cov B, as shown in Figure 3-22, thus increasing the contribution of 1H to ck when using Cov A. The increased changes seen in the covariance data for other nuclides, reactions, and specific energy ranges are responsible for the decreased similarity in a few cases when using Cov A compared to Cov B.

3-27 Figure 3-20 Uncertainty in 235U Reaction for Cov A and Cov B [19]

Figure 3-21 Uncertainty in 239Pu Reaction for Cov A and Cov B [20]

3-28 Figure 3-22 Uncertainty in 1H Elastic Scattering Reaction for Cov A and Cov B [20]

Following the same methodology described in previously published work for previous covariance library updates [21], an additional study was performed to confirm the claim that the change in similarity between the results previously shown with ENDF/B-VII.1 and ENDF/B-VIII.0 is primarily caused by the covariance data changes. As previously noted, the GBC-32 application model that corresponds to a 4 wt% 235U initial enrichment and 10 GWd/MTU burnup was selected to be tested with four different parameter combinations, as shown in Table 3-7.

TSUNAMI-IP was then used to calculate the similarity coefficients of those four applications with a small set of 26 experiments from LEU-COMP-THERM-078 (cases 1 to 15) and LEU-COMP-THERM-080 (cases 1 to 11). This approach isolates the impact of changing covariance libraries and provides insight into the most significant influences on the calculated ck values.

Table 3-7 Nuclear Data Library and Covariance Data Combinations Used in the Study The results of the TSUNAMI-IP calculations are given in Table 3-8. Using the ENDF/B-VII.1 nuclear data library with the SCALE 6.2 covariance data gives ck values between 0.49 and 0.53 and using the ENDF/B-VIII.0 nuclear data library with the SCALE 6.2 covariance library results in ck values between 0.52 and 0.56. This result shows that the nuclear data used for the depletion and sensitivity calculations has a small influence on ck in these two application cases, with similarity slightly higher when using the ENDF/B-VIII.0 data. Using the ENDF/B-VII.1 nuclear data library with ENDF/B-VIII.0 covariances and the ENDF/B-VIII.0 nuclear data library Nuclear data parameter Combination 1 Combination 2 Combination 3 Combination 4 Nuclear data library used for depletion, criticality, and sensitivity calculations ENDF/B-VII.1 (E7.1)

ENDF/B-VIII.0 (E8.0)

ENDF/B-VII.1 (E7.1)

ENDF/B-VIII.0 (E8.0)

Covariance data used for TSUNAMI-IP similarity assessment SCALE 6.2 (Cov B)

SCALE 6.2 (Cov B)

ENDF/B-VIII.0 (Cov A)

ENDF/B-VIII.0 (Cov A)

3-29 with ENDF/B-VIII.0 covariances both give ck values of 0.81 for all cases. The differences between the ck values resulting from swapping the source of the sensitivity data file are less than 0.01 in all cases. This proves that the nuclear data used for the depletion, criticality, and sensitivity calculations do not have a significant influence on ck values in that case. The ck results obtained with the SCALE 6.2 covariance library are between 0.49 and 0.56, which is much lower than the ck results obtained with the ENDF/B-VIII.0 covariance data of 0.81. In this example, those two sets of experiments do not appear to be very similar using ENDF/B-VII.1 covariance data, but they are marginally similar when using the ENDF/B-VIII.0 covariance data.

As mentioned in Section 3.2.3, the same behavior is the reason for the large difference in the number of similar experiments when comparing ENDF/B-VII.1 and ENDF/B-VIII.0. Specifically, the increase in 235U uncertainty leads to a large increase in the number of LEU experiments that are marginally similar for low burnup applications with ENDF/B-VIII.0, as shown in Figure 3-23.

Table 3-8 Similarity Coefficients Between 4 wt% 235U Initial Enrichment and 10 GWd/

MTU Burnup Case and 26 Critical Experiments Using Different Nuclear Data Experiment Similarity coefficient ck E7.1 with Cov B E8.0 with Cov B E7.1 with Cov A E8.0 with Cov A LEU-COMP-THERM-078-001 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-002 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-003 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-004 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-005 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-006 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-007 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-008 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-009 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-010 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-011 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-012 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-013 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-014 0.50 0.53 0.81 0.81 LEU-COMP-THERM-078-015 0.49 0.52 0.81 0.81 LEU-COMP-THERM-080-001 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-002 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-003 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-004 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-005 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-006 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-007 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-008 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-009 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-010 0.53 0.56 0.81 0.81 LEU-COMP-THERM-080-011 0.51 0.54 0.81 0.81

3-30 The extended ck output in TSUNAMI-IP describes the contribution of each individual nuclide-reaction combination pair to the calculated ck value. The highest three individual nuclide-reaction ck contributions from the 26 experiments for each of the four cases are shown in Table 3-9. The following observations can be made regarding Table 3-9:

The main ck contributors are the same for the cases using the same covariance data.

The second and third largest ck contributors differ when comparing the cases that use different covariance data.

235U nubar is the main contributor in all cases, increasing from a ck contribution of about 0.26 with the SCALE 6.2 covariance data to about 0.45 for the ENDF/B-VIII.0 covariance data. This is a result of the increased uncertainty in the ENDF/B-VIII.0 covariance for this distribution [20], as discussed previously.

The 1H elastic and (n,) reactions are the top contributors to ck for the ENDF/B-VIII.0 covariance data cases and are negligible for the SCALE 6.2 covariance cases. These differences are also explained by the increase of the reaction uncertainties in the ENDF/B-VIII.0 covariance data [20].

These results indicate that the covariance data update is the primary reason for the similarity assessment differences seen when comparing ENDF/B-VII.1 and ENDF/B-VIII.0, as demonstrated by showing the influence of specific nuclide-reaction uncertainty updates. In the next sections, the ck differences between the nuclear datasets are analyzed in detail for selected cases.

Table 3-9 Main Contributors to the Similarity Coefficient of Four Application Cases Combination and total ck Covariance data Nuclide Reaction Nuclide Reaction E7.1 with Cov B:

Total average - ck = 0.51 235U nubar 235U nubar 238U n,gamma 238U n,gamma 235U n,gamma 235U n,gamma E8.0 with Cov B:

Total average - ck=0.54 235U nubar 235U nubar 238U n,gamma 238U n,gamma 235U n,gamma 235U n,gamma E7.1 with Cov A:

Total average - ck = 0.81 235U nubar 235U nubar 1H elastic 1H elastic 1H n,gamma 1H n,gamma E8.0 with Cov A:

Total average - ck=0.81 235U nubar 235U nubar 1H elastic 1H elastic 1H n,gamma 1H n,gamma

3-31 3.2.4.2 Case-by-Case Differences The experiments with ck 0.8 for either ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 4 wt% 235U initial enrichment and 10 GWd/MTU burnup case are shown below in Figure 3-23. Significantly more experiments are similar with the ENDF/B-VIII.0 nuclear dataset, with 927 experiments that are at least marginally similar of which 17 are highly similar experiments, than with the ENDF/B-VII.1 data, with 234 experiments that are at least marginally similar of which 23 are highly similar experiments. Although the ENDF/B-VII.1 covariance library leads to higher maximum ck values for a few cases from the LEU-COMP-THERM (LCT) and LEU-MET-THERM (LMT) experiments, a larger number of these experiments exceed the 0.8 threshold with the ENDF/B-VIII.0 data. The ck values for the HTC experiments with the ENDF/B-VIII.0 dataset are higher compared to the ENDF/B-VII.1 dataset.

Figure 3-23 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 10 GWd/MTU Burnup with E7.1 and E8.0 (Applications 1 and 7)

The experiments with ck 0.8 for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 8 wt% 235U initial enrichment and 10 GWd/MTU burnup case are shown in Figure 3-24. More experiments are similar with the ENDF/B-VIII.0 nuclear dataset, with 1,205 experiments that are at least marginally similar of which 330 are highly similar experiments, than with the ENDF/B-VII.1 data, with 279 experiments that are at least marginally similar of which 46 are highly similar. Apart from a few LCT experiments, the ck values are higher when using ENDF/B-VIII.0 data compared with the ENDF/B-VII.1.

3-32 Figure 3-24 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 10 GWd/MTU Burnup GBC-32 Application Model with E7.1 and E8.0 (Applications 4 and 10)

The experiments with ck0.8 for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 4 wt% 235U initial enrichment and 40 GWd/MTU burnup case are shown in Figure 3-25. More experiments have ck values above 0.8, and the ck values are higher with the ENDF/B-VIII.0 nuclear dataset compared to those with the ENDF/B-VII.1 nuclear dataset. Using the ENDF/B-VIII.0 nuclear dataset, 273 experiments are at least marginally similar, of which 135 are highly similar. Using the ENDF/B-VII.1 dataset, 152 experiments are at least marginally similar, of which 86 are highly similar.

3-33 Figure 3-25 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 40 GWd/MTU Burnup with E7.1 and E8.0 (Applications 2 and 8)

The experiments with ck 0.8 for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 8 wt% 235U initial enrichment and 40 GWd/MTU burnup case are shown in Figure 3-26. More experiments are similar from different experiment types with the ENDF/B-VII.1 nuclear dataset, with 376 experiments that are at least marginally similar, of which 0 are highly similar than with the ENDF/B-VIII.0 data, with 143 experiments that are at least marginally similar, of which 31 are highly similar. Compared to 10 GWd/MTU at 8 wt% 235U initial enrichment, the similarity of LCT experiments with the application decreases at a faster rate at 40 GWd/MTU when using the ENDF/B-VIII.0 data compared to the ENDF/B-VII.1 data due to the decrease of 235U in the system and the higher influence of 235U covariance data on ck with the ENDF/B-VIII.0 data. This results in a decrease of ck to be slightly below the 0.8 threshold for most of the LCT experiments with the ENDF/B-VIII.0 data, thus excluded from this figure. While only some of the HTC experiments are highly similar with ENDF/B-VIII.0, these values are still higher than those with ENDF/B-VII.1.

3-34 Figure 3-26 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 40 GWd/MTU Burnup with E7.1 and E8.0 (Applications 5 and 11)

The experiments with ck0.8 for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 4 wt% 235U initial enrichment and 80 GWd/MTU burnup case are shown in Figure 3-27. The number of marginally similar experiments is similar between both nuclear datasets, but the number of highly similar experiments is way higher with the ENDF/B-VIII.0 data compared to the ENDF/B-VII.1 data. The number of marginally similar experiments is 208 that are at least marginally similar of which 3 are highly similar for the ENDF/B-VII.1 data and 285 experiments that are at least marginally similar, of which 96 are highly similar with the ENDF/B-VIII.0 data.

The ck values of those highly similar experiments are also higher when using ENDF/B-VIII.0 data.

3-35 Figure 3-27 Experiments with ck 0.8 for the 4 wt% 235U-Enriched 80 GWd/MTU Burnup with E7.1 and E8.0 (Applications 3 and 9)

The experiments with ck0.8 for both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets for the 8 wt% 235U initial enrichment and 80 GWd/MTU burnup case are shown in Figure 3-28. More experiments are similar, including a significant number of MCT experiments, when using the ENDF/B-VIII.0 nuclear dataset. There are 239 experiments that are at least marginally similar, of which 107 are highly similar compared with the ENDF/B-VII.1 data, with 147 experiments that are at least marginally similar, of which 67 are highly similar. The ck values are mostly higher when using ENDF/B-VIII.0 compared to ENDF/B-VII.1.

3-36 Figure 3-28 Experiments with ck 0.8 for the 8 wt% 235U-Enriched 80 GWd/MTU Burnup with E7.1 and E8.0 (Applications 6 and 12)

Additional effects of nuclear data library and covariance data combinations are discussed in Section 4.3.1.

3.2.5 Similarity Assessment Conclusion The similarity study results show that sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235U and burnups up to 80 GWd/MTU. Both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation, because enough marginally similar experiments to the application always exist (143 minimum). As in previous BUC validation studies, the French HTC experiments are the most similar in most application cases studied, especially from representative discharge burnups (40 to 80 GWd/MTU). These results support the conclusions presented in NUREG/CR-7109 [1]

regarding the validation of the primary actinides in PWR BUC analyses.

A few validation gaps exist in which highly similar experiments are not available, such as the 8 wt% 235U case at 40 GWd/MTU burnup using ENDF/B-VII.1 nuclear data, and the 5 wt% 235U case at 10 GWd/MTU burnup using ENDF/B-VIII.0 nuclear data, but there are enough marginally similar experiments available to proceed with validation for all initial enrichments and burnup combinations considered here. It was also shown that the covariance data updates primarily cause the differences in similarity assessment between ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear datasets. Major nuclides uncertainty estimates were increased between ENDF/B-VII.1 and ENDF/B-VIII.0, such as 235U nubar and 1H scatter and capture. These changes increased the shared nuclear data uncertainty between application and critical experiments, thus increasing the similarity coefficient ck.

3-37 While it is impossible to define a generic limit on the minimum number of benchmarks necessary to perform validation, it is clear that the number of experiments identified with ck 0.8 in all cases examined here is sufficient for validation. There are likely initial enrichment and burnup combinations with fewer clearly applicable experiments, so additional studies and appropriate justification may be needed to investigate and account for these combinations.

There are also cases in which all the applicable experiments identified originate from the same set of experiments, HTCs as an example. This is potentially problematic as it increases the risk of a systematic bias in the single set of benchmarks remaining undetected.

4 BIAS AND BIAS UNCERTAINTY ANALYSES 4.1 Bias and Bias Uncertainty Estimates One of the main steps in implementing the BUC methodology is determining the bias and bias uncertainty to use for the validation of application models using specific LCEs as a function of different trend parameters. The bias and bias uncertainty estimates are unique for each application, calculational method, and nuclear data combination. The addition of bias and bias uncertainty corresponds to the calculational margin (+ ). Before calculating the bias and bias uncertainty, keff and the applicable critical benchmark experiments for each application case must be determined.

This section presents the keff, bias, and bias uncertainty results of select GBC-32 application cases involving different initial enrichments and burnups, along with their corresponding applicable critical benchmarks. Table 4-1 summarizes the cases considered in this analysis.

Table 4-1 Application Cases in Bias and Bias Uncertainty Analysis Parameter Value Enrichment (wt% 235U) 4, 8 Burnup (GWd/MTU) 10, 40, 75, 80 Nuclear data library ENDF/B-VII.1 (E7.1), ENDF/B-VIII.0 (E8.0)

Nuclide set AFP Figure 4-1 shows the keff results for the application cases in Table 4-1. The stochastic uncertainty in each of these values was approximately 10 pcm. As expected, the keff is higher for the 8 wt% 235U fuel and decreases with increasing burnup. Utilizing different cross-section libraries did not significantly affect the keff results. The percentage difference in keff between the two nuclear data libraries for the 4 wt% 235U cases ranged from 0.2 to 0.7, whereas the percent difference in keff for the 8 wt% 235U cases ranged from 0.2 to 0.3.

Figure 4-1 keff of Application Cases in Bias and Bias Uncertainty Analysis 4-1

4-2 As discussed in Section 3.1, a ck threshold of 0.8 was used to select applicable critical experiments for use in the validation studies documented here. Table 4-2 shows the number of applicable experiments for each application case presented in Table 4-1 and further identifies how many of the experiments are found in the VALID library or in the HTC data set.

Only the benchmarks from the VALID library and the HTC set are used in these validation studies because there are available calculated keff values for these models using SCALE 6.3.1 and ENDF/B-VII.1 and ENDF/B-VIII.0 data. Benchmarks with SDFs generated and distributed with DICE were not calculated using the same computational method and, therefore, are not used.

Table 4-2 Number of Applicable Critical Experiments in Bias and Bias Uncertainty Analysis 4 wt% Enrichment 235U 8 wt% enrichment 235U Burnup (GWd/MTU) 10 40 75 80 10 40 75 80 Library Criteria E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 ck 0.8 (from section

3) 234 928 152 273 216 284 208 285 280 1,205 376 143 147 236 147 239 ck 0.8 (VALID) 36 90 0

17 12 17 10 17 45 169 46 0

0 11 0

12 ck 0.8 (HTC) 68 147 151 152 149 151 143 151 0

0 143 143 147 149 147 150 Experiments used in the and determination 104 237 151 169 161 168 153 168 45 169 189 143 147 160 147 162 The bias and bias uncertainty were calculated by using the different techniques available in the VADER tool in SCALE 6.3.2. The main techniques can be grouped into three categories:

trending analysis using a specific experimental parameter with the single-sided Lower Tolerance Limit method from NUREG/CR-6698 [22], nontrending parametric analysis, and nontrending nonparametric analysis. To select the best methodology to use for each case, it is recommended to start by using a trend on physical parameters such as EALF, ck, or enrichment because the derived bias and bias uncertainty would correspond to the system analyzed.

Statistical methods must test the significance of the trend, and if no trend exists, it is recommended to test the normality of the data. If the data is normal, nontrending parametric analysis is recommended. If the data is not normal, it is recommended to use nontrending nonparametric analysis. The process for calculating the bias and bias uncertainty is detailed in Reference 6. In the validation study, the experimental uncertainties from the critical experiment evaluations are extremely variable, so the uncertainty-weighted approach wont be used, and all bias and bias uncertainty calculations will be performed with the unweighted approach [6].

The trending analysis is performed by calculating a linear fit of the critical experiment results as a function of the trend parameters [22], incorporating unweighted calculated-to-expected (C/E) ratios based on the expected benchmark model keff values. The uncertainty in the C/E ratio propagates both the uncertainty in the benchmark model keff and the stochastic uncertainty in the CSAS calculation. The ck and the EALF parameters were used in the trending analysis. If

4-3 the trending is significant for the parameters, two methods can be used to determine the calculational margin. The first method uses the bias and bias uncertainty values from the trending parameter with the highest t-test value. The second one is to use the bias and bias uncertainty values, resulting in the most conservative estimates that are equivalent to the highest values. The second method will be used in this validation study.

In the following tables, the trend significance is evaluated by performing a t-test on the data, and the normality of the data is evaluated by performing the Anderson-Darling Normality Test. The numeric results of the tests will not be explicitly detailed, but the correct analysis methodology for determining the bias and bias uncertainty will be shown for each case between ck Trending using LTL from NUREG/CR-6698 [22] (ck T), EALF Trending using LTL from NUREG/CR-6698 (EALF T), Nontrending Parametric (NT P), and Nontrending Nonparametric (NT NP).

The ck and EALF trending plots for a few application cases using the ENDF/B-VII.1 nuclear data library are shown in Figure 4-2 through Figure 4-9. If the trend is significant, the trending equation is shown on the plot. Otherwise, only the data points are shown.

Figure 4-2 Significant ck Trending for 4 wt% 235U at 40 GWd/MTU Using ENDF/

B-VII.1 Nuclear Data Library

4-4 Figure 4-3 Nonsignificant EALF Trending for 4 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library Figure 4-4 Nonsignificant ck Trending for 4 wt% 235U at 75 GWd/MTU Using ENDF/

B-VII.1 Nuclear Data Library

4-5 Figure 4-5 Nonsignificant EALF Trending for 4 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library Figure 4-6 Significant ck Trending for 8 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1Nuclear Data Library

4-6 Figure 4-7 Nonsignificant EALF Trending for 8 wt% 235U at 40 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library Figure 4-8 Significant ck Trending for 8 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library

4-7 Figure 4-9 Nonsignificant EALF Trending for 8 wt% 235U at 75 GWd/MTU Using ENDF/B-VII.1 Nuclear Data Library The bias and bias uncertainties for the selected application cases generated using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries are shown in Table 4-3 for the results using the ck and EALF trending methods, and in Table 4-4 for the results using the nontrending parametric and nontrending nonparametric method. To determine which set of results from the four methods should be selected for the calculational margin of each case, the process described previously was used. In the 16 cases shown, the chosen methods are displayed in bold in Table 4-3 and Table 4-4. The bias, bias uncertainty, and calculational margin results variation between the four techniques and different parameters are mainly due to the distribution of keff C/E values for the chosen experiments, as seen in Figure 4-2 through Figure 4-9. Several observations are made based on the information presented in Table 4-3 and Table 4-4:

Out of the 16 cases, 12 have a significant trend. Among the 12 significant trends, 7 are ck and 5 are EALF. Among the 4 nontrending results, one is parametric, and the three others are nonparametric. At 4 wt% 235U, the two nonsignificant are obtained at the high burnup points, 75 and 80 GWd/MTU, both with ENDF/B-VII.1 library. At 8 wt% 235U, the two nonsignificant are obtained at the low burnup point of 10 GWd/MTU with ENDF/B-VII.1 library, and at the medium burnup point of 40 GWd/MTU with ENDF/B-VIII.0 library.

There is no apparent correlation to explain the nonsignificance.

The trend is significant for both ck and EALF in 3 out of the 16 cases. Those cases are 4 wt% 235U and 10 GWd/MTU with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries, and 8 wt% 235U 10 GWd/MTU with ENDF/B-VIII.0 library.

Out of the 16 cases, the data normality test passed only once, for the 8 wt% 235U 10 GWd/MTU with ENDF/B-VII.1 library, even though most of the cases have a significant trend.

In most cases, the ck trending appears to be significant at lower burnup for both libraries and enrichments.

4-8 All the cases where the EALF trending appears significant are with the ENDF/B-VIII.0 library.

In most cases, the calculational margin calculated using EALF trending is lower than the calculational margin calculated using ck trending. The difference ranges from 27 to 403 pcm, with an average of 212 pcm. This is mainly due to the lower bias using EALF trending compared to ck trending.

The calculational margin calculated using EALF and ck trending is different between using ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The difference ranges from 98 to 663 pcm for the ck trending results, with an average of 281 pcm. There is less variability for the EALF trending results, with 4 to 289 pcm, and an average of 88 pcm.

In most cases, the calculational margin calculated by nontrending parametric and nonparametric are similar, with an average difference of around 150 pcm.

In all the cases, the calculational margin calculated by nontrending techniques is lower than by trending techniques. The biases are similar between the four techniques, but the bias uncertainty is about 200 pcm lower in the nontrending techniques.

With all four techniques, the 8 wt% 235U 10 GWd/MTU with ENDF/B-VIII.0 case has a low bias and a high bias uncertainty. This is due to a particularly high distribution of high and low keff C/E values for the experiments used in this case.

There is no apparent correlation between calculational margin results and increasing enrichment or increasing burnup, for any nuclear data library or any method used.

From the chosen results using the appropriate technique for each case, the calculated bias is between 0 and -332 pcm, with an average of -173 pcm, and the calculated bias uncertainty is between 498 and 1223 pcm, with an average of 904 pcm. The corresponding calculational margin is between 625 and 1477 pcm, with an average of 1079 pcm.

4-9 Table 4-3 Bias and Bias Uncertainty Results for all Application Cases Obtained with the ck and EALF Trending Methods Method ck trending EALF trending Enrichment wt% 235U Burnup (GWd/MTU)

Library Trend significance

+

Trend significance

+

4 10 E7.1 Yes 0*

0.01005 0.01005 Yes

-0.00103 0.00758 0.00861 E8.0 Yes

-0.00298 0.00842 0.01140 Yes

-0.00133 0.00685 0.00818 40 E7.1 Yes

-0.00292 0.00958 0.01250 No

-0.00179 0.00885 0.01064 E8.0 No

-0.00176 0.00898 0.01074 Yes

-0.00102 0.00905 0.01007 75 E7.1 No

-0.00238 0.01064 0.01302 No

-0.00120 0.00914 0.01034 E8.0 No

-0.00216 0.00985 0.01201 Yes

-0.00083 0.00932 0.01015 80 E7.1 No

-0.00296 0.01120 0.01416 No

-0.00089 0.00924 0.01013 E8.0 No

-0.00195 0.00986 0.01181 Yes

-0.00080 0.00937 0.01017 8

10 E7.1 No

-0.00063 0.00751 0.00814 No

-0.00106 0.00735 0.00841 E8.0 Yes

-0.00254 0.01223 0.01477 Yes

-0.00144 0.00986 0.01130 40 E7.1 Yes

-0.00332 0.00980 0.01312 No

-0.00129 0.00829 0.00958 E8.0 No

-0.00309 0.01101 0.01410 No

-0.00096 0.01028 0.01124 75 E7.1 Yes

-0.00296 0.01078 0.01374 No

-0.00168 0.00980 0.01148 E8.0 No

-0.00183 0.00933 0.01116 Yes

-0.00074 0.00970 0.01044 80 E7.1 Yes

-0.00293 0.01070 0.01363 No

-0.00168 0.00979 0.01147 E8.0 No

-0.00195 0.00930 0.01125 Yes

-0.00067 0.00971 0.01038

• Positive bias was not credited (converted to 0).

4-10 Table 4-4 Bias and Bias Uncertainty Results for all Application Cases Obtained with the Nontrending Parametric and Nontrending Nonparametric Methods Method Nontrending parametric Nontrending nonparametric Enrichment wt% 235U Burnup (GWd/MTU)

Library Normality

+

+

4 10 E7.1 No

-0.00182 0.00643 0.00825

-0.00182 0.00576 0.00758 E8.0 No

-0.00153 0.00625 0.00778

-0.00153 0.00592 0.00745 40 E7.1 No

-0.00140 0.00709 0.00849

-0.00140 0.00610 0.00750 E8.0 No

-0.00187 0.00751 0.00938

-0.00187 0.00651 0.00838 75 E7.1 No

-0.00144 0.00730 0.00874

-0.00144 0.00841 0.00985 E8.0 No

-0.00186 0.00751 0.00937

-0.00186 0.00651 0.00837 80 E7.1 No

-0.00143 0.00718 0.00861

-0.00143 0.00624 0.00767 E8.0 No

-0.00186 0.00751 0.00937

-0.00186 0.00651 0.00837 8

10 E7.1 Yes

-0.00127 0.00498 0.00625

-0.00127 0.00570 0.00697 E8.0 No

-0.00057 0.00852 0.00909

-0.00057 0.01806 0.01863 40 E7.1 No

-0.00141 0.00640 0.00781

-0.00141 0.00560 0.00701 E8.0 No

-0.00176 0.00734 0.00910

-0.00176 0.00631 0.00807 75 E7.1 No

-0.00145 0.00711 0.00856

-0.00145 0.00605 0.00750 E8.0 No

-0.00183 0.00742 0.00925

-0.00183 0.00654 0.00837 80 E7.1 No

-0.00145 0.00711 0.00856

-0.00145 0.00605 0.00750 E8.0 No

-0.00181 0.00744 0.00925

-0.00181 0.00656 0.00837 To better understand the effect of the HTC experiments, the bias and bias uncertainty were calculated with and without the HTC experiments.

Table 4-5 shows the results of this comparison using the EALF trending parameter, where L is the nuclear data library, E is enrichment in wt% 235U, and BU is the burnup in GWd/MTU. The results without the HTC experiments are compared to the nominal results for each case, previously identified in bold in Table 4-3 and Table 4-4. The method used to determine the calculational margin is displayed as ck Trending using LTL from NUREG/CR-6698 [22] (ck T),

EALF Trending using LTL from NUREG/CR-6698 (EALF T), Nontrending Parametric (NT P),

and Nontrending Nonparametric (NT NP). In four cases, there were no experiments in VALID with ck 0.8, so bias determination was not possible. In two other cases, there was no HTC experiment with ck 0.8, so the calculational margin did not update. Of the 10 cases that could be analyzed using this methodology, 7 have their calculational margin increase without using the HTC experiments, and 3 have their calculational margin decrease. The calculational margin decreases are at 4 wt% 235U low burnup and 8 wt% 235U medium burnup, corresponding to the

4-11 applications where the HTC experiments are the least applicable. The calculational margin increases are explained by the removal of most of the experiments from the analysis. For example, when not using the HTC experiments, the 4 wt % enriched case at 75 GWd/MTU goes from 149 to 12 marginally applicable experiments. This change causes an increase of 5111 pcm when not using the HTC experiments. The use of a lower number of experiments increases the calculational margin. In most cases, the bias does not significantly change between using the HTC experiments or not, and most of the calculational margin differences are due to the bias uncertainty changes. The results show that the effect of the HTC experiments is significant and more observable at higher burnup because of the limited availability of applicable critical benchmarks (Table 4-2).

Table 4-5 Bias and Bias Uncertainty with and Without HTC Experiments wt% 235U

• Positive bias was not credited (converted to 0).

Additional cases were studied to show the effect of the ck cutoff methodology on the bias and bias uncertainty determination. The additional cases used the ENDF/B-VII.1 nuclear data libraries with trending on EALF. Trends were performed on the natural logarithm of the EALF to With HTC Without HTC Library Enrichment Burnup (GWd/MTU)

+

Analysis Method

+

Analysis Method (B+B)

(pcm)

ENDF/B-VII.1 4

10 0*

0.01005 0.01005 ck T

-0.00126 0.00478 0.00604 NT P

-401 40

-0.00292 0.00958 0.01250 ck T N/A: only HTC experiments 75

-0.00144 0.00841 0.00985 NT NP -0.00167 0.05818 0.05985 NT NP 5000 80

-0.00143 0.00624 0.00767 NT NP -0.00135 0.05743 0.05878 NT NP 5111 8

10

-0.00127 0.00498 0.00625 NT P

-0.00127 0.00498 0.00625 NT P 0

40

-0.00332 0.00980 0.01312 ck T

-0.00131 0.00502 0.00633 NT P

-679 75

-0.00296 0.01078 0.01374 ck T N/A: only HTC experiments 80

-0.00293 0.01070 0.01363 ck T N/A: only HTC experiments ENDF/B-VIII.0 4

10

-0.00298 0.00842 0.01140 ck T

-0.00116 0.00560 0.00676 NT NP -464 40

-0.00102 0.00905 0.01007 EALF T -0.00118 0.01614 0.01732 EALF T 725 75

-0.00083 0.00932 0.01015 EALF T -0.00079 0.01659 0.01738 EALF T 723 80

-0.00080 0.00937 0.01017 EALF T -0.00073 0.01667 0.01740 EALF T 723 8

10

-0.00254 0.01223 0.01477 ck T

-0.00254 0.01223 0.01477 ck T 0

40

-0.00176 0.00631 0.00807 NT NP N/A: only HTC experiments 75

-0.00074 0.00970 0.01044 EALF T -0.00052 0.01981 0.02033 EALF T 989 80

-0.00067 0.00971 0.01038 EALF T -0.00014 0.01969 0.01983 EALF T 945

4-12 allow linear trending, given the logarithmic behavior of neutron moderation. Two sets of critical experiments were considered: (1) 271 experiments in VALID from the categories of interest (all experiments from the IEU-SOL-THERM [IST], LCT, LEU-SOL-THERM [LST], LMT, MCT, and MIX-SOL-THERM [MST], categories), and 155 HTC experiments for a total of 426 experiments, and (2) only experiments from VALID and HTC with ck 0.8. Table 4-6 shows the bias and bias uncertainty for these additional cases. For the four cases evaluated, the use of all experiments allows for the EALF trending technique as the trend is significant. This is due to the higher number of experiments used than when using the ck cutoff. The bias values are usually lower, and the bias uncertainties are higher when using all experiments compared to when using the ck cutoff. Overall, for those four cases, the calculational margins are relatively similar when using the ck cutoff or not, with a difference between 44 and 242 pcm. Using more experiments helps reduce the bias, but at the cost of more bias uncertainty due to more variability of the datapoints. This analysis was only performed on four enrichment/burnup/library combinations, so no conclusion should be extended to all cases.

The EALF trending plot for the 4 wt% 235U 40 GWd/MTU with all experiments using ENDF/B-VII.1 nuclear data library is shown in Figure 4-10. The EALF trending plot for the same case using the ck cutoff methodology is shown in Figure 4-2 It is clear from the comparison of both figures that the slope and intercept of the trend line are affected by the set of critical experiments used, which can lead to incorrect results in the calculated bias and bias uncertainty.

Table 4-6 Effect of Critical Experiments Selection on Bias and Bias Uncertainty Using EALF Trending and ENDF/B-VII-1 Enrichment (wt% 235U)

Burnup (GWd/MTU)

Experiments used

+

Analysis method 4

40 151 (ck 0.8)

-0.00292 0.00958 0.01250 ck T 426 (All)

-0.00164 0.01041 0.01205 EALF T 75 161 (ck 0.8)

-0.00144 0.00841 0.00985 NT NP 426 (All)

-0.00176 0.01051 0.01227 EALF T 8

40 189 (ck 0.8)

-0.00332 0.00980 0.01312 ck T 426 (All)

-0.00198 0.01070 0.01268 EALF T 75 147 (ck 0.8)

-0.00296 0.01078 0.01374 ck T 426 (All)

-0.00194 0.01067 0.01261 EALF T

4-13 Figure 4-10 Significant EALF Trending with 4 wt% 235U 40 GWd/MTU with All Experiments Using ENDF/B-VII.1 Nuclear Data Library 4.2 Impacts on Bias and Bias Uncertainty Estimates 4.2.1 Cross-Section and Covariance Data As mentioned in Section 2.1.3, the SCALE TSUNAMI-3D sequence is used to determine the sensitivity of the calculated keff values in the application cases to each constituent nuclear data component used in the calculation. The covariance data libraries within SCALE and sensitivities for each application case from the SDFs can also be used in TSUNAMI-IP to calculate the uncertainty in keff due to the uncertainty in the nuclear data. The uncertainties provided are based on the model consideration of the major and minor actinides and 16 of the primary fission products, as described in Section 2.1.1. The nuclear data-induced uncertainty of keff is determined by the combined relative uncertainty of reactions in the model obtained from TSUNAMI-IP by subtracting the sum of the squares of negative values from the sum of the square of positive values. Negative values are used to indicate correlations in the uncertainty data that exist between reactions for a nuclide or between different nuclides. Neglecting these correlations would result in an overestimate of the data-induced uncertainty. The uncertainty of each group of nuclides (e.g., major actinides (AC)) is added in quadrature to obtain nuclear data-induced uncertainty in keff resulting from each group. Table 4-7 through Table 4-11 present the nuclear data-induced uncertainty in keff (in pcm) for the 4, 5, 6, 7, and 8 wt% 235U application cases using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries and 56 neutron energy group covariance data. It is worth noting that the nuclear data-induced uncertainty from all nuclides includes some not listed in the tables (see Section 4.3.1). Table 4-12 through Table 4-16 show the percent contribution in the nuclear data-induced uncertainty from the group (or nuclide) for the 4, 5, 6, 7, and 8 wt% 235U application cases. The results of two application cases are presented using 5 and 7 wt% 235U and the ENDF/B-VII.1 nuclear data libraries. Using the ENDF/B-VII.1 nuclear data libraries, Figure 4-11 shows the percent contribution from major actinides, minor actinides, and fission products, and Figure 4-12 shows the percent contribution from 235U, 238U, and 239Pu.

