CNS-15-078, WCAP-18060-NP, Revision 1, Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (Mur) Power Uprate Fluence Evaluations
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| Site: | Catawba  |
| Issue date: | 11/30/2015 |
| From: | Chen J Westinghouse |
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| CNS-15-078, TAC MF4526, TAC MF4527 WCAP-18060-NP, Rev. 1 | |
| Download: ML15324A084 (44) | |
Text
Westinghouse Non-Proprietary Class 3 WCAP-1 8060-N P November 2015 Revision 1 Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit I Measurement Uncertainty Recapture (MUR)
Power Uprate Fluence Evaluations SWestinghouse
Westinghouse Non-Proprietary Class 3 WCAP-1 8060-NP Revision 1 Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate Fluence Evaluations Jianwei Chen*
Nuclear Operations & Radiation Analysis November 2015 Reviewers: Stanwood L. Andeirson*
Nuclear Operations & Radiation Analysis Greg A. Fischer*
Nuclear Operations & Radiation Analysis Approved: Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis
- Electronically approved records are authenticated in the electronic document management system.
Westinghouse Electric Company LLC I000 Westinghouse Drive Cranberry Township, PA 16066, USA
© 2015 Westinghouse Electric Company LLC All Rights Reserved WCAP-18060-NPRL~docx-1 10315
ii TABLE OF CONTENTS LIST OF TABLES.....................................................................................................1i LIST OF FIGURES................................................................................................... v 1 INTRODUCTION ........................................................................................ i1-i 2 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION ............................ 2-1 3 REFERENCES ............................................................................................ 3-1 APPENDIX A VALIDATION OF RAPTOR-M3 G TRANSPORT CALCULATIONS................... A-i APPENDIX B PCA SIMULATOR BENCHMARK RESULTS ............................................. B-i APPENDIX C H. B. RONBINSON BENCHMARK RESULTS............................................ C-i1 APPENDIX D ANALYTIC UNCERTAINTY ANALYSIS .................................................. D-1 APPENDIX E OPERATING REACTOR MEASUREMENTS ............................................. E-1 APPENDIX F HIGHER ORDER SN SENSITIVITY STUDY............................................... F-i WCAP- i18060-NP November 2015 Revision 1
111°.
LIST OF TABLES Table A-I: Calculational Uncertainties for Surveillance Capsule Neutron Fluence Rate.................. A-2 Table A-2: Calculational Uncertainties for Circumferential Welds Neutron Fluence Rate................ A-3 Table A-3: Calculational Uncertainties for Cavity Capsule Neutron Fluence Rate ........................ A-4 Table B.2-1 : PCA Experimental Measurement Locations.................................................... B-2 Table B.2-2: M/C Comparisons for the PCA 12/13 Blind Test Experiment ................................ B-2 Table B.3-l: Summary of Simulator Benchmark M!C Comparisons........................................ B-3 Table B.3-1: M/C Comparisons for the H. B. Robinson Dosimetry Benchmark Performed with RAPTOR-M3G using DTW with P3 and $S.............................................................C-i Table D.2-I" Summary of Core Neutron Source Sensitivity Study.......................................... D-2 Table D.2-2: Difference of Source Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Surveillance Capsules ........................................................................ D-2 Table D.2-3" Difference of Source Permutation-to-Nominal Fast Neutron (EB> 1.0 MeV) Fluence Rate for Circumferential Welds........................................................................ D-3 Table D.2-4: Difference of Source Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Cavity Capsules............................................................................... D-3 Table D.2-5:_Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties................................................................ D-4 Table D.2-6: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties ......................................................... D-4 Table D.2-7: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties ......................................................... D-5 Table D.3-1: Summary of Geometry and Temperature Sensitivity Study ................................... D-7 Table D.3-2: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Surveillance Capsule .............................................. D-7 Table D.3-3: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (EB> 1.0 MeV) Fluence Rate for Circumferential Welds ............................................ D-8 Table D.3-4: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Cavity Capsules.................................................... D-9 Table D.3-5: Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties .................................................. D-9 Table D.3-6: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties................................................. D- 10 Table D.3-7: Summary of Cavity Capsules Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties............................................................. D- 10 WCAP- 18060-NP November 2015 Revision 1
4¢ iv Table D.4-l: Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties................... D-ll Table D.4-2: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties ................. D-12 Table D.4-3: Summary of Cavity Capsule Neutron Eluence Rate Uncertainties.......................... D-12 Table E-l : In-Vessel and Ex-Vessel Capsules Threshold Reactions MIC Ratios........................... E-2 Table E-2: In-Vessel Surveillance Capsules BE/C Ratios................................................... E-3 Table E-3: EVND Core Midplane BE/C Ratios ............................................................. E-4 Table E-4: EVND Off-Midplane BE/C Ratios ............................................................... E-5 Table E-5: EVND Off-Midplane BE/C Ratios with Outliners Rejected...................................E-6 Table F-i: Fast Neutron Fluence Rates Comparison between using Ss versus S12 Quadrature Sets ..... F-i WCAP-1 8060-NP November 2015 Revision 1
V LIST OF FIGURES No Figures are listed in this report.
November 2015 WCAP-18060-NP WCAP- 18060-NP Revision 1
I-1 1 INTRODUCTION As part of the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) power uprate project, Westinghouse performed an applicability assessment of the Catawba Unit 1 Pressure-Temperature (P-T) limit curves, and this information was then submitted to the U. S. Nuclear Regulatory Commission (NRC) for review and approval. The evaluations performed by Westinghouse are described in detail in Reference 1. These evaluations included a determination of neutron fluence for several reactor vessel materials.
The NRC has issued Requests for Additional Information (RA~s) concemning the use of RAPTOR-M3G in Reference 2. This document provides responses to the RAIs, as requested.
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2-1 2 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION The responses to the RAIs in Reference 2 issued by NRC are proposed below:
SRXB - RAI 11 The basis of review for RAPTOR-M3G is Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." RG 1.190 intends to ensure the accuracy and reliability of the neutron fluence determination required by General Design Criteria (GDC) related to the reactor coolant pressure boundary. RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence. RG 1.190 specifies that the neutron transport methods should be benchmarked to a statistically significant data base of measurement-to-calculation ratios (M/C) and that an overall neutron fluence bias and uncertainty be estimated. The NRC staff requests that the licensee provide an explanation for how RAPTOR-M3G satisfies this portion of RG 1.190.
Response
APPENDIX E summarizes the measurement-to-calculationratios (M/C) that have been evaluated with RAPTOR-M3G. There are total of 69 in-vessel surveillancecapsules with 295 high-quality thresholdfoil measurement datapointsfrom 18 nuclearpower plants that have been analyzed and compared against RAPTOR-M3G calculations. In addition to the in-vessel surveillance capsules, there are 87 Ex- Vessel Neutron Dosimetry (EVND) capsules (i. e., cavity capsules) with 454 high-quality thresholdfoil measurement datapoints at core midplane locations, and 44 EVND capsules with 22 7 high-quality thresholdfoil measurement data pointsfrom off-rnidplane locations that have been analyzed and compared against RAPTOR-M3G calculations. The results arepresented in Table E-1 through Table E-5.
The results show that the BE/C and M/C ratios are well within the 4-20% acceptance criterionfor the in-vessel capsules and 4-30% for the cavity capsules.
WCAP-1 8060-NPNoebr21 Revision 1
2-2 SRXB - RAI 12 In Section IV.l1.C.ii of the application, the licensee states that Westinghouse has evaluated the latest-available ENDF/B-VII-based cross-section data contained in the BUGLE-B7 library and the results of this evaluation indicate no significant differences between the results of analyses performed using two different cross-section data. This evaluation can be found in Appendix B of WCAP-16083-NP, Revision 1, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," dated April 2013. Appendix B in WCAP-16083-NP, Revision 1, discusses the evaluation of the two cross-section sets by Oak Ridge National Laboratory for applications related to fluence.
However, the Westinghouse evaluation only refers to work related to low-energy neutrons. There is an apparent conflict between statements made in the application and Appendix B of WCAP- 16083-NP, Revision 1, with respect to the Westinghouse evaluation of two multi-group cross section libraries. The NRC staff requests that the licensee clarify the statements or provide a revision of the relevant paragraph in Section IV.1.C.ii of the application and/or Appendix B of WCAP- 16083-NP, Revision 1.
