ML24089A135

10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes

ML24089A135
Person / Time
Site:	Wolf Creek
Issue date:	03/29/2024
From:	Hamman D Wolf Creek
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
000402
Download: ML24089A135 (1)

Text

Dustin T. Hamman Director Nuclear and Regulatory Affairs

March 29, 2024 000402

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

Reference:

Westinghouse Letter SAP-LOCA -TM -A5 -000005, dated February 12, 2024,

Wolf Creek Unit 1 10 CFR 50.46 Annual Notification and Reporting for 2023

Subject:

Docket No. 50-482: 10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes

Commissioners and Staff:

In accordance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the annual reporting requirement for the Wolf Creek Generating Station (WCGS).

WCNOC has reviewed the above Reference, which addresses 10 CFR 50.46 reporting information pertaining to the Emergency Core Cooling System (ECCS) Evaluation Model changes that were implemented by Westinghouse for 2023. The Evaluation Model changes and errors (except any plant-specific errors in the application of the model) have been provided to the NRC via Westinghouse letter. The review concludes that the effect of changes to, or errors in, the Evaluation Models on the limiting transient peak cladding temperature (PCT) is not significant for 2023. Therefore, changes to the ECCS Evaluation Models are being reported as an annual report.

Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2023. These model changes and enhancements do not have impacts on the PCT and, generally, will not be represented on the PCT rack -up forms.

Attachment II provides PCT rack-up forms for the calculated Large Break Loss -of-Coolant Accident (LOCA) and Small Break LOCA PCT margin allocations in effect for the 2023 WCGS Evaluation Models. The PCT values determined in the Large Break and Small Break LOCA analyses of record, combined with all of the PCT allocations, remain below the 10 CFR 50.46(b)(1) regulatory limit of 2200 °F. Therefore, WCGS is in compliance with 10 CFR 50.46 requirements and no reanalysis or other action is required.

P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 000402 Page 2 of 2

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204.

Sincerely,

Dustin T. Hamman

DTH/ jkt

Attachments : I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss -of-Coolant Accidents (LOCA)

II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms

cc: S. S. Lee (NRC), w/a J. D. Monninger (NRC), w/a G. E. Werner (NRC), w/a Senior Resident Inspector (NRC), w/a WC Licensing Correspondence, w/a - RA 24-000402

Attachment I to 000402 Page 1 of 1

ASSESSMENT OF CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS FOR LARGE AND SMALL BREAK LOSS-OF -

COOLANT ACCIDENTS (LOCA)

MINIMUM INJECTABLE REFUELING WATER STORAGE TANK VOLUME REDUCTION

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Background===

The minimum injectable refueling water storage tank (RWST) volume is modeled in the small break loss-of -coolant accident (SBLOCA) analysis to determine the time of switchover to sump recirculation. An updated input value resulted in a reduction to the volume modeled in the SBLOCA analysis of record, and therefore an earlier time of switchover to sump recirculation.

This change was qualitatively evaluated for impact on the SBLOCA analysis results. This item represents a Change in Plant Configuration or Set Points distinguished from an evaluation model change in Section 4 of WCAP-13451.

Affected Evaluaton Model(s)

1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP

Estimated Effect

The impact of the reduction in minimum injectable RWST volume was evaluated to be small, and the change in the time of sump recirculation would have a negligible effect on the SBLOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

GENERAL CODE MAINTENACE

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Background===

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.

Affected Evaluation Model(s)

2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM

Estimated Effect

The nature of these changes leads to an estimated PCT impact of 0°F.

Attachment II to 000402 Page 1 of 3

Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms

LOCA Peak Cladding Temperature (PCT) Summary Plant Name: WOLF CREEK EM: NOTRUMP AOR

Description:

Appendix K Small Break Summary Sheet Status: Current WOLF CREEK SAP_LOCA-50.46_SAP_Base_Appendix_K_SBLOCA - 1.1 V.V PCT (°F) Reference # Note #

ANALYSIS-OF-RECORD 936 1

Delta PCT Reporting ASSESSMENTS* (°F) Reference # Note # Year**

1. Loose Part Evaluation 45 2 (a) 1990

AOR + ASSESSMENTS PCT = 981.0 °F

• The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.

• • The Reporting Year refers to the annual reporting year in which this assessment was included.

REFERENCES 1 WCAP-16717-P, Rev. 0, Wolf Creek Generating Station (SAP), MSIV/MFIV Replacement Project, Small Break Loss of Coolant Accident Analysis Engineering Report, January 2007.

2 SAP 148/NS-OPLS-OPL-I-90-239, Wolf Creek Nuclear Operating Corporation, RCS Loose Part Evaluation, April 1990.

NOTES:

(a) This penalty will be carried to track the loose part which has not been recovered.

Attachment II to 000402 Page 2 of 3

LOCA Peak Cladding Temperature (PCT) Summary Plant Name: WOLF CREEK EM: ASTRUM (2004)

AOR

Description:

Best Estimate Large Break Summary Sheet Status: Current WOLF CREEK SAP_LOCA-50.46_SAP_Base_ASTRUM - 1.2 V.V PCT (°F) Reference # Note #

ANALYSIS-OF-RECORD 1900 1

Delta PCT Reporting ASSESSMENTS* (°F) Reference # Note # Year**

1. Containment Fan Cooler Capacity 0 2,4 (a) 2014

2. Decay Group Uncertainty Factors 3 2014 Errors -10

AOR + ASSESSMENTS PCT = 1890.0 °F

• The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.

• • The Reporting Year refers to the annual reporting year in which this assessment was included.

REFERENCES 1 WCAP-17107-P, Revision 1, Best -Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology, January 2014.

2 LTR-LIS-14-400, 10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity, August 2014.

3 LTR-LIS-14-492, Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors, November 2014.

4 LTR-LIS-19-282, Wolf Creek 10 CFR 50.46 PCT Summary Sheet Updates for Replacement Fan Cooler Tube Bundles Installation and Planned Retirement of Cycle 23 Sheets, August 2019.

NOTES:

(a) The estimated effect includes the corrected fan cooler heat removal rates and implementation of replacement tube bundles in the containment fan coolers, which were installed for Cycle 24.

Attachment II to 000402 Page 3 of 3

10 CFR 50.46 Reporting SharePoint Site Check:

EMs applicable to Wolf Creek:

Realistic Large Break - ASTRUM (2004)

Appendix K Small Break - NOTRUMP

2023 Issues Transmittal Letter Issue Description SAP-LOCA-TM-A5-000004 Wolf Creek 10 CFR 50.46 Report for the Reduction in Minimum Injectable Refueling Water Storage Tank Volume.