4-14 A few observations can be made from the tables and figures:

1. The results using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries differ, where ENDF/B-VII.1 consistently results in lower total nuclear data-induced uncertainty.

2. The major actinides are the main contributors to the difference in the nuclear data-induced uncertainty.

3. The contribution of fission products to the nuclear data-induced uncertainty is very small and increases with increasing burnup and enrichment.

4. With increasing burnup, the nuclear data-induced uncertainty from 235U and 238U decreases, whereas 239Pu increases.

The results show that the validation of the major actinides continues to be essential. In addition, comparing the bias values in Table 4-3 and Table 4-4 to the total nuclear data-induced uncertainties in Table 4-7 and Table 4-11 shows that the biases are less significant than the nuclear data-induced uncertainties.

Table 4-7 Nuclear Data-Induced Uncertainty (pcm) in k eff for 4 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 All nuc 456.0 569.6 392.6 552.5 380.6 544.6 378.4 541.6 378.5 541.4 378.2 540.8 Maj AC 434.0 506.8 365.4 491.9 352.5 487.2 350.5 486.6 350.8 487.5 350.7 487.7 U-234 0.0 0.0 0.1 0.1 0.1 0.1 0.2 0.2 0.2 0.2 0.2 0.2 U-235 339.9 379.2 202.0 232.7 149.2 173.2 127.0 148.7 116.4 137.1 107.2 126.9 U-238 226.6 163.5 201.3 146.2 188.2 138.1 181.7 134.8 179.0 133.3 175.5 131.8 Pu-238 0.7 0.7 8.9 8.8 17.9 18.1 23.6 23.8 27.0 27.0 29.9 30.1 Pu-239 145.4 291.9 215.4 400.4 235.4 420.5 243.8 426.8 247.9 430.2 251.2 432.4 Pu-240 8.2 30.1 20.8 31.7 25.2 32.6 26.9 33.2 27.7 33.5 28.3 33.8 Pu-241 11.4 11.0 55.5 54.3 79.1 78.6 90.4 89.6 96.1 95.3 101.0 99.9 Pu-242 1.5 1.4 16.4 15.1 25.5 23.8 29.6 27.5 31.8 29.4 33.4 30.9 Am-241 10.3 9.9 44.0 43.0 57.7 57.0 63.1 62.1 65.6 64.5 67.4 66.4 Min AC 12.0 12.0 21.2 21.1 25.3 25.4 27.2 27.3 28.3 28.2 29.1 29.0 Am-243 0.0 0.0 1.8 1.7 3.9 3.8 5.2 5.0 5.9 5.7 6.6 6.4 Np-237 4.1 4.1 13.7 13.6 18.5 18.6 20.6 20.7 21.8 21.7 22.6 22.6 U-236 11.2 11.3 16.1 16.1 16.8 16.9 17.0 17.0 17.1 17.1 17.1 17.1 FP 34.8 34.7 54.7 54.6 63.3 63.5 67.1 67.3 69.3 69.2 70.8 70.8

4-15 Table 4-7 Nuclear Data-Induced Uncertainty (pcm) in keff for 4 wt% 235U Cases (Continued)

Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Mo-95 2.2 2.2 4.2 4.2 4.8 4.8 5.1 5.0 5.2 5.2 5.3 5.3 Tc-99 4.3 4.3 8.6 8.5 10.6 10.6 11.3 11.3 11.8 11.6 12.1 12.0 Ru-101 3.3 3.3 7.3 7.3 8.9 8.9 9.7 9.7 10.1 10.1 10.5 10.4 Rh-103 10.4 10.3 19.0 18.8 21.0 20.9 21.7 21.6 22.1 21.9 22.3 22.1 Ag-109 0.7 0.7 2.4 2.4 3.3 3.3 3.7 3.7 4.0 3.9 4.2 4.1 Cs-133 8.1 8.1 15.2 15.1 17.4 17.3 18.2 18.1 18.7 18.5 19.0 18.8 Sm-147 2.8 2.8 4.9 5.0 5.3 5.4 5.4 5.5 5.5 5.5 5.5 5.5 Sm-149 22.3 22.2 19.1 18.9 17.6 17.5 17.0 16.9 16.8 16.7 16.5 16.4 Sm-150 1.7 1.7 4.6 4.5 5.9 5.9 6.5 6.6 6.9 6.9 7.2 7.2 Sm-151 9.7 9.6 11.8 11.7 12.5 12.5 12.9 12.8 13.1 13.0 13.2 13.1 Sm-152 3.1 3.0 5.8 5.7 6.5 6.4 6.7 6.6 6.8 6.7 6.9 6.8 Nd-143 17.5 17.5 33.9 33.8 39.0 39.1 40.8 40.8 41.8 41.6 42.3 42.2 Nd-145 7.8 8.2 15.3 15.8 17.8 18.4 18.8 19.4 19.4 20.0 19.8 20.3 Eu-151 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 Eu-153 2.5 2.4 8.4 8.3 11.3 11.3 12.6 12.5 13.3 13.2 13.8 13.8 Gd-155 3.8 3.8 14.7 14.7 22.3 22.7 26.2 26.6 28.4 28.6 30.0 30.4

4-16 Table 4-8 Nuclear Data-Induced Uncertainty (pcm) in keff for 5 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 All nuc 473.3 579.3 401.0 559.8 384.7 550.1 379.8 545.8 378.0 544.0 376.2 542.1 Maj AC 452.8 519.0 374.2 497.8 355.7 489.2 350.2 486.5 348.2 485.0 346.4 484.1 U-234 0.0 0.0 0.1 0.1 0.1 0.1 0.2 0.2 0.2 0.2 0.2 0.2 U-235 377.0 419.2 242.5 276.9 189.6 218.9 167.1 193.6 156.5 182.4 146.7 171.5 U-238 217.8 158.2 197.9 143.6 188.8 136.8 183.8 133.7 181.3 132.3 177.8 131.0 Pu-238 0.6 0.5 7.7 7.6 16.0 15.7 21.0 20.9 23.8 23.6 26.7 26.4 Pu-239 123.8 260.1 194.9 382.1 215.6 405.2 223.7 412.8 227.4 415.2 231.0 417.8 Pu-240 6.7 29.3 18.6 31.5 23.1 31.9 24.8 32.3 25.6 32.5 26.3 32.7 Pu-241 7.8 7.4 44.3 43.1 64.9 63.5 74.6 73.6 79.3 77.9 83.8 82.4 Pu-242 1.0 0.9 14.0 13.0 22.8 21.0 26.7 24.6 28.6 26.2 30.2 27.7 Am-241 7.7 7.4 39.3 38.2 53.9 52.5 59.5 58.3 62.1 60.6 64.3 62.9 Min AC 13.1 13.0 23.3 23.2 27.8 27.6 29.8 29.7 30.9 30.6 31.7 31.5 Am-243 0.0 0.0 1.4 1.3 3.2 3.0 4.2 4.1 4.8 4.6 5.4 5.1 Np-237 3.9 3.8 14.2 14.0 19.5 19.3 21.9 21.7 23.0 22.7 24.0 23.8 U-236 12.5 12.5 18.4 18.4 19.6 19.5 19.9 19.9 20.0 20.0 20.1 20.1 FP 35.1 34.9 56.3 56.2 65.3 64.9 68.9 68.8 70.7 70.3 72.2 71.9 Mo-95 2.4 2.4 4.6 4.6 5.2 5.2 5.5 5.4 5.6 5.6 5.7 5.7 Tc-99 4.3 4.2 8.7 8.6 10.8 10.6 11.5 11.3 11.8 11.6 12.1 11.9 Ru-101 3.5 3.5 7.9 7.8 9.6 9.5 10.4 10.3 10.7 10.6 11.0 10.9 Rh-103 10.6 10.5 20.1 19.9 22.4 22.1 23.1 22.8 23.4 23.1 23.6 23.3 Ag-109 0.6 0.6 2.3 2.2 3.1 3.0 3.5 3.4 3.7 3.6 3.9 3.7 Cs-133 8.5 8.4 16.4 16.2 18.8 18.4 19.5 19.3 20.0 19.6 20.3 19.9 Sm-147 2.9 3.0 5.3 5.4 5.9 6.0 6.1 6.2 6.1 6.2 6.2 6.2 Sm-149 22.8 22.7 20.1 19.9 18.7 18.5 18.0 17.9 17.8 17.5 17.4 17.2 Sm-150 1.7 1.6 4.5 4.5 5.8 5.8 6.4 6.4 6.7 6.6 7.0 6.9 Sm-151 9.6 9.6 12.1 12.0 12.9 12.8 13.2 13.1 13.4 13.2 13.5 13.4 Sm-152 3.0 3.0 6.0 6.0 6.8 6.6 7.0 6.9 7.1 6.9 7.1 7.0 Nd-143 17.0 16.9 34.7 34.5 40.7 40.4 42.8 42.6 43.8 43.5 44.5 44.3 Nd-145 8.3 8.6 16.7 17.3 19.5 20.0 20.5 21.1 21.0 21.7 21.4 22.0 Eu-151 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.3 0.2 0.3 0.2 Eu-153 2.3 2.3 8.2 8.1 11.1 10.9 12.3 12.2 12.9 12.7 13.4 13.3 Gd-155 3.2 3.2 12.4 12.4 19.4 19.3 22.8 23.0 24.6 24.6 26.2 26.3

4-17 Table 4-9 Nuclear Data-Induced Uncertainty (pcm) in keff for 6 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 All nuc 488.0 587.9 411.1 567.1 390.9 556.4 384.6 551.7 381.8 549.4 379.5 547.4 Maj AC 469.1 530.6 385.1 505.5 361.8 494.5 354.5 490.0 351.3 488.4 348.5 486.9 U-234 0.0 0.0 0.1 0.1 0.2 0.1 0.2 0.2 0.2 0.2 0.2 0.2 U-235 405.6 448.4 275.2 311.7 223.0 255.6 201.5 231.7 191.6 219.9 181.6 208.9 U-238 209.6 153.4 193.8 141.1 186.5 135.1 183.0 132.3 181.1 131.2 178.8 129.9 Pu-238 0.4 0.4 6.7 6.8 14.3 14.1 18.8 18.6 21.1 21.2 23.8 23.9 Pu-239 107.4 236.7 178.9 367.1 199.6 392.7 207.4 400.4 210.7 404.1 214.2 407.2 Pu-240 5.5 28.3 16.8 31.5 21.3 31.5 22.9 31.7 23.6 31.9 24.4 32.1 Pu-241 5.5 5.4 36.4 35.7 54.6 53.4 63.0 61.7 66.8 66.0 70.9 70.1 Pu-242 0.7 0.7 12.2 11.5 20.6 19.0 24.2 22.3 25.8 24.0 27.5 25.6 Am-241 5.9 5.8 35.2 34.6 50.0 48.6 55.6 54.3 58.0 57.2 60.6 59.7 Min AC 13.9 14.1 25.2 25.2 30.1 29.9 32.1 31.9 33.0 33.1 34.0 34.1 Am-243 0.0 0.0 1.1 1.1 2.7 2.6 3.6 3.4 4.0 3.9 4.5 4.4 Np-237 3.7 3.7 14.5 14.5 20.3 20.0 22.7 22.4 23.7 23.7 24.9 24.9 U-236 13.4 13.6 20.5 20.6 22.1 22.0 22.5 22.4 22.6 22.7 22.8 22.8 FP 35.1 35.4 57.9 58.1 67.3 66.9 70.6 70.4 72.0 72.3 73.7 73.9 Mo-95 2.5 2.5 4.9 4.9 5.7 5.6 5.9 5.9 6.0 6.0 6.1 6.1 Tc-99 4.2 4.2 8.8 8.8 10.9 10.7 11.5 11.4 11.8 11.8 12.1 12.1 Ru-101 3.6 3.7 8.4 8.4 10.3 10.2 11.0 10.9 11.3 11.3 11.7 11.7 Rh-103 10.7 10.8 21.1 21.0 23.7 23.4 24.3 24.1 24.5 24.4 24.8 24.7 Ag-109 0.5 0.5 2.1 2.1 3.0 2.9 3.3 3.2 3.5 3.4 3.6 3.6 Cs-133 8.7 8.8 17.4 17.3 20.0 19.7 20.8 20.5 21.1 20.9 21.4 21.3 Sm-147 3.0 3.1 5.8 5.9 6.5 6.6 6.6 6.8 6.7 6.9 6.8 6.9 Sm-149 23.2 23.3 21.0 20.9 19.7 19.5 18.9 18.8 18.6 18.5 18.3 18.2 Sm-150 1.6 1.6 4.4 4.5 5.8 5.7 6.3 6.3 6.6 6.6 6.8 6.8 Sm-151 9.4 9.4 12.3 12.3 13.2 13.0 13.4 13.3 13.6 13.5 13.7 13.7 Sm-152 3.0 3.0 6.3 6.2 7.1 6.9 7.3 7.1 7.3 7.2 7.4 7.3 Nd-143 16.5 16.6 35.3 35.3 41.9 41.6 44.1 43.9 45.0 45.1 46.0 46.0 Nd-145 8.5 9.0 18.0 18.8 21.1 21.7 22.2 22.8 22.6 23.4 23.0 23.8 Eu-151 0.2 0.2 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 Eu-153 2.2 2.2 8.1 8.0 11.0 10.9 12.1 12.0 12.6 12.6 13.2 13.2 Gd-155 2.7 2.7 10.6 10.8 17.0 16.9 20.0 20.0 21.4 21.7 23.0 23.3

4-18 Table 4-10 Nuclear Data-Induced Uncertainty (pcm) in keff for 7 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 All nuc 500.9 595.0 420.7 572.5 398.8 562.2 390.4 557.2 387.9 555.0 384.7 552.7 Maj AC 483.3 540.7 395.8 511.8 370.2 500.3 360.4 495.2 357.2 492.8 353.4 490.8 U-234 0.0 0.0 0.1 0.1 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 U-235 428.3 473.1 302.0 341.1 252.1 286.4 230.5 263.2 220.4 252.4 211.1 241.6 U-238 202.7 149.1 188.8 138.3 184.0 133.1 180.0 130.8 179.4 129.6 177.6 128.5 Pu-238 0.4 0.3 6.0 5.9 12.7 12.7 16.9 16.8 19.3 19.0 21.5 21.4 Pu-239 94.8 213.5 165.9 351.3 185.8 380.8 193.8 389.6 197.6 393.0 200.5 396.5 Pu-240 4.7 26.8 15.3 31.4 19.5 31.2 21.2 31.3 22.1 31.3 22.7 31.4 Pu-241 4.1 4.0 30.7 29.8 46.4 45.8 54.1 53.1 57.9 56.6 61.2 60.3 Pu-242 0.5 0.5 10.6 10.0 18.4 17.2 22.0 20.5 23.9 21.9 25.3 23.5 Am-241 4.6 4.5 31.7 31.0 45.8 45.0 51.8 50.8 54.8 53.4 57.0 56.1 Min AC 14.6 14.6 26.8 26.7 31.8 31.8 34.1 34.1 35.3 35.0 36.1 36.1 Am-243 0.0 0.0 0.9 0.9 2.3 2.2 3.1 3.0 3.5 3.3 3.9 3.8 Np-237 3.5 3.5 14.7 14.6 20.6 20.5 23.2 23.1 24.5 24.3 25.6 25.6 U-236 14.2 14.2 22.4 22.4 24.2 24.2 24.8 24.8 25.1 25.0 25.2 25.3 FP 35.2 35.3 59.4 59.4 68.6 68.7 72.2 72.4 74.2 73.9 75.5 75.5 Mo-95 2.5 2.5 5.2 5.2 6.0 6.0 6.3 6.3 6.4 6.4 6.5 6.5 Tc-99 4.1 4.1 8.9 8.8 10.9 10.8 11.6 11.5 12.0 11.8 12.2 12.1 Ru-101 3.7 3.7 8.9 8.9 10.8 10.8 11.6 11.6 12.0 11.9 12.3 12.3 Rh-103 10.8 10.8 22.0 21.8 24.6 24.5 25.5 25.3 25.9 25.5 26.0 25.8 Ag-109 0.5 0.5 2.0 1.9 2.8 2.7 3.2 3.1 3.3 3.2 3.5 3.4 Cs-133 8.9 8.9 18.3 18.1 21.0 20.8 21.9 21.7 22.3 22.0 22.6 22.4 Sm-147 3.1 3.2 6.1 6.3 6.9 7.1 7.2 7.4 7.3 7.4 7.3 7.5 Sm-149 23.6 23.6 21.8 21.7 20.4 20.3 19.8 19.7 19.5 19.4 19.2 19.1 Sm-150 1.5 1.5 4.4 4.4 5.7 5.7 6.3 6.3 6.6 6.5 6.8 6.8 Sm-151 9.1 9.1 12.3 12.3 13.2 13.2 13.6 13.5 13.8 13.7 13.9 13.8 Sm-152 2.9 2.9 6.4 6.3 7.3 7.2 7.5 7.4 7.6 7.5 7.7 7.6 Nd-143 16.0 16.0 35.7 35.7 42.5 42.6 45.1 45.1 46.4 46.2 47.3 47.2 Nd-145 8.8 9.2 19.2 19.9 22.5 23.2 23.6 24.5 24.3 24.9 24.6 25.4 Eu-151 0.2 0.2 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 Eu-153 2.1 2.1 8.0 7.9 10.8 10.8 12.0 12.0 12.6 12.5 13.1 13.0 Gd-155 2.4 2.4 9.2 9.3 14.8 15.0 17.7 17.8 19.2 19.2 20.5 20.6

4-19 Table 4-11 Nuclear Data-Induced Uncertainty (pcm) in keff for 8 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 All nuc 512.0 601.9 429.9 576.9 407.4 566.4 397.9 562.3 394.3 559.7 391.0 558.1 Maj AC 495.4 548.7 406.0 517.6 379.4 505.0 368.4 499.8 364.0 497.6 359.9 495.3 U-234 0.0 19.2 0.1 0.1 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 U-235 446.6 493.5 323.9 365.9 276.6 313.7 255.0 291.0 245.6 279.8 236.2 269.6 U-238 196.7 139.2 184.2 135.6 181.4 130.9 178.0 128.9 176.7 127.8 175.6 127.0 Pu-238 0.3 0.3 5.4 5.3 11.4 11.3 15.4 15.0 17.4 17.2 19.5 19.4 Pu-239 85.4 192.5 155.4 336.3 174.4 367.3 182.7 377.6 185.9 382.6 189.0 386.2 Pu-240 4.1 24.9 14.1 31.3 18.1 30.9 19.8 30.9 20.5 30.9 21.2 31.0 Pu-241 3.2 2.9 26.4 25.4 40.2 39.3 47.3 45.9 50.4 49.5 53.7 52.7 Pu-242 0.4 5.9 9.4 8.8 16.7 15.4 20.3 18.6 21.8 20.3 23.5 21.7 Am-241 3.7 3.8 28.9 27.8 42.3 41.1 48.5 47.0 51.0 50.0 53.7 52.6 Min AC 15.3 22.6 28.3 28.1 33.5 33.3 36.0 35.6 37.0 36.9 38.1 38.0 Am-243 0.0 15.3 0.8 0.8 2.0 1.9 2.7 2.5 3.1 2.9 3.4 3.3 Np-237 3.4 3.2 14.9 14.6 20.8 20.6 23.7 23.4 24.9 24.8 26.2 26.0 U-236 14.9 16.3 24.0 23.9 26.1 26.1 26.9 26.8 27.2 27.2 27.5 27.5 FP 35.5 39.4 60.9 60.6 70.2 69.9 74.2 73.8 75.7 75.7 77.3 77.3 Mo-95 2.6 2.6 5.5 5.4 6.3 6.3 6.6 6.6 6.8 6.7 6.9 6.8 Tc-99 4.1 4.0 9.0 8.8 11.0 10.8 11.7 11.5 12.0 11.9 12.3 12.2 Ru-101 3.8 3.7 9.3 9.3 11.3 11.3 12.3 12.2 12.6 12.6 13.0 12.9 Rh-103 10.9 10.5 22.8 22.5 25.6 25.3 26.6 26.2 26.9 26.7 27.2 27.0 Ag-109 0.4 0.7 1.9 1.8 2.7 2.6 3.0 2.9 3.2 3.1 3.3 3.2 Cs-133 9.1 8.9 19.2 18.9 21.9 21.6 23.0 22.6 23.4 23.1 23.7 23.5 Sm-147 3.2 3.2 6.5 6.6 7.4 7.5 7.7 7.8 7.8 8.0 7.9 8.1 Sm-149 24.1 23.1 22.5 22.4 21.2 21.0 20.6 20.4 20.3 20.1 20.0 19.9 Sm-150 1.5 1.9 4.5 4.4 5.7 5.7 6.3 6.2 6.5 6.5 6.8 6.7 Sm-151 8.9 8.9 12.3 12.3 13.3 13.2 13.7 13.6 13.8 13.7 14.0 13.9 Sm-152 2.9 2.7 6.6 6.4 7.4 7.3 7.7 7.6 7.8 7.7 7.9 7.8 Nd-143 15.8 15.4 36.3 36.0 43.3 43.0 46.2 45.8 47.2 47.1 48.4 48.3 Nd-145 9.0 9.1 20.3 20.8 23.7 24.4 25.1 25.9 25.6 26.4 26.2 26.9 Eu-151 0.2 14.7 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 Eu-153 2.0 2.0 8.0 7.8 10.8 10.7 12.1 11.9 12.6 12.5 13.1 13.0 Gd-155 2.2 12.2 8.2 8.1 13.1 13.1 15.9 15.7 17.1 17.1 18.4 18.5

4-20 Table 4-12 Percent Contribution in Nuclear Data-Induced Uncertainty for 4 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Maj AC 90.58 79.16 86.65 79.26 85.79 80.01 85.82 80.71 85.90 81.07 85.95 81.31 Min AC 0.07 0.04 0.29 0.15 0.44 0.22 0.52 0.25 0.56 0.27 0.59 0.29 FP 0.58 0.37 1.94 0.98 2.76 1.36 3.14 1.54 3.35 1.63 3.50 1.71 U-235 55.58 44.32 26.48 17.75 15.36 10.12 11.27 7.54 9.45 6.42 8.03 5.50 U-238 24.69 8.23 26.29 7.00 24.46 6.43 23.06 6.19 22.35 6.06 21.53 5.94 Pu-239 10.17 26.25 30.11 52.52 38.24 59.62 41.50 62.09 42.90 63.15 44.11 63.92 Other 8.77 20.43 11.11 19.61 11.01 18.41 10.52 17.50 10.19 17.02 9.95 16.68 Table 4-13 Percent Contribution in Nuclear Data-Induced Uncertainty for 5 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Maj AC 91.56 80.27 87.10 79.08 85.49 79.08 84.99 79.44 84.88 79.49 84.78 79.74 Min AC 0.08 0.05 0.34 0.17 0.52 0.25 0.62 0.30 0.67 0.32 0.71 0.34 FP 0.55 0.36 1.97 1.01 2.88 1.39 3.29 1.59 3.50 1.67 3.68 1.76 U-235 63.47 52.36 36.57 24.47 24.29 15.83 19.35 12.58 17.15 11.24 15.21 10.00 U-238 21.18 7.46 24.35 6.58 24.10 6.19 23.41 6.00 23.00 5.92 22.35 5.84 Pu-239 6.84 20.16 23.63 46.58 31.41 54.26 34.69 57.20 36.20 58.26 37.71 59.38 Other 7.81 19.32 10.59 19.74 11.10 19.28 11.10 18.68 10.95 18.52 10.82 18.17

4-21 Table 4-14 Percent Contribution in Nuclear Data-Induced Uncertainty for 6 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Maj AC 92.40 81.45 87.77 79.44 85.65 78.99 84.94 78.91 84.64 79.02 84.33 79.13 Min AC 0.08 0.06 0.37 0.20 0.59 0.29 0.70 0.33 0.75 0.36 0.80 0.39 FP 0.52 0.36 1.99 1.05 2.96 1.45 3.37 1.63 3.56 1.73 3.77 1.82 U-235 69.07 58.18 44.81 30.20 32.54 21.10 27.46 17.64 25.19 16.02 22.91 14.56 U-238 18.45 6.81 22.22 6.19 22.75 5.90 22.64 5.75 22.49 5.71 22.19 5.64 Pu-239 4.84 16.21 18.94 41.91 26.06 49.81 29.08 52.69 30.44 54.09 31.86 55.35 Other 7.00 18.13 9.87 19.32 10.80 19.28 10.99 19.13 11.06 18.89 11.10 18.66 Table 4-15 Percent Contribution in Nuclear Data-Induced Uncertainty for 7 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Maj AC 93.08 82.58 88.52 79.90 86.17 79.20 85.22 78.97 84.82 78.83 84.40 78.83 Min AC 0.08 0.06 0.40 0.22 0.64 0.32 0.76 0.37 0.83 0.40 0.88 0.43 FP 0.49 0.35 1.99 1.08 2.96 1.49 3.42 1.69 3.66 1.77 3.85 1.87 U-235 73.11 63.21 51.52 35.50 39.95 25.96 34.85 22.31 32.29 20.68 30.11 19.10 U-238 16.37 6.28 20.13 5.84 21.29 5.61 21.25 5.51 21.41 5.45 21.31 5.40 Pu-239 3.58 12.88 15.56 37.65 21.70 45.88 24.65 48.88 25.96 50.14 27.16 51.45 Other 6.34 17.00 9.09 18.81 10.24 18.99 10.59 18.96 10.69 18.99 10.87 18.87

4-22 Table 4-16 Percent Contribution in Nuclear Data-Induced Uncertainty for 8 wt% 235U Cases Burnup (GWd/MTU) 10 40 60 70 75 80 ENDF Nuclide E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 E7.1 E8.0 Maj AC 93.63 83.09 89.21 80.49 86.76 79.51 85.71 79.00 85.21 79.05 84.73 78.75 Min AC 0.09 0.14 0.43 0.24 0.68 0.35 0.82 0.40 0.88 0.43 0.95 0.46 FP 0.48 0.43 2.01 1.10 2.97 1.52 3.48 1.72 3.68 1.83 3.91 1.92 U-235 76.07 67.22 56.78 40.24 46.10 30.69 41.06 26.79 38.81 25.00 36.51 23.34 U-238 14.76 5.35 18.36 5.52 19.84 5.34 20.01 5.26 20.09 5.21 20.18 5.17 Pu-239 2.78 10.23 13.07 33.98 18.34 42.06 21.08 45.11 22.23 46.72 23.38 47.88 Other 5.80 16.34 8.35 18.17 9.60 18.62 9.99 18.88 10.22 18.69 10.41 18.86 Figure 4-11 Percent Contribution of Nuclear Data-Induced Uncertainty in keff from Major AC, Minor AC, and FP for 5 and 7 wt% 235U Cases

4-23 Figure 4-12 Percent Contribution of Nuclear Data-Induced Uncertainty in keff from 235U, 238U, and 239Pu for 5 and 7 wt% 235U Cases For each application studied, the worth of the minor actinides and fission products was calculated by taking the difference between each keff of the actinides and 16 fission products (AFP) cases and their respective AO counterpart. The ratio of the minor actinides and fission products nuclear data-induced uncertainty to their worth was also calculated and included in Table 4-17. It can be seen from the table that the ratio of the minor actinides and fission products nuclear data-induced uncertainty to their worth ranges from 0.73 to 1.05% of their worth. This ratio is similar for both the ENDF/B-VII.1 and ENDF/B-VIII.0 libraries and decreases with burnup. With enrichment, both libraries see an increase in the ratio. Consistent with NUREG/CR-7109 [1], an upper value of 1.5% for the AFP worth can be applied to conservatively estimate their associated biases. APPENDIX C contains the detailed keff results used to generate Table 4-17.

4-24 Table 4-17 Worth and Uncertainty to Worth Ratios of Minor Actinides and Fission Products Worth (k)

Uncertainty/Worth (%)

Enrichment (wt% 235U)

Library Burnup (GWd/MTU)

E7.1 E8.0 E7.1 E8.0 4

10

-0.04590

-0.04549 0.80%

0.81%

40

-0.07784

-0.07713 0.75%

0.76%

60

-0.09217

-0.09121 0.74%

0.75%

70

-0.09880

-0.09755 0.74%

0.74%

75

-0.10173

-0.10115 0.74%

0.74%

80

-0.10520

-0.10392 0.73%

0.74%

5 10

-0.04554

-0.04518 0.82%

0.82%

40

-0.07881

-0.07798 0.77%

0.78%

60

-0.09250

-0.09180 0.77%

0.77%

70

-0.09879

-0.09787 0.77%

0.77%

75

-0.10143

-0.10091 0.76%

0.76%

80

-0.10442

-0.10363 0.76%

0.76%

6 10

-0.04491

-0.04454 0.84%

0.85%

40

-0.07997

-0.07888 0.79%

0.80%

60

-0.09332

-0.09298 0.79%

0.79%

70

-0.09944

-0.09884 0.79%

0.78%

75

-0.10267

-0.10188 0.77%

0.78%

80

-0.10484

-0.10434 0.77%

0.78%

7 10

-0.04439

-0.04408 0.86%

0.87%

40

-0.08046

-0.07996 0.81%

0.81%

60

-0.09448

-0.09353 0.80%

0.81%

70

-0.10036

-0.09952 0.80%

0.80%

75

-0.10297

-0.10236 0.80%

0.80%

80

-0.10570

-0.10512 0.79%

0.80%

8 10

-0.04405

-0.04335 0.88%

1.05%

40

-0.08122

-0.08106 0.83%

0.82%

60

-0.09507

-0.09478 0.82%

0.82%

70

-0.10103

-0.10021 0.82%

0.82%

75

-0.10387

-0.10307 0.81%

0.82%

80

-0.10651

-0.10570 0.81%

0.82%

4-25 4.2.2 Extended Enrichment Fuel This section presents the keff results for the range of burnup and enrichment in the GBC-32 application cases for the three nuclide sets in Section 2.1.1 and using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries. Figure 4-13 shows the AO cases, Figure 4-14 shows the AFP cases, and Figure 4-15 shows the ALL keff results. The figures also show a target keff value of 0.94, which represents the 0.95 limit [23] with a 0.01 reduction in keff to account for uncertainties. Figure 4-16 shows the difference in keff results between the two nuclear data libraries for AFP. A few observations can be made regarding the figures:

1. The ENDF/B-VII.1 nuclear data libraries consistently produce more conservative results compared to ENDF/B-VIII.0.

2. The difference between the AFP and ALL results is not significant as compared to the difference between the AO and ALL results.

3. Crediting the actinides only is not sufficient to demonstrate subcritical conditions for all enrichments and burnup up to 80 GWd/MTU.

4. Crediting more than AFP will be necessary to demonstrate subcritical conditions for all enrichments at burnup of 80 GWd/MTU.

5. In general, the difference in keff results between the two nuclear data libraries increases with burnup, and it is less significant with higher enrichments.

The detailed keff results can be found in APPENDIX C.

Figure 4-13 keff Results for AO Cases

4-26 Figure 4-14 keff Results for AFP Cases Figure 4-15 keff Results for ALL Cases

4-27 Figure 4-16 Difference in keff Results Between E7.1 and E8.0 for AFP Cases One of the main reasons for the differences observed in Figure 4-16 is 239Pu. Figure 4-17 shows the 239Pu absolute number density difference between the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries in axial zone 09. More 239Pu inventory is predicted by using the ENDF/B-VII.1 library compared to ENDF/B-VIII.0. The increased inventory is attributed to the following factors:

The difference in the 238U cross section between the two libraries. The ENDF/B-VII.1 capture cross section for 238U in the thermal neutron energy range is slightly higher than ENDF/B-VIII.0.

The difference in the 239Pu cross section between the two libraries. Figure 4-18 shows a large reduction in the 239Pu capture cross section around 0.3 eV in the ENDF/B-VII.1 library compared to ENDF/B-VIII.0. It is noted that 0.3 eV is the approximate EALF for the GBC-32 models. As expected, the sensitivity of keff for 239Pu at these energies is high, as indicated in Figure 4-19.

4-28 Figure 4-17 Difference in Number Density of 239Pu in Zone 09 Between E7.1 and E8.0 Figure 4-18 Difference in 239Pu Cross-Sections Between E7.1 and E8.0

4-29 Figure 4-19 239Pu Fission Cross-Section Sensitivity Using E7.1 and E8.0 4.2.3 High Burnup Fuel Some of the impacts of high burnup on keff are discussed above in Section 4.2.2. It is also important to study the effect of high-burnup fuel on the BUC loading curve to understand the allowable fuel enrichment and burnup to meet the target of 0.94 keff [1].

A comparison was made with the BUC loading curve reported in NUREG/CR-7109 [1] and that produced using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries. The comparison was made using AO, AFP, and ALL for enrichments up to 8 wt% 235U. Note that NUREG/CR-7109 used ENDF/B-VII.0 and only considered enrichments up to 5 wt% 235U. Figure 4-20 and Figure 4-21 show the results for the ENDF/B-VII.1 and ENDF/B-VIII.0 libraries, respectively.

The loading curves show the acceptable and unacceptable fuel storage configurations that meet the target keff limit. Any combination of enrichment and burnup in the unacceptable region would not meet the target of 0.94 keff. The figures illustrate that the results obtained in this work compare well with those published in NUREG/CR-7109 [1]. The figures also show that using the fission products in the fuel inventory allows for additional acceptable storage configurations.