Response
Neutron fluence calculations that support reactor vessel integrity evaluationsfocus primarily on high-energy ('fast') portions of the neutron spectrum. The responsefunctions that are of the greatestconcern are neutronfluence (E > 1.0 MeV) and iron atom displacements (DPA). Neutronfluence (E > 1.0 MeV) only considers neutrons above energies of l. OOE+06 eV. Iron DPA is predominately influenced by neutrons above 1.OOE+05 eV.
The BUGLE-B7 cross-section library is an update to the BUGLE-96 library. The primary difference between the two librariesis that the BUGLE-B 7 library is derivedfrom ENDF-B/VII nuclear data, whereas BUGLE-96 is derivedfrom older ENDF/B- VI nuclear data. The energy group boundaries and techniques used to construct both librariesare the same.
IIn the documentation releasedwith BUGLE-B7 (Reference 15), Oak Ridge NationalLaboratory (ORNL) analyzed the H. B. Robinson, PCA, and VENUS-3 benchmarks using BUGLE-96 and BUGLE-B7.
Calculationswith both librarieswere compared to measurementsfrom reactions that cover high-energy portions of the neutron spectrum. The ORNL report demonstrates that the differences between BUGLE-96 and BUGLE-B 7 for high-energy neutron applicationsare minor. This finding is corroboratedby comparisons that Westinghouse has performed internally. Thus, for applications that concern Regulatory Guide 1.190, the differences are not significant.
In Section IV.]. C. ii of Reference 16, it was stated that "Westinghouse has evaluated the latest-available ENDF/B-VII-based cross-section data contained in the BUGLE-B 7 library.... The results of the evaluation indicate that no signi~ficant differences exist between the results of analysesperformed using BUGLE-B7 cross-section data versus BUGLE-96 cross-section data." This statement meant that Westinghouse has evaluated the differences between BUGLE-B7 and BUGLE-96 by reviewing the comparisonsbetween BUGLE-B7 and BUGLE-96 using measured data from H. B. Robinson-2 and VENUS-3 benchmarks that have done by ORNL in Reference 15. Westinghouse has completed an additionalcomparative study that has not been done by ORNL that revealed differences between BUGLE-B7 and BUGLE-96for low energy neutrons that resultedfrom discrepancies/errorsin the upscatter-rem oved BUGLE-B 7 library in lower energy range. However, this discovery is irrelevant to the Catawba Unit 1 MUR application, and demonstrates that using BUGLE-96 cross-section library is a better choice, ifanything.
WCAP-1 8060-NP November 2015 Revision 1
2-3 SRXB - RAI 13 The use of $8 angular quadrature is generally acceptable according to RG 1.190. However, RG 1.190 states that when off-midplane locations are analyzed, the adequacy of the Sg quadrature to determine the streaming component must be demonstrated with higher-order Sn, calculations. It is noted for Catawba Unit 1 that one of the limiting locations is near the extended beltline. The analysis of Catawba Unit 1 for the MUR includes an angular discretization "modeled with an $8 order of angular quadrature or higher."
Since some of the locations of interest are on the extended beltline, it must be shown that the angular quadrature for those calculations is adequate. The NRC staff requests that the licensee provide the angular quadrature used for the locations that are off-midplane. Additionally, the NRC staff requests that the licensee provide a justification for the angular quadrature.
Response
The comparison study between different orders of S,, calculation was performed and the results are presented in APPENDIX F. The same Catawba Unit 1 RAPTOR-M3G model in Reference 1 was usedfor the comparative calculations by changing only the S, orderfrom S8 to S12. Therefore, the S12 calculation was also performed using P3 Legendre expansion and DirectionalTheta Weighted (DTW) differencing scheme, as with calculationsperformed in Reference 1.
The results show thatfor extended beltline region materials at off-midplane locations, the fast neutron fluence rates are within 1% for S8 versus $13 level symmetric quadraturesets. Therefore, the use of Ss quadraturesets isjustified.
WCAP- 18060-NPNoebr21 Revision 1
2-4 SRXB -RAI 14 RG 1.190 requires that the neutron fluence methodology be qualified, and flux uncertainty estimates be determined. One component of the uncertainty is from the analytical sensitivity studies. The important sources of "analytical uncertainty" are listed in RG 1.190), Section 1.4.1, "Analytic Uncertainty Analysis."
The analytical sensitivity studies using RAPTOR-M3G are important in determining the overall uncertainty in neutron fluence calculations. The NRC staff requests that the licensee provide justification for the parameters that are important to performing neutron fluence analysis. Additionally, the NRC staff requests that the licensee provide the resulting uncertainties for each parameter and all parameters combined.
Response
The analytical uncertainty analysis has been performed and documented in APPENDJXD. The results presented in APPENDIXD show that the analytical uncertainty is less than 4-8.5%for the surveillance capsules, and +/- 11% for the vessel inner radius and cavity capsules at the core midplane. The analytical uncertaintyfor the off-midplane vessel inner radius (extended beltline region) is less than 4-15%, and is less than +/-: 17% for the off-midplane cavity capsules. For detailed descriptions of the components of these uncertaintiesplease see APPENDJY D.
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2-5 SRXB -RAI 15 The Pool Critical Assembly (PCA) benchmark is a full-scale section mockup of a pressure vessel with dosimetry measurements at the inner surface of the pressure vessel, and at locations within the pressure vessel wall. The PCA benchmark is one of the three suggested RG 1.190 benchmarks for pressure vessel simulator measurements. The PCA benchmark has been calculated using both TORT and RAPTOR-M3G with identical assumptions and the codes produced identical results. The spatial differencing in RAPTOR-M3G can be performed using either theta-weighted (TW) or directional theta-weighted (DTW) options. For the direct comparisons with TORT, the TW option is used in RAPTOR-M3G as TORT does not have the DTW option. The DTW is the normal option used in the Catawba neutron fluence analysis using RAPTOR-M3G. In order to determine the M/C ratios based on the PCA measurements, RAPTOR-M3G should be used with the DTW option. The NRC staff requests that the licensee provide RAPTOR-M3G results for the PCA benchmark, as part of the code validation, using the same modeling assumptions used for the Catawba neutron fluence calculations.
Response
The PCA benchmark has been performed using RAPTOR-M3G with P3, S78, and DTW cifferencing scheme in APPENDIX B. The results show an averageM/C ratio of. .01 with a 1 c-standarddeviation of2. 3%.
WCAP- 18060-NP November 2015 Revision 1
2-6 SRXB -RAI 16 The analysis of the H.B. Robinson benchmark using RAPTOR-M3G is provided in Appendix A of WCAP-16083-NP, Revision 1. The analysis is performed using P5 expansion of the scattering kernel and an $12 angular quadrature as compared to P3 and S8 quadratures used in the Catawba neutron fluence analysis. In Section A.2 of WCAP-16083-NP, Revision 1, it is stated that the results in Appendix A made no attempt to address power redistribution effects over the course of the fuel cycle when calculating measured reaction rates at-saturation, whereas the results presented in Section 3.3 did consider power redistribution. The analysis of the benchmark is atypical, although there is good agreement with the data.
The NRC staff requests that the licensee provide the results for the H.B. Robinson benchmark using the standard methodology used for the Catawba neutron fluence analysis for qualification of RAPTOR-M3G.
Response
The H. B. Robinson benchmark has been performed using RAPTOR-M3G with P3 Legendre expansion, an S8 level-symmetric quadrature set, and the DTW spatial differencing scheme in APPENDIX C. The results show an average M/C ratio ofli.07 with a I or standarddeviation of 5.9% for in-vessel capsules, and 1.01 with a 1 o- standarddeviation of 7.5%for ex-vessel capsules.
