RA-24-0083, Annual Report of Changes Pursuant to 10 CFR 50.46

Annual Report of Changes Pursuant to 10 CFR 50.46

ML24116A175
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire
Issue date:	04/25/2024
From:	Ellis K Duke Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-24-0083
Download: ML24116A175 (1)

Text

Kevin M. Ellis General Manager Nuclear Regulatory Affairs, Policy &

Emergency Preparedness

Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 843-951-1329 Kevin.Ellis@duke-energy.com

10 CFR 50.46 Serial: RA-24-0083 April 25, 2024

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324

Catawba Nuclear Station, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. NPF-35 and NPF-52 Docket Nos. 50-413 and 50-414

Shearon Harris Nuclear Power Plant, Unit 1 Renewed Facility Operating License No. NPF-63 Docket No. 50-400

McGuire Nuclear Station, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. NPF-9 and NPF-17 Docket Nos. 50-369 and 50-370

Oconee Nuclear Station, Unit Nos. 1, 2 and 3 Renewed Facility Operating License Nos. DPR-38, DPR-47 and DPR-55 Docket Nos. 50-269, 50-270 and 50-287

H. B. Robinson Steam Electric Plant, Unit 2 Renewed Facility Operating License No. DPR-23 Docket No. 50-261

SUBJECT:

Annual Report of Changes Pursuant to 10 CFR 50.46

Ladies and Gentlemen:

Pursuant to 10 CFR 50.46(a)(3)(ii), Duke Energy hereby submits the enclosed annual reports of changes to, or errors in, Emergency Core Cooling System (ECCS) evaluation models. These reports cover the period from January 1, 2023 to December 31, 2023 for the Brunswick Steam Electric Plant (BNP), Catawba Nuclear Station (CNS), Shearon Harris Nuclear Power Plant (HNP), McGuire Nuclear Station (MNS), Oconee Nuclear Station (ONS), and H.B. Robinson Steam Electric Plant (RNP) and are provided in Enclosures 1 through 6, respectively.

No regulatory commitments are contained in this submittal.

U.S. Nuclear Regulatory Commission RA-24-0083 Page 3 cc:

L. Dudes, USNRC, Region II Regional Administrator L. Haeg, USNRC NRR Project Manager for BNP and RNP S. Williams, USNRC NRR Project Manager for CNS and ONS M. Mahoney, USNRC NRR Project Manager for HNP J. Klos, USNRC NRR Project Manager for MNS G. Smith, USNRC Senior Resident Inspector for BNP D. Rivard, USNRC Senior Resident Inspector for CNS P. Boguszewski, USNRC Senior Resident Inspector for HNP C. Safouri, USNRC Senior Resident Inspector for MNS N. Smalley, USNRC Senior Resident Inspector for ONS J. Zeiler, USNRC Senior Resident Inspector for RNP

Page 1 of 4 RA-24-0083

ENCLOSURE 1: BNP 10 CFR 50.46 Annual Report

Brunswick Steam Electric Plant, Units 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62

Page 2 of 4 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for Brunswick Steam Electric Plant Units 1 and 2

During this reporting period there was one error report on the EXEM BWR-2000 evaluation model for ATRIUM 10XM, and one error report on the AURORA-B LOCA evaluation for ATRIUM 11. The A10XM error notice (Reference 2) documents an evaluation of the RODEX4 Pellet Rim Porosity Model error with an impact of +0°F, thus maintaining the A10XM Licensing PCT at 1925°F. The ATRIUM 11 error notice (Reference 1) documents an evaluation of the RODEX4 Pellet Rim Porosity Model error with an impact of

+0°F, thus maintaining the ATRIUM 11 Licensing PCT at 1897°F.

There were no revisions to the Licensing Basis AOR for A10XM nor to the ATRIUM 11 LOCA evaluations for this reporting period.

References

1. FS1-0060612 Revision 4.0, 10 CFR 50.46 PCT Error Report for Brunswick Units 1 and 2 ATRIUM 11 Fuel, Framatome Inc., October 2023.

2. FS1-0040060 Revision 4.0, 10 CFR 50.46 PCT Error Report for Brunswick Units 1 and 2 for MELLLA+ Operation, Framatome Inc., October 2023.