4-30 Figure 4-20 Comparison of Burnup Loading Curves Using ENDF/B-VII.1 Figure 4-21 Comparison of Burnup Loading Curves Using ENDF/B-VIII.0 4.3 Sensitivity 4.3.1 Nuclear Data Libraries The influence of the nuclear data library and covariance data on the similarity coefficient is presented in Section 3.2.4. In the present section, the influence of these libraries on nuclear data-induced uncertainty is addressed in further detail. Application cases of 5 and 7 wt% 235U at burnup of 60 GWd/MTU were used in this study.

Acceptable Not Acceptable Not Acceptable Acceptable

4-31 Table 4-18 and Table 4-19 show the results of this study. A few observations can be made from the data presented in the tables:

1. ENDF/B-VIII.0 and ENDF/B-VII.1, with 252 neutron energy groups, produce very similar nuclear data-induced uncertainty in keff for most of the nuclides. This indicates that the main cause of the different results is the covariance data.

2. The nuclear data-induced uncertainty from minor AC and major FP are very similar for all nuclear data library and covariance data combinations.

3. The SCALE 6.2 56-group (COV B) or 252-group (COV C) covariance data based on ENDF/B-VII.1 produce very similar results. This indicates that using the 56-group covariance data is sufficient, and there is no appreciable benefit in using finer energy group data.

4. Using the ENDF/B-VIII.0 56-group (COV A) covariance data, which was added to SCALE 6.3, produces a significantly higher uncertainty than using the other two covariance libraries.

5. The main nuclides causing the difference when using the ENDF/B-VIII.0 56-group covariance data are a.

235U, 238U, 239Pu, and 240Pu, mainly due to the (n,gamma) reaction b.

1H, 56Fe, and 16O, mainly due to the elastic scattering reaction, and c.

10B, mainly due to the (n, alpha) reaction.

Another study was performed with the same application cases to investigate the effect of using multigroup and continuous neutron energy libraries. Figure 4-22 shows that relative differences in keff have a maximum of 0.06% (corresponding to 58 pcm).

Figure 4-22 k eff Difference Using Multigroup and Continuous Neutron Energy Libraries

4-32 Table 4-18 Nuclear Data-Induced Uncertainty in keff for 5 wt% 235U, 60 GWd/MTU (pcm)

ENDF/B library E8.0 E7.1 Covariance library COV A COV B COV C COV A COV B COV C Total 550.09 383.57 384.69 554.49 384.70 385.83 Major AC 489.17 354.34 355.45 493.78 355.70 356.82 U-234 0.14 0.14 0.14 0.14 0.14 0.14 U-235 218.85 191.14 191.28 218.04 189.59 189.75 U-238 136.83 187.66 189.26 137.27 188.85 190.47 Pu-238 15.72 15.72 15.72 15.98 15.98 15.97 Pu-239 405.19 214.06 214.12 410.47 215.59 215.64 Pu-240 31.86 22.66 24.57 32.55 23.12 25.19 Pu-241 63.55 63.55 63.62 64.92 64.92 65.01 Pu-242 20.97 20.97 20.97 22.85 22.85 22.85 Am-241 52.45 52.45 52.45 53.88 53.88 53.88 Minor AC 27.62 27.62 27.63 27.85 27.85 27.86 Am-243 3.02 3.02 3.02 3.17 3.17 3.18 Np-237 19.29 19.29 19.27 19.54 19.53 19.52 U-236 19.53 19.53 19.56 19.59 19.59 19.62 Major FP 64.93 64.92 64.92 65.33 65.32 65.32 Mo-95 5.20 5.20 5.20 5.24 5.24 5.24 Tc-99 10.61 10.61 10.62 10.78 10.78 10.78 Ru-101 9.54 9.54 9.54 9.64 9.64 9.64 Rh-103 22.09 22.09 22.09 22.39 22.38 22.38 Ag-109 3.01 3.01 3.01 3.13 3.13 3.13 Cs-133 18.45 18.45 18.45 18.75 18.75 18.75 Sm-147 6.03 6.03 6.02 5.91 5.91 5.91 Sm-149 18.47 18.46 18.46 18.69 18.68 18.68 Sm-150 5.76 5.76 5.75 5.82 5.82 5.82 Sm-151 12.79 12.78 12.78 12.93 12.93 12.93 Sm-152 6.64 6.64 6.64 6.77 6.77 6.77 Nd-143 40.44 40.44 40.43 40.71 40.71 40.71 Nd-145 20.00 20.00 20.00 19.51 19.51 19.51 Eu-151 0.24 0.24 0.24 0.25 0.25 0.25

4-33 Table 4-18 Nuclear Data-Induced Uncertainty in k eff for 5 wt% 235U, 60 GWd/MTU (pcm)

(Continued)

ENDF/B library E8.0 E7.1 Covariance library COV A COV B COV C COV A COV B COV C Eu-153 10.95 10.95 10.95 11.11 11.11 11.10 Gd-155 19.34 19.34 19.36 19.39 19.39 19.41 Other nuclides 241.52 128.78 129.09 242.06 128.19 128.44 H-1 235.78 92.31 92.36 236.55 91.56 91.59 Fe-56 28.59 66.75 67.27 28.41 66.43 66.83 O-16 20.49 50.00 49.92 16.71 49.86 49.83 Zr-91 21.95 21.96 22.01 22.30 22.30 22.37 Zr-92 19.48 19.49 19.49 19.78 19.79 19.80 Zr-90 8.58 8.58 8.61 8.75 8.76 8.79 B-10 20.62 5.42 5.82 20.76 5.46 5.87 Ni-58 6.66 6.66 6.64 6.69 6.69 6.66 Cr-52 4.99 4.96 5.02 5.02 4.99 5.04 Mn-55 4.47 4.47 4.52 4.51 4.51 4.56 Zr-94 4.31 4.31 4.38 4.31 4.31 4.37 Fe-57 1.58 1.58 1.58 3.55 3.55 3.54 Cr-53 3.57 3.57 3.57 3.57 3.57 3.58 Ni-62 2.81 2.81 2.35 2.81 2.81 2.35 Fe-54 2.29 2.70 2.71 2.37 2.73 2.74 Zr-96 1.60 1.60 1.60 1.64 1.64 1.64 Cr-50 1.07 1.07 1.07 1.07 1.07 1.07 Al-27 1.23 1.23 1.21 1.22 1.22 1.19 Ni-60 0.92 0.92 0.95 0.93 0.93 0.96 Cr-54 0.41 0.41 0.41 0.43 0.43 0.43 Ni-61 0.44 0.44 0.41 0.44 0.44 0.41 Fe-58 0.31 0.37 0.35 0.21 0.23 0.19

4-34 Table 4-19 Nuclear Data-Induced Uncertainty in keff for 7 wt% 235U, 60 GWd/MTU (pcm)

ENDF/B library E8.0 E7.1 Covariance library COV A COV B COV C COV A COV B COV C Total 562.20 398.16 399.36 565.51 398.84 400.06 Major AC 500.32 369.60 370.77 503.21 370.24 371.44 U-234 0.16 0.16 0.16 0.16 0.16 0.16 U-235 286.43 253.58 253.78 285.99 252.10 252.34 U-238 133.10 182.16 184.00 133.36 184.01 185.86 Pu-238 12.71 12.71 12.71 12.69 12.69 12.69 Pu-239 380.79 184.85 184.90 384.55 185.81 185.84 Pu-240 31.21 19.20 20.95 31.78 19.52 21.38 Pu-241 45.80 45.80 45.83 46.44 46.44 46.48 Pu-242 17.25 17.25 17.25 18.40 18.40 18.40 Am-241 45.00 45.00 45.00 45.81 45.81 45.81 Minor AC 31.84 31.84 31.85 31.82 31.82 31.84 Am-243 2.20 2.20 2.20 2.27 2.27 2.27 Np-237 20.54 20.54 20.53 20.56 20.56 20.55 U-236 24.22 24.22 24.26 24.18 24.18 24.22 Major FP 68.74 68.73 68.74 68.59 68.58 68.59 Mo-95 5.98 5.98 5.98 5.99 5.99 5.99 Tc-99 10.84 10.84 10.85 10.92 10.92 10.93 Ru-101 10.84 10.84 10.84 10.83 10.83 10.82 Rh-103 24.51 24.51 24.51 24.65 24.65 24.65 Ag-109 2.72 2.72 2.72 2.80 2.80 2.80 Cs-133 20.80 20.80 20.80 20.96 20.96 20.96 Sm-147 7.11 7.11 7.11 6.93 6.93 6.93 Sm-149 20.32 20.31 20.31 20.43 20.42 20.42 Sm-150 5.72 5.72 5.72 5.71 5.71 5.71 Sm-151 13.18 13.18 13.18 13.23 13.23 13.23 Sm-152 7.16 7.16 7.16 7.26 7.26 7.26 Nd-143 42.56 42.56 42.56 42.55 42.55 42.55 Nd-145 23.22 23.22 23.22 22.46 22.46 22.46 Eu-151 0.28 0.28 0.28 0.28 0.28 0.28

4-35 Table 4-19 Nuclear Data-Induced Uncertainty in keff for 7 wt% 235U, 60 GWd/MTU (pcm) (Continued)

ENDF/B library E8.0 E7.1 Covariance library COV A COV B COV C COV A COV B COV C Eu-153 10.81 10.81 10.81 10.83 10.83 10.83 Gd-155 14.96 14.96 14.98 14.80 14.80 14.81 Other nuclides 244.97 127.23 127.56 246.69 127.61 127.91 H-1 239.02 88.42 88.48 241.07 88.72 88.78 Fe-56 30.12 70.43 70.96 29.93 69.95 70.38 O-16 22.08 49.09 49.00 17.88 49.83 49.79 Zr-91 20.17 20.17 20.24 20.51 20.51 20.60 Zr-92 17.55 17.56 17.56 17.81 17.82 17.83 Zr-90 9.02 9.02 9.05 9.16 9.16 9.20 B-10 22.92 6.11 6.57 22.99 6.12 6.58 Ni-58 7.16 7.16 7.14 7.20 7.20 7.18 Cr-52 5.26 5.25 5.31 5.28 5.26 5.30 Mn-55 4.99 4.99 5.04 4.95 4.95 5.01 Zr-94 4.47 4.47 4.54 4.45 4.45 4.51 Fe-57 1.67 1.68 1.67 3.71 3.71 3.70 Cr-53 3.78 3.78 3.78 3.78 3.78 3.78 Ni-62 2.94 2.94 2.43 2.94 2.94 2.43 Fe-54 2.52 2.88 2.88 2.49 2.88 2.89 Zr-96 1.75 1.75 1.75 1.79 1.79 1.79 Cr-50 1.15 1.15 1.15 1.15 1.15 1.16 Al-27 1.38 1.38 1.36 1.35 1.35 1.36 Ni-60 0.99 0.99 1.01 0.97 0.97 0.99 Cr-54 0.45 0.45 0.45 0.46 0.46 0.46 Ni-61 0.48 0.48 0.46 0.48 0.48 0.46 Fe-58 0.33 0.38 0.35 0.22 0.24 0.20 4.3.2 Decay Time The effect of the decay time on the bias and bias uncertainty was studied for a limited set of application cases for 5 and 7 wt% 235U, burnup of 60 GWd/MTU, and the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries. The decay times considered were 0, 1, 5, 10, 20, and 40

4-36 years. Figure 4-23 shows the keff and EALF for the different application cases. For each case, the keff decreases with increasing decay time, and EALF increases; thus, the initial point in the series (0 days decay time) is always the upper leftmost value. The differences in the actinide cross sections between the nuclear data libraries (for example, see Figure 4-18) result in a higher keff and harder neutron energy spectrum using ENDF/B-VII.1.

Figure 4-23 Effect of Decay Time on k eff and EALF Using the EALF trending analysis approach described in Section 4.1, the bias and bias uncertainty were calculated for the different application cases and are shown in Table 4-20.

Applicable experiments, as determined by Section 3.2.2 burnup and enrichment values, were used rather than redetermining applicable experiments for the decay time perturbation. This decision and the resulting changes in compositions and spectrum are expected to result in possible variations in bias and bias uncertainties from differing experiment sets. As the decay time increases, the combined bias and bias uncertainty (+ ) slightly increases. The slight increase is mainly driven by the bias uncertainty as a result of decreasing applicable experiment sample size with increasing EALF.

4-37 Table 4-20 Effect of Decay Time on Bias and Bias Uncertainty Simple average Fission density-weighted Decay time Enrichment (wt% 235U)

Final burnup (GWd/MTU)

Library

+

(pcm)

Final enrichment (wt% 235U)

Pu/(Pu+U)

(wt%)

Final enrichment (wt% 235U)

Pu/(Pu+U)

(wt%)

0 days 5

60 E7.1 141 884 1025 1.051 1.563 1.779 1.293 E8.0 95 917 1012 1.016 1.557 1.763 1.282 7

60 E7.1 184 925 1109 2.227 1.597 3.157 1.324 E8.0 82 945 1027 2.182 1.591 3.134 1.315 80 E7.1 177 943 1120 1.365 1.860 2.544 1.507 E8.0 79 945 1024 1.318 1.853 2.521 1.496 1 year 5

60 E7.1 141 889 1030 1.051 1.572 1.814 1.284 E8.0 91 922 1013 1.016 1.565 1.796 1.275 7

60 E7.1 185 930 1115 2.227 1.602 3.206 1.312 E8.0 80 949 1029 2.182 1.597 3.184 1.303 80 E7.1 178 948 1126 1.365 1.868 2.589 1.496 E8.0 76 950 1026 1.318 1.860 2.565 1.486 5 years 5

60 E7.1 142 893 1035 1.051 1.533 1.873 1.234 E8.0 89 926 1015 1.016 1.527 1.855 1.225 7

60 E7.1 185 934 1119 2.227 1.562 3.281 1.262 E8.0 78 953 1031 2.182 1.557 3.260 1.253 80 E7.1 179 954 1133 1.365 1.821 2.666 1.439 E8.0 73 955 1028 1.319 1.814 2.643 1.429 10 years 5

60 E7.1 142 896 1038 1.051 1.492 1.921 1.188 E8.0 87 929 1016 1.016 1.487 1.904 1.179 7

60 E7.1 186 936 1122 2.227 1.521 3.344 1.216 E8.0 76 955 1031 2.182 1.516 3.321 1.209 80 E7.1 180 958 1138 1.365 1.772 2.728 1.386 E8.0 71 958 1029 1.319 1.766 2.704 1.377 20 years 5

60 E7.1 142 899 1041 1.052 1.435 1.974 1.130 E8.0 85 932 1017 1.016 1.430 1.954 1.123 7

60 E7.1 186 938 1124 2.227 1.462 3.408 1.160 E8.0 76 957 1033 2.182 1.458 3.387 1.154 80 E7.1 180 961 1141 1.366 1.702 2.796 1.319 E8.0 69 962 1031 1.319 1.697 2.772 1.311 40 years 5

60 E7.1 142 900 1042 1.052 1.374 2.016 1.076 E8.0 84 933 1017 1.016 1.370 1.997 1.070 7

60 E7.1 186 939 1125 2.228 1.401 3.461 1.108 E8.0 75 958 1033 2.182 1.397 3.438 1.102 80 E7.1 181 963 1144 1.366 1.627 2.850 1.256 E8.0 68 963 1031 1.319 1.622 2.825 1.249

5 CONCLUSIONS AND RECOMMENDATIONS A validation study was performed to investigate the effect of high burnup and extended enrichment fuels on criticality safety analyses using BUC. The study was mainly divided into two sections: the performance of a similarity assessment to determine the appropriate criticality benchmark experiments for each application considered and the investigation of the bias and bias uncertainty for the different application cases. Both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries were used in the study.

The similarity study results showed that, in general, sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235U and burnups up to 80 GWd/MTU, with the following notes:

1. The French HTC critical benchmark experiments have the most similarity with the application cases.

2. Some application cases (for example: 8 wt% 235U case at 40 GWd/MTU burnup and 5 wt% 235U case at 10 GWd/MTU burnup) have limited, or no, highly similar benchmark experiments, but they have sufficient marginally similar experiments.

Using the applicable critical benchmark experiments, the bias and bias uncertainty were calculated for different application cases. Both the ck and EALF trending parameters were used in the calculations. The calculational margin was found to range from 625 to 1477 pcm, with an average of 1079 pcm. In addition, the nuclear data-induced uncertainty in keff was calculated and found to range from 376 to 512 pcm using the ENDF/B-VII.1 library and from 541 to 602 pcm using the ENDF/B-VIII.0 library. The minor actinide and fission product nuclear data-induced uncertainty ranged from 0.7 to 1.1% of their worth. Finally, the BUC loading curve was presented for some application cases.

For both the similarity assessment and the bias and bias uncertainty studies, the following conclusions can be made:

1. Both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used effectively.

2. The choice of the covariance library can have a significant impact on similarity assessment. The identification of a larger number of applicable experiments with the ENDF/B-VIII.0 covariance can alter the bias uncertainty, resulting in a different USL compared to what is obtained by using ENDF/B-VII.1.

5.1 Continued Validity of NUREG/CR-7109 Conclusions The study showed that the conclusions of NUREG/CR-7109 continue to be valid, specifically:

1. For the application cases considered, there were sufficient benchmark experiments available to confidently validate BUC criticality safety analysis.

2. A conservative estimate for the bias associated with minor actinide and FP nuclides of 1.5% of their worth may be used to account for the lack of adequate critical benchmarks containing these nuclides.

5-1

5-2 5.2 Additional Recommendations The validation study performed can be complemented in the future with the following additional research:

1. Validation studies for high-assay low-enriched uranium (HALEU fuel) (enrichments extending to 20 wt% 235U).

2. Guidance on validation gaps where limited marginally similar critical benchmark experiments exist.

3. The impact of structural material and neutron absorbers on BUC validation studies.

6 REFERENCES

1. Scaglione, J. M., Mueller, D. E., Wagner, J. C., Marshall, W. J., An Approach for Validating Actinide and Fission Product Burnup Credit Criticality AnalysesCriticality (keff) Predictions, NUREG/CR-7109 (ORNL/TM-2011/514), prepared for the U.S.

Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, April 2012.

2. Wieselquist, W. A. and Lefebvre, R. A., eds., SCALE 6.3.1 User Manual, ORNL/TM-SCALE-6.3.1, Oak Ridge National Laboratory, Oak Ridge, TN, February 2023, https://info.ornl.gov/sites/publications/Files/Pub191420.pdf.

3. Chadwick, M. B., et al., ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data, Nuclear Data Sheets, Vol. 112, No. 12, pp. 2887-2996, December 2011, https://doi.org/10.1016/j.nds.2011.11.002.

4. Brown, D. A., et al., ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data, Nuclear Data Sheets, Vol. 148,pp. 1-142, February 2018, https://doi.org/10.1016/j.nds.2018.02.001.

5. Wagner, J. C., Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit, NUREG/CR-6747 (ORNL/TM-2000/306), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, October 2001.

6. Clarity, J.C., Marshall, W.J., Mueller, D.E., et al., Determination of Bias and Bias Uncertainty for Criticality Safety Computational Methods, ORNL/TM-2024/3, prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, June 2024.

7. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations, ORNL/TM-2005/39, Version 6, Vols. I-III, January 2009.

8. Little, R. C., Kawano, T., Hale, G. D., Pigni, M. T., Herman, M., Oblozinsky, P., Williams, M. L., Dunn, M. E., Arbanas, G., and Wiarda, D., Low-Fidelity Covariance Project.

Nuclear Data Sheets, 109(12):2828-2833, 2008.

9. American National Standards Institute / American Nuclear Society, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, ANSI/ANS-8.1-1998; R2007, American Nuclear Society, LaGrange Park, IL, 1998.

10. Broadhead, B. L., Rearden, B. T., Hopper, C. M., Wagschal, J. J., and Parks, C. V.,

Sensitivity-and Uncertainty-Based Criticality Safety Validation Techniques, Nuclear Science and Engineering, vol 146, pp. 340-366, 2004.

11. Rearden, B. T., Williams, M. L., Jessee, M. A., Mueller, D. E., and Wiarda, D. A.,

Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE, Nuclear Technology, vol. 174, pp. 236-288, 2011.

6-1

6-2

12. Bostelmann, F., Holcomb, A. M., Clarity, J. B., Marshall, W. J., Sobes, V., and Rearden, B. T., Nuclear Data Performance Assessment for Advanced Reactors, ORNL/TM-2018/1033, Oak Ridge National Laboratory, Oak Ridge, TN, March 2019, https://doi.org/10.2172/1506806.

13. International Handbook of Evaluated Criticality Safety Benchmark Experiments. NEA 7497, OECD NEA, Paris, France, 2021.

14. Marshall, W. J., Rearden, B. T., The SCALE Verified, Archived Library of Inputs and Data-VALID, Proceedings of the American Nuclear Society Nuclear Criticality Safety Division Conference, 2013 (NCSD 2013), Wilmington, NC, USA, September 29 -

October 03, 2013.

15. Hill, I., Gulliford, J., Briggs, J. B., Rearden, B. T., and Ivanova, T., Generation of 1800 New Sensitivity Data Files for ICSBEP Using SCALE 6.0, Transactions of the American Nuclear Society, vol. 109(1), pp. 867-869, 2013.

16. Fernex, F., Programme HTC-Phase 1: Réseaux de Crayons dans lEau Pure (Water-Moderated and Reflected Simple Arrays) Reevaluation des Expériences, DSU/SEC/T/2005-33/D.R, Valduc, France, IRSN, 2006. PROPRIETARY document.

17. Fernex, F., Programme HTC-Phase 2: Réseaux Simples en Eau Empoisonnée (Bore et Gadolinium) (Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or Boron) Réévaluation des Experiences, DSU/SEC/T/2005-38/D.R, Valduc, France, IRSN, 2006. PROPRIETARY document.

18. Fernex, F., Programme HTC-Phase 3: Configurations Stockage en Piscine (Pool Storage) Réévaluation des Experiences, DSU/SEC/T/2005-37/D.R, Valduc, France, IRSN, 2006. PROPRIETARY document.

19. Fernex, F., Programme HTC-Phase 4: Configurations Châteaux de Transport (Shipping Cask) Réévaluation des Experiences, DSU/SEC/T/2005-36/D.R., Valduc, France, IRSN, 2006. PROPRIETARY document.

20. Sobes, V., et al., ENDF/B-VIII.0 Covariance Data Development and Testing for Advanced Reactors, ORNL/TM-2018/1037, Oak Ridge National Laboratory, Oak Ridge, TN, 2019.

21. Marshall, W. J., et al., Development and Testing of Neutron Cross-section Covariance Data for SCALE 6.2, Proceedings of the International Conference on Nuclear Criticality Safety 2015 (ICNC 2015), Charlotte, NC, Sept 13-17, 2015.

22. Dean, J. C., Tayloe, Jr., R. W., and Morey, D., Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, prepared for the US Nuclear Regulatory Commission by Science Applications International Corporation, Oak Ridge, TN, January 2001.

23. Code of Federal Regulations, Title 10 CFR 50.68, Criticality Accident Requirements, U.S. Nuclear Regulatory Commission.

APPENDIX A SAMPLE INPUT FILES Example input files are provided in this appendix. Note that in the TSUNAMI-IP input file, the experiment listing is abbreviated due to length.

A.1 TRITON

=t-depl parm=(centrm,addnux=3)

PWR Westinghouse OFA 17x17 with WABA, 1/4 assembly model, 5.0 wt% enriched v7-252 read comp

'fuel 5.0 wt%

uo2 1 den=10.5216 1 1100 92235 5.0 92238 95.0 end

'clad zirc4 2 1 620 end

'water moderator with 1000 ppm soluble boron h2o 3 den=0.63 1 610 end arbmb 0.63 1 1 0 0 5000 100 3 1000e-06 610 end

'gap n 4 den=0.00125 1 620 end

'guide tube zirc4 5 1 610 end

'WABA inner water moderator with 1000 ppm soluble boron h2o 10 den=0.63 1 610 end arbmb 0.63 1 1 0 0 5000 100 10 1000e-06 610 end

'WABA inner cladding zirc4 11 1 610 end

'WABA inner gap n 12 den=0.00125 1 610 end

'WABA Al2O3-B4C b-10 13 0 3.0697E-3 610.0 end b-11 13 0 1.2753E-2 610.0 end c 13 0 3.9521E-3 610.0 end al 13 0 2.6344E-2 610.0 end o 13 0 3.9506E-2 610.0 end

'WABA Al2O3-B4C b-10 14 0 3.0697E-3 610.0 end b-11 14 0 1.2753E-2 610.0 end c 14 0 3.9521E-3 610.0 end al 14 0 2.6344E-2 610.0 end o 14 0 3.9506E-2 610.0 end

'WABA Al2O3-B4C b-10 15 0 3.0697E-3 610.0 end b-11 15 0 1.2753E-2 610.0 end c 15 0 3.9521E-3 610.0 end al 15 0 2.6344E-2 610.0 end o 15 0 3.9506E-2 610.0 end

'WABA Al2O3-B4C b-10 16 0 3.0697E-3 610.0 end b-11 16 0 1.2753E-2 610.0 end c 16 0 3.9521E-3 610.0 end al 16 0 2.6344E-2 610.0 end o 16 0 3.9506E-2 610.0 end

'WABA Al2O3-B4C b-10 17 0 3.0697E-3 610.0 end b-11 17 0 1.2753E-2 610.0 end c 17 0 3.9521E-3 610.0 end al 17 0 2.6344E-2 610.0 end o 17 0 3.9506E-2 610.0 end

'WABA outer gap n 18 den=0.00125 1 610 end

'WABA outer cladding zirc4 19 1 610 end

'WABA outer water moderator with 1000 ppm soluble boron h2o 20 den=0.63 1 610 end arbmb 0.63 1 1 0 0 5000 100 20 1000e-06 610 end

'WABA guide tube zirc4 21 1 610 end

'WABA pitch water moderator with 1000 ppm soluble boron h2o 22 den=0.63 1 610 end arbmb 0.63 1 1 0 0 5000 100 22 1000e-06 610 end

'ring around WABA for centrm 5.0 wt%

uo2 23 den=10.5216 1 1100 92235 5.0 92238 95.0 end end comp

' Cell data read celldata latticecell squarepitch pitch=1.2598 3 fueld=0.7844 1 gapd=0.8001 4 cladd=0.9144 2 end multiregion cylindrical right_bdy=white end 10 0.28575 11 0.33910 12 0.35305 13 0.35767 14 0.36690 15 0.38999 16 0.39923 17 0.40385 18 0.41785 19 0.48387 20 0.56135 21 0.60200 22 0.71077 23 0.80 end zone end celldata

' Depletion data read depletion 1 flux 13 14 15 16 17 end depletion

' Burn data A-1

A-2 read burndata power=60.0 burn=1500 down=0 nlib=30 end end burndata

' NEWT model data read model Westinghouse OFA 17x17, 4.0 wt% uo2 read parm run=yes end parm read materials 1 1 ! fuel ! end 2 1 ! clad ! end 3 2 ! water ! end 4 0 ! gap ! end 5 1 ! guide tube ! end 10 2 ! WABA inner water ! end 11 1 ! WABA clad ! end 12 0 ! WABA inner gap ! end 13 2 ! WABA poison rod ! end 14 2 ! WABA poison rod ! end 15 2 ! WABA poison rod ! end 16 2 ! WABA poison rod ! end 17 2 ! WABA poison rod ! end 18 0 ! WABA gap ! end 19 1 ! WABA outer clad ! end 20 2 ! WABA outer water ! end 21 1 ! WABA guide tube ! end 22 2 ! WABA pitch water ! end end materials read geom unit 1 com='fuel rod' cylinder 10.3922 cylinder 20.40005 cylinder 30.4572 cuboid 40 4p0.6299 media 1 1 10 media 4 1 20 -10 media 2 1 30 -20 media 3 1 40 -30 boundary 40 4 4 unit 5 com='guide tube' cylinder 10.28575 cylinder 20.33910 cylinder 30.35305 cylinder 40.35767 cylinder 50.36690 cylinder 60.38999 cylinder 70.39923 cylinder 80.40385 cylinder 90.41785 cylinder 100.48387 cylinder 110.56135 cylinder 120.602 cuboid 130 4p0.6299 media 10 1 10 media 11 1 20 -10 media 12 1 30 -20 media 13 1 40 -30 media 14 1 50 -40 media 15 1 60 -50 media 16 1 70 -60 media 17 1 80 -70 media 18 1 90 -80 media 19 1 100 -90 media 20 1 110 -100 media 21 1 120 -110 media 22 1 130 -120 boundary 130 4 4 unit 11 com='right half of fuel rod' cylinder 10.3922 chord +x=0 cylinder 20.40005 chord +x=0 cylinder 30.4572 chord +x=0 cuboid 40 0.6299 0.0 2p0.6299 media 1 1 10 media 4 1 20 -10 media 2 1 30 -20 media 3 1 40 -30 boundary 40 2 4 unit 12 com='top half of fuel rod' cylinder 10.3922 chord +y=0 cylinder 20.40005 chord +y=0 cylinder 30.4572 chord +y=0 cuboid 40 2p0.6299 0.6299 0.0 media 1 1 10 media 4 1 20 -10 media 2 1 30 -20 media 3 1 40 -30 boundary 40 4 2 unit 51 com='right half of guide tube' cylinder 10.28575 chord +x=0.0 cylinder 20.33910 chord +x=0.0 cylinder 30.35305 chord +x=0.0 cylinder 40.35767 chord +x=0.0 cylinder 50.36690 chord +x=0.0 cylinder 60.38999 chord +x=0.0 cylinder 70.39923 chord +x=0.0 cylinder 80.40385 chord +x=0.0 cylinder 90.41785 chord +x=0.0 cylinder 100.48387 chord +x=0.0 cylinder 110.56135 chord +x=0.0 cylinder 120.602 chord +x=0.0 cuboid 130 0.6299 0.0 2p0.6299 media 10 1 10 media 11 1 20 -10

A-3 media 12 1 30 -20 media 13 1 40 -30 media 14 1 50 -40 media 15 1 60 -50 media 16 1 70 -60 media 17 1 80 -70 media 18 1 90 -80 media 19 1 100 -90 media 20 1 110 -100 media 21 1 120 -110 media 22 1 130 -120 boundary 130 2 4 unit 52 com='top half of guide tube' cylinder 10.28575 chord +y=0.0 cylinder 20.33910 chord +y=0.0 cylinder 30.35305 chord +y=0.0 cylinder 40.35767 chord +y=0.0 cylinder 50.36690 chord +y=0.0 cylinder 60.38999 chord +y=0.0 cylinder 70.39923 chord +y=0.0 cylinder 80.40385 chord +y=0.0 cylinder 90.41785 chord +y=0.0 cylinder 100.48387 chord +y=0.0 cylinder 110.56135 chord +y=0.0 cylinder 120.602 chord +y=0.0 cuboid 130 2p0.6299 0.6299 0.0 media 10 1 10 media 11 1 20 -10 media 12 1 30 -20 media 13 1 40 -30 media 14 1 50 -40 media 15 1 60 -50 media 16 1 70 -60 media 17 1 80 -70 media 18 1 90 -80 media 19 1 100 -90 media 20 1 110 -100 media 21 1 120 -110 media 22 1 130 -120 boundary 130 4 2 unit 53 com='1/4 instrument tube' cylinder 10.56135 chord +x=0 chord +y=0 cylinder 20.602 chord +x=0 chord +y=0 cuboid 40 0.6299 0.0 0.6299 0.0 media 3 1 10 media 5 1 20 -10 media 3 1 40 -20 boundary 40 2 2 global unit 10 cuboid 10 10.7083 0.0 10.7083 0.0 array 1 10 place 1 1 0 0 media 3 1 10 boundary 10 34 34 end geom read array ara=1 nux=9 nuy=9 typ=cuboidal fill 53 12 12 52 12 12 52 12 12 11 1 1 1 1 1 1 1 1 11 1 1 1 1 1 1 1 1 51 1 1 5 1 1 5 1 1 11 1 1 1 1 1 1 1 1 11 1 1 1 1 5 1 1 1 51 1 1 5 1 1 1 1 1 11 1 1 1 1 1 1 1 1 11 1 1 1 1 1 1 1 1 end fill end array read bounds all=refl end bounds end model end

=shell cp ft33f001.cmbined $RTNDIR/origenlib-4.0 end A.2 ARP DATA

!w17x17_waba_89gwd 7 1 31 3.0 4.0 5.0 6.0 7.0 8.0 9.0 0.63

'e3_endf7.mix0001.f33' 'e4_endf7.mix0001.f33'

'e5_endf7.mix0001.f33' 'e6_endf7.mix0001.f33'

'e7_endf7.mix0001.f33' 'e8_endf7.mix0001.f33'

'e9_endf7.mix0001.f33' 0.0 1491.3 4474.6 7459.4 10446.3 13434.3 16423.1 19412.9 22402.9 25393.3 28383.9 31374.6 34365.1 37355.7 40346.0 43336.6 46326.3 49316.6 52306.3 55295.7 58285.3 61274.7 64264.0 67252.9 70241.7 73230.4 76218.9 79207.0 82195.3 85183.4 88171.1 A.3 ORIGAMI

=origami title="5wt%_40GWD" options{ mtu=1.0 ft71=all stdcomp=yes fdens=10.5216 relnorm=no nz=18

}

libs=[ "w17x17_waba_89gwd" ]

A-4 fuelcomp{

mix(1) { stdcomp(fuel){dens=10.5216 base=uo2 iso[92235=5 92238=95]

}

}

}

pz=[ 0.652 0.967 1.074 1.103 1.108 1.106 1.102 1.097 1.094 1.094 1.095 1.096 1.095 1.086 1.059 0.971 0.738 0.462