WCAP- 18060-NP November 2015 Revision 1
I-2-7 SRXB -RAI 17 The comparisons of RAPTOR-M3G calculations with capsule data are discussed in Appendix A of WCAP- 16083-NP, Revision 1, and in Appendix C of WCAP- 17669-NP, Revision 0, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations," dated June 2013 (ML14353A029). A requirement of RG 1.190 is that the uncertainty and bias in the neutron fluence calculations be assessed based on three components: surveillance measurements, benchmarks, and analytical sensitivity studies.
a) The NRC staff requests that the licensee provide an explanation for how each component of uncertainty will be weighted to arrive at the uncertainty and bias for fast neutron fluence at the limiting vessel locations.
b) The component of uncertainty and bias from comparisons of calculations with surveillance capsule data should be from a statistically significant number of measurements. The NRC staff requests that the licensee provide a justification of the use of the database, used to assess RAPTOR-M3G, based on its statistical significance.
Response
APPENDIXA documents how the PCA and H.B. Robinson benchmarks and the analyticalsensitivity study are combined to determine the total calculationaluncertaintiesfor the reactorvessel inner radius, surveillance capsules, and cavity capsules. The calculationaluncertaintyfor the surveillance capsule location is +/- 13%.
In Table E-2, the comparisons of the adjusted results with the originalcalculations of the neutronfluence rate (E > 1.0 Me V) are provided as [Best Estimate]/[Calculated](BE/C) ratiosfor each of the 69 surveillance capsule dosimetry sets included in the database. Also included in the tabulation are average BE/C values for the 18 individual reactors andfor the overall database. From the data listed in Table E-2, the overall database average BE/C is 0.98 with an associatedstandarddeviation of 6% for fast neutronfluence rate, and 0.99 with an associatedstandarddeviation of 5% for iron atom displacement rate (DPA/s). This data shows that the stand-alone transportcalculations are essentially unbiased.
Further, the 6% standarddeviation associatedwith the BE/C database is approximately half of the 13%
uncertainty assigned to the calculation alone.
a) Each component of uncertainty is summed in quadrature to arrive at the total uncertaintyfor the fast neutronfluence at the limiting vessel locations. The total uncertaintiesfor all the key vessel materials are summarized in APPENDIXA.
b) APPENDIXE summarizes the databasethat has been used to benchmark RAPTOR-M3G. There are total of 69 in-vessel capsules with 295 thresholdfoil measurements, 87 Ex- Vessel Neutron Dosimetry (EVND) capsules at core midplane with 454 thresholdfoil measurements, and 44 EVND capsules at off-midplane locations with 227 thresholdfoil measurements. Due to large number ofmeasurement datapoints, this database is deemed to be statistically signi~ficant.
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3-1 3 REFERENCES
- 1. Westinghouse Report WCAP- 17669-NP, Rev. 0, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations,"
June 2013.
- 2. NRC Agencywide Documents Access and Management System (ADAMS) MLl5201A5 12, "Catawba Nuclear Station, Units 1 and 2: Request for Additional Information Regarding License Amendment Request to Support a Measurement Uncertainty Recapture Power Uprate for Catawba, Unit I (TAC Nos. MF4526 and MF4527)." August 2015.
- 3. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.
- 4. ORNL Report ORNL/TM-13204, "Pool Critical Assembly Pressure Vessel Facility Benchmark,"
(NLUREG/CR-6454), October 1997.
- 5. RSICC Data Library Collection DLC-185, "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.
- 6. RSICC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," July 1994.
- 7. Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
- 8. Westinghouse Report WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for the Least Squares Evaluation of Light Water Reactor Dosimetry," May 2006.
- 9. Westinghouse Report WCAP-13348, "Consumers Power Company Palisades Nuclear Plant Reactor Vessel Fluence Analysis," May 1992.
- 10. Westinghouse Report WCAP-133 62, "Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation," May 1992.
- 11. ASTM Designation E844, 2014, "Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance," ASTM International, West Conshohocken, PA, 2014.
- 12. NUREG/CR-6453, "H. B. Robinson-2 Pressure Vessel Benchmark," U. S. Nuclear Regulatory Commission, October 1997.
- 13. ASTM Designation E944, 2013, "Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance," ASTM International, West Conshohocken, PA, 2014.
WCAP- 18060-NP November 2015 Revision I
3-2 I14. R. E. Maerker, "Application of LEPRICON Methodology to LWR Pressure Vessel Surveillance Dosimetry," Reactor Dosimetry, Proc. 6 h ASTM-Euratom Symposium, Jackson Hole, Wyoming, May 31 - June 5, 1987, American Society for Testing and Materials (1989).
- 15. RSJCC Data Library Collection DLC-245, "BUGLE-B7/VITAMIN-B7, Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data," Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL),
October 2011.
- 16. NRC Agencywide Documents Access and Management System (ADAMS) ML14176A109, "Duke Energy Carolinas, LLC (Duke Energy) Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR)
Power Uprate," June 2014.
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A-i APPENDIX A VALIDATION OF RAPTOR-M3G TRANSPORT CALCULATIONS The transport calculation models used to calculate the Catawba Unit 1 MUR fast neutron fluence used the RAPTOR-M3G code with $8 level-symmetric quadrature set, anisotropic scattering cross section treatment with P3 Legendre expansion, and Directional Theta Weighted (DTW) differencing scheme.
This Appendix provides the validation of the transport methodology based on the guidance provided in Regulatory Guide 1.190 (Reference 3). Similar to the approach used in References 7 and 8, the validation consists of the following stages:
- 1. Comparisons of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) using P3 Legendre expansion, $8 level-symmetric quadrature sets, and DTW differencing scheme (APPENDIX B).
- 2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment using P3 Legendre expansion, $8level-symmetric quadrature sets, and DTW differencing scheme (APPENDIX C).
- 3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant specific transport calculations used in the exposure assessments using P3 Legendre expansion, S8 level-symmetric quadrature sets, and DTW differencing scheme (APPENDIX D).
- 4. Comparisons of calculations with a measurement database obtained from a large number of surveillance capsules withdrawn from a variety of pressurized water reactors (APPENDIX E).
In Catawba Unit 1 MUJR-specific application of the methodology, comparisons are made with Catawba-specific dosimetry results in Appendix C of Reference 1 and demonstrate that the plant specific transport calculations are consistent with the uncertainties derived from the methods qualification presented in this Appendix.
The first stage of the methods validation (PCA benchmark) addresses the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, does not test the accuracy of commercial core neutron source calculations nor does it address uncertainties in operational or geometric variables that affect power reactor calculations. The second stage of the validation (H. B. Robinson 2 benchmark) addresses uncertainties that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third stage of the validation (Analytical Uncertainty Analysis) identifies the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific parameters. The overall calculational uncertainty is established from the results of these three stages of the validation process.