Page 3 of 4 RA-24-0083

A10XM Summary

10 CFR 50.46 Report for Brunswick Steam Electric Plant Units 1 and 2

Plant: Brunswick Steam Electric Plant, Units 1 and 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable):

Evaluation Model: EMF-2361(P)(A), Revision 0 EXEM BWR-2000 ECCS Evaluation Model, May 2001 Fuel: ATRIUM 10XM (A10XM)

A. Analysis of Record PCT 1923 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +2 °F 2 °F C. Baseline PCT for assessing new 1925 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. RODEX4 Pellet Rim Porosity +0 °F Model E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + net E) 1925 °F

Page 4 of 4 RA-24-0083

ATRIUM 11 Summary

10 CFR 50.46 Report for Brunswick Steam Electric Plant Units 1 and 2

Plant: Brunswick Steam Electric Plant, Units 1 and 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable):

Evaluation Model: ANP-10332P-A, Revision 0 AURORA-B LOCA Evaluation Model, March 2019 Fuel: ATRIUM 11 (A11)

A. Analysis of Record PCT 1897 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1897 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. RODEX4 Pellet Rim Porosity +0 °F Model E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + net E) 1897 °F

Page 1 of 6 RA-24-0083

ENCLOSURE 2: CNS 10 CFR 50.46 Annual Report

Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 / Renewed License Nos. NPF-35 and NPF-52

Page 2 of 6 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for Catawba Units 1 and 2

Evaluation of Off-Specification Uranium

The use of off-specification enriched uranium product (OSEUP) began with Catawba Unit 2 Cycle 26 in October 2022. The resulting impact on PCT was reported to the NRC in the spring of 2023 for the annual 10 CFR 50.46 reporting for the 2022 calendar year for Catawba Unit 2 only. OSEUP was first used at Catawba Unit 1 in 2023, so the resulting impact on PCT is included in the reporting for the 2023 calendar year.

Westinghouse evaluated the use of OSEUP in Catawba Unit 1 for impact to the Small Break LOCA PCT.

The licensing basis small-break loss-of-coolant accident (SBLOCA) analysis for Catawba Unit 1 is based on the NOTRUMP Evaluation Model which uses the decay heat model described in Section I.A.4 of Appendix K to 10 CFR 50. A change is being made to the fuel assemblies manufactured for the Catawba units in that off-spec enriched uranium product will be utilized for the fuel pellets. This item represents a change in plant configuration or associated setpoints, distinguished from a change to the evaluation model documented in Section 4 of WCAP-13451. On a best estimate basis, the off-spec enriched uranium product will have a very minor impact on the decay heat generation rate; however, this does not impact the degree of conservatism inherent in the use of the Appendix K decay heat model. The impact of the off-spec enriched uranium product on the SBLOCA analysis of record is negligible, leading to an estimated PCT impact of 0°F for Catawba Unit 1.

Westinghouse evaluated the use of OSEUP in Catawba Unit 1 for impact to the Large Break LOCA PCT.

The licensing basis large-break loss-of-coolant accident (LBLOCA) analysis for Catawba Unit 1 is based on the Code Qualification Document (CQD) methodology. The decay heat modeled in the analysis followed the NRC-approved method for standard UO2 fuel pellets. A change is being made to the fuel assemblies manufactured for the Catawba Unit 1 in that off-spec enriched uranium product will be utilized for the fuel pellets. This change represents a change in plant configuration or associated setpoints, distinguished from a change to the evaluation model documented in Section 4 of WCAP-13451. The impact of the off-spec enriched uranium product on the analysis of record (AOR) is negligible, leading to an estimated PCT impact of 0°F for all transient phases for Catawba Unit 1.

Error in Flow Area and Volume of Thimble Components

The reactor vessel thimble bypass flow is modeled in best-estimate (BE) loss-of-coolant accident (LOCA) analyses using three PIPE components, which represent the thimble bypass for peripheral low power assemblies, interior assemblies located under guide tubes, and interior assemblies not located under guide tubes. Westinghouse discovered that the number of assemblies modeled is inconsistent with the number of assemblies represented by one or more of the thimble components, leading to an incorrect flow area and volume for the affected thimble component(s). The correction of this error represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. The error was evaluated to have a negligible impact on the calculated results, leading to an estimated PCT impact of 0°F.

General Code Maintenance

Westinghouse identified and communicated that general code maintenance changes in the Best Estimate LBLOCA evaluation model were made in 2023. Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451. The nature of these changes leads to an estimated peak cladding temperature impact of 0°F.

Page 3 of 6 RA-24-0083

10 CFR 50.46 Report for Catawba Unit 1 - Large Break LOCA

Plant: Catawba Nuclear Station, Unit 1 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA

A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +58 °F 378 °F C. Baseline PCT for assessing new 2086 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Error in Flow Area and Volume +0 °F of Thimble Components

2. Use of Off Specification +0 °F Enriched Uranium Product E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 2086 °F

Page 4 of 6 RA-24-0083

10 CFR 50.46 Report for Catawba Unit 1 - Small Break LOCA

Plant: Catawba Nuclear Station, Unit 1 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA

A. Analysis of Record PCT 1323 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1323 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Use of Off Specification +0 °F Enriched Uranium Product E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1323 °F