]

hist[

cycle{ power=60 burn=666.6666667 nlib=30 down=1826.25} ]

end A.4 CSAS5 Note: only three representative compositions are listed

=csas5 parm=()

gbc-32 w17x17, 5 wt% U-235, 40 gwd/mtu v7.1-252 read comp o-16 101 0 4.6833E-02 293.0 end u-234 101 0 1.3549E-07 293.0 end u-235 101 0 6.1725E-04 293.0 end u-236 101 0 1.0874E-04 293.0 end u-238 101 0 2.1874E-02 293.0 end np-237 101 0 8.1160E-06 293.0 end pu-238 101 0 1.7839E-06 293.0 end pu-239 101 0 1.5492E-04 293.0 end pu-240 101 0 3.6209E-05 293.0 end pu-241 101 0 1.8555E-05 293.0 end pu-242 101 0 3.9047E-06 293.0 end am-241 101 0 5.5839E-06 293.0 end am-243 101 0 4.9870E-07 293.0 end mo-95 101 0 3.7790E-05 293.0 end tc-99 101 0 3.6942E-05 293.0 end ru-101 101 0 3.4032E-05 293.0 end rh-103 101 0 2.1724E-05 293.0 end ag-109 101 0 2.2050E-06 293.0 end cs-133 101 0 3.9797E-05 293.0 end nd-143 101 0 3.0918E-05 293.0 end nd-145 101 0 2.2321E-05 293.0 end sm-147 101 0 7.5994E-06 293.0 end sm-149 101 0 1.9538E-07 293.0 end sm-150 101 0 8.2052E-06 293.0 end sm-151 101 0 5.8471E-07 293.0 end sm-152 101 0 3.1331E-06 293.0 end eu-151 101 0 2.3841E-08 293.0 end eu-153 101 0 2.8049E-06 293.0 end gd-155 101 0 8.7674E-08 293.0 end o-16 109 0 4.6831E-02 293.0 end u-234 109 0 3.2722E-07 293.0 end u-235 109 0 3.7098E-04 293.0 end u-236 109 0 1.4766E-04 293.0 end u-238 109 0 2.1561E-02 293.0 end np-237 109 0 1.5667E-05 293.0 end pu-238 109 0 5.8868E-06 293.0 end pu-239 109 0 1.7632E-04 293.0 end pu-240 109 0 6.0919E-05 293.0 end pu-241 109 0 3.3938E-05 293.0 end pu-242 109 0 1.3337E-05 293.0 end am-241 109 0 1.0062E-05 293.0 end am-243 109 0 2.7628E-06 293.0 end mo-95 109 0 5.9768E-05 293.0 end tc-99 109 0 5.8643E-05 293.0 end ru-101 109 0 5.6202E-05 293.0 end rh-103 109 0 3.4600E-05 293.0 end ag-109 109 0 4.6966E-06 293.0 end cs-133 109 0 6.2075E-05 293.0 end nd-143 109 0 4.4687E-05 293.0 end nd-145 109 0 3.4426E-05 293.0 end sm-147 109 0 9.7756E-06 293.0 end sm-149 109 0 2.4441E-07 293.0 end sm-150 109 0 1.4884E-05 293.0 end sm-151 109 0 6.9160E-07 293.0 end sm-152 109 0 4.4997E-06 293.0 end eu-151 109 0 2.7710E-08 293.0 end eu-153 109 0 5.3299E-06 293.0 end gd-155 109 0 2.0017E-07 293.0 end o-16 118 0 4.6834E-02 293.0 end u-234 118 0 8.0427E-08 293.0 end u-235 118 0 7.5312E-04 293.0 end u-236 118 0 8.4556E-05 293.0 end u-238 118 0 2.1998E-02 293.0 end np-237 118 0 5.0472E-06 293.0 end pu-238 118 0 7.7902E-07 293.0 end pu-239 118 0 1.3420E-04 293.0 end pu-240 118 0 2.4036E-05 293.0 end pu-241 118 0 1.0953E-05 293.0 end pu-242 118 0 1.5419E-06 293.0 end am-241 118 0 3.3105E-06 293.0 end am-243 118 0 1.3856E-07 293.0 end mo-95 118 0 2.7499E-05 293.0 end tc-99 118 0 2.6779E-05 293.0 end ru-101 118 0 2.4236E-05 293.0 end rh-103 118 0 1.5589E-05 293.0 end ag-109 118 0 1.3080E-06 293.0 end cs-133 118 0 2.9038E-05 293.0 end nd-143 118 0 2.3292E-05 293.0 end nd-145 118 0 1.6407E-05 293.0 end sm-147 118 0 6.0408E-06 293.0 end sm-149 118 0 1.7413E-07 293.0 end sm-150 118 0 5.5401E-06 293.0 end sm-151 118 0 5.3031E-07 293.0 end sm-152 118 0 2.3390E-06 293.0 end eu-151 118 0 2.1969E-08 293.0 end eu-153 118 0 1.7441E-06 293.0 end

A-5 gd-155 118 0 5.1627E-08 293.0 end zr-90 2 0 2.1763E-02 293.0 end zr-91 2 0 4.7461E-03 293.0 end zr-92 2 0 7.2544E-03 293.0 end zr-94 2 0 7.3517E-03 293.0 end zr-96 2 0 1.1844E-03 293.0 end o-16 3 0 3.3370E-02 293.0 end h-1 3 0 6.6740E-02 293.0 end o-16 4 0 3.3370E-02 293.0 end h-1 4 0 6.6740E-02 293.0 end cr-50 5 0 7.5733E-04 293.0 end cr-52 5 0 1.4605E-02 293.0 end cr-53 5 0 1.6558E-03 293.0 end cr-54 5 0 4.1222E-04 293.0 end mn-55 5 0 1.7400E-03 293.0 end fe-54 5 0 3.5022E-03 293.0 end fe-56 5 0 5.4445E-02 293.0 end fe-57 5 0 1.2466E-03 293.0 end fe-58 5 0 1.6621E-04 293.0 end ni-58 5 0 5.2704E-03 293.0 end ni-60 5 0 2.0149E-03 293.0 end ni-61 5 0 8.7236E-05 293.0 end ni-62 5 0 2.7715E-04 293.0 end ni-64 5 0 7.0252E-05 293.0 end b-10 6 0 6.5795E-03 293.0 end b-11 6 0 2.7260E-02 293.0 end c

6 0 8.4547E-03 293.0 end al-27 6 0 4.1795E-02 293.0 end cr-50 7 0 7.5733E-04 293.0 end cr-52 7 0 1.4605E-02 293.0 end cr-53 7 0 1.6558E-03 293.0 end cr-54 7 0 4.1222E-04 293.0 end mn-55 7 0 1.7400E-03 293.0 end fe-54 7 0 3.5022E-03 293.0 end fe-56 7 0 5.4445E-02 293.0 end fe-57 7 0 1.2466E-03 293.0 end fe-58 7 0 1.6621E-04 293.0 end ni-58 7 0 5.2704E-03 293.0 end ni-60 7 0 2.0149E-03 293.0 end ni-61 7 0 8.7236E-05 293.0 end ni-62 7 0 2.7715E-04 293.0 end ni-64 7 0 7.0252E-05 293.0 end al-27 8 0 6.0200E-02 293.0 end o-16 9 0 3.3370E-02 293.0 end h-1 9 0 6.6740E-02 293.0 end o-16 10 0 3.3370E-02 293.0 end h-1 10 0 6.6740E-02 293.0 end zr-90 11 0 2.1763E-02 293.0 end zr-91 11 0 4.7461E-03 293.0 end zr-92 11 0 7.2544E-03 293.0 end zr-94 11 0 7.3517E-03 293.0 end zr-96 11 0 1.1844E-03 293.0 end zr-90 209 0 2.1763E-02 293.0 end zr-91 209 0 4.7461E-03 293.0 end zr-92 209 0 7.2544E-03 293.0 end zr-94 209 0 7.3517E-03 293.0 end zr-96 209 0 1.1844E-03 293.0 end o-16 309 0 3.3370E-02 293.0 end h-1 309 0 6.6740E-02 293.0 end o-16 409 0 3.3370E-02 293.0 end h-1 409 0 6.6740E-02 293.0 end zr-90 218 0 2.1763E-02 293.0 end zr-91 218 0 4.7461E-03 293.0 end zr-92 218 0 7.2544E-03 293.0 end zr-94 218 0 7.3517E-03 293.0 end zr-96 218 0 1.1844E-03 293.0 end o-16 318 0 3.3370E-02 293.0 end h-1 318 0 6.6740E-02 293.0 end o-16 418 0 3.3370E-02 293.0 end h-1 418 0 6.6740E-02 293.0 end end comp read celldata latticecell squarepitch pitch= 1.25980 3 fueld=

0.78440 101 cladd= 0.91440 2 gapd= 0.80020 4 end latticecell squarepitch pitch= 1.25980 302 fueld=

0.78440 102 cladd= 0.91440 202 gapd= 0.80020 402 end latticecell squarepitch pitch= 1.25980 303 fueld=

0.78440 103 cladd= 0.91440 203 gapd= 0.80020 403 end latticecell squarepitch pitch= 1.25980 304 fueld=

0.78440 104 cladd= 0.91440 204 gapd= 0.80020 404 end latticecell squarepitch pitch= 1.25980 305 fueld=

0.78440 105 cladd= 0.91440 205 gapd= 0.80020 405 end latticecell squarepitch pitch= 1.25980 306 fueld=

0.78440 106 cladd= 0.91440 206 gapd= 0.80020 406 end latticecell squarepitch pitch= 1.25980 307 fueld=

0.78440 107 cladd= 0.91440 207 gapd= 0.80020 407 end latticecell squarepitch pitch= 1.25980 308 fueld=

0.78440 108 cladd= 0.91440 208 gapd= 0.80020 408 end latticecell squarepitch pitch= 1.25980 309 fueld=

0.78440 109 cladd= 0.91440 209 gapd= 0.80020 409 end latticecell squarepitch pitch= 1.25980 310 fueld=

0.78440 110 cladd= 0.91440 210 gapd= 0.80020 410 end latticecell squarepitch pitch= 1.25980 311 fueld=

0.78440 111 cladd= 0.91440 211 gapd= 0.80020 411 end latticecell squarepitch pitch= 1.25980 312 fueld=

0.78440 112 cladd= 0.91440 212 gapd= 0.80020 412 end latticecell squarepitch pitch= 1.25980 313 fueld=

0.78440 113 cladd= 0.91440 213 gapd= 0.80020 413 end

A-6 latticecell squarepitch pitch= 1.25980 314 fueld=

0.78440 114 cladd= 0.91440 214 gapd= 0.80020 414 end latticecell squarepitch pitch= 1.25980 315 fueld=

0.78440 115 cladd= 0.91440 215 gapd= 0.80020 415 end latticecell squarepitch pitch= 1.25980 316 fueld=

0.78440 116 cladd= 0.91440 216 gapd= 0.80020 416 end latticecell squarepitch pitch= 1.25980 317 fueld=

0.78440 117 cladd= 0.91440 217 gapd= 0.80020 417 end latticecell squarepitch pitch= 1.25980 318 fueld=

0.78440 118 cladd= 0.91440 218 gapd= 0.80020 418 end end celldata

' ********** GC-32: Generic 32-Assembly Cask

' * -GC-32 Characteristics-

'

• Basket Cell ID: 22.0 cm

'

• Basket Cell OD: 23.5 cm, basket wall thickness = 0.75 cm

'

• Boral Thickness: 0.2565 cm (0.101 in)

'

• Boral Width: 19.05 cm (7.5 in)

'

• Boral B-10 Loading: 0.0225 g/sqcm (75% of 0.030)

'

• Cask ID: 175.0 cm

'

• Cask OD: 215.0 cm

'

• Cask Top & Bottom Thickness: 30.0 cm

' * -Assembly Characteristics-

'

• Assembly Type: Westinghouse 17x17 OFA/V5

' ********** GC-32: Generic 32-Assembly Cask read param

'Parameters set to run until sig is reached htm=no gen=10000 nsk=100 npg=20000 sig=0.0001 end parm read geom unit 1 cylinder 101 1 0.3922 20.32 0 cylinder 4 1 0.4001 20.32 0 cylinder 2 1 0.4572 20.32 0 cuboid 3 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 2 cylinder 102 1 0.3922 20.32 0 cylinder 402 1 0.4001 20.32 0 cylinder 202 1 0.4572 20.32 0 cuboid 302 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 3 cylinder 103 1 0.3922 20.32 0 cylinder 403 1 0.4001 20.32 0 cylinder 203 1 0.4572 20.32 0 cuboid 303 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 4 cylinder 104 1 0.3922 20.32 0 cylinder 404 1 0.4001 20.32 0 cylinder 204 1 0.4572 20.32 0 cuboid 304 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 5 cylinder 105 1 0.3922 20.32 0 cylinder 405 1 0.4001 20.32 0 cylinder 205 1 0.4572 20.32 0 cuboid 305 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 6 cylinder 106 1 0.3922 20.32 0 cylinder 406 1 0.4001 20.32 0 cylinder 206 1 0.4572 20.32 0 cuboid 306 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 7 cylinder 107 1 0.3922 20.32 0 cylinder 407 1 0.4001 20.32 0 cylinder 207 1 0.4572 20.32 0 cuboid 307 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 8 cylinder 108 1 0.3922 20.32 0 cylinder 408 1 0.4001 20.32 0 cylinder 208 1 0.4572 20.32 0 cuboid 308 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 9 cylinder 109 1 0.3922 20.32 0 cylinder 409 1 0.4001 20.32 0 cylinder 209 1 0.4572 20.32 0 cuboid 309 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 10 cylinder 110 1 0.3922 20.32 0 cylinder 410 1 0.4001 20.32 0 cylinder 210 1 0.4572 20.32 0 cuboid 310 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 11 cylinder 111 1 0.3922 20.32 0 cylinder 411 1 0.4001 20.32 0 cylinder 211 1 0.4572 20.32 0 cuboid 311 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 12 cylinder 112 1 0.3922 20.32 0 cylinder 412 1 0.4001 20.32 0 cylinder 212 1 0.4572 20.32 0 cuboid 312 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 13 cylinder 113 1 0.3922 20.32 0 cylinder 413 1 0.4001 20.32 0 cylinder 213 1 0.4572 20.32 0

A-7 cuboid 313 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 14 cylinder 114 1 0.3922 20.32 0 cylinder 414 1 0.4001 20.32 0 cylinder 214 1 0.4572 20.32 0 cuboid 314 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 15 cylinder 115 1 0.3922 20.32 0 cylinder 415 1 0.4001 20.32 0 cylinder 215 1 0.4572 20.32 0 cuboid 315 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 16 cylinder 116 1 0.3922 20.32 0 cylinder 416 1 0.4001 20.32 0 cylinder 216 1 0.4572 20.32 0 cuboid 316 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 17 cylinder 117 1 0.3922 20.32 0 cylinder 417 1 0.4001 20.32 0 cylinder 217 1 0.4572 20.32 0 cuboid 317 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 18 cylinder 118 1 0.3922 20.32 0 cylinder 418 1 0.4001 20.32 0 cylinder 218 1 0.4572 20.32 0 cuboid 318 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

' Fuel Pin unit 19 array 2 -0.6299 -0.6299 0

' Fuel Pin

' Guide Thimble/Instrument Tube (assumed to be same) unit 20 cylinder 10 1 0.5613 365.76 0 cylinder 11 1 0.602 365.76 0 cuboid 10 1 0.6299 -0.6299 0.6299 -0.6299 365.76 0

' Top Half Horizontal Boral Panel unit 40 cuboid 8 1 9.525 -9.525 0.0254 0 365.76 0 cuboid 6 1 9.525 -9.525 0.12827 0 365.76 0

' Right-Hand Side Half Vertical Boral Panel unit 50 cuboid 8 1 0.0254 0 9.525 -9.525 365.76 0 cuboid 6 1 0.12827 0 9.525 -9.525 365.76 0

' Bottom Half Horizontal Boral Panel unit 60 cuboid 8 1 9.525 -9.525 0 -0.0254 365.76 0 cuboid 6 1 9.525 -9.525 0 -0.12827 365.76 0

' Left-Hand Side Half Vertical Boral Panel unit 70 cuboid 8 1 0 -0.0254 9.525 -9.525 365.76 0 cuboid 6 1 0 -0.12827 9.525 -9.525 365.76 0

' Assembly Basket Cell unit 101 array 1 -10.7086 -10.7086 0 cuboid 9 1 11 -11 11 -11 365.76 0 cuboid 5 1 11.75 -11.75 11.75 -11.75 365.76 0 cuboid 9 1 11.87827 -11.87827 11.87827 -

11.87827 365.76 0 hole 40 0 11.75 0 hole 50 11.75 0 0 hole 60 0 -11.75 0 hole 70 -11.75 0 0

' Top Boral/Basket Plate unit 110 cuboid 6 1 9.525 -9.525 0.10287 0 365.76 0 cuboid 8 1 9.525 -9.525 0.12827 0 365.76 0 cuboid 9 1 11.75 -11.75 0.12827 0 365.76 0 cuboid 5 1 11.75 -11.75 0.87827 0 365.76 0

' Left-Hand Side Boral/Basket Plate unit 112 cuboid 6 1 0 -0.10287 9.525 -9.525 365.76 0 cuboid 8 1 0 -0.12827 9.525 -9.525 365.76 0 cuboid 9 1 0 -0.12827 10.9999 -10.9999 365.76 0 cuboid 5 1 0 -0.87827 10.9999 -10.9999 365.76 0

' Right-Hand Side Boral/Basket Plate unit 113 cuboid 6 1 0.10287 0 9.525 -9.525 365.76 0 cuboid 8 1 0.12827 0 9.525 -9.525 365.76 0 cuboid 9 1 0.12827 0 10.9999 -10.9999 365.76 0 cuboid 5 1 0.87827 0 10.9999 -10.9999 365.76 0 unit 114 array 4 -47.51308 0 0

' Cask Inner Volume global unit 200 array 3 -71.26962 0 0 zhemicyl+y 9 1 87.5 375.76 -5 hole 114 0 47.51309 0

' Exterior Half Boral Panels

' Top Plates hole 110 -35.63481 71.26964 0 hole 110 -11.87827 71.26964 0 hole 110 11.87827 71.26964 0 hole 110 35.63481 71.26964 0 hole 110 59.39135 47.51309 0 hole 110 -59.39135 47.51309 0

' Left-Hand Side Plates

A-8 hole 112 -47.51310 59.39135 0 hole 112 -71.26964 35.63481 0 hole 112 -71.26964 11.87827 0

' Right-Hand Side Plates hole 113 47.51310 59.39135 0 hole 113 71.26964 35.63481 0 hole 113 71.26964 11.87827 0 zhemicyl+y 9 1 87.5 385.76 -10 zhemicyl+y 9 1 87.5 395.76 -15

' Steel Cask/Overpack zhemicyl+y 7 1 92.5 400.76 -25 zhemicyl+y 7 1 97.5 405.76 -35 zhemicyl+y 7 1 107.5 415.76 -45

' Cube Surrounding Cask cuboid 0 1 108 -108 108 0 415.76 -45 end geom

' Assembly Type: Westinghouse 17x17 OFA/V5 read array ara= 1 nux= 17 nuy= 17 nuz= 1 fill 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 end fill ara=2 nux=1 nuy=1 nuz=18 fill 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 end fill ara=3 nux=6 nuy=2 nuz=1 fill 12r101 end fill ara=4 nux=4 nuy=1 nuz=1 fill 4r101 end fill end array read bounds

+xb=specular

-xb=specular

+yb=specular

+zb=void

-zb=void

' the -yb must be kept at specular because only half of cask is modeled

-yb=specular end bounds read start nst=7 xsm=-72 xsp=72 ysm=0 ysp=72 zsm=0 zsp=365.76 end start end data end A.5 TSUNAMI-3D Note: only three representative compositions are listed

=tsunami-3d gbc-32 w17x17, 5 wt% U-235, 40 gwd/mtu v7.1-252 read comp o-16 101 0 4.6833E-02 293.0 end u-234 101 0 1.3549E-07 293.0 end u-235 101 0 6.1725E-04 293.0 end u-236 101 0 1.0874E-04 293.0 end u-238 101 0 2.1874E-02 293.0 end np-237 101 0 8.1160E-06 293.0 end pu-238 101 0 1.7839E-06 293.0 end pu-239 101 0 1.5492E-04 293.0 end pu-240 101 0 3.6209E-05 293.0 end pu-241 101 0 1.8555E-05 293.0 end pu-242 101 0 3.9047E-06 293.0 end am-241 101 0 5.5839E-06 293.0 end am-243 101 0 4.9870E-07 293.0 end mo-95 101 0 3.7790E-05 293.0 end tc-99 101 0 3.6942E-05 293.0 end ru-101 101 0 3.4032E-05 293.0 end rh-103 101 0 2.1724E-05 293.0 end ag-109 101 0 2.2050E-06 293.0 end cs-133 101 0 3.9797E-05 293.0 end nd-143 101 0 3.0918E-05 293.0 end nd-145 101 0 2.2321E-05 293.0 end sm-147 101 0 7.5994E-06 293.0 end sm-149 101 0 1.9538E-07 293.0 end sm-150 101 0 8.2052E-06 293.0 end sm-151 101 0 5.8471E-07 293.0 end sm-152 101 0 3.1331E-06 293.0 end eu-151 101 0 2.3841E-08 293.0 end eu-153 101 0 2.8049E-06 293.0 end gd-155 101 0 8.7674E-08 293.0 end o-16 109 0 4.6831E-02 293.0 end u-234 109 0 3.2722E-07 293.0 end u-235 109 0 3.7098E-04 293.0 end u-236 109 0 1.4766E-04 293.0 end u-238 109 0 2.1561E-02 293.0 end

A-9 np-237 109 0 1.5667E-05 293.0 end pu-238 109 0 5.8868E-06 293.0 end pu-239 109 0 1.7632E-04 293.0 end pu-240 109 0 6.0919E-05 293.0 end pu-241 109 0 3.3938E-05 293.0 end pu-242 109 0 1.3337E-05 293.0 end am-241 109 0 1.0062E-05 293.0 end am-243 109 0 2.7628E-06 293.0 end mo-95 109 0 5.9768E-05 293.0 end tc-99 109 0 5.8643E-05 293.0 end ru-101 109 0 5.6202E-05 293.0 end rh-103 109 0 3.4600E-05 293.0 end ag-109 109 0 4.6966E-06 293.0 end cs-133 109 0 6.2075E-05 293.0 end nd-143 109 0 4.4687E-05 293.0 end nd-145 109 0 3.4426E-05 293.0 end sm-147 109 0 9.7756E-06 293.0 end sm-149 109 0 2.4441E-07 293.0 end sm-150 109 0 1.4884E-05 293.0 end sm-151 109 0 6.9160E-07 293.0 end sm-152 109 0 4.4997E-06 293.0 end eu-151 109 0 2.7710E-08 293.0 end eu-153 109 0 5.3299E-06 293.0 end gd-155 109 0 2.0017E-07 293.0 end o-16 118 0 4.6834E-02 293.0 end u-234 118 0 8.0427E-08 293.0 end u-235 118 0 7.5312E-04 293.0 end u-236 118 0 8.4556E-05 293.0 end u-238 118 0 2.1998E-02 293.0 end np-237 118 0 5.0472E-06 293.0 end pu-238 118 0 7.7902E-07 293.0 end pu-239 118 0 1.3420E-04 293.0 end pu-240 118 0 2.4036E-05 293.0 end pu-241 118 0 1.0953E-05 293.0 end pu-242 118 0 1.5419E-06 293.0 end am-241 118 0 3.3105E-06 293.0 end am-243 118 0 1.3856E-07 293.0 end mo-95 118 0 2.7499E-05 293.0 end tc-99 118 0 2.6779E-05 293.0 end ru-101 118 0 2.4236E-05 293.0 end rh-103 118 0 1.5589E-05 293.0 end ag-109 118 0 1.3080E-06 293.0 end cs-133 118 0 2.9038E-05 293.0 end nd-143 118 0 2.3292E-05 293.0 end nd-145 118 0 1.6407E-05 293.0 end sm-147 118 0 6.0408E-06 293.0 end sm-149 118 0 1.7413E-07 293.0 end sm-150 118 0 5.5401E-06 293.0 end sm-151 118 0 5.3031E-07 293.0 end sm-152 118 0 2.3390E-06 293.0 end eu-151 118 0 2.1969E-08 293.0 end eu-153 118 0 1.7441E-06 293.0 end gd-155 118 0 5.1627E-08 293.0 end zr-90 2 0 2.1763E-02 293.0 end zr-91 2 0 4.7461E-03 293.0 end zr-92 2 0 7.2544E-03 293.0 end zr-94 2 0 7.3517E-03 293.0 end zr-96 2 0 1.1844E-03 293.0 end o-16 3 0 3.3370E-02 293.0 end h-1 3 0 6.6740E-02 293.0 end o-16 4 0 3.3370E-02 293.0 end h-1 4 0 6.6740E-02 293.0 end cr-50 5 0 7.5733E-04 293.0 end cr-52 5 0 1.4605E-02 293.0 end cr-53 5 0 1.6558E-03 293.0 end cr-54 5 0 4.1222E-04 293.0 end mn-55 5 0 1.7400E-03 293.0 end fe-54 5 0 3.5022E-03 293.0 end fe-56 5 0 5.4445E-02 293.0 end fe-57 5 0 1.2466E-03 293.0 end fe-58 5 0 1.6621E-04 293.0 end ni-58 5 0 5.2704E-03 293.0 end ni-60 5 0 2.0149E-03 293.0 end ni-61 5 0 8.7236E-05 293.0 end ni-62 5 0 2.7715E-04 293.0 end ni-64 5 0 7.0252E-05 293.0 end b-10 6 0 6.5795E-03 293.0 end b-11 6 0 2.7260E-02 293.0 end c 6 0 8.4547E-03 293.0 end al-27 6 0 4.1795E-02 293.0 end cr-50 7 0 7.5733E-04 293.0 end cr-52 7 0 1.4605E-02 293.0 end cr-53 7 0 1.6558E-03 293.0 end cr-54 7 0 4.1222E-04 293.0 end mn-55 7 0 1.7400E-03 293.0 end fe-54 7 0 3.5022E-03 293.0 end fe-56 7 0 5.4445E-02 293.0 end fe-57 7 0 1.2466E-03 293.0 end fe-58 7 0 1.6621E-04 293.0 end ni-58 7 0 5.2704E-03 293.0 end ni-60 7 0 2.0149E-03 293.0 end ni-61 7 0 8.7236E-05 293.0 end ni-62 7 0 2.7715E-04 293.0 end ni-64 7 0 7.0252E-05 293.0 end al-27 8 0 6.0200E-02 293.0 end o-16 9 0 3.3370E-02 293.0 end h-1 9 0 6.6740E-02 293.0 end o-16 10 0 3.3370E-02 293.0 end h-1 10 0 6.6740E-02 293.0 end zr-90 11 0 2.1763E-02 293.0 end zr-91 11 0 4.7461E-03 293.0 end zr-92 11 0 7.2544E-03 293.0 end zr-94 11 0 7.3517E-03 293.0 end zr-96 11 0 1.1844E-03 293.0 end zr-90 209 0 2.1763E-02 293.0 end zr-91 209 0 4.7461E-03 293.0 end zr-92 209 0 7.2544E-03 293.0 end zr-94 209 0 7.3517E-03 293.0 end zr-96 209 0 1.1844E-03 293.0 end o-16 309 0 3.3370E-02 293.0 end h-1 309 0 6.6740E-02 293.0 end o-16 409 0 3.3370E-02 293.0 end

A-10 h-1 409 0 6.6740E-02 293.0 end zr-90 218 0 2.1763E-02 293.0 end zr-91 218 0 4.7461E-03 293.0 end zr-92 218 0 7.2544E-03 293.0 end zr-94 218 0 7.3517E-03 293.0 end zr-96 218 0 1.1844E-03 293.0 end o-16 318 0 3.3370E-02 293.0 end h-1 318 0 6.6740E-02 293.0 end o-16 418 0 3.3370E-02 293.0 end h-1 418 0 6.6740E-02 293.0 end end comp read celldata latticecell squarepitch pitch= 1.25980 3 fueld=

0.78440 101 cladd= 0.91440 2 gapd= 0.80020 4 end latticecell squarepitch pitch= 1.25980 302 fueld=

0.78440 102 cladd= 0.91440 202 gapd= 0.80020 402 end latticecell squarepitch pitch= 1.25980 303 fueld=

0.78440 103 cladd= 0.91440 203 gapd= 0.80020 403 end latticecell squarepitch pitch= 1.25980 304 fueld=

0.78440 104 cladd= 0.91440 204 gapd= 0.80020 404 end latticecell squarepitch pitch= 1.25980 305 fueld=

0.78440 105 cladd= 0.91440 205 gapd= 0.80020 405 end latticecell squarepitch pitch= 1.25980 306 fueld=

0.78440 106 cladd= 0.91440 206 gapd= 0.80020 406 end latticecell squarepitch pitch= 1.25980 307 fueld=

0.78440 107 cladd= 0.91440 207 gapd= 0.80020 407 end latticecell squarepitch pitch= 1.25980 308 fueld=

0.78440 108 cladd= 0.91440 208 gapd= 0.80020 408 end latticecell squarepitch pitch= 1.25980 309 fueld=

0.78440 109 cladd= 0.91440 209 gapd= 0.80020 409 end latticecell squarepitch pitch= 1.25980 310 fueld=

0.78440 110 cladd= 0.91440 210 gapd= 0.80020 410 end latticecell squarepitch pitch= 1.25980 311 fueld=

0.78440 111 cladd= 0.91440 211 gapd= 0.80020 411 end latticecell squarepitch pitch= 1.25980 312 fueld=

0.78440 112 cladd= 0.91440 212 gapd= 0.80020 412 end latticecell squarepitch pitch= 1.25980 313 fueld=

0.78440 113 cladd= 0.91440 213 gapd= 0.80020 413 end latticecell squarepitch pitch= 1.25980 314 fueld=

0.78440 114 cladd= 0.91440 214 gapd= 0.80020 414 end latticecell squarepitch pitch= 1.25980 315 fueld=

0.78440 115 cladd= 0.91440 215 gapd= 0.80020 415 end latticecell squarepitch pitch= 1.25980 316 fueld=

0.78440 116 cladd= 0.91440 216 gapd= 0.80020 416 end latticecell squarepitch pitch= 1.25980 317 fueld=

0.78440 117 cladd= 0.91440 217 gapd= 0.80020 417 end latticecell squarepitch pitch= 1.25980 318 fueld=

0.78440 118 cladd= 0.91440 218 gapd= 0.80020 418 end end celldata

' ********** GC-32: Generic 32-Assembly Cask

' * -GC-32 Characteristics-

'

• Basket Cell ID: 22.0 cm

'

• Basket Cell OD: 23.5 cm, basket wall thickness =

0.75 cm

'

• Boral Thickness: 0.2565 cm (0.101 in)

'

• Boral Width: 19.05 cm (7.5 in)

'

• Boral B-10 Loading: 0.0225 g/sqcm (75% of 0.030)

'

• Cask ID: 175.0 cm

'

• Cask OD: 215.0 cm

'

• Cask Top & Bottom Thickness: 30.0 cm

' * -Assembly Characteristics-

'

• Assembly Type: Westinghouse 17x17 OFA/V5

' ********** GC-32: Generic 32-Assembly Cask read param gen=10000 nsk=100 npg=10000 sig=0.0001 htm=no

' tsunami parameters:

agn=5100 apg=100000 ask=100 asg=0.001 tfm=no pnm=3 mfx=yes end parm read geom unit 1 cylinder 101 1 0.3922 20.32 0 cylinder 4 1 0.4001 20.32 0 cylinder 2 1 0.4572 20.32 0 cuboid 3 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0 unit 2 cylinder 102 1 0.3922 20.32 0

A-11 cylinder 402 1 0.4001 20.32 0 cylinder 202 1 0.4572 20.32 0 cuboid 302 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 3 cylinder 103 1 0.3922 20.32 0 cylinder 403 1 0.4001 20.32 0 cylinder 203 1 0.4572 20.32 0 cuboid 303 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 4 cylinder 104 1 0.3922 20.32 0 cylinder 404 1 0.4001 20.32 0 cylinder 204 1 0.4572 20.32 0 cuboid 304 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 5 cylinder 105 1 0.3922 20.32 0 cylinder 405 1 0.4001 20.32 0 cylinder 205 1 0.4572 20.32 0 cuboid 305 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 6 cylinder 106 1 0.3922 20.32 0 cylinder 406 1 0.4001 20.32 0 cylinder 206 1 0.4572 20.32 0 cuboid 306 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 7 cylinder 107 1 0.3922 20.32 0 cylinder 407 1 0.4001 20.32 0 cylinder 207 1 0.4572 20.32 0 cuboid 307 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 8 cylinder 108 1 0.3922 20.32 0 cylinder 408 1 0.4001 20.32 0 cylinder 208 1 0.4572 20.32 0 cuboid 308 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 9 cylinder 109 1 0.3922 20.32 0 cylinder 409 1 0.4001 20.32 0 cylinder 209 1 0.4572 20.32 0 cuboid 309 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 10 cylinder 110 1 0.3922 20.32 0 cylinder 410 1 0.4001 20.32 0 cylinder 210 1 0.4572 20.32 0 cuboid 310 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 11 cylinder 111 1 0.3922 20.32 0 cylinder 411 1 0.4001 20.32 0 cylinder 211 1 0.4572 20.32 0 cuboid 311 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 12 cylinder 112 1 0.3922 20.32 0 cylinder 412 1 0.4001 20.32 0 cylinder 212 1 0.4572 20.32 0 cuboid 312 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 13 cylinder 113 1 0.3922 20.32 0 cylinder 413 1 0.4001 20.32 0 cylinder 213 1 0.4572 20.32 0 cuboid 313 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 14 cylinder 114 1 0.3922 20.32 0 cylinder 414 1 0.4001 20.32 0 cylinder 214 1 0.4572 20.32 0 cuboid 314 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 15 cylinder 115 1 0.3922 20.32 0 cylinder 415 1 0.4001 20.32 0 cylinder 215 1 0.4572 20.32 0 cuboid 315 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 16 cylinder 116 1 0.3922 20.32 0 cylinder 416 1 0.4001 20.32 0 cylinder 216 1 0.4572 20.32 0 cuboid 316 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 17 cylinder 117 1 0.3922 20.32 0 cylinder 417 1 0.4001 20.32 0 cylinder 217 1 0.4572 20.32 0 cuboid 317 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

unit 18 cylinder 118 1 0.3922 20.32 0 cylinder 418 1 0.4001 20.32 0 cylinder 218 1 0.4572 20.32 0 cuboid 318 1 0.6299 -0.6299 0.6299 -0.6299 20.32 0