Table A-i through Table A-3 summarize the uncertainties for the fast neutron fluence rate at different vessel locations determined from the results of the first three stages of the validation process:
WCAP- 18060-NP November 2015 Revision 1
A-2 Table A-i: Calculational Uncertainties for Surveillance Capsule Neutron Fluence Rate Uncertainty Component Dual Surveillance Dual Surveillance Capsule at 290 Capsule at 31.5° PCA Benchmark Comparisons +5 -5 H. B. Robinson Benchmark Comparisons 4-7% 4-7%
Peripheral Assembly Source Strength 4-4.46% +-4.49%
Pin Power Distribution + 0.65% +/-- 0.65%
Peripheral Assembly Burnup (+/-5000MWD/MTU) 4-0.89% +/-- 0.88%
Axial Power Distribution 4- 1.26% 4-1.26%
Internals Dimensions +/-- 1.07% +/--0.86%
Vessel JR +/-- 0.03% +/--0.03%
Vessel Thickness +/--0.00% +/- 0.00%
Dosimetry Position +/-- 1.74% +/-- 1.77%
Coolant Temperature 4-4.62% +/-- 4.48%
Core Periphery Modeling 4-4.61% 4-4.61%
Analytical Sensitivity Studies +/-- 8.34% 4-8.25%
Other Factors ++/-5% +/--5%
Calculational Uncertainties 4- 12.98% +/-- 12.92%
WCAP- 18060-NP November 2015 Revision 1
A-3 Table A-2: Calculational Uncertainties for Circumferential Welds Neutron Fluence Rate Uncertainty Component Lower Shell to Intermediate Shell Upper Shell to Bottom Head Ring to Lower Shell Circ. Intermediate Shell Circ. Weld W04 Weld WO5 Circ. Weld W06 PCA Benchmark Comparisons 4-5% 4-5% 4-5%
H. B. Robinson Benchmark Comparisons 4-7% 4-7% 4-7%
Peripheral Assembly Source Strength +-4.37% +/- 4.59% +/--3.66%
Pin Power Distribution + 0.29% 4-0.93% 4-0.32%
Peripheral Assembly Burnup
(+5000MWD/MTU) 4- 1.06% +-0.97% 4- 1.14%
Axial Power Distribution +-7.43% 4- 1.25% +-9.59%
Internals Dimensions +- 1.42% 4- 1.41% 4- 1.61%
Vessel JR +-3.39% +-4.21% +-2.96%
Vessel Thickness 4-0.00% +-0.00% +-0.00%
Dosimetry Position N/A N/A N/A Coolant Temperature 4-7.09% 4-5.92% 4-8.84%
Core Periphery Modeling +/- 4.61% +-4.61% 4-4.61%
Analytical Sensitivity Studies +-12.67% 10.02% 14.75%
Other Factors +-5% +-5% +-5%
Calculational Uncertainties +-16.11% +-14.12% +-17.79%
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A-4 Table A-3: Calculational Uncertainties for Cavity Capsule Neutron Fluence Rate Uncertainty Component Cavity Capsule Cavity Capsule Cavity Capsule (Top) (Midplane) (Bottom)
PCA Benchmark Comparisons 4-5% +/--5% +/--5%
H. B. Robinson Benchmark Comparisons 4-7% +/--7% 4-7%
Peripheral Assembly Source Strength 4-4.43% +-4.54% 4-4.5 1%
Pin Power Distribution +-0.71% +/--0.90% + 0.80%
Peripheral Assembly Burnup
(+/--5000MWD/MTU) +/--1.02% +/--1.00% +/--1.01%
Axial Power Distribution +/--3.06% 4-1.26% +/--1.87%
Internals Dimensions +/- 1.40% 4-1.40% +/-- 1.39%
Vessel JR +/- 2.76% +/--2.72% +/--2.79%
Vessel Thickness +/--2.26% +/--2.19% ++/-2.28%
Dosimetry Position 4- 11.73% +/--4.20% 4- 13.65%
Coolant Temperature +/--6.24% +/- 5.98% +/- 6.08%
Core Periphery Modeling +/--4.6 1% +/- 4.6 1% +/--4.61%
Analytical Sensitivity Studies 15.59% 10.62% 16.88%
Other Factors +/--5% +/--5% +/--5%
Calculational Uncertainties +/--18.49% 4-14.55% 4-19.59%
The category designated "Other Factors" is intended to attribute an additional uncertainty to other geometrical or operational variables that individually have an insignificant effect on the overall uncertainty, but collectively should be accounted for in the assessment.
The uncertainty components tabulated above represent percent uncertainty at the 1c* level. In the tabulation, the net uncertainty from the analytical sensitivity studies has been broken down into its individual components. When the four uncertainty values listed above are combined in quadrature, the resultant overall 1* calculational uncertainty is estimated to be bounded by 13% for surveillance capsules and 15% for vessel inner radius within the beltline region and cavity capsules at core midplane. The overall 1 cr calculational uncertainty is estimated to be bounded by 18% for off-midplane vessel inner radius within the extended beltline region, and bounded by 20% for off-midplane cavity capsules.
To date, the methodology described in Reference 1 combined with the BUGLE-96 cross-section library has been used in the evaluation of dosimetry sets from 69 in-vessel surveillance capsules, 87 ex-vessel neutron dosimetry (EVND) at core midplane, and 44 EVND capsules at off-midplane, from 18 pressurized water reactors. The comparisons of the measurement to calculation results for RAPTOR-M3G have been summarized in APPENDIX E. The comparisons of the plant specific calculations with WCAP-18060-NP Revision 1
A-5 the results of the capsule dosimetry (Appendix C in Reference 1) are used to further validate the calculational methodology within the context of a 1 ci calculational uncertainty at corresponding vessel locations.
WCAP- 18060-NPNoebr21 Revision 1
B-i APPENDIX B PCA SIMULATOR BENCHMARK RESULTS B.1 INTRODUCTION Several simulator benchmark experiments have been performed for the purpose of providing a qualification basis for neutron fluence analysis methods. The experiments were performed in laboratory settings, and simulate the configuration of an operating nuclear reactor on a smaller scale. This appendix provides the results of comparisons of the PCA simulator benchmark measurement results with calculations performed with RAPTOR-M3G.
B.2 POOL CRITICAL ASSEMBLY (PCA)
The Pool Critical Assembly (PCA) Pressure Vessel Facility Benchmark (Reference 4) is an industry-standard benchmark that can be used to partially qualify a fluence determination methodology according to Regulatory Guide 1.190. The PCA facility provides a small-scale simulation of the configuration of a pressurized water reactor (PWR). The geometry, material compositions, and neutron source for this experiment were all well-characterized, and accurate dosimetry measurements were collected at several locations of interest. A complete description of the benchmark is available in Reference 4. Table B.2-1 shows the distribution of measurement locations.
The RAPTOR-M3G analysis of the PCA problem is modeled on a 67 x 139 x 102 Cartesian mesh grid.
Angular quadrature is modeled with an S8 level-symmetric quadrature set, and anisotropic scattering is treated with a P3 Legendre expansion. The transport cross-section set was constructed from the BUGLE-96 library (Reference 5), and dosimetry reaction rate cross-sections are taken from the SNLRML library (Reference 6). The PCA problem was analyzed in RAPTOR-M3G using the directional theta-weighted (DTW) differencing scheme. (See Reference 34 of Regulatory Guide 1.190 for more information about the DTW scheme.)
Results of the benchmark comparisons using RAPTOR-M3G are presented in Table B.2-2.
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B-2 Table B.2-1: PCA Experimental Measurement Locations Position Y (cm) Location Description Al 12.0 Thermal Shield (Front)
A2 23.8 Thermal Shield (Back)
A3 29.7 Pressure Vessel (Front)
A4 39.5 Pressure Vessel (1/4 T)
A5 44.7 Pressure Vessel (1/2 T)
A6 50.1 Pressure Vessel (3/4 T)
A7 59.1 Void Box Table B.2-2: M/C Comparisons for the PCA 12/13 Blind Test Experiment MIC Ratio for Dosimetry Position Noted Reaction Al A2 A3 A4 A5 A6 A7 27A1 (n,cL) 24Na (Cd) 0.98 0.99 0.95 0.97/ 0.99 1.00 58 58 Ni (n,p) Co (Cd) 1.02 1.02 0.99 1.01 1.02 0.98 "5 ~n (n,n') llSmInl (Cd) 1.02 1.02 0.96 0.97 1.00 1.00 1.06
'°3 Rh (n,n') lo3mmh (Cd) 1.00 0.97 0.98 0.98 1.04 1.07 1.08 23 8 U (n,f) FP (Cd) 0.92 1.01 1.04 1.06 1.07 23 7 Np (n,f) FP (Cd) 1.06 1.01 1.00 1.03 0.99 Average 1.01 1.00 0.98 0.99 1.02 1.02 1.05
% std dev 1.9 2.4 4.9 2.1 2.1 3.5 3.9 WCAP- 18060-NP November 2015 Revision I
B-3 B.3
SUMMARY
OF SIMULATOR BENCHMARK RESULTS Results of the PCA simulator benchmark, grouped by reaction, are summarized in Table B.3-1. Also included in Table B.3-1 are the energies between which 90% of activity is produced in a U-235 fission spectrum, taken from ASTM E844 (Reference 11). The U-238 and Np-237 measurements exhibit slightly worse agreement with the calculations; however, per Reference 11, the U-23 8 and Np-237 reactions are subject to higher uncertainties in the measurement process.
The PCA simulator benchmark test the adequacy of the transport and dosimetry evaluation techniques, and the underlying nuclear data. The simulator benchmark comparison results demonstrate that, when the configuration of the system is well-known, the level of agreement between RAPTOR-M3G calculations and measurements is within the uncertainties associated with the measurements, themselves. The calculational uncertainty determined from PCA benchmark is 5% to be conservative.