Page 5 of 6 RA-24-0083

10 CFR 50.46 Report for Catawba Unit 2 - Large Break LOCA

Plant: Catawba Nuclear Station, Unit 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA

A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +84 °F 404 °F C. Baseline PCT for assessing new 2112 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Error in Flow Area and Volume +0 °F of Thimble Components E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 2112 °F

Page 6 of 6 RA-24-0083

10 CFR 50.46 Report for Catawba Unit 2 - Small Break LOCA

Plant: Catawba Nuclear Station, Unit 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA

A. Analysis of Record PCT 1243 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1243 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1243 °F

Page 1 of 4 RA-24-0083

ENCLOSURE 3: HNP 10 CFR 50.46 Annual Report

Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63

Page 2 of 4 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for Shearon Harris Unit 1

One error for large break loss-of-coolant accident (LBLOCA) was identified in 2023. This error was in the default value used for fuel open porosity in the COPERNIC fuel model that is used in the Realistic LBLOCA uncertainty analysis performed under the EMF-2103(P)(A) Revision 3 methodology (Reference 1). The default value was incorrectly selected from a different fuel type when the fuel performance code COPERNIC was incorporated into the RLBLOCA methodology. The error in the fuel open porosity impacts the LOCA fuel initial conditions determined by COPERNIC and used in the RLBLOCA uncertainty analysis. The impact of this error on PCT is 0°F. The small break LOCA analysis is not impacted as the error is unique to EMF-2103(P)(A) Revision 3 applications.

References

1. AREVA Topical Report EMF-2103(P)(A), Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.

Page 3 of 4 RA-24-0083

10 CFR 50.46 Report for Shearon Harris Unit 1 - Large Break LOCA

Plant: Shearon Harris, Unit 1 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: EMF-2103(P)(A), Revision 3 Realistic Large Break LOCA for PWRs Fuel: 17x17 GAIA

A. Analysis of Record PCT 1820 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F None C. Baseline PCT for assessing new 1820 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Fuel model open porosity input +0 °F error E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1820 °F

Page 4 of 4 RA-24-0083

10 CFR 50.46 Report for Shearon Harris Unit 1 - Small Break LOCA

Plant: Shearon Harris, Unit 1 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: EMF-2328(P)(A), Rev. 0, and EMF-2328(P)(A), Rev. 0, Supplement 1, Rev. 0 PWR Small Break LOCA Evaluation Model Fuel: 17x17 GAIA and HTP

A. Analysis of Record PCT 1832 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1832 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1832 °F

Page 1 of 4 RA-24-0083

ENCLOSURE 4: MNS 10 CFR 50.46 Annual Report

McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 / Renewed License Nos. NPF-9 and NPF-17

Page 2 of 4 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for McGuire Units 1 and 2

Error in Flow Area and Volume of Thimble Components

The reactor vessel thimble bypass flow is modeled in best-estimate (BE) loss-of-coolant accident (LOCA) analyses using three PIPE components, which represent the thimble bypass for peripheral low power assemblies, interior assemblies located under guide tubes, and interior assemblies not located under guide tubes. Westinghouse discovered that the number of assemblies modeled is inconsistent with the number of assemblies represented by one or more of the thimble components, leading to an incorrect flow area and volume for the affected thimble component(s). The correction of this error represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451. The error was evaluated to have a negligible impact on the calculated results, leading to an estimated PCT impact of 0°F.

General Code Maintenance

Westinghouse identified and communicated that general code maintenance changes in the Best Estimate LBLOCA evaluation model were made in 2023. Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451. The nature of these changes leads to an estimated peak cladding temperature impact of 0°F.

Page 3 of 4 RA-24-0083

10 CFR 50.46 Report for McGuire Units 1 & 2 - Large Break LOCA

Plant: McGuire Nuclear Station, Units 1 & 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: WCAP-12945-P-A, Revision 0 Code Qualification Document for Best Estimate LOCA Analysis Fuel: 17x17 RFA

A. Analysis of Record PCT 2028 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +58 °F 378 °F C. Baseline PCT for assessing new 2086 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Error in Flow Area and Volume +0 °F of Thimble Components E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 2086 °F

Page 4 of 4 RA-24-0083

10 CFR 50.46 Report for McGuire Units 1 & 2 - Small Break LOCA

Plant: McGuire Nuclear Station, Units 1 & 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: WCAP-10054-P-A, Revision 0 NOTRUMP Fuel: 17x17 RFA

A. Analysis of Record PCT 1323 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1323 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1323 °F

Page 1 of 4 RA-24-0083

ENCLOSURE 5: ONS 10 CFR 50.46 Annual Report

Oconee Nuclear Station, Units 1, 2 and 3 Docket Nos. 50-269, 50-270 and 50-287 Renewed License Nos. DPR-38, DPR-47 and DPR-55

Page 2 of 4 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for Oconee Units 1, 2, & 3

There are no new errors or changes that affect the Oconee large break loss-of-coolant accident (LBLOCA) nor small break LOCA analyses of record for the 2023 calendar year.