' Fuel Pin unit 19 array 2 -0.6299 -0.6299 0

' Fuel Pin

' Guide Thimble/Instrument Tube (assumed to be same) unit 20 cylinder 10 1 0.5613 365.76 0 cylinder 11 1 0.602 365.76 0

A-12 cuboid 10 1 0.6299 -0.6299 0.6299 -0.6299 365.76 0

' Top Half Horizontal Boral Panel unit 40 cuboid 8 1 9.525 -9.525 0.0254 0 365.76 0 cuboid 6 1 9.525 -9.525 0.12827 0 365.76 0

' Right-Hand Side Half Vertical Boral Panel unit 50 cuboid 8 1 0.0254 0 9.525 -9.525 365.76 0 cuboid 6 1 0.12827 0 9.525 -9.525 365.76 0

' Bottom Half Horizontal Boral Panel unit 60 cuboid 8 1 9.525 -9.525 0 -0.0254 365.76 0 cuboid 6 1 9.525 -9.525 0 -0.12827 365.76 0

' Left-Hand Side Half Vertical Boral Panel unit 70 cuboid 8 1 0 -0.0254 9.525 -9.525 365.76 0 cuboid 6 1 0 -0.12827 9.525 -9.525 365.76 0

' Assembly Basket Cell unit 101 array 1 -10.7086 -10.7086 0 cuboid 9 1 11 -11 11 -11 365.76 0 cuboid 5 1 11.75 -11.75 11.75 -11.75 365.76 0 cuboid 9 1 11.87827 -11.87827 11.87827 -

11.87827 365.76 0 hole 40 0 11.75 0 hole 50 11.75 0 0 hole 60 0 -11.75 0 hole 70 -11.75 0 0

' Top Boral/Basket Plate unit 110 cuboid 6 1 9.525 -9.525 0.10287 0 365.76 0 cuboid 8 1 9.525 -9.525 0.12827 0 365.76 0 cuboid 9 1 11.75 -11.75 0.12827 0 365.76 0 cuboid 5 1 11.75 -11.75 0.87827 0 365.76 0

' Left-Hand Side Boral/Basket Plate unit 112 cuboid 6 1 0 -0.10287 9.525 -9.525 365.76 0 cuboid 8 1 0 -0.12827 9.525 -9.525 365.76 0 cuboid 9 1 0 -0.12827 10.9999 -10.9999 365.76 0 cuboid 5 1 0 -0.87827 10.9999 -10.9999 365.76 0

' Right-Hand Side Boral/Basket Plate unit 113 cuboid 6 1 0.10287 0 9.525 -9.525 365.76 0 cuboid 8 1 0.12827 0 9.525 -9.525 365.76 0 cuboid 9 1 0.12827 0 10.9999 -10.9999 365.76 0 cuboid 5 1 0.87827 0 10.9999 -10.9999 365.76 0 unit 114 array 4 -47.51308 0 0

' Cask Inner Volume global unit 200 array 3 -71.26962 0 0 zhemicyl+y 9 1 87.5 375.76 -5 hole 114 0 47.51309 0

' Exterior Half Boral Panels

' Top Plates hole 110 -35.63481 71.26964 0 hole 110 -11.87827 71.26964 0 hole 110 11.87827 71.26964 0 hole 110 35.63481 71.26964 0 hole 110 59.39135 47.51309 0 hole 110 -59.39135 47.51309 0

' Left-Hand Side Plates hole 112 -47.51310 59.39135 0 hole 112 -71.26964 35.63481 0 hole 112 -71.26964 11.87827 0

' Right-Hand Side Plates hole 113 47.51310 59.39135 0 hole 113 71.26964 35.63481 0 hole 113 71.26964 11.87827 0 zhemicyl+y 9 1 87.5 385.76 -10 zhemicyl+y 9 1 87.5 395.76 -15

' Steel Cask/Overpack zhemicyl+y 7 1 92.5 400.76 -25 zhemicyl+y 7 1 97.5 405.76 -35 zhemicyl+y 7 1 107.5 415.76 -45

' Cube Surrounding Cask cuboid 0 1 108 -108 108 0 415.76 -45 end geom

' Assembly Type: Westinghouse 17x17 OFA/V5 read array ara= 1 nux= 17 nuy= 17 nuz= 1 fill 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 20 19 19 20 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 19 20 19 19 19 19 19 19 19 19 20 19 19 20 19 19 20 19 19 19 19 19

A-13 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 end fill ara=2 nux=1 nuy=1 nuz=18 fill 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 end fill ara=3 nux=6 nuy=2 nuz=1 fill 12r101 end fill ara=4 nux=4 nuy=1 nuz=1 fill 4r101 end fill end array read bounds

+xb=specular

-xb=specular

+yb=specular

+zb=void

-zb=void

' the -yb must be kept at specular because only half of cask is modeled

-yb=specular end bounds read grid

' the fine mesh grid below provides 108 intervals in the x-direction,

' 54 intervals in the y-direction and 28 intervals in the z-direction.

' total mesh regions = 163,296.

1 xplanes=0.0

-0.12827 -0.87827 -1.16997 -2.42977 -4.94937 -

7.46897 -9.98857

-12.50817 -15.02777 -17.54737 -20.06697 -

22.58657 -22.87827 -23.62827

-23.75654 -23.88481 -24.63481 -24.92651 -

26.18631 -28.70591 -31.22551

-33.74511 -36.26471 -38.78431 -41.30391 -

43.82351 -46.34311 -46.63481

-47.38481 -47.51308 -47.64135 -48.39135 -

48.68305 -49.94285 -52.46245

-54.98205 -57.50165 -60.02125 -62.54085 -

65.06045 -67.58005 -70.09965

-70.39135 -71.14135 -71.26962 -71.39789 -

72.14789 -75.0 -80.0

-85.0 -90.0 -95.0 -100.0 -108.0 0.12827 0.87827 1.16997 2.42977 4.94937 7.46897 9.98857 12.50817 15.02777 17.54737 20.06697 22.58657 22.87827 23.62827 23.75654 23.88481 24.63481 24.92651 26.18631 28.70591 31.22551 33.74511 36.26471 38.78431 41.30391 43.82351 46.34311 46.63481 47.38481 47.51308 47.64135 48.39135 48.68305 49.94285 52.46245 54.98205 57.50165 60.02125 62.54085 65.06045 67.58005 70.09965 70.39135 71.14135 71.26962 71.39789 72.14789 75.0 80.0 85.0 90.0 95.0 100.0 108.0 end yplanes=0.0 0.12827 0.87827 1.16997 2.42977 4.94937 7.46897 9.98857 12.50817 15.02777 17.54737 20.06697 22.58657 22.87827 23.62827 23.75654 23.88481 24.63481 24.92651 26.18631 28.70591 31.22551 33.74511 36.26471 38.78431 41.30391 43.82351 46.34311 46.63481 47.38481 47.51308 47.64135 48.39135 48.68305 49.94285 52.46245 54.98205 57.50165 60.02125 62.54085 65.06045 67.58005 70.09965 70.39135 71.14135 71.26962 71.39789 72.14789 75.0 80.0 85.0 90.0 95.0 100.0 108.0 end zplanes=0.0

-45.0 -20.0 -15.0 -5.0 0.0 20.32 40.64 60.96 81.28 101.6 121.92 142.24 162.56 182.88 203.20 223.52 243.84 264.16 284.48 304.80 325.12 345.44 365.76 370.00 380.00 395.00 400.00 416.00 end end grid read volume

'1 billion points used for monte carlo determination of mesh volumes.

type=random points=1000000 batches=1000 xp=108.1 xm=-108.1 yp=108.1 ym=-0.1 zp=417 zm=-46.0 end volume end data read sams nohtml pn=5 prtimp nocovar end sams end A.6 TSUNAMI-IP

=tsunami-ip Title Card - 1 Run read parameter cov_fix udcov_therm=0.05 udcov_inter=0.15 udcov_fast=0.40 use_dcov

A-14 usename html uncert uncert_long c

values end parameter read apps application.sdf end apps read exps LEU-COMP-THERM-001-001.sdf end exps End A.7 VADER

=vader data=[-1.57690 0.99771 0.00309

-1.70141 0.99937 0.00310

-1.75618 0.99920 0.00310

-1.74052 0.99870 0.00280

-1.76903 0.99984 0.00280

-1.77812 0.99839 0.00280

-1.79458 0.99877 0.00280

-1.80422 0.99763 0.00279

-1.81340 0.99767 0.00279

-1.81992 0.99688 0.00279

-1.75496 0.99896 0.00280

-1.78271 0.99981 0.00280

-1.82794 0.99779 0.00279

-0.96730 0.99582 0.00279

-1.12013 0.99781 0.00279

-1.25122 0.99847 0.00280

-1.36349 0.99881 0.00280

-2.65026 1.00009 0.00180

-2.69367 1.00001 0.00180

-2.69623 1.00143 0.00181

-2.45020 0.99976 0.00180

-2.47520 0.99972 0.00180

-2.48393 0.99930 0.00180

-2.26386 0.99948 0.00180

-2.27911 0.99928 0.00180

-2.28902 0.99918 0.00180

-1.94575 0.99949 0.00180

-1.98020 0.99873 0.00180

-1.99398 0.99873 0.00180

-1.34433 0.99850 0.00180

-1.43546 0.99835 0.00180

-1.45207 0.99826 0.00180

-2.27141 0.99902 0.00180

-2.29218 0.99914 0.00180

-2.26931 0.99631 0.00180

-1.38065 0.99909 0.00247

-1.39110 0.99792 0.00247

-1.34864 0.99832 0.00247

-1.31575 0.99916 0.00247

-1.27555 0.99916 0.00247

-1.28806 0.99927 0.00247

-1.25759 1.00044 0.00247

-1.23133 0.99943 0.00247

-1.77732 1.00003 0.00247

-1.81003 0.99867 0.00247

-1.83745 1.00002 0.00247

-1.87906 1.00013 0.00247

-1.94832 1.00208 0.00248

-2.25517 1.00305 0.00248

-2.21659 1.00154 0.00248

-2.18704 1.00349 0.00248

-2.13962 0.99479 0.00246

-2.23968 1.00005 0.00247

-2.39534 0.99295 0.00245

-2.43595 0.99669 0.00246

-1.35840 0.99852 0.00499

-1.37228 0.99832 0.00499

-1.29097 0.99834 0.00499

-1.30286 0.99829 0.00499

-1.31701 0.99823 0.00499

-1.23903 0.99733 0.00499

-1.26305 0.99781 0.00499

-1.20276 0.99692 0.00499

-1.21551 0.99717 0.00499

-1.85387 0.99888 0.00500

-1.86541 0.99896 0.00500

-1.89914 0.99998 0.00500

-1.91952 0.99998 0.00500

-2.22461 1.00094 0.00501

-2.41577 1.00106 0.00501

-2.22240 1.00073 0.00500

-2.07254 0.99721 0.00321

-1.94261 0.99924 0.00322

-2.03007 0.99696 0.00321

-2.07124 0.99698 0.00321

-1.98583 0.99622 0.00321

-2.02476 0.99961 0.00322

-2.03381 0.99548 0.00321

-1.95664 1.00268 0.00323

-1.99765 0.99591 0.00321

-2.02577 0.99593 0.00321

-1.96921 0.99522 0.00321

-2.16675 0.99892 0.00254

-2.17640 0.99892 0.00254

-2.17548 0.99921 0.00254

-2.18211 0.99922 0.00254

-2.18703 0.99920 0.00254

-2.20534 0.99943 0.00254

A-15

-2.22328 0.99961 0.00254

-2.24482 1.00022 0.00254

-2.26634 1.00025 0.00254

-2.24067 0.99945 0.00254

-2.21696 1.00000 0.00254

-2.14636 0.99854 0.00254

-1.87627 0.99865 0.00254

-2.04936 0.99949 0.00254

-2.14347 0.99893 0.00254

-1.85595 1.00091 0.00483

-1.87657 0.99748 0.00328

-1.90104 0.99700 0.00328

-1.93224 0.99733 0.00328

-1.95397 0.99753 0.00220

-1.96191 0.99649 0.00219

-1.97087 0.99600 0.00219

-1.97952 0.99623 0.00219

-1.98490 0.99565 0.00219

-1.99006 0.99740 0.00220

-2.00656 0.99780 0.00293

-1.98125 1.00049 0.00376

-1.99215 0.99598 0.00214

-1.88224 1.00347 0.00618

-1.92552 0.99723 0.00405

-1.97152 0.99706 0.00190

-1.97972 0.99511 0.00189

-1.99060 0.99489 0.00189

-1.99875 0.99439 0.00189

-2.00492 0.99431 0.00189

-1.99373 0.99770 0.00237

-1.73034 0.99903 0.00432

-1.77661 0.99953 0.00432

-1.82649 0.99874 0.00470

-1.83504 0.99858 0.00469

-1.84427 0.99851 0.00469

-1.85075 0.99810 0.00469

-1.85760 0.99821 0.00469

-1.91944 0.99856 0.00469

-1.99101 0.99918 0.00090

-2.03348 0.99891 0.00090

-2.05988 0.99905 0.00090

-2.07630 0.99896 0.00090

-1.86170 1.00072 0.00499

-1.88261 0.99801 0.00305

-1.90846 0.99797 0.00305

-1.93358 0.99833 0.00305

-1.95392 0.99827 0.00318

-1.99076 0.99831 0.00318

-2.00630 0.99867 0.00318

-1.96214 0.99759 0.00232

-1.96938 0.99698 0.00232

-1.97631 0.99680 0.00231

-1.98138 0.99657 0.00231

-1.98444 1.00059 0.00511

-1.98508 1.00056 0.00511

-1.99023 0.99968 0.00511

-2.01941 0.99605 0.00246

-1.99665 0.99623 0.00246

-1.88879 1.00457 0.00482

-1.93141 0.99802 0.00226

-1.97871 0.99809 0.00245

-2.00070 0.99899 0.00245

-1.98715 0.99684 0.00244

-1.99581 0.99639 0.00244

-2.00270 0.99609 0.00218

-2.00877 0.99584 0.00218

-2.01957 0.99854 0.00180

-2.03800 0.99914 0.00180

-1.73671 0.99880 0.00581

-1.78699 0.99945 0.00582

-1.83634 0.99908 0.00582

-1.92898 0.99966 0.00582

-1.99430 0.99983 0.00136

-2.03639 1.00012 0.00136

-2.06248 1.00015 0.00136

-2.07863 1.00029 0.00136

-1.84679 0.99956 0.00493

-1.85426 0.99962 0.00493

-1.86148 0.99966 0.00493

-1.86702 0.99926 0.00493

]

trend_values=[-1.24122

]

tests{

anderson_darling{confidence=0.95}

t-test{}

}

methods{

USL1{ fit_confidence=0.95 admin_margin=0 extrapolate=no}

CR6698{f_p=0.95 normal_p=0.95 chi_p=0.95 weighted=no}

parametric{ proportion_of_pop=0.95 confidence=0.95 weighted=no}

nonparametric{ proportion_of_pop=0.95 confidence=0.95 use_npm=yes weighted=no}

}

end

APPENDIX B BENCHMARK EXPERIMENTS The list of all 2,104 critical experiments considered in performing the similarity assessments is provided in Table B-1, regardless of the source of the sensitivity data used. The sources of the evaluations are either the International Criticality Safety Benchmark Evaluation Project Handbook [13] or the proprietary HTC evaluation reports generated by IRSN [16-19].

Table B-1 Benchmark Experiments Used in Similarity Assessments B-1 IEU-COMP-THERM-001-002 LEU-COMP-THERM-033-009 LEU-COMP-THERM-101-010 IEU-COMP-THERM-001-003 LEU-COMP-THERM-033-010 LEU-COMP-THERM-101-011 IEU-COMP-THERM-001-004 LEU-COMP-THERM-033-011 LEU-COMP-THERM-101-012 IEU-COMP-THERM-001-005 LEU-COMP-THERM-033-012 LEU-COMP-THERM-101-013 IEU-COMP-THERM-001-006 LEU-COMP-THERM-033-013 LEU-COMP-THERM-101-014 IEU-COMP-THERM-001-007 LEU-COMP-THERM-033-014 LEU-COMP-THERM-101-015 IEU-COMP-THERM-001-008 LEU-COMP-THERM-033-015 LEU-COMP-THERM-101-016 IEU-COMP-THERM-001-009 LEU-COMP-THERM-033-016 LEU-COMP-THERM-101-017 IEU-COMP-THERM-001-010 LEU-COMP-THERM-033-017 LEU-COMP-THERM-101-018 IEU-COMP-THERM-001-011 LEU-COMP-THERM-033-018 LEU-COMP-THERM-101-019 IEU-COMP-THERM-001-012 LEU-COMP-THERM-033-019 LEU-COMP-THERM-101-020 IEU-COMP-THERM-001-013 LEU-COMP-THERM-033-020 LEU-COMP-THERM-101-021 IEU-COMP-THERM-001-014 LEU-COMP-THERM-033-021 LEU-COMP-THERM-101-022 IEU-COMP-THERM-001-015 LEU-COMP-THERM-033-022 LEU-MET-THERM-001-001 IEU-COMP-THERM-001-016 LEU-COMP-THERM-033-023 LEU-MET-THERM-002-001 IEU-COMP-THERM-001-018 LEU-COMP-THERM-033-024 LEU-MET-THERM-002-002 IEU-COMP-THERM-001-019 LEU-COMP-THERM-033-025 LEU-MET-THERM-002-003 IEU-COMP-THERM-001-020 LEU-COMP-THERM-033-026 LEU-MET-THERM-002-004 IEU-COMP-THERM-001-023 LEU-COMP-THERM-033-027 LEU-MET-THERM-002-005 IEU-COMP-THERM-001-024 LEU-COMP-THERM-033-028 LEU-MET-THERM-002-006 IEU-COMP-THERM-001-025 LEU-COMP-THERM-033-029 LEU-MET-THERM-002-007 IEU-COMP-THERM-001-026 LEU-COMP-THERM-033-030 LEU-MET-THERM-002-008 IEU-COMP-THERM-001-027 LEU-COMP-THERM-033-031 LEU-MET-THERM-002-009 IEU-COMP-THERM-001-028 LEU-COMP-THERM-033-032 LEU-MET-THERM-002-010 IEU-COMP-THERM-001-029 LEU-COMP-THERM-033-033 LEU-MET-THERM-002-011

B-2 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

IEU-COMP-THERM-002-001 LEU-COMP-THERM-033-034 LEU-MET-THERM-002-012 IEU-COMP-THERM-002-002 LEU-COMP-THERM-033-035 LEU-MET-THERM-004-001 IEU-COMP-THERM-002-003 LEU-COMP-THERM-033-036 LEU-MET-THERM-004-002 IEU-COMP-THERM-002-004 LEU-COMP-THERM-033-037 LEU-MET-THERM-004-003 IEU-COMP-THERM-002-005 LEU-COMP-THERM-033-038 LEU-MET-THERM-004-004 IEU-COMP-THERM-002-006 LEU-COMP-THERM-033-039 LEU-MET-THERM-004-005 IEU-COMP-THERM-011-013 LEU-COMP-THERM-033-040 LEU-MET-THERM-004-006 IEU-COMP-THERM-011-014 LEU-COMP-THERM-033-041 LEU-MET-THERM-004-007 IEU-COMP-THERM-015-002 LEU-COMP-THERM-033-042 LEU-MET-THERM-004-008 IEU-COMP-THERM-015-003 LEU-COMP-THERM-033-043 LEU-MET-THERM-006-001 IEU-COMP-THERM-015-006 LEU-COMP-THERM-033-044 LEU-MET-THERM-006-002 IEU-COMP-THERM-015-007 LEU-COMP-THERM-033-045 LEU-MET-THERM-006-003 IEU-COMP-THERM-015-008 LEU-COMP-THERM-033-046 LEU-MET-THERM-006-004 IEU-COMP-THERM-015-009 LEU-COMP-THERM-033-047 LEU-MET-THERM-006-005 IEU-COMP-THERM-015-010 LEU-COMP-THERM-033-048 LEU-MET-THERM-006-006 IEU-COMP-THERM-015-011 LEU-COMP-THERM-033-049 LEU-MET-THERM-006-007 IEU-COMP-THERM-015-012 LEU-COMP-THERM-033-050 LEU-MET-THERM-006-008 IEU-COMP-THERM-015-013 LEU-COMP-THERM-033-051 LEU-MET-THERM-006-009 IEU-COMP-THERM-015-014 LEU-COMP-THERM-033-052 LEU-MET-THERM-006-010 IEU-COMP-THERM-015-015 LEU-COMP-THERM-034-001 LEU-MET-THERM-006-011 IEU-COMP-THERM-015-016 LEU-COMP-THERM-034-002 LEU-MET-THERM-006-012 IEU-COMP-THERM-015-017 LEU-COMP-THERM-034-003 LEU-MET-THERM-006-013 IEU-COMP-THERM-015-018 LEU-COMP-THERM-034-004 LEU-MET-THERM-006-014 IEU-COMP-THERM-015-023 LEU-COMP-THERM-034-005 LEU-MET-THERM-006-015 IEU-COMP-THERM-015-024 LEU-COMP-THERM-034-006 LEU-MET-THERM-006-016 IEU-COMP-THERM-015-025 LEU-COMP-THERM-034-007 LEU-MET-THERM-006-017 IEU-COMP-THERM-015-026 LEU-COMP-THERM-034-008 LEU-MET-THERM-006-018 IEU-COMP-THERM-015-027 LEU-COMP-THERM-034-009 LEU-MET-THERM-006-019 IEU-COMP-THERM-015-028 LEU-COMP-THERM-034-010 LEU-MET-THERM-006-020 IEU-COMP-THERM-015-029 LEU-COMP-THERM-034-011 LEU-MET-THERM-006-021 IEU-COMP-THERM-015-030 LEU-COMP-THERM-034-012 LEU-MET-THERM-006-022 IEU-COMP-THERM-015-031 LEU-COMP-THERM-034-013 LEU-MET-THERM-006-023

B-3 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

IEU-COMP-THERM-015-032 LEU-COMP-THERM-034-014 LEU-MET-THERM-006-024 IEU-COMP-THERM-016-001 LEU-COMP-THERM-034-015 LEU-MET-THERM-006-025 IEU-COMP-THERM-016-002 LEU-COMP-THERM-034-016 LEU-MET-THERM-006-026 IEU-COMP-THERM-016-003 LEU-COMP-THERM-034-017 LEU-MET-THERM-006-027 IEU-COMP-THERM-016-004 LEU-COMP-THERM-034-018 LEU-MET-THERM-006-028 IEU-COMP-THERM-016-005 LEU-COMP-THERM-034-019 LEU-MET-THERM-006-029 IEU-COMP-THERM-016-006 LEU-COMP-THERM-034-020 LEU-MET-THERM-006-030 IEU-COMP-THERM-016-007 LEU-COMP-THERM-034-021 LEU-MET-THERM-007-001 IEU-COMP-THERM-016-008 LEU-COMP-THERM-034-022 LEU-MET-THERM-007-002 IEU-COMP-THERM-016-009 LEU-COMP-THERM-034-023 LEU-MET-THERM-007-003 IEU-COMP-THERM-016-010 LEU-COMP-THERM-034-024 LEU-MET-THERM-007-004 IEU-COMP-THERM-016-011 LEU-COMP-THERM-035-001 LEU-MET-THERM-007-005 IEU-COMP-THERM-016-012 LEU-COMP-THERM-035-002 LEU-MET-THERM-007-006 IEU-COMP-THERM-016-013 LEU-COMP-THERM-035-003 LEU-MET-THERM-015-001 IEU-COMP-THERM-016-014 LEU-COMP-THERM-036-001 LEU-MET-THERM-015-002 IEU-COMP-THERM-016-015 LEU-COMP-THERM-036-002 LEU-MET-THERM-015-003 IEU-COMP-THERM-016-016 LEU-COMP-THERM-036-003 LEU-MET-THERM-015-004 IEU-COMP-THERM-016-017 LEU-COMP-THERM-036-004 LEU-MET-THERM-015-005 IEU-COMP-THERM-016-018 LEU-COMP-THERM-036-005 LEU-MET-THERM-015-006 IEU-COMP-THERM-016-019 LEU-COMP-THERM-036-006 LEU-MET-THERM-015-007 IEU-COMP-THERM-016-020 LEU-COMP-THERM-036-007 LEU-MET-THERM-015-008 IEU-COMP-THERM-016-021 LEU-COMP-THERM-036-008 LEU-MET-THERM-015-009 IEU-COMP-THERM-016-022 LEU-COMP-THERM-036-009 LEU-MET-THERM-015-010 IEU-COMP-THERM-016-023 LEU-COMP-THERM-036-010 LEU-MET-THERM-015-011 IEU-COMP-THERM-016-024 LEU-COMP-THERM-036-011 LEU-MET-THERM-015-012 IEU-COMP-THERM-016-025 LEU-COMP-THERM-036-012 LEU-MET-THERM-015-013 IEU-COMP-THERM-016-026 LEU-COMP-THERM-036-013 LEU-MET-THERM-015-014 IEU-COMP-THERM-016-027 LEU-COMP-THERM-036-014 LEU-MET-THERM-015-015 IEU-COMP-THERM-016-028 LEU-COMP-THERM-036-015 LEU-MET-THERM-015-016 IEU-COMP-THERM-016-029 LEU-COMP-THERM-036-016 LEU-MET-THERM-015-017 IEU-COMP-THERM-016-030 LEU-COMP-THERM-036-017 LEU-MET-THERM-015-018 IEU-COMP-THERM-016-031 LEU-COMP-THERM-036-018 LEU-MET-THERM-015-019

B-4 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

IEU-COMP-THERM-016-032 LEU-COMP-THERM-036-019 LEU-MET-THERM-015-020 IEU-COMP-THERM-016-033 LEU-COMP-THERM-036-020 LEU-MET-THERM-015-021 IEU-COMP-THERM-016-034 LEU-COMP-THERM-036-021 LEU-MET-THERM-015-022 IEU-COMP-THERM-016-035 LEU-COMP-THERM-036-022 LEU-MISC-THERM-001-001 IEU-COMP-THERM-016-036 LEU-COMP-THERM-036-023 LEU-MISC-THERM-001-002 IEU-COMP-THERM-016-037 LEU-COMP-THERM-036-024 LEU-MISC-THERM-001-003 IEU-COMP-THERM-016-038 LEU-COMP-THERM-036-025 LEU-MISC-THERM-001-004 IEU-COMP-THERM-016-039 LEU-COMP-THERM-036-026 LEU-MISC-THERM-001-005 IEU-COMP-THERM-016-040 LEU-COMP-THERM-036-027 LEU-MISC-THERM-002-001 IEU-COMP-THERM-016-041 LEU-COMP-THERM-036-028 LEU-MISC-THERM-002-002 IEU-COMP-THERM-016-042 LEU-COMP-THERM-036-029 LEU-MISC-THERM-002-003 IEU-COMP-THERM-016-043 LEU-COMP-THERM-036-030 LEU-MISC-THERM-002-004 IEU-COMP-THERM-016-044 LEU-COMP-THERM-036-031 LEU-MISC-THERM-002-005 IEU-COMP-THERM-016-045 LEU-COMP-THERM-036-032 LEU-MISC-THERM-002-006 IEU-SOL-THERM-001-001 LEU-COMP-THERM-036-033 LEU-MISC-THERM-003-001 IEU-SOL-THERM-001-002 LEU-COMP-THERM-036-034 LEU-MISC-THERM-003-002 IEU-SOL-THERM-001-003 LEU-COMP-THERM-036-035 LEU-MISC-THERM-003-003 IEU-SOL-THERM-001-004 LEU-COMP-THERM-036-036 LEU-MISC-THERM-003-004 IEU-SOL-THERM-002-001 LEU-COMP-THERM-036-037 LEU-MISC-THERM-003-005 IEU-SOL-THERM-002-002 LEU-COMP-THERM-036-038 LEU-MISC-THERM-003-006 IEU-SOL-THERM-002-003 LEU-COMP-THERM-036-039 LEU-MISC-THERM-003-007 IEU-SOL-THERM-002-004 LEU-COMP-THERM-036-040 LEU-MISC-THERM-003-008 IEU-SOL-THERM-002-005 LEU-COMP-THERM-036-041 LEU-MISC-THERM-003-009 IEU-SOL-THERM-002-006 LEU-COMP-THERM-036-042 LEU-MISC-THERM-003-010 IEU-SOL-THERM-002-007 LEU-COMP-THERM-036-043 LEU-MISC-THERM-003-011 IEU-SOL-THERM-002-008 LEU-COMP-THERM-036-044 LEU-MISC-THERM-003-012 IEU-SOL-THERM-002-009 LEU-COMP-THERM-036-045 LEU-MISC-THERM-003-013 IEU-SOL-THERM-002-010 LEU-COMP-THERM-036-046 LEU-MISC-THERM-003-014 IEU-SOL-THERM-002-011 LEU-COMP-THERM-036-047 LEU-MISC-THERM-003-015 IEU-SOL-THERM-003-001 LEU-COMP-THERM-036-048 LEU-MISC-THERM-005-001 IEU-SOL-THERM-003-002 LEU-COMP-THERM-036-049 LEU-MISC-THERM-005-002 IEU-SOL-THERM-003-003 LEU-COMP-THERM-036-050 LEU-MISC-THERM-005-003

B-5 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

IEU-SOL-THERM-003-004 LEU-COMP-THERM-036-051 LEU-MISC-THERM-005-004 IEU-SOL-THERM-003-005 LEU-COMP-THERM-036-052 LEU-MISC-THERM-005-005 IEU-SOL-THERM-003-006 LEU-COMP-THERM-036-053 LEU-MISC-THERM-005-006 IEU-SOL-THERM-003-007 LEU-COMP-THERM-036-054 LEU-MISC-THERM-005-007 IEU-SOL-THERM-003-008 LEU-COMP-THERM-036-055 LEU-MISC-THERM-005-008 IEU-SOL-THERM-003-009 LEU-COMP-THERM-036-056 LEU-MISC-THERM-005-009 IEU-SOL-THERM-003-010 LEU-COMP-THERM-036-057 LEU-MISC-THERM-005-010 IEU-SOL-THERM-003-011 LEU-COMP-THERM-036-058 LEU-MISC-THERM-005-011 IEU-SOL-THERM-003-012 LEU-COMP-THERM-036-059 LEU-MISC-THERM-005-012 IEU-SOL-THERM-003-013 LEU-COMP-THERM-036-060 LEU-MISC-THERM-006-001 IEU-SOL-THERM-003-014 LEU-COMP-THERM-036-061 LEU-MISC-THERM-006-002 IEU-SOL-THERM-003-015 LEU-COMP-THERM-036-062 LEU-MISC-THERM-006-003 IEU-SOL-THERM-003-016 LEU-COMP-THERM-036-063 LEU-MISC-THERM-006-004 IEU-SOL-THERM-003-017 LEU-COMP-THERM-036-064 LEU-MISC-THERM-006-005 IEU-SOL-THERM-003-018 LEU-COMP-THERM-036-065 LEU-MISC-THERM-006-006 IEU-SOL-THERM-003-019 LEU-COMP-THERM-036-066 LEU-MISC-THERM-006-007 IEU-SOL-THERM-003-020 LEU-COMP-THERM-036-067 LEU-MISC-THERM-006-008 IEU-SOL-THERM-003-021 LEU-COMP-THERM-036-068 LEU-MISC-THERM-006-009 IEU-SOL-THERM-003-022 LEU-COMP-THERM-036-069 LEU-MISC-THERM-006-010 IEU-SOL-THERM-003-023 LEU-COMP-THERM-037-001 LEU-SOL-THERM-001-001 IEU-SOL-THERM-003-024 LEU-COMP-THERM-037-002 LEU-SOL-THERM-002-001 IEU-SOL-THERM-003-025 LEU-COMP-THERM-037-003 LEU-SOL-THERM-002-002 IEU-SOL-THERM-003-026 LEU-COMP-THERM-037-004 LEU-SOL-THERM-002-003 IEU-SOL-THERM-003-027 LEU-COMP-THERM-037-005 LEU-SOL-THERM-003-001 IEU-SOL-THERM-003-028 LEU-COMP-THERM-037-006 LEU-SOL-THERM-003-002 IEU-SOL-THERM-003-029 LEU-COMP-THERM-037-007 LEU-SOL-THERM-003-003 IEU-SOL-THERM-003-030 LEU-COMP-THERM-037-008 LEU-SOL-THERM-003-004 IEU-SOL-THERM-003-031 LEU-COMP-THERM-037-009 LEU-SOL-THERM-003-005 IEU-SOL-THERM-003-032 LEU-COMP-THERM-037-010 LEU-SOL-THERM-003-006 IEU-SOL-THERM-003-033 LEU-COMP-THERM-037-011 LEU-SOL-THERM-003-007 IEU-SOL-THERM-003-034 LEU-COMP-THERM-038-001 LEU-SOL-THERM-003-008 IEU-SOL-THERM-003-035 LEU-COMP-THERM-038-002 LEU-SOL-THERM-003-009