Table B.3-1: Summary of Simulator Benchmark M/C Comparisons Neutron Energy Number of Reaction Response Observations Average M/C % std dev 27 24 A1 (n,cL) Na (Cd) 6.45 -11.9 MeV 6 0.98 1.8 58Ni (n,p) 58Co (Cd) 1.98 7.51 MeV 6 1.01
- 1.7 "l5 In (n,n') *lsmIn (Cd) 1.12 -5.86 MeV 7 1.00 2.5 103 03 Rh~ (n,n') ' 'mRh (Cd) 0.731 - 5.73 MeV 7 1.02 3.9 23 8 U (n,f) FP (Cd) 1.44 -6.69 MeV 5 1.02 6.1 2 37 Np (n,f) FP (Cd) 0.684-5.61 MeV 5 1.02 2.6 Total 36 1.01 2.3 WCAP-18060-NPNoebr21 Revision 1
c-1 APPENDIX C H. B. RONBINSON BENCHMARK RESULTS H. B. Robinson Unit 2 is a Westinghouse 3-1oop pressurized light-water reactor (PWR). As part of the USNRC-sponsored light water reactor (LWR) Pressure Vessel Surveillance Dosimetry Improvement Program, a comprehensive set of surveillance capsule and ex-vessel neutron dosimetry measurements were performed during Cycle 9 (Reference 12).
For the Cycle 9 benchmark, a replacement surveillance capsule was installed in a vacant surveillance capsule holder at the 200 azimuth with respect to the nearest cardinal axis. The dosimetry sets were placed at the geometric center of the surveillance capsule such that all measurements were taken within 30 cm of the core midplane.
The ex-vessel neutron dosimetry was installed in the reactor cavity, between the concrete biological shield and the reactor vessel insulation. Multiple foil sensor sets were placed in capsules that were attached to gradient wires. The gradient wires were installed in the reactor cavity and axially spanned the length of the reactor core. Only the midplane capsule from the Cycle 9 ex-vessel neutron dosimetry set was analyzed.
H. B. Robinson Unit 2 was modeled in RZ geometry with 158x 136x 172 mesh RAPTOR-M3G model.
Angular quadrature is modeled with an S* level-symmetric quadrature set, and anisotropic scattering is treated with a P3 Legendre expansion. The H. B. Robinson 2 problem was analyzed in RAPTOR-M3G using the DTW differencing scheme. Detailed geometry data for H. B. Robinson Unit 2 can be found in Reference 12. The M/C results for H. B. Robinson are presented in Table B.3-1. The calculational uncertainty determined from H. B. Robinson benchmark is 7%.
Table B.3-1: M/C Comparisons for the H. B. Robinson Dosimetry Benchmark Performed with RAPTOR-M3G using DTW with P3 and Ss MIC Ratio Reaction In-Vessel Ex-Vessel 63 6 Cu (n,ax) °Co 1.05 0.94 46 Ti (n,p) 46 5c 1.19 1.12 54 Fe (n,p) 54Mn 1.04 0.94 58 Ni (n,p) 58Co 1.07 1.02 238 U (n,p) FP 1.05 1.03 237 Np (n,p) FP 1.03 Average 1.07 1.01
% std dev 5.9 7.5 November 2015 8060-NP WCAP-118060-NP WCAP-Revision 1
D-1 APPENDIX D ANALYTIC UNCERTAINTY ANALYSIS D.1 INTRODUCTION Operating reactors are subject to several uncertainties that may influence the validity of the calculated neutron fluence results. The most significant among these are:
- Uncertainties in the core neutron source
- Uncertainties in the as-built thicknesses and locations of the reactor vessel and internal components
- Uncertainties in the at-power coolant temperatures (water density)
This listing of parameters is consistent with the findings of other neutron fluence uncertainty studies (References 9, 10, and 14). This appendix presents the results of a sensitivity study performed using RAPTOR-M3G that evaluate the impacts of variations in the parameters listed above on calculated neutron fluence values.
D.2 CORE NEUTRON SOURCE UNCERTAINTIES To assess the impact of uncertainties in the core neutron source on calculated neutron fluence results, changes in the following parameters were evaluated:
- Absolute source strength of peripheral fuel assemblies- Studies have shown that the neutron flux in ex-core regions is dominated by the neutron source from fuel assemblies on the core periphery. Measurements from in-core flux maps indicate that a source magnitude uncertainty of 5% is bounding.
- Pin-by-pin spatial distributions of neutron source at the core periphery - Core management studies indicate that uncertainties in the relative pin powers in peripheral fuel assemblies can be on the order of 10%.
- Burnnp of the peripheral fuel assemblies - Perturbations in fuel assembly bumup impact the fission spectrum, neutron yield per fission, and energy released per fission for each peripheral fuel assembly. A 5000 MWD/MTU uncertainty in the peripheral fuel assembly burnups is considered conservative. The sensitivity study is performed using a series of calculations starting with mid-cycle burnup at 3000 MWD/MTU, and 5000 MWD/MTU to 50,000 MWD/MTU with 5000 MWD/MTU delta mid-cycle bumup between each run. The results show less than 1.5%
uncertainties are introduced by a 5000 MWD/MTU uncertainty in burnup.
- Axial power distribution - Based on variations in axial peaking factors over the course of a fuel cycle, a 10% uncertainty in the shape of the axial power distribution is considered conservative.
Each case evaluated as part of the sensitivity study is described in Table D.2-1. The base case consisted of a low-leakage power distribution cycle (Cycle 21) from Catawba Unit 1. Table D.2-2 through Table D.2-4 provides the differences between calculated fast neutron (E > 1.0 MeV) fluence rate results at several locations for each permutation case, each normalized to the corresponding base case result. The overall uncertainty estimates are summarized in Table D.2-5 through Table D.2-7.
WCAP- 18060-NP November 2015 Revision 1
D-2 Table D.2-1: Summary of Core Neutron Source Sensitivity Study Case Description Number 1 Peripheral source strength biased by a factor of 0.95 2 Peripheral source strength biased by a factor of 1.05 Pin power distribution gradient diminished according to:
3 Pm ,n s = [(P - 1.0) X 0.9] + 1.0 Pin power distribution gradient intensified according to:
4 Pus= [(P -- 1.0) x 1.11 + 1.0 5 Mid-cycle burnup at 3000 MWD/MTU 6 Mid-cycle burnup at 50,000 MWD/MTU Axial power distribution gradient intensified according to:
Axialvtus [(Axial - 1.0) x 1.11 + 1.0 Axial power distribution gradient diminished according to:
8
_______Axialminus = [(Axial - 1.0) X 0.9] + 1.0 Table D.2-2: Difference of Source Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Surveillance Capsules Case Dual Surveillance Dual Surveillance Number Capsule at 29° Capsule at 31.50 1 -4.46% -4.49%
2 4.46% 4.49%
3 0.65% 0.65%
4 -0.65% -0.65%
5 -7.34% -7.14%
6 1.06% 1.15%
7 1.27% 1.27%
8 -1.25% -1.25%
WCAP-1 8060-NP November 2015 Revision 1
D-3 Table D.2-3: Difference of Source Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Circumferential Welds Lower Shell to Intermediate Shell Upper Shell to Cae Bottom Head Ring to Lower Shell Intermediate Shell Nubr Circ. Weld W04 Circ. Weld W05 Circ. Weld W06 1 -4.37% -4.59% -3.66%
2 4.37% 4.58% 3.66%
3 0.29% 0.92% -0.32%
4 -0.29% -0.93% 0.32%
5 -7.18% -6.93% -6.90%
6 2.76% 2.17% 3.80%
7 -7.48% 1.26% -9.64%
8 7.38%/. -1.25% 9.53%
Table D.2-4: Difference of Source Permutation-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate for Cavity Capsules Case Cavity Capsule Cavity Capsule Cavity Capsule Number (Top) (Midplane) (Bottom) 1 -4.42% -4.54% -4.49%
2 4.42% 4.55% 4.48%
3 0.72% 0.91% 0.80%
4 -0.72% -0.92% -0.81%
5 -7.30% -7.37% -7.36%
6 2.24% 2.02% 2.13%
7 -3.26% 1.27% -2.04%
8 3.20% -1.26% 2.01%
WCAP- 18060-NP November 2015 Revision 1
D-4 Table D.2-5: Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties Dual Surveillance Dual Surveillance Uncertainty Component Capsule at 290 Capsule at 31.50 Peripheral Assembly Source Strength 4-4.46% ++/-4.49%
Pin Power Distribution 4- 0.65% + 0.65%
Peripheral Assembly Burnup 4-0.89%* +-0.88%*
(4-5000MWD/MTU)
Axial Power Distribution +/--1.26% +- 1.26%
- Note the burnup uncertainty only assumed +/-- 5000 MWD/MTU, based on the difference between 3000 MWD/MTU case and 50,000 MWD/MTU case.
