ML25090A334
| ML25090A334 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/31/2025 |
| From: | Hamman D Wolf Creek |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 000906 | |
| Download: ML25090A334 (1) | |
Text
P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Dustin T. Hamman Director Nuclear and Regulatory Affairs March 31, 2025 000906 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Reference:
Westinghouse Letter SAP-LOCA-TM-A5-000006, dated February 13, 2025, Wolf Creek Unit 1 10 CFR 50.46 Annual Notification and Reporting for 2024
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 2024. 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 2024. 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 2024. 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 2024 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.
000906 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
Attachment:
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:
A. N. Agrawal (NRC), w/a S. S. Lee (NRC), w/a J. D. Monninger (NRC), w/a Senior Resident Inspector (NRC), w/a WCNOC Licensing Correspondence RA 25-000906, w/a
Attachment I to 000906 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)
Barrel/Baffle Region Volume Error
=
Background===
An error was identified during the course of a recent Best Estimate Large Break LOCA analysis in which the volume between the core barrel and the baffle plates within the active fuel length was modeled incorrectly. The corrected values have been evaluated for impact on the current licensing-basis analysis results. The correction of this error represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM.
Estimated Effect The error was evaluated to have a negligible impact on the calculated results, leading to an estimated peak cladding temperature (PCT) impact of 0°F.
CCFL Global Volume Error
=
Background===
An error was identified during the course of a recent Best Estimate Large Break LOCA analysis in which the volume in the global channel in the CCFL region above the active fuel length was modeled incorrectly. The corrected values have been evaluated for impact on the current licensing-basis analysis results. This error represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM.
Estimated Effect The CCFL global volume modeling error has been generically evaluated to have a negligible impact on PCT for affected analyses, leading to an estimate PCT impact of 0°F.
Attachment II to 000906 Page 1 of 4 Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms
Attachment II to 000906 Page 2 of 4 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
ASSESSMENTS*
Delta PCT
(°F)
Reference #
Note #
Reporting 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-90-148/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 000906 Page 3 of 4 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
WOLF CREEK EM:
ASTRUM (2004)
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
ASSESSMENTS*
Delta PCT
(°F)
Reference #
Note #
Reporting Year**
1.
Containment Fan Cooler Capacity 0
2,4 (a) 2014 2.
Decay Group Uncertainty Factors Errors
-10 3
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 00906 Page 4 of 4 10 CFR 50.46 Reporting SharePoint Site Check:
EMs applicable to Wolf Creek Unit 1:
Realistic Large Break - ASTRUM (2004)
Appendix K Small Break - NOTRUMP 2024 Issues Transmittal Letter Issue Description None None