B-6 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

IEU-SOL-THERM-003-036 LEU-COMP-THERM-038-003 LEU-SOL-THERM-004-001 IEU-SOL-THERM-003-037 LEU-COMP-THERM-038-004 LEU-SOL-THERM-004-002 IEU-SOL-THERM-003-038 LEU-COMP-THERM-038-005 LEU-SOL-THERM-004-003 IEU-SOL-THERM-003-039 LEU-COMP-THERM-038-006 LEU-SOL-THERM-004-004 IEU-SOL-THERM-003-040 LEU-COMP-THERM-038-007 LEU-SOL-THERM-004-005 IEU-SOL-THERM-003-041 LEU-COMP-THERM-038-008 LEU-SOL-THERM-004-006 IEU-SOL-THERM-003-042 LEU-COMP-THERM-038-009 LEU-SOL-THERM-004-007 IEU-SOL-THERM-003-043 LEU-COMP-THERM-038-010 LEU-SOL-THERM-005-001 IEU-SOL-THERM-003-044 LEU-COMP-THERM-038-011 LEU-SOL-THERM-005-002 IEU-SOL-THERM-003-045 LEU-COMP-THERM-038-012 LEU-SOL-THERM-005-003 IEU-SOL-THERM-003-046 LEU-COMP-THERM-038-013 LEU-SOL-THERM-006-001 IEU-SOL-THERM-004-001 LEU-COMP-THERM-038-014 LEU-SOL-THERM-006-002 IEU-SOL-THERM-005-001 LEU-COMP-THERM-039-001 LEU-SOL-THERM-006-003 LEU-COMP-THERM-001-001 LEU-COMP-THERM-039-002 LEU-SOL-THERM-006-004 LEU-COMP-THERM-001-002 LEU-COMP-THERM-039-003 LEU-SOL-THERM-006-005 LEU-COMP-THERM-001-003 LEU-COMP-THERM-039-004 LEU-SOL-THERM-007-001 LEU-COMP-THERM-001-004 LEU-COMP-THERM-039-005 LEU-SOL-THERM-007-002 LEU-COMP-THERM-001-005 LEU-COMP-THERM-039-006 LEU-SOL-THERM-007-003 LEU-COMP-THERM-001-006 LEU-COMP-THERM-039-007 LEU-SOL-THERM-007-004 LEU-COMP-THERM-001-007 LEU-COMP-THERM-039-008 LEU-SOL-THERM-007-005 LEU-COMP-THERM-001-008 LEU-COMP-THERM-039-009 LEU-SOL-THERM-008-001 LEU-COMP-THERM-002-001 LEU-COMP-THERM-039-010 LEU-SOL-THERM-008-002 LEU-COMP-THERM-002-002 LEU-COMP-THERM-039-011 LEU-SOL-THERM-008-003 LEU-COMP-THERM-002-003 LEU-COMP-THERM-039-012 LEU-SOL-THERM-008-004 LEU-COMP-THERM-002-004 LEU-COMP-THERM-039-013 LEU-SOL-THERM-009-001 LEU-COMP-THERM-002-005 LEU-COMP-THERM-039-014 LEU-SOL-THERM-009-002 LEU-COMP-THERM-003-001 LEU-COMP-THERM-039-015 LEU-SOL-THERM-009-003 LEU-COMP-THERM-003-002 LEU-COMP-THERM-039-016 LEU-SOL-THERM-011-001 LEU-COMP-THERM-003-003 LEU-COMP-THERM-039-017 LEU-SOL-THERM-011-002 LEU-COMP-THERM-003-004 LEU-COMP-THERM-040-001 LEU-SOL-THERM-011-003 LEU-COMP-THERM-003-005 LEU-COMP-THERM-040-002 LEU-SOL-THERM-011-004 LEU-COMP-THERM-003-006 LEU-COMP-THERM-040-003 LEU-SOL-THERM-011-005

B-7 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-003-007 LEU-COMP-THERM-040-004 LEU-SOL-THERM-011-006 LEU-COMP-THERM-003-008 LEU-COMP-THERM-040-005 LEU-SOL-THERM-011-007 LEU-COMP-THERM-003-009 LEU-COMP-THERM-040-006 LEU-SOL-THERM-011-008 LEU-COMP-THERM-003-010 LEU-COMP-THERM-040-007 LEU-SOL-THERM-011-009 LEU-COMP-THERM-003-011 LEU-COMP-THERM-040-008 LEU-SOL-THERM-011-011 LEU-COMP-THERM-003-012 LEU-COMP-THERM-040-009 LEU-SOL-THERM-011-012 LEU-COMP-THERM-003-013 LEU-COMP-THERM-040-010 LEU-SOL-THERM-011-013 LEU-COMP-THERM-003-014 LEU-COMP-THERM-042-001 LEU-SOL-THERM-012-001 LEU-COMP-THERM-003-015 LEU-COMP-THERM-042-002 LEU-SOL-THERM-012-002 LEU-COMP-THERM-003-016 LEU-COMP-THERM-042-003 LEU-SOL-THERM-013-001 LEU-COMP-THERM-003-017 LEU-COMP-THERM-042-004 LEU-SOL-THERM-016-001 LEU-COMP-THERM-003-018 LEU-COMP-THERM-042-005 LEU-SOL-THERM-016-002 LEU-COMP-THERM-003-019 LEU-COMP-THERM-042-006 LEU-SOL-THERM-016-003 LEU-COMP-THERM-003-020 LEU-COMP-THERM-042-007 LEU-SOL-THERM-016-004 LEU-COMP-THERM-003-021 LEU-COMP-THERM-043-001 LEU-SOL-THERM-016-005 LEU-COMP-THERM-003-022 LEU-COMP-THERM-043-002 LEU-SOL-THERM-016-006 LEU-COMP-THERM-004-001 LEU-COMP-THERM-043-003 LEU-SOL-THERM-016-007 LEU-COMP-THERM-004-002 LEU-COMP-THERM-043-004 LEU-SOL-THERM-017-001 LEU-COMP-THERM-004-003 LEU-COMP-THERM-043-005 LEU-SOL-THERM-017-002 LEU-COMP-THERM-004-004 LEU-COMP-THERM-043-006 LEU-SOL-THERM-017-003 LEU-COMP-THERM-004-005 LEU-COMP-THERM-043-007 LEU-SOL-THERM-017-004 LEU-COMP-THERM-004-006 LEU-COMP-THERM-043-008 LEU-SOL-THERM-017-005 LEU-COMP-THERM-004-007 LEU-COMP-THERM-043-009 LEU-SOL-THERM-017-006 LEU-COMP-THERM-004-008 LEU-COMP-THERM-044-001 LEU-SOL-THERM-018-001 LEU-COMP-THERM-004-009 LEU-COMP-THERM-044-002 LEU-SOL-THERM-018-002 LEU-COMP-THERM-004-010 LEU-COMP-THERM-044-003 LEU-SOL-THERM-018-003 LEU-COMP-THERM-004-011 LEU-COMP-THERM-044-004 LEU-SOL-THERM-018-004 LEU-COMP-THERM-004-012 LEU-COMP-THERM-044-005 LEU-SOL-THERM-018-005 LEU-COMP-THERM-004-013 LEU-COMP-THERM-044-006 LEU-SOL-THERM-018-006 LEU-COMP-THERM-004-014 LEU-COMP-THERM-044-007 LEU-SOL-THERM-019-001 LEU-COMP-THERM-004-015 LEU-COMP-THERM-044-008 LEU-SOL-THERM-019-002 LEU-COMP-THERM-004-016 LEU-COMP-THERM-044-009 LEU-SOL-THERM-019-003

B-8 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-004-017 LEU-COMP-THERM-044-010 LEU-SOL-THERM-019-004 LEU-COMP-THERM-004-018 LEU-COMP-THERM-045-001 LEU-SOL-THERM-019-005 LEU-COMP-THERM-004-019 LEU-COMP-THERM-045-002 LEU-SOL-THERM-019-006 LEU-COMP-THERM-004-020 LEU-COMP-THERM-045-003 LEU-SOL-THERM-020-001 LEU-COMP-THERM-005-001 LEU-COMP-THERM-045-004 LEU-SOL-THERM-020-002 LEU-COMP-THERM-005-002 LEU-COMP-THERM-045-005 LEU-SOL-THERM-020-003 LEU-COMP-THERM-005-003 LEU-COMP-THERM-045-006 LEU-SOL-THERM-020-004 LEU-COMP-THERM-005-004 LEU-COMP-THERM-045-007 LEU-SOL-THERM-021-001 LEU-COMP-THERM-005-005 LEU-COMP-THERM-045-008 LEU-SOL-THERM-021-002 LEU-COMP-THERM-005-006 LEU-COMP-THERM-045-009 LEU-SOL-THERM-021-003 LEU-COMP-THERM-005-007 LEU-COMP-THERM-045-010 LEU-SOL-THERM-021-004 LEU-COMP-THERM-005-008 LEU-COMP-THERM-045-011 LEU-SOL-THERM-022-001 LEU-COMP-THERM-005-009 LEU-COMP-THERM-045-012 LEU-SOL-THERM-022-002 LEU-COMP-THERM-005-010 LEU-COMP-THERM-045-013 LEU-SOL-THERM-022-003 LEU-COMP-THERM-005-011 LEU-COMP-THERM-045-014 LEU-SOL-THERM-022-004 LEU-COMP-THERM-005-012 LEU-COMP-THERM-045-015 LEU-SOL-THERM-023-001 LEU-COMP-THERM-005-013 LEU-COMP-THERM-045-016 LEU-SOL-THERM-023-002 LEU-COMP-THERM-005-014 LEU-COMP-THERM-045-017 LEU-SOL-THERM-023-003 LEU-COMP-THERM-005-015 LEU-COMP-THERM-045-018 LEU-SOL-THERM-023-004 LEU-COMP-THERM-005-016 LEU-COMP-THERM-045-019 LEU-SOL-THERM-023-005 LEU-COMP-THERM-006-001 LEU-COMP-THERM-045-020 LEU-SOL-THERM-023-006 LEU-COMP-THERM-006-002 LEU-COMP-THERM-045-021 LEU-SOL-THERM-023-007 LEU-COMP-THERM-006-003 LEU-COMP-THERM-046-001 LEU-SOL-THERM-023-008 LEU-COMP-THERM-006-004 LEU-COMP-THERM-046-002 LEU-SOL-THERM-023-009 LEU-COMP-THERM-006-005 LEU-COMP-THERM-046-003 LEU-SOL-THERM-024-001 LEU-COMP-THERM-006-006 LEU-COMP-THERM-046-004 LEU-SOL-THERM-024-002 LEU-COMP-THERM-006-007 LEU-COMP-THERM-046-005 LEU-SOL-THERM-024-003 LEU-COMP-THERM-006-008 LEU-COMP-THERM-046-006 LEU-SOL-THERM-024-004 LEU-COMP-THERM-006-009 LEU-COMP-THERM-046-007 LEU-SOL-THERM-024-005 LEU-COMP-THERM-006-010 LEU-COMP-THERM-046-008 LEU-SOL-THERM-024-006 LEU-COMP-THERM-006-011 LEU-COMP-THERM-046-009 LEU-SOL-THERM-024-007 LEU-COMP-THERM-006-012 LEU-COMP-THERM-046-010 LEU-SOL-THERM-025-001

B-9 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-006-013 LEU-COMP-THERM-046-011 LEU-SOL-THERM-025-002 LEU-COMP-THERM-006-014 LEU-COMP-THERM-046-012 LEU-SOL-THERM-025-003 LEU-COMP-THERM-006-015 LEU-COMP-THERM-046-013 LEU-SOL-THERM-025-004 LEU-COMP-THERM-006-016 LEU-COMP-THERM-046-014 LEU-SOL-THERM-025-005 LEU-COMP-THERM-006-017 LEU-COMP-THERM-046-015 LEU-SOL-THERM-025-006 LEU-COMP-THERM-006-018 LEU-COMP-THERM-046-016 LEU-SOL-THERM-025-007 LEU-COMP-THERM-008-001 LEU-COMP-THERM-046-017 MIX-COMP-THERM-001-001 LEU-COMP-THERM-008-002 LEU-COMP-THERM-046-018 MIX-COMP-THERM-001-002 LEU-COMP-THERM-008-003 LEU-COMP-THERM-046-019 MIX-COMP-THERM-001-003 LEU-COMP-THERM-008-004 LEU-COMP-THERM-046-020 MIX-COMP-THERM-001-004 LEU-COMP-THERM-008-005 LEU-COMP-THERM-046-021 MIX-COMP-THERM-002-001S LEU-COMP-THERM-008-006 LEU-COMP-THERM-046-022 MIX-COMP-THERM-002-002S LEU-COMP-THERM-008-007 LEU-COMP-THERM-047-001 MIX-COMP-THERM-002-003S LEU-COMP-THERM-008-008 LEU-COMP-THERM-047-002 MIX-COMP-THERM-002-004S LEU-COMP-THERM-008-009 LEU-COMP-THERM-047-003 MIX-COMP-THERM-002-005S LEU-COMP-THERM-008-010 LEU-COMP-THERM-048-001 MIX-COMP-THERM-002-006S LEU-COMP-THERM-008-011 LEU-COMP-THERM-048-002 MIX-COMP-THERM-004-001 LEU-COMP-THERM-008-012 LEU-COMP-THERM-048-003 MIX-COMP-THERM-004-002 LEU-COMP-THERM-008-013 LEU-COMP-THERM-048-004 MIX-COMP-THERM-004-003 LEU-COMP-THERM-008-014 LEU-COMP-THERM-048-005 MIX-COMP-THERM-004-004 LEU-COMP-THERM-008-015 LEU-COMP-THERM-049-001 MIX-COMP-THERM-004-005 LEU-COMP-THERM-008-016 LEU-COMP-THERM-049-002 MIX-COMP-THERM-004-006 LEU-COMP-THERM-008-017 LEU-COMP-THERM-049-003 MIX-COMP-THERM-004-007 LEU-COMP-THERM-009-001 LEU-COMP-THERM-049-004 MIX-COMP-THERM-004-008 LEU-COMP-THERM-009-002 LEU-COMP-THERM-049-005 MIX-COMP-THERM-004-009 LEU-COMP-THERM-009-003 LEU-COMP-THERM-049-006 MIX-COMP-THERM-004-010 LEU-COMP-THERM-009-004 LEU-COMP-THERM-049-007 MIX-COMP-THERM-004-011 LEU-COMP-THERM-009-005 LEU-COMP-THERM-049-008 MIX-COMP-THERM-006-001 LEU-COMP-THERM-009-006 LEU-COMP-THERM-049-009 MIX-COMP-THERM-006-002 LEU-COMP-THERM-009-007 LEU-COMP-THERM-049-010 MIX-COMP-THERM-006-003 LEU-COMP-THERM-009-008 LEU-COMP-THERM-049-011 MIX-COMP-THERM-006-004 LEU-COMP-THERM-009-009 LEU-COMP-THERM-049-012 MIX-COMP-THERM-006-005

B-10 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-009-010 LEU-COMP-THERM-049-013 MIX-COMP-THERM-006-006 LEU-COMP-THERM-009-011 LEU-COMP-THERM-049-014 MIX-COMP-THERM-006-007 LEU-COMP-THERM-009-012 LEU-COMP-THERM-049-015 MIX-COMP-THERM-006-008 LEU-COMP-THERM-009-013 LEU-COMP-THERM-049-016 MIX-COMP-THERM-006-009 LEU-COMP-THERM-009-014 LEU-COMP-THERM-049-017 MIX-COMP-THERM-006-010 LEU-COMP-THERM-009-015 LEU-COMP-THERM-049-018 MIX-COMP-THERM-006-011 LEU-COMP-THERM-009-016 LEU-COMP-THERM-050-001 MIX-COMP-THERM-006-012 LEU-COMP-THERM-009-017 LEU-COMP-THERM-050-002 MIX-COMP-THERM-006-013 LEU-COMP-THERM-009-018 LEU-COMP-THERM-050-003 MIX-COMP-THERM-006-014 LEU-COMP-THERM-009-019 LEU-COMP-THERM-050-004 MIX-COMP-THERM-006-015 LEU-COMP-THERM-009-020 LEU-COMP-THERM-050-005 MIX-COMP-THERM-006-016 LEU-COMP-THERM-009-021 LEU-COMP-THERM-050-006 MIX-COMP-THERM-006-017 LEU-COMP-THERM-009-022 LEU-COMP-THERM-050-007 MIX-COMP-THERM-006-018 LEU-COMP-THERM-009-023 LEU-COMP-THERM-050-008 MIX-COMP-THERM-006-019 LEU-COMP-THERM-009-024 LEU-COMP-THERM-050-009 MIX-COMP-THERM-006-020 LEU-COMP-THERM-009-025 LEU-COMP-THERM-050-010 MIX-COMP-THERM-006-021 LEU-COMP-THERM-009-026 LEU-COMP-THERM-050-011 MIX-COMP-THERM-006-022 LEU-COMP-THERM-009-027 LEU-COMP-THERM-050-012 MIX-COMP-THERM-006-023 LEU-COMP-THERM-010-001 LEU-COMP-THERM-050-013 MIX-COMP-THERM-006-024 LEU-COMP-THERM-010-002 LEU-COMP-THERM-050-014 MIX-COMP-THERM-006-025 LEU-COMP-THERM-010-003 LEU-COMP-THERM-050-015 MIX-COMP-THERM-006-026 LEU-COMP-THERM-010-004 LEU-COMP-THERM-050-016 MIX-COMP-THERM-006-027 LEU-COMP-THERM-010-005 LEU-COMP-THERM-050-017 MIX-COMP-THERM-006-028 LEU-COMP-THERM-010-006 LEU-COMP-THERM-050-018 MIX-COMP-THERM-006-029 LEU-COMP-THERM-010-007 LEU-COMP-THERM-051-001 MIX-COMP-THERM-006-030 LEU-COMP-THERM-010-008 LEU-COMP-THERM-051-002 MIX-COMP-THERM-006-031 LEU-COMP-THERM-010-009 LEU-COMP-THERM-051-003 MIX-COMP-THERM-006-032 LEU-COMP-THERM-010-010 LEU-COMP-THERM-051-004 MIX-COMP-THERM-006-033 LEU-COMP-THERM-010-011 LEU-COMP-THERM-051-005 MIX-COMP-THERM-006-034 LEU-COMP-THERM-010-012 LEU-COMP-THERM-051-006 MIX-COMP-THERM-006-035 LEU-COMP-THERM-010-013 LEU-COMP-THERM-051-007 MIX-COMP-THERM-006-036 LEU-COMP-THERM-010-014 LEU-COMP-THERM-051-008 MIX-COMP-THERM-006-037

B-11 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-010-015 LEU-COMP-THERM-051-009 MIX-COMP-THERM-006-038 LEU-COMP-THERM-010-016 LEU-COMP-THERM-051-010 MIX-COMP-THERM-006-039 LEU-COMP-THERM-010-017 LEU-COMP-THERM-051-011 MIX-COMP-THERM-006-040 LEU-COMP-THERM-010-018 LEU-COMP-THERM-051-012 MIX-COMP-THERM-006-041 LEU-COMP-THERM-010-019 LEU-COMP-THERM-051-013 MIX-COMP-THERM-006-042 LEU-COMP-THERM-010-020 LEU-COMP-THERM-051-014 MIX-COMP-THERM-006-043 LEU-COMP-THERM-010-021 LEU-COMP-THERM-051-015 MIX-COMP-THERM-006-044 LEU-COMP-THERM-010-022 LEU-COMP-THERM-051-016 MIX-COMP-THERM-006-045 LEU-COMP-THERM-010-023 LEU-COMP-THERM-051-017 MIX-COMP-THERM-006-046 LEU-COMP-THERM-010-024 LEU-COMP-THERM-051-018 MIX-COMP-THERM-006-047 LEU-COMP-THERM-010-025 LEU-COMP-THERM-051-019 MIX-COMP-THERM-006-048 LEU-COMP-THERM-010-026 LEU-COMP-THERM-052-001 MIX-COMP-THERM-006-049 LEU-COMP-THERM-010-027 LEU-COMP-THERM-052-002 MIX-COMP-THERM-006-050 LEU-COMP-THERM-010-028 LEU-COMP-THERM-052-003 MIX-COMP-THERM-007-001 LEU-COMP-THERM-010-029 LEU-COMP-THERM-052-004 MIX-COMP-THERM-007-002 LEU-COMP-THERM-010-030 LEU-COMP-THERM-052-005 MIX-COMP-THERM-007-003 LEU-COMP-THERM-011-001 LEU-COMP-THERM-052-006 MIX-COMP-THERM-007-004 LEU-COMP-THERM-011-002 LEU-COMP-THERM-053-001 MIX-COMP-THERM-007-005 LEU-COMP-THERM-011-003 LEU-COMP-THERM-053-002 MIX-COMP-THERM-007-006 LEU-COMP-THERM-011-004 LEU-COMP-THERM-053-003 MIX-COMP-THERM-007-007 LEU-COMP-THERM-011-005 LEU-COMP-THERM-053-004 MIX-COMP-THERM-007-008 LEU-COMP-THERM-011-006 LEU-COMP-THERM-053-005 MIX-COMP-THERM-007-009 LEU-COMP-THERM-011-007 LEU-COMP-THERM-053-006 MIX-COMP-THERM-007-010 LEU-COMP-THERM-011-008 LEU-COMP-THERM-053-007 MIX-COMP-THERM-007-011 LEU-COMP-THERM-011-009 LEU-COMP-THERM-053-008 MIX-COMP-THERM-007-012 LEU-COMP-THERM-011-010 LEU-COMP-THERM-053-009 MIX-COMP-THERM-007-013 LEU-COMP-THERM-011-011 LEU-COMP-THERM-053-010 MIX-COMP-THERM-007-014 LEU-COMP-THERM-011-012 LEU-COMP-THERM-053-011 MIX-COMP-THERM-007-015 LEU-COMP-THERM-011-013 LEU-COMP-THERM-053-012 MIX-COMP-THERM-007-016 LEU-COMP-THERM-011-014 LEU-COMP-THERM-053-013 MIX-COMP-THERM-007-017 LEU-COMP-THERM-011-015 LEU-COMP-THERM-053-014 MIX-COMP-THERM-007-018 LEU-COMP-THERM-012-001 LEU-COMP-THERM-054-001 MIX-COMP-THERM-007-019

B-12 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-012-002 LEU-COMP-THERM-054-002 MIX-COMP-THERM-007-020 LEU-COMP-THERM-012-003 LEU-COMP-THERM-054-003 MIX-COMP-THERM-007-021 LEU-COMP-THERM-012-004 LEU-COMP-THERM-054-004 MIX-COMP-THERM-007-022 LEU-COMP-THERM-012-005 LEU-COMP-THERM-054-005 MIX-COMP-THERM-007-023 LEU-COMP-THERM-012-006 LEU-COMP-THERM-054-006 MIX-COMP-THERM-007-024 LEU-COMP-THERM-012-007 LEU-COMP-THERM-054-007 MIX-COMP-THERM-007-025 LEU-COMP-THERM-012-008 LEU-COMP-THERM-054-008 MIX-COMP-THERM-007-026 LEU-COMP-THERM-012-009 LEU-COMP-THERM-055-001 MIX-COMP-THERM-007-027 LEU-COMP-THERM-012-010 LEU-COMP-THERM-055-002 MIX-COMP-THERM-008-001 LEU-COMP-THERM-013-001 LEU-COMP-THERM-057-001 MIX-COMP-THERM-008-002 LEU-COMP-THERM-013-002 LEU-COMP-THERM-057-002 MIX-COMP-THERM-008-003 LEU-COMP-THERM-013-003 LEU-COMP-THERM-057-003 MIX-COMP-THERM-008-004 LEU-COMP-THERM-013-004 LEU-COMP-THERM-057-004 MIX-COMP-THERM-008-005 LEU-COMP-THERM-013-005 LEU-COMP-THERM-057-005 MIX-COMP-THERM-008-006 LEU-COMP-THERM-013-006 LEU-COMP-THERM-057-006 MIX-COMP-THERM-008-007 LEU-COMP-THERM-013-007 LEU-COMP-THERM-057-007 MIX-COMP-THERM-008-008 LEU-COMP-THERM-014-001 LEU-COMP-THERM-057-008 MIX-COMP-THERM-008-009 LEU-COMP-THERM-014-002 LEU-COMP-THERM-057-009 MIX-COMP-THERM-008-010 LEU-COMP-THERM-014-005 LEU-COMP-THERM-057-010 MIX-COMP-THERM-008-011 LEU-COMP-THERM-014-006 LEU-COMP-THERM-057-011 MIX-COMP-THERM-008-012 LEU-COMP-THERM-014-007 LEU-COMP-THERM-057-012 MIX-COMP-THERM-008-013 LEU-COMP-THERM-015-001 LEU-COMP-THERM-057-013 MIX-COMP-THERM-008-014 LEU-COMP-THERM-015-002 LEU-COMP-THERM-057-014 MIX-COMP-THERM-008-015 LEU-COMP-THERM-015-003 LEU-COMP-THERM-057-015 MIX-COMP-THERM-008-016 LEU-COMP-THERM-015-004 LEU-COMP-THERM-057-016 MIX-COMP-THERM-008-017 LEU-COMP-THERM-015-005 LEU-COMP-THERM-057-017 MIX-COMP-THERM-008-018 LEU-COMP-THERM-015-006 LEU-COMP-THERM-057-018 MIX-COMP-THERM-008-019 LEU-COMP-THERM-015-007 LEU-COMP-THERM-057-019 MIX-COMP-THERM-008-020 LEU-COMP-THERM-015-008 LEU-COMP-THERM-057-020 MIX-COMP-THERM-008-021 LEU-COMP-THERM-015-009 LEU-COMP-THERM-057-021 MIX-COMP-THERM-008-022 LEU-COMP-THERM-015-010 LEU-COMP-THERM-057-022 MIX-COMP-THERM-008-023 LEU-COMP-THERM-015-011 LEU-COMP-THERM-057-023 MIX-COMP-THERM-008-024

B-13 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-015-012 LEU-COMP-THERM-057-024 MIX-COMP-THERM-008-025 LEU-COMP-THERM-015-013 LEU-COMP-THERM-057-025 MIX-COMP-THERM-008-026 LEU-COMP-THERM-015-014 LEU-COMP-THERM-057-026 MIX-COMP-THERM-008-027 LEU-COMP-THERM-015-015 LEU-COMP-THERM-057-027 MIX-COMP-THERM-008-028 LEU-COMP-THERM-015-016 LEU-COMP-THERM-057-028 MIX-COMP-THERM-012-001 LEU-COMP-THERM-015-017 LEU-COMP-THERM-057-029 MIX-COMP-THERM-012-002 LEU-COMP-THERM-015-018 LEU-COMP-THERM-057-030 MIX-COMP-THERM-012-003 LEU-COMP-THERM-015-019 LEU-COMP-THERM-057-031 MIX-COMP-THERM-012-004 LEU-COMP-THERM-015-020 LEU-COMP-THERM-057-032 MIX-COMP-THERM-012-005 LEU-COMP-THERM-015-021 LEU-COMP-THERM-057-033 MIX-COMP-THERM-012-006 LEU-COMP-THERM-015-022 LEU-COMP-THERM-057-034 MIX-COMP-THERM-012-007 LEU-COMP-THERM-015-023 LEU-COMP-THERM-057-035 MIX-COMP-THERM-012-008 LEU-COMP-THERM-015-024 LEU-COMP-THERM-057-036 MIX-COMP-THERM-012-009 LEU-COMP-THERM-015-025 LEU-COMP-THERM-058-001 MIX-COMP-THERM-012-010 LEU-COMP-THERM-015-026 LEU-COMP-THERM-058-002 MIX-COMP-THERM-012-011 LEU-COMP-THERM-015-027 LEU-COMP-THERM-058-003 MIX-COMP-THERM-012-012 LEU-COMP-THERM-015-028 LEU-COMP-THERM-058-004 MIX-COMP-THERM-012-013 LEU-COMP-THERM-015-029 LEU-COMP-THERM-058-005 MIX-COMP-THERM-012-014 LEU-COMP-THERM-015-030 LEU-COMP-THERM-058-006 MIX-COMP-THERM-012-015 LEU-COMP-THERM-015-031 LEU-COMP-THERM-058-007 MIX-COMP-THERM-012-016 LEU-COMP-THERM-015-032 LEU-COMP-THERM-058-008 MIX-COMP-THERM-012-017 LEU-COMP-THERM-015-033 LEU-COMP-THERM-058-009 MIX-COMP-THERM-012-018 LEU-COMP-THERM-015-034 LEU-COMP-THERM-061-001 MIX-COMP-THERM-012-019 LEU-COMP-THERM-015-035 LEU-COMP-THERM-061-002 MIX-COMP-THERM-012-020 LEU-COMP-THERM-015-036 LEU-COMP-THERM-061-003 MIX-COMP-THERM-012-021 LEU-COMP-THERM-015-037 LEU-COMP-THERM-061-004 MIX-COMP-THERM-012-022 LEU-COMP-THERM-015-038 LEU-COMP-THERM-061-005 MIX-COMP-THERM-012-023 LEU-COMP-THERM-015-039 LEU-COMP-THERM-061-006 MIX-COMP-THERM-012-024 LEU-COMP-THERM-015-040 LEU-COMP-THERM-061-007 MIX-COMP-THERM-012-025 LEU-COMP-THERM-015-041 LEU-COMP-THERM-061-008 MIX-COMP-THERM-012-026 LEU-COMP-THERM-015-042 LEU-COMP-THERM-061-009 MIX-COMP-THERM-012-027 LEU-COMP-THERM-015-043 LEU-COMP-THERM-061-010 MIX-COMP-THERM-012-028

B-14 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-015-044 LEU-COMP-THERM-062-001 MIX-COMP-THERM-012-029 LEU-COMP-THERM-015-045 LEU-COMP-THERM-062-002 MIX-COMP-THERM-012-030 LEU-COMP-THERM-015-046 LEU-COMP-THERM-062-003 MIX-COMP-THERM-012-031 LEU-COMP-THERM-015-047 LEU-COMP-THERM-062-004 MIX-COMP-THERM-012-032 LEU-COMP-THERM-015-048 LEU-COMP-THERM-062-005 MIX-COMP-THERM-012-033 LEU-COMP-THERM-015-049 LEU-COMP-THERM-062-006 MIX-COMP-THERM-013-001 LEU-COMP-THERM-015-050 LEU-COMP-THERM-062-007 MIX-COMP-THERM-013-002 LEU-COMP-THERM-015-051 LEU-COMP-THERM-062-008 MIX-COMP-THERM-013-003 LEU-COMP-THERM-015-052 LEU-COMP-THERM-062-009 MIX-COMP-THERM-013-004 LEU-COMP-THERM-015-053 LEU-COMP-THERM-062-010 MIX-COMP-THERM-013-005 LEU-COMP-THERM-015-054 LEU-COMP-THERM-062-011 MIX-COMP-THERM-013-006 LEU-COMP-THERM-015-055 LEU-COMP-THERM-062-012 MIX-COMP-THERM-013-007 LEU-COMP-THERM-015-056 LEU-COMP-THERM-062-013 MIX-COMP-THERM-013-008 LEU-COMP-THERM-015-057 LEU-COMP-THERM-062-014 MIX-COMP-THERM-013-009 LEU-COMP-THERM-015-058 LEU-COMP-THERM-062-015 MIX-COMP-THERM-013-010 LEU-COMP-THERM-015-059 LEU-COMP-THERM-065-001 MIX-COMP-THERM-013-011 LEU-COMP-THERM-015-060 LEU-COMP-THERM-065-002 MIX-COMP-THERM-013-012 LEU-COMP-THERM-015-061 LEU-COMP-THERM-065-003 MIX-COMP-THERM-013-013 LEU-COMP-THERM-015-062 LEU-COMP-THERM-065-004 MIX-COMP-THERM-013-014 LEU-COMP-THERM-015-063 LEU-COMP-THERM-065-005 MIX-COMP-THERM-013-015 LEU-COMP-THERM-015-064 LEU-COMP-THERM-065-006 MIX-COMP-THERM-013-016 LEU-COMP-THERM-015-065 LEU-COMP-THERM-065-007 MIX-COMP-THERM-013-017 LEU-COMP-THERM-015-066 LEU-COMP-THERM-065-008 MIX-COMP-THERM-013-018 LEU-COMP-THERM-015-067 LEU-COMP-THERM-065-009 MIX-COMP-THERM-013-019 LEU-COMP-THERM-015-068 LEU-COMP-THERM-065-010 MIX-COMP-THERM-013-020 LEU-COMP-THERM-015-069 LEU-COMP-THERM-065-011 MIX-COMP-THERM-013-021 LEU-COMP-THERM-015-070 LEU-COMP-THERM-065-012 MIX-COMP-THERM-013-022 LEU-COMP-THERM-015-071 LEU-COMP-THERM-065-013 MIX-COMP-THERM-013-023 LEU-COMP-THERM-015-072 LEU-COMP-THERM-065-014 MIX-COMP-THERM-013-024 LEU-COMP-THERM-015-073 LEU-COMP-THERM-065-015 MIX-COMP-THERM-013-025 LEU-COMP-THERM-015-074 LEU-COMP-THERM-065-016 MIX-COMP-THERM-013-026 LEU-COMP-THERM-015-075 LEU-COMP-THERM-065-017 MIX-COMP-THERM-013-027