Table D.2-6: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties Lower Shell to Intermediate Shell Upper Shell to Uncertainty Component Bottom Head Ring to Lower Shell Intermediate Shell Circ. Weld W04 Circ. Weld W05 Circ. Weld W06 Peripheral Assembly Source Strength +/--4.37% +/--4.59% +/--3.66%
Pin Power Distribution + 0.29% +/--0.93% 4-0.32%
Peripheral Assembly Burnup
(+/--5000MWD/MTU) 4- 1.06%* +/--0.97%* 4- 1.14%*
Axial Power Distribution 4-7.43% + 1.25% 4-9.59%
- Note the burnup uncertainty only assumed 4-5000 MWD/MTU, based on the difference between 3000 MWD/MTU case and 50,000 MWD/MTU case.
WCAP- 18060-NPNoebr21 Revision l
D-5 Table D.2-7: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Core Neutron Source Uncertainties Unetany opoet Cavity Capsule Cavity Capsule Cavity Capsule UcranyCmoet(Top) (Midplane) (Bottom)
Peripheral Assembly Source Strength +/--4.43% +/--4.54% +/--4.51%
Pin Power Distribution 4- 0.7 1% 4- 0.90% 4-0.80%
Peripheral Assembly Burnup
(+/--5000MWD/MTU) 4- 1.02%* 4- 1.00%* +/-- 1.01%*
Axial Power Distribution +/--3.06% 4- 1.26% 4-1.87%
- Note the burnup uncertainty only assumed +/-- 5000 MWD/MTU, based on the difference between 3000 MWD/MTU case and 50,000 MWD/MTU case.
WCAP- 18060-NP November 2015 Revision 1
D-6 D.3 GEOMETRIC AND TEMPERATURE UNCERTAINTIES To assess the impact of uncertainties in the location and thickness of reactor components, as well as uncertainties in reactor coolant temperature, on calculated neutron fluence results, changes in the following parameters were evaluated:
- Reactor internals dimensions - Thickness tolerances on stainless steel reactor internals components (e.g., core baffle, core barrel, thermal shield/neutron pad) are typically specified as 1/16 inch or tighter.
- Reactor vessel inner radius - Reactor vessels typically specify an inner radius with tolerance bounds of -0.00 inches and +1/32 inches. A tolerance of+ 1/8 inch is considered.
- Reactor vessel thickness - Some techniques for fabricating reactor vessels result in larger-than-nominal reactor vessel base metal plate thicknesses. A tolerance of+= 1/16 inch is considered.
- Dosimetry Positioning - Surveillance capsules have a tolerance of+/- 1/16 inch associated with the positioning of the dosimetry in radial, azimuthal, and axial directions. A larger positioning uncertainty of+ 2 inches is associated with ex-vessel neutron dosimetry in radial azimuthal, and axial directions.
- Coolant Temperature - Variations in water temperature over the course of a fuel cycle are expected to be less than 4-10 0F.
- Core Peripheral Modeling - The modeling of the rectilinear core baffle in RZ geometry represents another potential source of uncertainty in the geometric modeling of the reactor. The sensitivity of the solution to the modeling approach is determined by a direct comparison of the results of an RZ computation with those of an XYZ calculation in which the baffle region and core periphery were modeled explicitly. The comparisons of interest were taken at various locations external to the core baffle, but inside the core barrel.
Each case evaluated as part of the sensitivity study is described in Table D.3-1. The base case consisted of a low-leakage power distribution cycle (Cycle 21) from Catawba Unit 1. Table D.3-2 through Table D.3-4 provides the differences between calculated fast neutron (E > 1.0 MeV) fluence rate results at several locations for each permutation case, each normalized to the corresponding base case result. The overall uncertainty estimates are summarized in Table D.3-5 through Table D.3-7.
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D-7 Table D.3-1: Summary of Geometry and Temperature Sensitivity Study Case Description Number 1 Baffle plates, core barrel, and neutron pad th~ickness decreased by 1/16 inch 2 Baffle plates, core barrel, and neutron pad thickness increased by 1/16 inch 3 Reactor coolant temperatures decreased by 10 °F 4 Reactor coolant temperatures increased by 10 °F 5 Reactor vessel radius decreased by 1/8 inch 6 Reactor vessel radius increased by 1/8 inch 7 Reactor vessel thickness decreased by 1/16 inch 8 Reactor vessel thickness increased by 1/16 inch Surveillance capsule position adjusted by 1/16 inch, ex-vessel dosimetry
_________position adjusted by 2 inches 10 XYZ versus RZ modeling difference in core periphery Table D.3-2: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (E >
1.0 MeV) Fluence Rate for Surveillance Capsule Case Dual Surveillance Dual Surveillance Number Capsule at 290 Capsule at 31.50 1 1.21% 0.96%
2 -0.92% -0.75%
3 -4.3 7% -4.23%
4 4.88% 4.72%
5 -0.01% 0.04%
6 -0.07% -0.02%
7 -0.04% 0.0 1%
8 -0.04% 0.0 1%
9 +/-: 1.74%(a + 1.77%(a 10 4 6 1 %(b) +/-46%b (a) Surveillance capsule positioning uncertainty includes radial, azimuthal, and axial position variations (b) Core periphery modeling uncertainty determined from direct comparison between ROZ and XYZ results in bypass region WCAP- 18060-NP November 2015 Revision 1
D-8 Table D.3-3: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (E >
1.0 MeV) Fluence Rate for Circumferential Welds Lower Shell to Intermediate Shell Upper Shell to Cae Bottom Head Ring to Lower Shell Intermediate Shell Nubr Circ. Weld W04 Circ. Weld W05 Circ. Weld W06 1 0.21% -0.52% 0.32%
2 -2.63% -3.35% -2.91%
3 -6.61% -5.56% -8.13%
4 7.56% 6.28% 9.55%
5 3.67% 3.95% 3.34%
6 -3.10% -4.47% -2.58%
7 0.21% -0.52% 0.32%
8 0.21% -0.52% 0.31%
9 N/A N/A N/A 10 + 4.61%* +-4.61%* 4-4.61%*
- Core periphery modeling uncertainty determined from direct comparison between RZ and XYZ results in bypass region WCAP-1 8060-NPNoebr21 Revision I
D-9 Table D.3-4: Difference of Geometry and Temperature Permutation-to-Nominal Fast Neutron (E >
1.0 MeV) Fluence Rate for Cavity Capsules Case Cavity Capsule Cavity Capsule Cavity Capsule Number (Top) (Midplane) (Bottom) 1 -0.34% -0.35% -0.36%
2 -3.14% -3.14% -3.15%
3 -5.84% -5.61% -5.70%
4 6.64% 6.35% 6.45%
5 1.35% 1.33% 1.35%
6 -4.17% -4.10% -4.24%
7 1.96% 1.87% 1.95%
8 -2.57% -2.5 1% -2.6 1%
9 + 11.73%<" 4.20%(a) + 13.65%<a 10 + 4 .6 1 %(b) 4 6. 1 %(b) - .1%b (a) Cavity capsule positioning uncertainty includes radial, azimuthal, and axial position variations (b) Core periphery modeling uncertainty determined from direct comparison between ROZ and XYZ results in bypass region Table D.3-5: Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties Dual Surveillance Dual Surveillance Uncertainty Component Capsule at 290 Capsule at 31.50 Internals Dimensions +/- 1.07% + 0.86%
Vessel JR +/- 0.03% + 0.03%
Vessel Thickness +/- 0.00% +/- 0.00%
Dosimetry Position + 1.74% + 1.77%
Coolant Temperature + 4.62% + 4.48%
Core Periphery Modeling +/- 4.6 1% + 4.6 1%
WCAP-1 8060-NPNoebr21 Revision 1
D- 10 Table D.3-6: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties Lower Shell to Intermediate Shell Upper Shell to Uncertainty Component Bottom Head Ring to Lower Shell Intermediate Shell Circ. Weld W04 Circ. Weld W05 Circ. Weld W06 Internals Dimensions 4-1.42% +/- 1.41% 4- 1.61%
Vessel JR 4-3.39% + 4.21% 4-2.96%
Vessel Thickness 4-0.00% 4-0.00% 4-0.00%
Coolant Temperature 4-7.09% 4-5.92% 4-8.84%
Core Periphery Modeling 4-4.61% 4-4.61% 4-4.61%
Table D.3-7: Summary of Cavity Capsules Neutron Fluence Rate Uncertainties Resulting from Geometry and Temperature Uncertainties Unetany opoet Cavity Capsule Cavity Capsule Cavity Capsule UcranyCmoet(Top) (Midplane) (Bottom)
Internals Dimensions +/-- 1.40% 4-1.40% 4-1.39%
Vessel JR 4-2.76% +/-- 2.72% 4-2.79%
Vessel Thickness +/--2.26% 4-2.19% 4-2.28%
Dosimetry Position 4- 11.73% 4-4.20% 4- 13.65%
Coolant Temperature 4-6.24% 4-5.98% 4-6.08%
Core Periphery Modeling 4-4.6 1% 4-4.6 1% 4-4.6 1%
WCAP-18060-NIPNoebr21 Revision 1
D- 11 D.4
SUMMARY
OF ANALYTIC UNCERTAINTY ANALYSIS Table D.4-1 through Table D.4-3 summarize the analytical uncertainties for Catawba Unit 1 RAPTOR-M3G calculations using P3 Legendre expansion anisotropic cross section treatment with S8 level symmetric angular quadrature sets, directional theta weighted differencing scheme. Note that only the dual capsules were investigated for the surveillance capsule uncertainties. However, the almost identical total uncertainties between 29° and 31.*50 dual surveillance capsules indicate their uncertainties are representative for single surveillance capsule at 31.,50.