Page 3 of 4 RA-24-0083

10 CFR 50.46 Report for Oconee Units 1, 2, & 3 - Large Break LOCA

Plant: Oconee Nuclear Station, Units 1, 2, & 3 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: BAW-10192P-A, Revision 0, BWNT LOCA Evaluation Model for Once-Through Steam Generator Plants and BAW-10192PA, Revision 0, Supplement 1P-A, Revision 0 Fuel: 15x15 Mark-B-HTP

A. Analysis of Record PCT; 1988 °F Reanalysis to fully incorporate fuel TCD, reported via ML21096A007 B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported N/A N/A C. Baseline PCT for assessing new 1988 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1988 °F

Page 4 of 4 RA-24-0083

10 CFR 50.46 Report for Oconee Units 1, 2, & 3 - Small Break LOCA

Plant: Oconee Nuclear Station, Units 1, 2, & 3 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: BAW-10192P-A, Revision 0, BWNT LOCA Evaluation Model for Once-Through Steam Generator Plants Fuel: 15x15 Mark-B-HTP

A. Analysis of Record PCT 1598 °F Full Power - 100% FP B. Net Cumulative 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new changes 1598 °F for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Error Net PCT Effect Absolute PCT Effect Corrections against Baseline PCT +0 °F 0 °F F. Licensing Basis PCT (C + E) 1598 °F

A. Analysis of Record PCT 1480 °F Reduced Power - 50% FP B. Net Cumulative 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new changes 1480 °F for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Error Net PCT Effect Absolute PCT Effect Corrections against Baseline PCT +0 °F 0 °F F. Licensing Basis PCT (C + E) 1480 °F

Page 1 of 4 RA-24-0083

ENCLOSURE 6: RNP 10 CFR 50.46 Annual Report

H. B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 / Renewed License No. DPR-23

Page 2 of 4 RA-24-0083

Summary of Errors Reported

10 CFR 50.46 Report for H.B. Robinson Unit 2

One error for large break loss-of-coolant accident (LBLOCA) was identified in 2023. This error was in the default value used for fuel open porosity in the COPERNIC fuel model that is used in the Realistic LBLOCA (RLBLOCA) uncertainty analysis performed under the EMF-2103(P)(A) Revision 3 methodology (Reference 1). The default value was incorrectly selected from a different fuel type when the fuel performance code COPERNIC was incorporated into the RLBLOCA methodology. The error in the fuel open porosity impacts the LOCA fuel initial conditions determined by COPERNIC and used in the RLBLOCA uncertainty analysis. The impact of this error on PCT is 0°F. The small break LOCA analysis is not impacted as the error is unique to EMF-2103(P)(A) Revision 3 applications.

References

1. AREVA Topical Report EMF-2103(P)(A), Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.

Page 3 of 4 RA-24-0083

10 CFR 50.46 Report for H.B. Robinson Unit 2 - Large Break LOCA

Plant: H.B. Robinson, Unit 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Large Break Evaluation Model: EMF-2103(P)(A), Revision 3 Realistic Large Break LOCA for PWRs Fuel: W15-LC, W15 HTP

A. Analysis of Record PCT 1771 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +0 °F 0 °F C. Baseline PCT for assessing new 1771 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. Fuel model open porosity input +0 °F error E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1771 °F

Page 4 of 4 RA-24-0083

10 CFR 50.46 Report for H.B. Robinson Unit 2 - Small Break LOCA

Plant: H.B. Robinson, Unit 2 Reporting Period: January 1, 2023 - December 31, 2023 LOCA Analysis Type (if applicable): Small Break Evaluation Model: EMF-2328(P)(A), Rev. 0 and EMF-2328(P)(A), Rev. 0, Supplement 1, Rev. 0 PWR Small Break LOCA Evaluation Model Fuel: W15-LC, W15 HTP

A. Analysis of Record PCT 1538 °F B. Net Cumulative 10 CFR 50.46 Net PCT Effect Absolute PCT Effect Changes and Error Corrections

- Previously Reported +3 °F 3 °F C. Baseline PCT for assessing new 1541 °F changes for significance (A + B)

D. Cumulative 10 CFR 50.46 Changes and Error Corrections

- This Reporting Period

1. None +0 °F E. Sum of 10 CFR 50.46 Changes and Net PCT Effect Absolute PCT Effect Error Corrections against Baseline +0 °F 0 °F PCT F. Licensing Basis PCT (C + E) 1541 °F