B-15 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-015-076 LEU-COMP-THERM-066-004 MIX-COMP-THERM-013-028 LEU-COMP-THERM-015-077 LEU-COMP-THERM-066-005 MIX-COMP-THERM-013-029 LEU-COMP-THERM-015-078 LEU-COMP-THERM-066-006 MIX-COMP-THERM-013-030 LEU-COMP-THERM-015-079 LEU-COMP-THERM-066-007 MIX-COMP-THERM-014-001 LEU-COMP-THERM-015-080 LEU-COMP-THERM-066-008 MIX-COMP-THERM-014-002 LEU-COMP-THERM-015-081 LEU-COMP-THERM-066-009 MIX-COMP-THERM-014-003 LEU-COMP-THERM-015-082 LEU-COMP-THERM-066-010 MIX-COMP-THERM-014-004 LEU-COMP-THERM-015-083 LEU-COMP-THERM-068-004 MIX-COMP-THERM-014-005 LEU-COMP-THERM-015-084 LEU-COMP-THERM-068-005 MIX-COMP-THERM-014-006 LEU-COMP-THERM-015-085 LEU-COMP-THERM-068-006 MIX-COMP-THERM-014-007 LEU-COMP-THERM-015-086 LEU-COMP-THERM-068-007 MIX-COMP-THERM-014-008 LEU-COMP-THERM-015-087 LEU-COMP-THERM-068-008 MIX-COMP-THERM-014-009 LEU-COMP-THERM-015-088 LEU-COMP-THERM-068-009 MIX-COMP-THERM-014-010 LEU-COMP-THERM-015-089 LEU-COMP-THERM-068-010 MIX-COMP-THERM-014-011 LEU-COMP-THERM-015-090 LEU-COMP-THERM-068-011 MIX-COMP-THERM-014-012 LEU-COMP-THERM-015-091 LEU-COMP-THERM-068-012 MIX-COMP-THERM-014-013 LEU-COMP-THERM-015-092 LEU-COMP-THERM-068-013 MIX-COMP-THERM-014-014 LEU-COMP-THERM-015-093 LEU-COMP-THERM-068-014 MIX-COMP-THERM-014-015 LEU-COMP-THERM-015-094 LEU-COMP-THERM-068-015 MIX-COMP-THERM-014-016 LEU-COMP-THERM-015-095 LEU-COMP-THERM-068-016 MIX-COMP-THERM-014-017 LEU-COMP-THERM-015-096 LEU-COMP-THERM-068-017 MIX-COMP-THERM-014-018 LEU-COMP-THERM-015-097 LEU-COMP-THERM-069-001 MIX-COMP-THERM-014-019 LEU-COMP-THERM-015-098 LEU-COMP-THERM-069-002 MIX-COMP-THERM-014-020 LEU-COMP-THERM-015-099 LEU-COMP-THERM-069-003 MIX-COMP-THERM-014-021 LEU-COMP-THERM-015-100 LEU-COMP-THERM-069-004 MIX-COMP-THERM-014-022 LEU-COMP-THERM-015-101 LEU-COMP-THERM-069-005 MIX-COMP-THERM-016-001 LEU-COMP-THERM-015-102 LEU-COMP-THERM-070-001 MIX-COMP-THERM-016-002 LEU-COMP-THERM-015-103 LEU-COMP-THERM-070-002 MIX-COMP-THERM-016-003 LEU-COMP-THERM-015-104 LEU-COMP-THERM-070-003 MIX-COMP-THERM-016-004 LEU-COMP-THERM-015-105 LEU-COMP-THERM-070-004 MIX-COMP-THERM-016-005 LEU-COMP-THERM-015-106 LEU-COMP-THERM-070-005 MIX-COMP-THERM-016-006 LEU-COMP-THERM-015-107 LEU-COMP-THERM-070-006 MIX-COMP-THERM-016-007

B-16 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-015-108 LEU-COMP-THERM-070-007 MIX-COMP-THERM-016-008 LEU-COMP-THERM-015-109 LEU-COMP-THERM-070-008 MIX-COMP-THERM-016-009 LEU-COMP-THERM-015-110 LEU-COMP-THERM-070-009 MIX-COMP-THERM-016-010 LEU-COMP-THERM-015-111 LEU-COMP-THERM-070-010 MIX-COMP-THERM-016-011 LEU-COMP-THERM-015-112 LEU-COMP-THERM-070-011 MIX-COMP-THERM-016-012 LEU-COMP-THERM-015-113 LEU-COMP-THERM-070-012 MIX-COMP-THERM-016-013 LEU-COMP-THERM-015-114 LEU-COMP-THERM-071-001 MIX-COMP-THERM-016-014 LEU-COMP-THERM-015-115 LEU-COMP-THERM-071-002 MIX-COMP-THERM-016-015 LEU-COMP-THERM-015-116 LEU-COMP-THERM-071-003 MIX-COMP-THERM-016-016 LEU-COMP-THERM-015-117 LEU-COMP-THERM-071-004 MIX-COMP-THERM-016-017 LEU-COMP-THERM-015-118 LEU-COMP-THERM-072-001 MIX-COMP-THERM-016-018 LEU-COMP-THERM-015-119 LEU-COMP-THERM-072-002 MIX-COMP-THERM-016-019 LEU-COMP-THERM-015-120 LEU-COMP-THERM-072-003 MIX-COMP-THERM-017-001 LEU-COMP-THERM-015-121 LEU-COMP-THERM-072-004 MIX-COMP-THERM-017-002 LEU-COMP-THERM-015-122 LEU-COMP-THERM-072-005 MIX-COMP-THERM-017-003 LEU-COMP-THERM-015-123 LEU-COMP-THERM-072-006 MIX-COMP-THERM-017-004 LEU-COMP-THERM-015-124 LEU-COMP-THERM-072-007 MIX-COMP-THERM-017-005 LEU-COMP-THERM-015-125 LEU-COMP-THERM-072-008 MIX-COMP-THERM-017-006 LEU-COMP-THERM-015-126 LEU-COMP-THERM-072-009 MIX-COMP-THERM-017-007 LEU-COMP-THERM-015-127 LEU-COMP-THERM-073-001 MIX-COMP-THERM-017-008 LEU-COMP-THERM-015-128 LEU-COMP-THERM-073-002 MIX-COMP-THERM-017-009 LEU-COMP-THERM-015-129 LEU-COMP-THERM-073-003 MIX-COMP-THERM-017-010 LEU-COMP-THERM-015-130 LEU-COMP-THERM-073-004 MIX-COMP-THERM-017-011 LEU-COMP-THERM-015-131 LEU-COMP-THERM-073-005 MIX-COMP-THERM-017-012 LEU-COMP-THERM-015-132 LEU-COMP-THERM-073-006 MIX-COMP-THERM-017-013 LEU-COMP-THERM-015-133 LEU-COMP-THERM-073-007 MIX-COMP-THERM-017-014 LEU-COMP-THERM-015-134 LEU-COMP-THERM-073-008 MIX-COMP-THERM-017-015 LEU-COMP-THERM-015-135 LEU-COMP-THERM-073-009 MIX-COMP-THERM-017-016 LEU-COMP-THERM-015-136 LEU-COMP-THERM-073-010 MIX-COMP-THERM-017-017 LEU-COMP-THERM-015-137 LEU-COMP-THERM-073-011 MIX-COMP-THERM-017-018 LEU-COMP-THERM-015-138 LEU-COMP-THERM-073-012 MIX-COMP-THERM-017-019 LEU-COMP-THERM-015-139 LEU-COMP-THERM-073-013 HTC1-001

B-17 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-015-140 LEU-COMP-THERM-073-014 HTC1-002 LEU-COMP-THERM-015-141 LEU-COMP-THERM-074-001 HTC1-003 LEU-COMP-THERM-015-142 LEU-COMP-THERM-074-002 HTC1-004 LEU-COMP-THERM-015-143 LEU-COMP-THERM-074-003 HTC1-005 LEU-COMP-THERM-015-144 LEU-COMP-THERM-074-004 HTC1-006 LEU-COMP-THERM-015-145 LEU-COMP-THERM-074-005 HTC1-007 LEU-COMP-THERM-015-146 LEU-COMP-THERM-074-006 HTC1-008 LEU-COMP-THERM-015-147 LEU-COMP-THERM-074-007 HTC1-009 LEU-COMP-THERM-015-148 LEU-COMP-THERM-074-008 HTC1-010 LEU-COMP-THERM-015-149 LEU-COMP-THERM-074-009 HTC1-011 LEU-COMP-THERM-015-150 LEU-COMP-THERM-074-010 HTC1-012 LEU-COMP-THERM-015-151 LEU-COMP-THERM-074-011 HTC1-013 LEU-COMP-THERM-015-152 LEU-COMP-THERM-074-012 HTC1-014 LEU-COMP-THERM-015-153 LEU-COMP-THERM-074-013 HTC1-015 LEU-COMP-THERM-015-154 LEU-COMP-THERM-074-014 HTC1-016 LEU-COMP-THERM-015-155 LEU-COMP-THERM-074-015 HTC1-017 LEU-COMP-THERM-015-156 LEU-COMP-THERM-074-016 HTC1-018 LEU-COMP-THERM-015-157 LEU-COMP-THERM-074-017 HTC2B-001 LEU-COMP-THERM-015-158 LEU-COMP-THERM-074-018 HTC2B-002 LEU-COMP-THERM-015-159 LEU-COMP-THERM-074-019 HTC2B-003 LEU-COMP-THERM-015-160 LEU-COMP-THERM-074-020 HTC2B-004 LEU-COMP-THERM-015-161 LEU-COMP-THERM-074-021 HTC2B-005 LEU-COMP-THERM-015-162 LEU-COMP-THERM-074-022 HTC2B-006 LEU-COMP-THERM-015-163 LEU-COMP-THERM-074-023 HTC2B-007 LEU-COMP-THERM-015-164 LEU-COMP-THERM-074-024 HTC2B-008 LEU-COMP-THERM-015-165 LEU-COMP-THERM-074-025 HTC2B-009 LEU-COMP-THERM-016-001 LEU-COMP-THERM-074-026 HTC2B-010 LEU-COMP-THERM-016-002 LEU-COMP-THERM-074-027 HTC2B-011 LEU-COMP-THERM-016-003 LEU-COMP-THERM-074-028 HTC2B-012 LEU-COMP-THERM-016-004 LEU-COMP-THERM-075-001 HTC2B-014 LEU-COMP-THERM-016-005 LEU-COMP-THERM-075-002 HTC2B-015 LEU-COMP-THERM-016-006 LEU-COMP-THERM-075-003 HTC2B-016

B-18 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-016-007 LEU-COMP-THERM-075-004 HTC2B-017 LEU-COMP-THERM-016-008 LEU-COMP-THERM-075-005 HTC2B-018 LEU-COMP-THERM-016-009 LEU-COMP-THERM-075-006 HTC2B-019 LEU-COMP-THERM-016-010 LEU-COMP-THERM-076-001 HTC2B-020 LEU-COMP-THERM-016-011 LEU-COMP-THERM-076-002 HTC2B-021 LEU-COMP-THERM-016-012 LEU-COMP-THERM-076-003 HTC2G-001 LEU-COMP-THERM-016-013 LEU-COMP-THERM-076-004 HTC2G-002 LEU-COMP-THERM-016-014 LEU-COMP-THERM-076-005 HTC2G-003 LEU-COMP-THERM-016-015 LEU-COMP-THERM-076-006 HTC2G-004 LEU-COMP-THERM-016-016 LEU-COMP-THERM-076-007 HTC2G-005 LEU-COMP-THERM-016-017 LEU-COMP-THERM-077-001 HTC2G-006 LEU-COMP-THERM-016-018 LEU-COMP-THERM-077-002 HTC2G-007 LEU-COMP-THERM-016-019 LEU-COMP-THERM-077-003 HTC2G-008 LEU-COMP-THERM-016-020 LEU-COMP-THERM-077-004 HTC2G-009 LEU-COMP-THERM-016-021 LEU-COMP-THERM-077-005 HTC2G-010 LEU-COMP-THERM-016-022 LEU-COMP-THERM-078-001 HTC2G-011 LEU-COMP-THERM-016-023 LEU-COMP-THERM-078-002 HTC2G-012 LEU-COMP-THERM-016-024 LEU-COMP-THERM-078-003 HTC2G-013 LEU-COMP-THERM-016-025 LEU-COMP-THERM-078-004 HTC2G-014 LEU-COMP-THERM-016-026 LEU-COMP-THERM-078-005 HTC2G-015 LEU-COMP-THERM-016-027 LEU-COMP-THERM-078-006 HTC2G-016 LEU-COMP-THERM-016-028 LEU-COMP-THERM-078-007 HTC2G-017 LEU-COMP-THERM-016-029 LEU-COMP-THERM-078-008 HTC2G-018 LEU-COMP-THERM-016-030 LEU-COMP-THERM-078-009 HTC2G-019 LEU-COMP-THERM-016-031 LEU-COMP-THERM-078-010 HTC2G-020 LEU-COMP-THERM-016-032 LEU-COMP-THERM-078-011 HTC3-001 LEU-COMP-THERM-017-001 LEU-COMP-THERM-078-012 HTC3-002 LEU-COMP-THERM-017-002 LEU-COMP-THERM-078-013 HTC3-003 LEU-COMP-THERM-017-003 LEU-COMP-THERM-078-014 HTC3-004 LEU-COMP-THERM-017-004 LEU-COMP-THERM-078-015 HTC3-005 LEU-COMP-THERM-017-005 LEU-COMP-THERM-079-001 HTC3-006 LEU-COMP-THERM-017-006 LEU-COMP-THERM-079-002 HTC3-007

B-19 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-017-007 LEU-COMP-THERM-079-003 HTC3-008 LEU-COMP-THERM-017-008 LEU-COMP-THERM-079-004 HTC3-009 LEU-COMP-THERM-017-009 LEU-COMP-THERM-079-005 HTC3-010 LEU-COMP-THERM-017-010 LEU-COMP-THERM-079-006 HTC3-011 LEU-COMP-THERM-017-011 LEU-COMP-THERM-079-007 HTC3-012 LEU-COMP-THERM-017-012 LEU-COMP-THERM-079-008 HTC3-013 LEU-COMP-THERM-017-013 LEU-COMP-THERM-079-009 HTC3-014 LEU-COMP-THERM-017-014 LEU-COMP-THERM-079-010 HTC3-015 LEU-COMP-THERM-017-015 LEU-COMP-THERM-080-001 HTC3-016 LEU-COMP-THERM-017-016 LEU-COMP-THERM-080-002 HTC3-017 LEU-COMP-THERM-017-017 LEU-COMP-THERM-080-003 HTC3-018 LEU-COMP-THERM-017-018 LEU-COMP-THERM-080-004 HTC3-019 LEU-COMP-THERM-017-019 LEU-COMP-THERM-080-005 HTC3-020 LEU-COMP-THERM-017-020 LEU-COMP-THERM-080-006 HTC3-021 LEU-COMP-THERM-017-021 LEU-COMP-THERM-080-007 HTC3-022 LEU-COMP-THERM-017-022 LEU-COMP-THERM-080-008 HTC3-023 LEU-COMP-THERM-017-023 LEU-COMP-THERM-080-009 HTC3-024 LEU-COMP-THERM-017-024 LEU-COMP-THERM-080-010 HTC3-025 LEU-COMP-THERM-017-025 LEU-COMP-THERM-080-011 HTC3-026 LEU-COMP-THERM-017-026 LEU-COMP-THERM-082-002 HTC4FE-001 LEU-COMP-THERM-017-027 LEU-COMP-THERM-082-003 HTC4FE-002 LEU-COMP-THERM-017-028 LEU-COMP-THERM-082-004 HTC4FE-003 LEU-COMP-THERM-017-029 LEU-COMP-THERM-082-005 HTC4FE-004 LEU-COMP-THERM-018-001 LEU-COMP-THERM-082-006 HTC4FE-005 LEU-COMP-THERM-020-001 LEU-COMP-THERM-083-001 HTC4FE-006 LEU-COMP-THERM-020-002 LEU-COMP-THERM-083-002 HTC4FE-007 LEU-COMP-THERM-020-003 LEU-COMP-THERM-083-003 HTC4FE-008 LEU-COMP-THERM-020-004 LEU-COMP-THERM-084-001 HTC4FE-009 LEU-COMP-THERM-020-005 LEU-COMP-THERM-085-001 HTC4FE-010 LEU-COMP-THERM-020-006 LEU-COMP-THERM-085-002 HTC4FE-011 LEU-COMP-THERM-020-007 LEU-COMP-THERM-085-003 HTC4FE-012 LEU-COMP-THERM-021-001 LEU-COMP-THERM-085-004 HTC4FE-013

B-20 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-021-002 LEU-COMP-THERM-085-005 HTC4FE-014 LEU-COMP-THERM-021-003 LEU-COMP-THERM-085-006 HTC4FE-015 LEU-COMP-THERM-021-004 LEU-COMP-THERM-085-007 HTC4FE-016 LEU-COMP-THERM-021-005 LEU-COMP-THERM-085-008 HTC4FE-017 LEU-COMP-THERM-021-006 LEU-COMP-THERM-085-009 HTC4FE-018 LEU-COMP-THERM-022-001 LEU-COMP-THERM-085-010 HTC4FE-019 LEU-COMP-THERM-022-002 LEU-COMP-THERM-085-011 HTC4FE-020 LEU-COMP-THERM-022-003 LEU-COMP-THERM-085-012 HTC4FE-021 LEU-COMP-THERM-022-004 LEU-COMP-THERM-085-013 HTC4FE-022 LEU-COMP-THERM-022-005 LEU-COMP-THERM-089-001 HTC4FE-023 LEU-COMP-THERM-022-006 LEU-COMP-THERM-089-002 HTC4FE-024 LEU-COMP-THERM-022-007 LEU-COMP-THERM-089-003 HTC4FE-025 LEU-COMP-THERM-023-001 LEU-COMP-THERM-089-004 HTC4FE-026 LEU-COMP-THERM-023-002 LEU-COMP-THERM-090-001 HTC4FE-027 LEU-COMP-THERM-023-003 LEU-COMP-THERM-090-002 HTC4FE-028 LEU-COMP-THERM-023-004 LEU-COMP-THERM-090-003 HTC4FE-029 LEU-COMP-THERM-023-005 LEU-COMP-THERM-090-004 HTC4FE-030 LEU-COMP-THERM-023-006 LEU-COMP-THERM-090-005 HTC4FE-031 LEU-COMP-THERM-024-001 LEU-COMP-THERM-090-006 HTC4FE-032 LEU-COMP-THERM-024-002 LEU-COMP-THERM-090-007 HTC4FE-033 LEU-COMP-THERM-025-001 LEU-COMP-THERM-090-008 HTC4PB-001 LEU-COMP-THERM-025-002 LEU-COMP-THERM-090-009 HTC4PB-002 LEU-COMP-THERM-025-003 LEU-COMP-THERM-091-001 HTC4PB-003 LEU-COMP-THERM-025-004 LEU-COMP-THERM-091-002 HTC4PB-004 LEU-COMP-THERM-026-001 LEU-COMP-THERM-091-003 HTC4PB-005 LEU-COMP-THERM-026-002 LEU-COMP-THERM-091-004 HTC4PB-006 LEU-COMP-THERM-026-003 LEU-COMP-THERM-091-005 HTC4PB-007 LEU-COMP-THERM-026-004 LEU-COMP-THERM-091-006 HTC4PB-008 LEU-COMP-THERM-026-005 LEU-COMP-THERM-091-007 HTC4PB-009 LEU-COMP-THERM-026-006 LEU-COMP-THERM-091-008 HTC4PB-010 LEU-COMP-THERM-027-001 LEU-COMP-THERM-091-009 HTC4PB-011 LEU-COMP-THERM-027-002 LEU-COMP-THERM-092-001 HTC4PB-012

B-21 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-027-003 LEU-COMP-THERM-092-002 HTC4PB-013 LEU-COMP-THERM-027-004 LEU-COMP-THERM-092-003 HTC4PB-014 LEU-COMP-THERM-028-001 LEU-COMP-THERM-092-004 HTC4PB-015 LEU-COMP-THERM-028-002 LEU-COMP-THERM-092-005 HTC4PB-016 LEU-COMP-THERM-028-003 LEU-COMP-THERM-092-006 HTC4PB-017 LEU-COMP-THERM-028-004 LEU-COMP-THERM-094-001 HTC4PB-018 LEU-COMP-THERM-028-005 LEU-COMP-THERM-094-002 HTC4PB-019 LEU-COMP-THERM-028-006 LEU-COMP-THERM-094-003 HTC4PB-020 LEU-COMP-THERM-028-007 LEU-COMP-THERM-094-004 HTC4PB-021 LEU-COMP-THERM-028-008 LEU-COMP-THERM-094-005 HTC4PB-022 LEU-COMP-THERM-028-009 LEU-COMP-THERM-094-006 HTC4PB-023 LEU-COMP-THERM-028-010 LEU-COMP-THERM-094-007 HTC4PB-024 LEU-COMP-THERM-028-011 LEU-COMP-THERM-094-008 HTC4PB-025 LEU-COMP-THERM-028-012 LEU-COMP-THERM-094-009 HTC4PB-026 LEU-COMP-THERM-028-013 LEU-COMP-THERM-094-010 HTC4PB-027 LEU-COMP-THERM-028-014 LEU-COMP-THERM-094-011 HTC4PB-028 LEU-COMP-THERM-028-015 LEU-COMP-THERM-096-001 HTC4PB-029 LEU-COMP-THERM-028-016 LEU-COMP-THERM-096-002 HTC4PB-030 LEU-COMP-THERM-028-017 LEU-COMP-THERM-096-003 HTC4PB-031 LEU-COMP-THERM-028-018 LEU-COMP-THERM-096-004 HTC4PB-032 LEU-COMP-THERM-028-019 LEU-COMP-THERM-096-005 HTC4PB-033 LEU-COMP-THERM-028-020 LEU-COMP-THERM-096-006 HTC4PB-034 LEU-COMP-THERM-029-001 LEU-COMP-THERM-096-007 HTC4PB-035 LEU-COMP-THERM-029-002 LEU-COMP-THERM-096-008 HTC4PB-036 LEU-COMP-THERM-029-003 LEU-COMP-THERM-096-009 HTC4PB-037 LEU-COMP-THERM-029-004 LEU-COMP-THERM-096-010 HTC4PB-038 LEU-COMP-THERM-029-005 LEU-COMP-THERM-096-011 MIX-SOL-THERM-001-001 LEU-COMP-THERM-029-006 LEU-COMP-THERM-096-012 MIX-SOL-THERM-001-002 LEU-COMP-THERM-029-007 LEU-COMP-THERM-096-013 MIX-SOL-THERM-001-003 LEU-COMP-THERM-029-008 LEU-COMP-THERM-096-014 MIX-SOL-THERM-001-004 LEU-COMP-THERM-029-009 LEU-COMP-THERM-096-015 MIX-SOL-THERM-001-005 LEU-COMP-THERM-029-010 LEU-COMP-THERM-096-016 MIX-SOL-THERM-001-006

B-22 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-029-011 LEU-COMP-THERM-096-017 MIX-SOL-THERM-001-007 LEU-COMP-THERM-029-012 LEU-COMP-THERM-096-018 MIX-SOL-THERM-001-008 LEU-COMP-THERM-030-001 LEU-COMP-THERM-096-019 MIX-SOL-THERM-001-009 LEU-COMP-THERM-030-002 LEU-COMP-THERM-097-001 MIX-SOL-THERM-001-010 LEU-COMP-THERM-030-003 LEU-COMP-THERM-097-002 MIX-SOL-THERM-001-011 LEU-COMP-THERM-030-004 LEU-COMP-THERM-097-003 MIX-SOL-THERM-001-012 LEU-COMP-THERM-030-005 LEU-COMP-THERM-097-004 MIX-SOL-THERM-001-013 LEU-COMP-THERM-030-006 LEU-COMP-THERM-097-005 MIX-SOL-THERM-002-001 LEU-COMP-THERM-030-007 LEU-COMP-THERM-097-006 MIX-SOL-THERM-002-002 LEU-COMP-THERM-030-008 LEU-COMP-THERM-097-007 MIX-SOL-THERM-002-003 LEU-COMP-THERM-030-009 LEU-COMP-THERM-097-008 MIX-SOL-THERM-003-001 LEU-COMP-THERM-030-010 LEU-COMP-THERM-097-009 MIX-SOL-THERM-003-002 LEU-COMP-THERM-030-011 LEU-COMP-THERM-097-010 MIX-SOL-THERM-003-003 LEU-COMP-THERM-030-012 LEU-COMP-THERM-097-011 MIX-SOL-THERM-003-004 LEU-COMP-THERM-031-001 LEU-COMP-THERM-097-012 MIX-SOL-THERM-003-005 LEU-COMP-THERM-031-002 LEU-COMP-THERM-097-013 MIX-SOL-THERM-003-006 LEU-COMP-THERM-031-003 LEU-COMP-THERM-097-014 MIX-SOL-THERM-003-007 LEU-COMP-THERM-031-004 LEU-COMP-THERM-097-015 MIX-SOL-THERM-003-008 LEU-COMP-THERM-031-005 LEU-COMP-THERM-097-016 MIX-SOL-THERM-003-009 LEU-COMP-THERM-031-006 LEU-COMP-THERM-097-017 MIX-SOL-THERM-003-010 LEU-COMP-THERM-032-001 LEU-COMP-THERM-097-018 MIX-SOL-THERM-007-001 LEU-COMP-THERM-032-002 LEU-COMP-THERM-097-019 MIX-SOL-THERM-007-002 LEU-COMP-THERM-032-003 LEU-COMP-THERM-097-020 MIX-SOL-THERM-007-003 LEU-COMP-THERM-032-004 LEU-COMP-THERM-097-021 MIX-SOL-THERM-007-004 LEU-COMP-THERM-032-005 LEU-COMP-THERM-097-022 MIX-SOL-THERM-007-005 LEU-COMP-THERM-032-006 LEU-COMP-THERM-097-023 MIX-SOL-THERM-007-006 LEU-COMP-THERM-032-007 LEU-COMP-THERM-097-024 MIX-SOL-THERM-007-007 LEU-COMP-THERM-032-008 LEU-COMP-THERM-101-001 MIX-SOL-THERM-010-001 LEU-COMP-THERM-032-009 LEU-COMP-THERM-101-002 MIX-SOL-THERM-010-002 LEU-COMP-THERM-033-001 LEU-COMP-THERM-101-003 MIX-SOL-THERM-010-003 LEU-COMP-THERM-033-002 LEU-COMP-THERM-101-004 MIX-SOL-THERM-010-004 LEU-COMP-THERM-033-003 LEU-COMP-THERM-101-005 MIX-SOL-THERM-010-005

B-23 Table B-1 Benchmark Experiments Used in Similarity Assessments (Continued)

LEU-COMP-THERM-033-004 LEU-COMP-THERM-101-006 MIX-SOL-THERM-010-006 LEU-COMP-THERM-033-005 LEU-COMP-THERM-101-007 MIX-SOL-THERM-010-007 LEU-COMP-THERM-033-006 LEU-COMP-THERM-101-008 MIX-SOL-THERM-010-008 LEU-COMP-THERM-033-007 LEU-COMP-THERM-101-009 MIX-SOL-THERM-010-009 LEU-COMP-THERM-033-008

APPENDIX C keff RESULTS FOR GBC-32 CASES C.1 GBC-32 keff Results with the ENDF/B-VII.1 Nuclear Data Library C-1 Table C-1 Results with Initial Enrichment at 4 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.104533 0.227 3.03 0.5 2.99 0.52 20 1.054207 0.232 2.28 0.82 2.31 0.81 30 1.011945 0.233 1.69 1.05 1.89 0.97 40 0.975837 0.235 1.23 1.24 1.57 1.1 50 0.943447 0.237 0.88 1.4 1.3 1.22 55 0.929207 0.239 0.74 1.47 1.18 1.27 60 0.915334 0.24 0.63 1.54 1.07 1.32 65 0.902421 0.242 0.53 1.6 0.96 1.38 70 0.890425 0.244 0.44 1.65 0.87 1.43 75 0.879139 0.246 0.37 1.71 0.77 1.48 80 0.869015 0.248 0.32 1.76 0.69 1.53 AFP 10 1.058632 0.245 3.02 0.50 3.02 0.50 20 0.994267 0.258 2.27 0.82 2.45 0.74 30 0.942432 0.266 1.68 1.05 2.06 0.90 40 0.897992 0.275 1.22 1.24 1.74 1.02 50 0.858457 0.286 0.88 1.39 1.47 1.14 55 0.840447 0.291 0.74 1.46 1.35 1.19 60 0.823168 0.296 0.62 1.53 1.24 1.24 65 0.806993 0.302 0.52 1.59 1.13 1.28 70 0.791630 0.308 0.44 1.65 1.03 1.33 75 0.777407 0.315 0.37 1.70 0.94 1.38 80 0.763816 0.321 0.31 1.75 0.85 1.42 ALL 10 1.052834 0.247 3.02 0.50 3.02 0.50 20 0.985957 0.260 2.27 0.82 2.48 0.73 30 0.932059 0.270 1.68 1.05 2.09 0.89 40 0.885490 0.280 1.22 1.24 1.77 1.01 50 0.844477 0.291 0.88 1.39 1.50 1.12 55 0.825613 0.297 0.74 1.46 1.38 1.17 60 0.807817 0.304 0.62 1.53 1.27 1.22 65 0.790866 0.310 0.52 1.59 1.16 1.27 70 0.775020 0.318 0.44 1.65 1.06 1.32 75 0.760242 0.325 0.37 1.70 0.96 1.36 80 0.745991 0.332 0.31 1.75 0.87 1.41

C-2 Table C-2 Results with Initial Enrichment at 4.5 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final Enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.133407 0.241 3.51 0.49 3.47 0.5 20 1.085774 0.243 2.72 0.81 2.74 0.8 30 1.044087 0.242 2.07 1.05 2.26 0.98 40 1.007915 0.242 1.55 1.24 1.91 1.11 50 0.975876 0.242 1.14 1.41 1.61 1.22 55 0.960872 0.242 0.97 1.48 1.48 1.27 60 0.946749 0.243 0.83 1.54 1.36 1.32 65 0.933456 0.244 0.71 1.61 1.25 1.37 70 0.920279 0.245 0.6 1.67 1.13 1.42 75 0.908299 0.246 0.51 1.72 1.03 1.47 80 0.897195 0.247 0.43 1.78 0.94 1.52 AFP 10 1.087633 0.260 3.50 0.48 3.49 0.49 20 1.025157 0.270 2.71 0.80 2.88 0.74 30 0.973882 0.277 2.06 1.04 2.45 0.90 40 0.929800 0.283 1.54 1.24 2.10 1.03 50 0.890261 0.291 1.13 1.40 1.81 1.14 55 0.871995 0.294 0.97 1.47 1.68 1.19 60 0.854494 0.299 0.82 1.53 1.55 1.24 65 0.837893 0.303 0.70 1.60 1.43 1.28 70 0.821976 0.308 0.59 1.65 1.32 1.33 75 0.806949 0.313 0.50 1.71 1.22 1.37 80 0.792631 0.318 0.43 1.76 1.12 1.42 ALL 10 1.082230 0.262 3.50 0.48 3.49 0.49 20 1.016303 0.273 2.71 0.80 2.90 0.73 30 0.963517 0.281 2.06 1.04 2.48 0.88 40 0.917335 0.288 1.54 1.24 2.14 1.01 50 0.876117 0.297 1.13 1.40 1.84 1.12 55 0.857056 0.301 0.97 1.47 1.71 1.17 60 0.839185 0.306 0.82 1.53 1.58 1.22 65 0.821708 0.312 0.70 1.60 1.46 1.27 70 0.805399 0.317 0.59 1.65 1.35 1.32 75 0.789791 0.323 0.50 1.71 1.25 1.36 80 0.774802 0.329 0.43 1.76 1.14 1.40

C-3 Table C-3 Results with Initial Enrichment at 5.0 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.158718 0.257 3.99 0.47 3.95 0.49 20 1.113129 0.256 3.16 0.8 3.16 0.8 30 1.072493 0.253 2.47 1.04 2.64 0.98 40 1.037058 0.25 1.89 1.24 2.25 1.12 50 1.005193 0.249 1.43 1.41 1.94 1.23 55 0.990122 0.248 1.23 1.48 1.79 1.28 60 0.975768 0.248 1.06 1.55 1.66 1.33 65 0.962098 0.248 0.91 1.62 1.54 1.38 70 0.949375 0.249 0.78 1.68 1.42 1.42 75 0.936643 0.249 0.67 1.74 1.31 1.47 80 0.924789 0.25 0.57 1.79 1.2 1.51 AFP 10 1.113181 0.277 3.98 0.47 3.96 0.48 20 1.051603 0.285 3.15 0.79 3.30 0.74 30 1.001628 0.289 2.46 1.04 2.85 0.90 40 0.958252 0.294 1.88 1.24 2.47 1.03 50 0.919083 0.299 1.42 1.40 2.15 1.14 55 0.900649 0.302 1.23 1.47 2.01 1.19 60 0.883265 0.305 1.05 1.54 1.87 1.24 65 0.866516 0.308 0.91 1.61 1.75 1.29 70 0.850590 0.312 0.78 1.67 1.63 1.33 75 0.835209 0.316 0.66 1.72 1.51 1.38 80 0.820373 0.320 0.57 1.78 1.40 1.42 ALL 10 1.107882 0.279 3.98 0.47 3.97 0.48 20 1.043170 0.288 3.15 0.79 3.33 0.72 30 0.991094 0.293 2.46 1.04 2.88 0.89 40 0.945898 0.299 1.88 1.24 2.51 1.02 50 0.904990 0.305 1.42 1.40 2.19 1.13 55 0.886238 0.309 1.23 1.47 2.05 1.18 60 0.868003 0.312 1.05 1.54 1.91 1.23 65 0.850760 0.316 0.91 1.61 1.78 1.27 70 0.834050 0.320 0.78 1.67 1.66 1.32 75 0.817954 0.325 0.66 1.72 1.54 1.36 80 0.802552 0.330 0.57 1.78 1.43 1.40