Table D.4-1: Summary of Surveillance Capsule Neutron Fluence Rate Uncertainties Dual Surveillance Dual Surveillance Uncertainty Component Capsule at 290 Capsule at 31.50 Peripheral Assembly Source Strength + 4.46% +/--4.49%
Pin Power Distribution +-0.65% + 0.65%
Peripheral Assembly Burnup +/-- 0.89% 4-0.88%
(+/-5000MWD/MTU)
Axial Power Distribution 4-1.26% +/- 1.26%
Internals Dimensions 4- 1.07% 4-0.86%
Vessel JR + 0.03% +/- 0.03%
Vessel Thickness +-0.00% 4-0.00%
Dosimetry Position +- 1.74% +/--1.77%
Coolant Temperature 4-4.62% +/--4.48%
Core Periphery Modeling 4-4.61% +-4.61%
Total Analytical Uncertainty 8.34% 8.25%
WCAP- 18060-NP November 2015 Revision 1
D-12 Table D.4-2: Summary of Circumferential Welds Neutron Fluence Rate Uncertainties Lower Shell to Intermediate Shell Upper Shell to Uncertainty Component Bottom Head Ring to Lower Shell Circ. Intermediate Shell Circ. Weld W04 Weld W05 Circ. Weld W06 Peripheral Assembly Source Strength +/- 4.37% +-4.59% 4-3.66%
Pin Power Distribution 4-0.29% 4-0.93% +-0.32%
Peripheral Assembly Burnup
(+5000MWD/MTU) 4-1.06% ++/-0.97% +/- 1.14%
Axial Power Distribution + 7.43% 4- 1.25% +/- 9.59%
Internals Dimensions +/- 1.42% +/- 1.41% 4-1.61%
Vessel JR 4-3.39% 4-4.21% +/- 2.96%
Vessel Thickness 4- 0.00% +/- 0.00% +/- 0.00%
Dosimetry Position N/A N/A N/A Coolant Temperature 4-7.09% +/- 5.92% 4-8.84%
Core Periphery Modeling 4-4.6 1% +/- 4.6 1% 4-4.61%
Total Analytical Uncertainty 12.67% 10.02% 14.75%
Table D.4-3: Summary of Cavity Capsule Neutron Fluence Rate Uncertainties Unetany opoetCavity Capsule Cavity Capsule Cavity Capsule UcranyCmoet(Top) (Midplane) (Bottom)
Peripheral Assembly Source Strength +/- 4.43% +-4.54% 4-4.5 1%
Pin Power Distribution +/- 0.7 1% +-0.90% 4-0.80%
Peripheral Assembly Burnup (4-5000MWD/MTU) +/- 1.02% +/--1.00% +/- 1.01%
Axial Power Distribution +-3.06% 4-1.26% +/- 1.87%
Internals Dimensions 4-1.40% +/- 1.40% +- 1.39%
Vessel JR +/- 2.76% 4-2.72% +/- 2.79%
Vessel Thickness +/- 2.26% +/- 2.19% +/- 2.28%
Dosimetry Position +/--11.73% -+/-4.20% +/- 13.65%
Coolant Temperature +/- 6.24% +/-- 5.98% +/-- 6.08%
Core Periphery Modeling +/--4.6 1% +-4.6 1% +/- 4.61%
Total Analytical Uncertainty 15.59% 10.62% 16.88%
WCAP- 18060-NP November 2015 Revision 1
E- 1 APPENDIX E OPERATING REACTOR MEASUREMENTS There are 69 in-vessel surveillance capsules with 295 high quality threshold foil measurement data points from 18 nuclear power plants that have been analyzed and compared against RAPTOR-M3G calculations.
In addition to the in-vessel surveillance capsules, there are 87 Ex-Vessel Neutron Dosimetry (EVND) capsules (i.e., cavity capsules) with 454 high quality threshold foil measurement data points at core midplane, and 44 EXTND capsules with 227 high quality threshold foil measurement data points from off-midplane locations that have been analyzed and compared against RAPTOR-M3G calculations.
Similar to the approach used in References 7 and 8, the comparisons between the plant specific calculations and the data base measurements are provided on two levels. In the first instance, the average measurement to calculation (M/C) ratio over all the fast neutron sensor reaction rates from each reactor is listed. This tabulation provides a direct comparison, on an absolute basis, of measurement and calculation. The results of this comparison for the surveillance capsule database are listed in Table E-1.
These comparisons show that the calculations and measurements for the surveillance capsule database, 1.03 4-5%, fall well within the 13% calculational uncertainty for the surveillance capsule fast neutron fluence rate.
The second comparison of calculations with the data base is based on the least squares adjustment of the individual surveillance capsule data sets. The least squares adjustment procedure provides a weighting of the individual sensor measurements based on spectral coverage and allows a comparison of the neutron flux (E > 1.0 MeV) before and after adjustment. The neutron flux/fluence (E > 1.0 MeV) is the primary parameter of interest in the overall pressure vessel exposure evaluations.
The least squares evaluations of the 69 surveillance capsule dosimetry sets followed the guidance provided in Section 1.4.2 of Regulatory Guide 1.190 (Reference 3) and in Reference 13.
The application of the least squares methodology requires the following input:
- 1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.
- 2. The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.
- 3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For the data base comparisons, the calculated neutron spectra were obtained from the results of plant specific neutron transport calculations applicable to each of the 69 surveillance capsules. The sensor reaction rates and dosimetry cross-sections were the same as those used in the direct M/C comparisons noted above. The results are presented in Table E-2 through Table E-5. The overall database average BE/C for surveillance capsules is 0.98 with an associated standard deviation of 6% for fast neutron fluence rate, and 0.99 with an associated standard deviation of 5% for iron atom displacement rate (DPA/s). The results show that the BE/C and M/C ratios are unbiased and well within the 4-20%
acceptance criteria for the in-vessel capsules and 4- 30% for the cavity capsules (i.e., EVND capsules).