C-4 Table C-4 Results with Initial Enrichment at 5.5 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.181002 0.274 4.47 0.46 4.43 0.48 20 1.137143 0.271 3.62 0.79 3.6 0.79 30 1.097791 0.266 2.88 1.04 3.03 0.99 40 1.063203 0.261 2.26 1.24 2.61 1.13 50 1.031906 0.258 1.74 1.42 2.27 1.24 55 1.017112 0.256 1.52 1.49 2.12 1.29 60 1.003113 0.255 1.32 1.56 1.97 1.34 65 0.989632 0.255 1.15 1.63 1.84 1.39 70 0.976085 0.254 0.99 1.69 1.71 1.43 75 0.96351 0.254 0.86 1.75 1.59 1.47 80 0.951245 0.254 0.74 1.81 1.48 1.52 AFP 10 1.135762 0.295 4.46 0.46 4.44 0.47 20 1.075732 0.302 3.60 0.78 3.74 0.73 30 1.026750 0.304 2.87 1.03 3.25 0.90 40 0.983976 0.307 2.25 1.23 2.85 1.03 50 0.945219 0.310 1.73 1.40 2.51 1.14 55 0.927461 0.312 1.51 1.48 2.36 1.20 60 0.909996 0.314 1.31 1.55 2.21 1.25 65 0.893155 0.316 1.14 1.62 2.07 1.29 70 0.877271 0.318 0.98 1.68 1.94 1.34 75 0.861572 0.321 0.85 1.74 1.82 1.38 80 0.846647 0.324 0.73 1.79 1.70 1.42 ALL 10 1.130258 0.297 4.46 0.46 4.45 0.47 20 1.067335 0.305 3.60 0.78 3.77 0.72 30 1.016003 0.308 2.87 1.03 3.29 0.89 40 0.971565 0.312 2.25 1.23 2.89 1.02 50 0.931531 0.316 1.73 1.40 2.55 1.13 55 0.912757 0.319 1.51 1.48 2.40 1.18 60 0.894735 0.321 1.31 1.55 2.25 1.23 65 0.877378 0.324 1.14 1.62 2.11 1.28 70 0.860327 0.327 0.98 1.68 1.98 1.32 75 0.844297 0.331 0.85 1.74 1.86 1.37 80 0.829163 0.334 0.73 1.79 1.74 1.41

C-5 Table C-5 Results with Initial Enrichment at 6.0 wt% 235U (ENDF/B-VII.1)

Set BU (GWd

/MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.200543 0.293 4.96 0.45 4.91 0.47 20 1.158805 0.288 4.08 0.78 4.05 0.79 30 1.120732 0.28 3.31 1.03 3.43 0.99 40 1.087297 0.273 2.65 1.24 2.98 1.13 50 1.056428 0.268 2.08 1.42 2.61 1.25 55 1.041821 0.266 1.83 1.5 2.45 1.3 60 1.027885 0.264 1.61 1.57 2.3 1.35 65 1.014256 0.263 1.41 1.64 2.15 1.4 70 1.001277 0.262 1.23 1.7 2.02 1.44 75 0.988821 0.26 1.07 1.76 1.89 1.49 80 0.976405 0.26 0.93 1.82 1.77 1.53 AFP 10 1.155634 0.315 4.95 0.45 4.92 0.46 20 1.097147 0.320 4.06 0.77 4.19 0.73 30 1.049255 0.321 3.29 1.03 3.67 0.90 40 1.007331 0.322 2.63 1.23 3.24 1.04 50 0.969330 0.323 2.06 1.41 2.88 1.15 55 0.951626 0.324 1.82 1.49 2.71 1.20 60 0.934565 0.325 1.60 1.56 2.56 1.25 65 0.917715 0.326 1.40 1.63 2.41 1.30 70 0.901837 0.328 1.22 1.69 2.27 1.34 75 0.886154 0.330 1.06 1.75 2.14 1.39 80 0.871568 0.332 0.92 1.81 2.01 1.43 ALL 10 1.149840 0.317 4.95 0.45 4.93 0.46 20 1.088495 0.323 4.06 0.77 4.21 0.72 30 1.038692 0.325 3.29 1.03 3.71 0.89 40 0.994832 0.327 2.63 1.23 3.29 1.02 50 0.955603 0.329 2.06 1.41 2.93 1.14 55 0.936918 0.331 1.82 1.49 2.76 1.19 60 0.919105 0.333 1.60 1.56 2.61 1.24 65 0.901889 0.334 1.40 1.63 2.46 1.28 70 0.885266 0.337 1.22 1.69 2.32 1.33 75 0.869256 0.339 1.06 1.75 2.18 1.37 80 0.853679 0.342 0.92 1.81 2.05 1.41

C-6 Table C-6 Results with Initial Enrichment at 6.5 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.217899 0.313 5.45 0.44 5.4 0.46 20 1.177939 0.305 4.55 0.77 4.51 0.78 30 1.141324 0.295 3.75 1.03 3.82 1

40 1.108637 0.287 3.05 1.24 3.35 1.15 50 1.078719 0.28 2.44 1.42 2.96 1.26 55 1.064379 0.277 2.17 1.5 2.79 1.31 60 1.05054 0.275 1.92 1.58 2.63 1.36 65 1.037442 0.272 1.7 1.65 2.48 1.41 70 1.02454 0.27 1.49 1.72 2.33 1.45 75 1.012087 0.269 1.31 1.78 2.2 1.5 80 0.999698 0.267 1.14 1.84 2.06 1.54 AFP 10 1.173484 0.337 5.44 0.44 5.41 0.45 20 1.116063 0.340 4.53 0.76 4.63 0.73 30 1.069145 0.340 3.73 1.02 4.08 0.90 40 1.028484 0.338 3.02 1.23 3.64 1.04 50 0.991266 0.338 2.42 1.41 3.26 1.16 55 0.973857 0.338 2.15 1.49 3.08 1.21 60 0.956838 0.338 1.90 1.56 2.92 1.26 65 0.940349 0.339 1.68 1.63 2.76 1.31 70 0.924688 0.339 1.48 1.70 2.61 1.35 75 0.909466 0.340 1.30 1.76 2.47 1.39 80 0.894423 0.341 1.13 1.82 2.33 1.44 ALL 10 1.167946 0.339 5.44 0.44 5.42 0.45 20 1.107687 0.343 4.53 0.76 4.66 0.72 30 1.058702 0.344 3.73 1.02 4.13 0.89 40 1.016122 0.344 3.02 1.23 3.69 1.03 50 0.977486 0.344 2.42 1.41 3.31 1.14 55 0.959165 0.345 2.15 1.49 3.14 1.19 60 0.941668 0.346 1.90 1.56 2.97 1.24 65 0.924885 0.347 1.68 1.63 2.81 1.29 70 0.908241 0.348 1.48 1.70 2.66 1.34 75 0.892196 0.350 1.30 1.76 2.52 1.38 80 0.876728 0.352 1.13 1.82 2.38 1.42

C-7 Table C-7 Results with Initial Enrichment at 7.0 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.233673 0.334 5.95 0.43 5.9 0.45 20 1.195442 0.325 5.02 0.76 4.98 0.77 30 1.160314 0.312 4.19 1.02 4.25 1

40 1.128224 0.302 3.46 1.24 3.73 1.16 50 1.098994 0.293 2.81 1.42 3.32 1.28 55 1.085128 0.289 2.52 1.51 3.13 1.33 60 1.07192 0.286 2.25 1.59 2.96 1.38 65 1.058747 0.283 2

1.66 2.8 1.43 70 1.046159 0.281 1.78 1.73 2.65 1.47 75 1.033636 0.278 1.57 1.79 2.51 1.51 80 1.0218 0.276 1.38 1.85 2.37 1.55 AFP 10 1.189287 0.360 5.93 0.43 5.90 0.44 20 1.133677 0.362 5.00 0.75 5.09 0.72 30 1.087607 0.359 4.17 1.01 4.51 0.90 40 1.047768 0.356 3.43 1.23 4.04 1.04 50 1.011258 0.354 2.79 1.41 3.64 1.16 55 0.993989 0.354 2.50 1.49 3.46 1.22 60 0.977443 0.353 2.23 1.57 3.28 1.27 65 0.961415 0.353 1.98 1.64 3.12 1.31 70 0.945803 0.353 1.76 1.71 2.96 1.36 75 0.930664 0.353 1.55 1.77 2.81 1.40 80 0.916102 0.353 1.37 1.83 2.67 1.44 ALL 10 1.183922 0.362 5.93 0.43 5.90 0.44 20 1.125042 0.365 5.00 0.75 5.12 0.71 30 1.077181 0.364 4.17 1.01 4.56 0.89 40 1.035345 0.362 3.43 1.23 4.10 1.03 50 0.997291 0.361 2.79 1.41 3.70 1.15 55 0.979533 0.361 2.50 1.49 3.51 1.20 60 0.962342 0.361 2.23 1.57 3.34 1.25 65 0.945496 0.361 1.98 1.64 3.17 1.30 70 0.929356 0.362 1.76 1.71 3.02 1.34 75 0.913548 0.362 1.55 1.77 2.86 1.39 80 0.898361 0.363 1.37 1.83 2.72 1.43

C-8 Table C-8 Results with Initial Enrichment at 7.5 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.247811 0.357 6.44 0.42 6.39 0.44 20 1.211508 0.345 5.5 0.75 5.45 0.77 30 1.177247 0.33 4.65 1.01 4.68 1

40 1.146397 0.318 3.88 1.23 4.13 1.16 50 1.117826 0.308 3.2 1.43 3.68 1.29 55 1.104276 0.303 2.89 1.51 3.49 1.34 60 1.091054 0.299 2.6 1.59 3.31 1.39 65 1.078522 0.295 2.33 1.67 3.14 1.44 70 1.065992 0.292 2.08 1.74 2.98 1.48 75 1.053833 0.289 1.86 1.8 2.83 1.53 80 1.041867 0.286 1.65 1.87 2.68 1.57 AFP 10 1.203733 0.384 6.43 0.42 6.39 0.43 20 1.149295 0.385 5.47 0.74 5.54 0.72 30 1.104389 0.381 4.62 1.01 4.94 0.91 40 1.065320 0.376 3.85 1.22 4.45 1.05 50 1.029723 0.373 3.17 1.41 4.03 1.17 55 1.012828 0.371 2.86 1.50 3.84 1.22 60 0.996515 0.370 2.58 1.57 3.66 1.27 65 0.980758 0.369 2.31 1.65 3.48 1.32 70 0.965334 0.368 2.06 1.72 3.32 1.37 75 0.950515 0.367 1.84 1.78 3.16 1.41 80 0.936045 0.367 1.63 1.84 3.01 1.46 ALL 10 1.198420 0.386 6.43 0.42 6.39 0.43 20 1.141000 0.388 5.47 0.74 5.58 0.71 30 1.093874 0.386 4.62 1.01 4.99 0.89 40 1.052818 0.382 3.85 1.22 4.51 1.03 50 1.015692 0.380 3.17 1.41 4.09 1.15 55 0.998357 0.379 2.86 1.50 3.90 1.21 60 0.981367 0.378 2.58 1.57 3.72 1.26 65 0.965076 0.378 2.31 1.65 3.55 1.31 70 0.948968 0.377 2.06 1.72 3.38 1.35 75 0.933292 0.377 1.84 1.78 3.22 1.40 80 0.918504 0.377 1.63 1.84 3.07 1.44

C-9 Table C-9 Results with Initial Enrichment at 8.0 wt% 235U (ENDF/B-VII.1)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.261066 0.381 6.94 0.41 6.88 0.43 20 1.225974 0.367 5.98 0.74 5.92 0.76 30 1.193007 0.35 5.11 1

5.12 1

40 1.162654 0.335 4.32 1.23 4.52 1.17 50 1.134903 0.323 3.6 1.43 4.05 1.3 55 1.121988 0.318 3.28 1.51 3.85 1.35 60 1.109005 0.313 2.97 1.6 3.66 1.41 65 1.096544 0.309 2.68 1.67 3.48 1.45 70 1.084412 0.304 2.41 1.75 3.31 1.5 75 1.072756 0.301 2.16 1.82 3.16 1.54 80 1.061019 0.297 1.93 1.88 3.01 1.58 AFP 10 1.217018 0.410 6.92 0.41 6.88 0.43 20 1.164013 0.409 5.95 0.74 6.01 0.72 30 1.119718 0.404 5.08 1.00 5.38 0.91 40 1.081438 0.397 4.28 1.22 4.87 1.05 50 1.046595 0.392 3.57 1.41 4.43 1.17 55 1.029986 0.390 3.24 1.50 4.23 1.23 60 1.013937 0.388 2.94 1.58 4.04 1.28 65 0.998932 0.386 2.65 1.65 3.86 1.33 70 0.983380 0.384 2.38 1.73 3.68 1.38 75 0.968889 0.383 2.14 1.79 3.52 1.42 80 0.954511 0.382 1.91 1.86 3.36 1.46 ALL 10 1.211905 0.413 6.92 0.41 6.88 0.43 20 1.155662 0.413 5.95 0.74 6.04 0.71 30 1.109460 0.409 5.08 1.00 5.43 0.89 40 1.069086 0.404 4.28 1.22 4.93 1.03 50 1.032613 0.400 3.57 1.41 4.50 1.16 55 1.015432 0.398 3.24 1.50 4.30 1.21 60 0.999021 0.397 2.94 1.58 4.11 1.26 65 0.982900 0.395 2.65 1.65 3.93 1.31 70 0.967100 0.394 2.38 1.73 3.75 1.36 75 0.951887 0.393 2.14 1.79 3.59 1.40 80 0.936829 0.393 1.91 1.86 3.43 1.45

C-10 C.2 GBC-32 keff Results with the ENDF/B-VIII.0 Nuclear Data Library Table C-10 Results with Initial Enrichment at 4.0 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.101823 0.225 3.02 0.5 2.98 0.51 20 1.051735 0.228 2.26 0.81 2.3 0.8 30 1.00873 0.23 1.67 1.05 1.88 0.96 40 0.971891 0.232 1.2 1.24 1.56 1.09 50 0.939245 0.234 0.85 1.39 1.29 1.2 55 0.924152 0.236 0.72 1.46 1.17 1.26 60 0.910129 0.238 0.6 1.53 1.06 1.31 65 0.896971 0.24 0.5 1.59 0.95 1.36 70 0.884509 0.242 0.42 1.65 0.85 1.41 75 0.873386 0.244 0.35 1.7 0.76 1.46 80 0.862603 0.246 0.3 1.76 0.67 1.52 AFP 10 1.056331 0.242 3.01 0.50 3.01 0.50 20 0.991902 0.254 2.26 0.81 2.45 0.73 30 0.939947 0.262 1.66 1.05 2.05 0.89 40 0.894762 0.271 1.20 1.23 1.73 1.02 50 0.854519 0.281 0.85 1.39 1.46 1.13 55 0.836341 0.287 0.71 1.46 1.34 1.18 60 0.818916 0.292 0.60 1.52 1.22 1.23 65 0.802496 0.298 0.50 1.58 1.12 1.27 70 0.786964 0.304 0.42 1.64 1.02 1.32 75 0.772238 0.311 0.35 1.69 0.92 1.37 80 0.758679 0.318 0.29 1.75 0.83 1.41 ALL 10 1.050762 0.243 3.01 0.50 3.01 0.49 20 0.983619 0.256 2.26 0.81 2.47 0.72 30 0.929488 0.266 1.66 1.05 2.08 0.88 40 0.882633 0.276 1.20 1.23 1.76 1.00 50 0.840845 0.287 0.85 1.39 1.49 1.11 55 0.821806 0.293 0.71 1.46 1.37 1.16 60 0.803617 0.300 0.60 1.52 1.25 1.21 65 0.786574 0.307 0.50 1.58 1.14 1.26 70 0.770517 0.313 0.42 1.64 1.04 1.31 75 0.755209 0.321 0.35 1.69 0.94 1.35 80 0.740984 0.329 0.29 1.75 0.86 1.40

C-11 Table C-11 Results with Initial Enrichment at 4.5 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.130903 0.238 3.5 0.48 3.46 0.5 20 1.08263 0.24 2.7 0.8 2.72 0.8 30 1.040798 0.239 2.04 1.05 2.24 0.97 40 1.004444 0.238 1.52 1.24 1.89 1.1 50 0.971803 0.239 1.11 1.4 1.6 1.21 55 0.956561 0.239 0.94 1.47 1.47 1.26 60 0.942051 0.24 0.8 1.54 1.35 1.31 65 0.928398 0.241 0.67 1.6 1.23 1.36 70 0.915268 0.242 0.57 1.66 1.12 1.41 75 0.902935 0.244 0.48 1.72 1.02 1.45 80 0.891394 0.245 0.41 1.77 0.92 1.5 AFP 10 1.085469 0.257 3.49 0.48 3.48 0.49 20 1.022680 0.266 2.69 0.80 2.86 0.73 30 0.971717 0.272 2.03 1.04 2.44 0.89 40 0.927073 0.279 1.51 1.23 2.09 1.02 50 0.886892 0.286 1.10 1.39 1.79 1.13 55 0.868383 0.290 0.94 1.46 1.66 1.18 60 0.850654 0.294 0.79 1.53 1.53 1.23 65 0.833658 0.299 0.67 1.59 1.41 1.27 70 0.817760 0.304 0.57 1.65 1.30 1.32 75 0.802365 0.309 0.48 1.70 1.19 1.36 80 0.787921 0.314 0.40 1.76 1.10 1.41 ALL 10 1.079778 0.258 3.49 0.48 3.48 0.48 20 1.014153 0.269 2.69 0.80 2.89 0.72 30 0.960910 0.276 2.03 1.04 2.47 0.88 40 0.914620 0.284 1.51 1.23 2.12 1.01 50 0.873045 0.292 1.10 1.39 1.82 1.12 55 0.853685 0.297 0.94 1.46 1.69 1.17 60 0.835305 0.302 0.79 1.53 1.56 1.21 65 0.817854 0.307 0.67 1.59 1.44 1.26 70 0.801330 0.313 0.57 1.65 1.33 1.31 75 0.785327 0.318 0.48 1.70 1.23 1.35 80 0.770248 0.325 0.40 1.76 1.12 1.39

C-12 Table C-12 Results with Initial Enrichment at 5.0 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.156075 0.254 3.98 0.47 3.94 0.49 20 1.110055 0.253 3.14 0.79 3.15 0.79 30 1.069613 0.249 2.44 1.04 2.62 0.98 40 1.033699 0.247 1.86 1.24 2.23 1.11 50 1.001409 0.245 1.39 1.4 1.92 1.22 55 0.986341 0.245 1.2 1.48 1.78 1.27 60 0.971708 0.245 1.03 1.55 1.65 1.32 65 0.957998 0.245 0.88 1.61 1.52 1.36 70 0.944693 0.245 0.75 1.67 1.4 1.41 75 0.931783 0.246 0.64 1.73 1.29 1.46 80 0.919623 0.247 0.54 1.78 1.19 1.5 AFP 10 1.105187 0.275 3.97 0.47 3.95 0.48 20 1.041041 0.284 3.13 0.79 3.29 0.73 30 0.988845 0.289 2.43 1.03 2.83 0.89 40 0.943459 0.294 1.85 1.23 2.46 1.02 50 0.902193 0.300 1.38 1.40 2.14 1.13 55 0.883009 0.304 1.19 1.47 1.99 1.18 60 0.864786 0.307 1.02 1.54 1.85 1.23 65 0.847184 0.311 0.87 1.60 1.73 1.28 70 0.830219 0.316 0.74 1.66 1.60 1.32 75 0.814074 0.320 0.63 1.72 1.49 1.37 80 0.798583 0.325 0.54 1.77 1.38 1.41 ALL 10 1.151037 0.256 3.97 0.47 3.96 0.47 20 1.102028 0.257 3.13 0.79 3.32 0.72 30 1.059673 0.255 2.43 1.03 2.87 0.88 40 1.022458 0.254 1.85 1.23 2.49 1.01 50 0.988584 0.254 1.38 1.40 2.17 1.12 55 0.972792 0.254 1.19 1.47 2.03 1.17 60 0.957801 0.255 1.02 1.54 1.89 1.22 65 0.943122 0.256 0.87 1.60 1.76 1.26 70 0.929393 0.256 0.74 1.66 1.64 1.31 75 0.915869 0.257 0.63 1.72 1.52 1.35 80 0.903098 0.259 0.54 1.77 1.41 1.39

C-13 Table C-13 Results with Initial Enrichment at 5.5 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.177996 0.271 4.47 0.46 4.42 0.48 20 1.134633 0.267 3.6 0.78 3.58 0.79 30 1.095006 0.262 2.86 1.03 3.01 0.98 40 1.060135 0.257 2.23 1.24 2.59 1.12 50 1.028588 0.254 1.71 1.41 2.26 1.23 55 1.013447 0.253 1.48 1.49 2.1 1.28 60 0.999162 0.252 1.28 1.56 1.96 1.33 65 0.985314 0.251 1.11 1.62 1.82 1.37 70 0.971938 0.251 0.95 1.68 1.7 1.42 75 0.958956 0.25 0.82 1.74 1.58 1.46 80 0.94643 0.251 0.7 1.8 1.46 1.5 AFP 10 1.133043 0.292 4.46 0.46 4.43 0.47 20 1.073295 0.297 3.58 0.78 3.72 0.73 30 1.024374 0.300 2.84 1.03 3.24 0.89 40 0.981719 0.302 2.21 1.23 2.84 1.03 50 0.942653 0.305 1.69 1.40 2.49 1.14 55 0.924458 0.307 1.47 1.47 2.34 1.19 60 0.906749 0.309 1.27 1.54 2.19 1.24 65 0.889846 0.311 1.10 1.61 2.05 1.28 70 0.873723 0.313 0.95 1.67 1.92 1.33 75 0.857881 0.316 0.81 1.73 1.80 1.37 80 0.842705 0.319 0.70 1.78 1.67 1.41 ALL 10 1.127658 0.293 4.46 0.46 4.44 0.46 20 1.064947 0.300 3.58 0.78 3.76 0.72 30 1.013992 0.303 2.84 1.03 3.27 0.88 40 0.969088 0.307 2.21 1.23 2.88 1.01 50 0.928808 0.311 1.69 1.40 2.54 1.12 55 0.909978 0.313 1.47 1.47 2.38 1.17 60 0.891573 0.316 1.27 1.54 2.23 1.22 65 0.873892 0.319 1.10 1.61 2.09 1.27 70 0.857298 0.322 0.95 1.67 1.96 1.31 75 0.840813 0.325 0.81 1.73 1.83 1.36 80 0.825358 0.329 0.70 1.78 1.71 1.40

C-14 Table C-14 Results with Initial Enrichment at 6.0 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.197645 0.29 4.95 0.45 4.91 0.47 20 1.155992 0.284 4.06 0.77 4.04 0.78 30 1.11804 0.276 3.28 1.03 3.4 0.99 40 1.084049 0.269 2.61 1.24 2.95 1.13 50 1.053203 0.264 2.04 1.41 2.59 1.24 55 1.038587 0.262 1.79 1.49 2.43 1.29 60 1.024484 0.261 1.57 1.57 2.28 1.34 65 1.010756 0.259 1.37 1.63 2.14 1.39 70 0.997531 0.258 1.19 1.7 2

1.43 75 0.984583 0.257 1.03 1.76 1.87 1.47 80 0.972075 0.256 0.89 1.81 1.75 1.51 AFP 10 1.153105 0.312 4.94 0.45 4.92 0.46 20 1.094584 0.316 4.04 0.77 4.17 0.73 30 1.046815 0.316 3.26 1.02 3.65 0.90 40 1.005166 0.316 2.59 1.23 3.22 1.03 50 0.966866 0.318 2.02 1.40 2.86 1.14 55 0.948949 0.319 1.78 1.48 2.69 1.19 60 0.931503 0.319 1.55 1.55 2.54 1.24 65 0.914829 0.321 1.35 1.62 2.39 1.29 70 0.898690 0.322 1.18 1.68 2.25 1.33 75 0.882704 0.324 1.02 1.74 2.11 1.38 80 0.867734 0.326 0.88 1.80 1.99 1.42 ALL 10 1.147432 0.313 4.94 0.45 4.92 0.45 20 1.086246 0.319 4.04 0.77 4.20 0.71 30 1.036236 0.320 3.26 1.02 3.69 0.88 40 0.992653 0.322 2.59 1.23 3.27 1.02 50 0.953009 0.324 2.02 1.40 2.91 1.13 55 0.934508 0.325 1.78 1.48 2.74 1.18 60 0.916378 0.327 1.55 1.55 2.58 1.23 65 0.899030 0.329 1.35 1.62 2.44 1.27 70 0.882035 0.331 1.18 1.68 2.29 1.32 75 0.865897 0.333 1.02 1.74 2.16 1.36 80 0.850269 0.336 0.88 1.80 2.03 1.40

C-15 Table C-15 Results with Initial Enrichment at 6.5 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.214811 0.309 5.45 0.44 5.4 0.46 20 1.175257 0.301 4.53 0.76 4.49 0.78 30 1.138724 0.291 3.72 1.02 3.81 0.99 40 1.105872 0.282 3.01 1.24 3.33 1.14 50 1.075593 0.276 2.4 1.42 2.94 1.25 55 1.061384 0.273 2.13 1.5 2.77 1.3 60 1.047342 0.271 1.88 1.57 2.61 1.35 65 1.03397 0.268 1.65 1.64 2.46 1.4 70 1.021019 0.266 1.45 1.71 2.31 1.44 75 1.008239 0.265 1.26 1.77 2.18 1.49 80 0.99555 0.263 1.1 1.83 2.05 1.53 AFP 10 1.170530 0.333 5.43 0.44 5.40 0.45 20 1.113736 0.336 4.51 0.76 4.62 0.72 30 1.066905 0.334 3.70 1.02 4.07 0.90 40 1.026272 0.333 2.99 1.23 3.62 1.03 50 0.988799 0.332 2.38 1.40 3.24 1.15 55 0.971347 0.332 2.11 1.48 3.06 1.20 60 0.954186 0.332 1.86 1.56 2.90 1.25 65 0.937467 0.333 1.63 1.63 2.74 1.30 70 0.921898 0.334 1.43 1.69 2.59 1.34 75 0.906309 0.335 1.25 1.75 2.45 1.39 80 0.891264 0.336 1.09 1.81 2.31 1.43 ALL 10 1.165181 0.335 5.43 0.44 5.41 0.45 20 1.105475 0.339 4.51 0.76 4.65 0.71 30 1.056524 0.339 3.70 1.02 4.11 0.88 40 1.013878 0.338 2.99 1.23 3.67 1.02 50 0.974986 0.339 2.38 1.40 3.29 1.13 55 0.956859 0.339 2.11 1.48 3.11 1.19 60 0.939007 0.340 1.86 1.56 2.95 1.24 65 0.922063 0.341 1.63 1.63 2.79 1.28 70 0.905406 0.342 1.43 1.69 2.64 1.33 75 0.889250 0.344 1.25 1.75 2.50 1.37 80 0.873628 0.345 1.09 1.81 2.36 1.41

C-16 Table C-16 Results with Initial Enrichment at 7.0 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.230468 0.331 5.94 0.43 5.89 0.45 20 1.19256 0.32 5

0.75 4.96 0.77 30 1.157343 0.308 4.17 1.02 4.23 1

40 1.125379 0.297 3.42 1.23 3.71 1.15 50 1.096245 0.289 2.77 1.42 3.3 1.27 55 1.082056 0.285 2.48 1.5 3.12 1.32 60 1.068561 0.282 2.21 1.58 2.95 1.37 65 1.055387 0.279 1.96 1.65 2.78 1.41 70 1.042709 0.276 1.73 1.72 2.64 1.46 75 1.030133 0.274 1.52 1.79 2.49 1.5 80 1.01799 0.272 1.34 1.85 2.36 1.54 AFP 10 1.181057 0.358 5.92 0.43 5.89 0.44 20 1.122404 0.360 4.98 0.75 5.07 0.72 30 1.074883 0.359 4.14 1.01 4.49 0.90 40 1.033187 0.357 3.40 1.22 4.02 1.04 50 0.995309 0.355 2.75 1.41 3.62 1.16 55 0.976994 0.355 2.45 1.49 3.44 1.21 60 0.960077 0.355 2.18 1.56 3.26 1.26 65 0.943006 0.355 1.94 1.64 3.10 1.31 70 0.926708 0.356 1.71 1.70 2.94 1.35 75 0.910867 0.356 1.51 1.77 2.79 1.39 80 0.895458 0.357 1.32 1.83 2.64 1.44 ALL 10 1.224866 0.334 5.92 0.43 5.90 0.44 20 1.183460 0.326 4.98 0.75 5.10 0.71 30 1.146106 0.316 4.14 1.01 4.54 0.89 40 1.112295 0.308 3.40 1.22 4.08 1.02 50 1.081432 0.301 2.75 1.41 3.68 1.14 55 1.066738 0.297 2.45 1.49 3.49 1.19 60 1.052732 0.295 2.18 1.56 3.32 1.24 65 1.038689 0.292 1.94 1.64 3.15 1.29 70 1.025164 0.291 1.71 1.70 2.99 1.33 75 1.012205 0.289 1.51 1.77 2.84 1.38 80 0.999469 0.287 1.32 1.83 2.70 1.42

C-17 Table C-17 Results with Initial Enrichment at 7.5 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.244415 0.353 6.43 0.42 6.38 0.44 20 1.208206 0.341 5.48 0.74 5.43 0.76 30 1.174415 0.326 4.62 1.01 4.66 1

40 1.143688 0.313 3.85 1.23 4.11 1.16 50 1.115103 0.303 3.16 1.42 3.66 1.28 55 1.101512 0.298 2.85 1.51 3.47 1.33 60 1.088178 0.294 2.55 1.59 3.29 1.38 65 1.07538 0.291 2.28 1.66 3.12 1.43 70 1.062665 0.288 2.03 1.73 2.96 1.47 75 1.050509 0.285 1.81 1.8 2.81 1.52 80 1.038427 0.282 1.6 1.86 2.67 1.56 AFP 10 1.200679 0.380 6.42 0.42 6.38 0.43 20 1.147136 0.380 5.46 0.74 5.54 0.72 30 1.102226 0.375 4.59 1.00 4.93 0.90 40 1.063135 0.370 3.82 1.22 4.43 1.04 50 1.027371 0.367 3.13 1.41 4.01 1.16 55 1.010455 0.365 2.82 1.49 3.82 1.22 60 0.994181 0.364 2.53 1.57 3.63 1.27 65 0.978199 0.362 2.26 1.64 3.46 1.31 70 0.962922 0.361 2.01 1.71 3.30 1.36 75 0.947910 0.361 1.78 1.78 3.13 1.40 80 0.933389 0.360 1.58 1.84 2.99 1.44 ALL 10 1.195435 0.383 6.42 0.42 6.38 0.43 20 1.138323 0.384 5.46 0.74 5.56 0.71 30 1.091615 0.380 4.59 1.00 4.98 0.89 40 1.050863 0.377 3.82 1.22 4.49 1.03 50 1.013641 0.374 3.13 1.41 4.07 1.14 55 0.995803 0.373 2.82 1.49 3.88 1.20 60 0.978947 0.372 2.53 1.57 3.70 1.25 65 0.962296 0.371 2.26 1.64 3.52 1.30 70 0.946673 0.371 2.01 1.71 3.36 1.34 75 0.930984 0.370 1.78 1.78 3.20 1.39 80 0.915551 0.371 1.58 1.84 3.04 1.43

C-18 Table C-18 Results with Initial Enrichment at 8.0 wt% 235U (ENDF/B-VIII.0)

Set BU (GWd/

MTU) keff EALF (eV)

Simple average Fission density weighted average Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

Final enr.

(w/o 235U)

Pu/(Pu+U)

(w/o Pu)

AO 10 1.257174 0.377 6.93 0.41 6.88 0.43 20 1.222501 0.362 5.97 0.74 5.9 0.76 30 1.190209 0.345 5.09 1

5.1 1

40 1.160162 0.33 4.28 1.23 4.5 1.16 50 1.132272 0.318 3.56 1.42 4.03 1.29 55 1.119141 0.313 3.23 1.51 3.83 1.34 60 1.106378 0.308 2.92 1.59 3.64 1.4 65 1.093613 0.304 2.63 1.67 3.46 1.44 70 1.081297 0.3 2.36 1.74 3.29 1.49 75 1.069394 0.296 2.11 1.81 3.13 1.53 80 1.057626 0.293 1.88 1.87 2.99 1.57 AFP 10 1.213826 0.406 6.91 0.41 6.88 0.42 20 1.161187 0.404 5.94 0.73 5.99 0.71 30 1.117491 0.398 5.05 0.99 5.37 0.90 40 1.079099 0.392 4.25 1.22 4.85 1.05 50 1.044467 0.386 3.53 1.41 4.40 1.17 55 1.027869 0.383 3.20 1.49 4.20 1.22 60 1.011599 0.381 2.89 1.57 4.01 1.27 65 0.996140 0.380 2.60 1.65 3.83 1.32 70 0.981091 0.378 2.33 1.72 3.66 1.37 75 0.966327 0.376 2.08 1.79 3.49 1.41 80 0.951924 0.375 1.85 1.85 3.34 1.45 ALL 10 1.208660 0.409 6.91 0.41 6.88 0.42 20 1.152644 0.408 5.94 0.73 6.03 0.70 30 1.107052 0.403 5.05 0.99 5.41 0.89 40 1.066994 0.398 4.25 1.22 4.91 1.03 50 1.030395 0.393 3.53 1.41 4.47 1.15 55 1.013373 0.391 3.20 1.49 4.27 1.20 60 0.996349 0.390 2.89 1.57 4.08 1.26 65 0.980518 0.389 2.60 1.65 3.90 1.30 70 0.964612 0.387 2.33 1.72 3.73 1.35 75 0.949494 0.386 2.08 1.79 3.56 1.39 80 0.934361 0.386 1.85 1.85 3.40 1.44

NUREG/CR-7309 W. Metwally, M. Dupont, W. Marshall, A. Lang, V. Karriem, C. Celik, K.

Fassino, A. Shaw Oak Ridge National Laboratory 1 Bethel Valley Road Oak Ridge, TN 37831 Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 L. Kyriazidis, NRC Project Manager Interest is growing in the use of extended enrichment (between 5 and 8 wt% 235U) and higher burnup fuels in nuclear reactors. Therefore, safe storage and transportation of the resulting spent nuclear fuel (SNF) must be demonstrated. This report describes an investigation of the effects of extended enrichment and high-burnup fuels on validation of burnup credit (BUC) criticality safety analyses of SNF storage systems.

The report presents results from a detailed similarity assessment study conducted to determine the appropriate criticality benchmark experiments for different application cases. In addition, the report presents the calculation of the bias and bias uncertainty, nuclear data-induced uncertainties, sensitivities, and BUC loading curves for selected application cases.

Extended enrichment High burnup Criticality safety Burnup credit April 2025 Technical Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses

NUREG/CR-7309 Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses April 2025