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E-2 Table E-l: In-Vessel and Ex-Vessel Capsules Threshold Reactions M/C Ratios EVND EVND Off-Plant Number In-Vessel M/C Midplane M/C Midplane M/C Domestic Plant #1 1.05 0.96 0.87 Domestic Plant #2 0.99 0.97 0.89 International Plant #1 1.13 1.03 0.92 International Plant #2 1.06 1.00 0.85 International Plant #3 N/A 0.97 0.92 International Plant #4 0.99 0.89 1 International Plant #5 1.09 0.88 0.84 International Plant #6 0.95 0.87 0.99 International Plant #7 0.95 0.86 1 Domestic Plant #3 1.02 0.89 0.99 Domestic Plant #4 1.01 0.89 0.93 Domestic Plant #5 1.00 0.93 0.88 International Plant #8 0.96 0.871 International Plant #9 1.08 0.83 0.95 Domestic Plant #6 1.01 0.90 0.85 Domestic Plant #7 1.07 N/A N/A Domestic Plant #8 1.11 N/A N/A Domestic Plant #9 1.07 N/A N/A Average 1.03 0.92 0.93 Std. Dev. % 5% 6% 7%
Total Number of Capsules 69 87 44 Total Number of Threshold Foils 295 454 227 WCAP- 18060-NPNoebr21 Revision 1
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E-3 Table E-2: In-Vessel Surveillance Capsules BE/C Ratios BE/C Fluence Number of Number of Plant Number (E>I.0 MeV) BE/C DPA Threshold Foils Capsules International Plant #1 1.04 1.05 8 3 International Plant #2 0.99 0.99 8 3 International Plant #3 N/A N/A N/A N/A Domestic Plant #1 1.03 1.04 20 4 International Plant #4 0.92 0.90 14 4 Domestic Plant #2 0.96 0.97 20 4 International Plant #5 1.05 1.06 19 4 International Plant #6 0.90 0.93 13 3 International Plant #7 0.89 0.92 12 3 Domestic Plant #3 0.97 0.99 30 6 Domestic Plant #4 0.96 0.96 29 6 Domestic Plant #5 0.93 0.95 25 5 International Plant #8 0.94 0.95 14 4 International Plant #9 1.04 1.04 16 4 Domestic Plant #6 0.98 0.99 19 5 Domestic Plant #7 1.02 1.03 13 3 Domestic Plant #8 1.08 1.08 11 3 Domestic Plant #9 1.02 1.02 24 5 Average 0.98 0.99 295 (Total) 69 (Total)
Std. Dev. % 6% 5%
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E-4 Table E-3: EVND Core Midplane BE/C Ratios Plant Number BE/C Fluence Number of Number of
(# of EVND Sets) (E>I.0 MeV) BE/C DPA Threshold Foils Capsules Domestic Plant #1 (1) 0.93 0.90 15 4 Domestic Plant #2 (1) 0.92 0.88 16 4 International Plant #1 (2) 1.03 1.03 40 8 International Plant #2 (2) 1.00 1.01 40 8 International Plant #3 (1) 0.99 1.02 28 4 International Plant #4 (1) 0.89 0.88 20 4 International Plant #5 (1) 0.92 0.97 15 3 International Plant #6 (2) 0.89 0.91 48 8 International Plant #7 (2) 0.85 0.85 32 8 Domestic Plant #3 (2) 0.89 0.90 48 8 Domestic Plant #4 (2) 0.89 0.92 48 8 Domestic Plant #5 (1) 0.96 0.98 24 4 International Plant #8 (2) 0.86 0.88 40 8 International Plant #9 (1) 0.84 0.84 20 4 Domestic Plant #6 (1) 0.93 0.94 20 4 Average 0.92 0.93 454 (Total) 87 (Total)
Std. Dev. % 6% 7%
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E-5 Table E-4: EVND Off-Midplane BE/C Ratios BE/C Fluence Number of Number of Plant Number (E>I.0 MeV) BE/C DPA Threshold Foils Capsules Domestic Plant #1 1.32 0.40 8 2 Domestic Plant #2 0.92 0.74 8 2 International Plant #1 0.97 1.18 10 2 International Plant #41* 0.90 1.06 10 2 International Plant #2 1.04 0.85 10 2 International Plant #2* 0.93 0.90 10 2 International Plant #3 1.28 1.03 14 2 International Plant #4 1.42 0.54 10 2 International Plant #5 0.82 1.03 10 2 International Plant #6 0.96 1.11 14 2 International Plant #6* 1.07 0.97 10 2 International Plant #7 0.98 0.99 7 2 International Plant #7* 1.56 0.56 10 2 Domestic Plant #3 1.01 0.85 10 2 Domestic Plant #3' 1.01 0.95 10 2 Domestic Plant #4 0.91 1.06 12 2 Domestic Plant #4* 0.86 1.09 12 2 Domestic Plant #5 1.02 0.79 12 2 International Plant #8 1.30 0.71 10 2 International Plant #8* 1.05 1.00 10 2 International Plant #9 0.85 1.08 10 2 Domestic Plant #6 1.06 0.70 10 2 Average 1.06 0.89 227 (Total) 44 (Total)
Std. Dev. % 19% 23%
- These plants have two sets of EVND capsules installed, they are listed separated in Table E-4, but listed together in Table E-3.
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E-6 Table E-5: EVND Off-Midplane BE/C Ratios with Outliners Rejected BE/C Fluence Number of Number of Plant Number (E>I.0 MeV) BE/C DPA Threshold Foils Capsules Domestic Plant #1 Rejected Rejected Rejected Rejected Domestic Plant #2 0.92 0.74 8 2 International Plant # 1 0.97 1.18 10 2 International Plant #1" 0.90 1.06 10 2 International Plant #2 1.04 0.85 10 2 International Plant #2* 0.93 0.90 10 2 International Plant #3 1.28 1.03 14 2 International Plant #4 Rejected Rejected Rejected Rejected International Plant #5 0.82 1.03 10 2 International Plant #6 0.96 1.11 14 2 International Plant #6* 1.07 0097 10 2 International Plant #7 0.98 0.99 7 2 International Plant #7* Rejected Rejected Rejected Rejected Domestic Plant #3 1.01 0.85 10 2 Domestic Plant #3' 1.01 0.95 10 2 Domestic Plant #4 0.91 1.06 12 2 Domestic Plant #4* 0.86 1.09 12 2 Domestic Plant #5 1.02 0.79 12 2 International Plant #8 1.30 0.71 10 2 International Plant #8* 1.05 1.00 10 2 International Plant #9 0.85 1.08 10 2 Domestic Plant #6 1.06 0.70 10 2 Average 1.00 0.95 199 (Total) 38 (Total)
Std. Dev. % 13% 15%
- These plants have two sets of EVND capsules installed, they are listed separated in Table E-5, but listed together in Table E-3.
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F-i APPENDIX F HIGHER ORDER SN SENSITIVITY STUDY In order to address SRXB-RAI 13, the comparison study has been done for RAPTOR-M3G code using different angular quadrature sets, namely S8 versus Sl 2 level symmetric quadrature sets. Cycle 21 RAPTOR-M3G model in Reference I was used for the comparative calculations. Both calculations were done using P3 Legendre expansion for anisotropic scattering cross section treatment and DTW differencing scheme. The results for the materials in the beltline and extended beitline regions are presented in Table F-I. The results show that the difference between the fast neutron fluence rates calculated using Ss level symmetric quadrature sets are within 1% of those calculated using Sj3 level symmetric quadrature sets.
Table F-l: Fast Neutron Fluence Rates Comparison between using S8 versus S12 Quadrature Sets Lower Shell to Intermediate Shell Upper Shell to Bottom Head Ring to Lower Shell Intermediate Shell Quadrature Sets Circ. Weld W04 Circ. Weld W05 Circ. Weld W06 S8 1.26E+09 1.54E+I10 6.98E+08 S12 1.27E+09 1.53E+ 10 6.94E+08 (Ss-Sl2)/Sl2 Ratio -0.8 1% 0.45% 0.67%
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