ML24226B211
| ML24226B211 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/22/2024 |
| From: | Turner Z Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Turner, Zachary, NRR/DORL/LPL2-1 | |
| References | |
| EPID L-2024-LLA-0098, TS 3.6.5 | |
| Download: ML24226B211 (1) | |
Text
August 22, 2024 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - REGULATORY AUDIT
SUMMARY
IN SUPPORT OF LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE, ACTIONS (EPID L-2024-LLA-0098)
Dear Jamie Coleman:
By application dated July 18, 2024, Agencywide Documents Access and Management System (ADAMS) Accession No. ML24201A108, as supplemented by letter dated August 16, 2024 (ML24229A245), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a license amendment request (LAR) to amend the Technical Specifications (TS) for Renewed Facility Operating License Nos. NPF-2 and NPF-8 for Joseph M. Farley Nuclear Plant (Farley),
Units 1 and 2, respectively. Specifically, the LAR would revise the Required Actions and Completion Times in TS 3.6.5, Containment Air Temperature.
The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the licensee's submittals and determined that a regulatory audit would assist in the timely completion of the licensing review process. The NRC staff issued the regulatory audit plan on July 25, 2024 (ML24200A139).
The NRC staff conducted a hybrid formatted regulatory audit from July 30 to August 16, 2024, to support its review of the LAR. During the regulatory audit, the NRC staff reviewed documents and held discussions with SNC and its representatives. The regulatory audit summary is enclosed with this letter.
J.
If you have any questions, please contact me at (301) 415-2258 or via email at Zachary.Turner@nrc.gov.
Sincerely,
/RA/
Zachary M. Turner, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
Regulatory Audit Summary cc: Listserv
Enclosure REGULATORY AUDIT
SUMMARY
BY THE OFFICE OF NUCLEAR REACTOR REGULATION IN SUPPORT OF THE LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE, ACTIONS JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By application dated July 18, 2024, Agencywide Documents Access and Management System (ADAMS) Accession No. ML24201A108, as supplemented by letter dated August 16, 2024 (ML24229A245), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a license amendment request (LAR) to amend the Technical Specifications (TS) for Renewed Facility Operating License Nos. NPF-2 and NPF-8 for Joseph M. Farley Nuclear Plant (FNP, Farley), Units 1 and 2, respectively. Specifically, the LAR would revise the Required Actions and Completion Times in TS 3.6.5, Containment Air Temperature.
The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the licensee's submittals and determined that a regulatory audit would assist in the timely completion of the licensing review process. The NRC staff issued the regulatory audit plan on July 25, 2024 (ML24200A139),
which detailed the logistics, scheduling, and audit request items.
The NRC staff conducted a hybrid (virtual and in-person) formatted regulatory audit from July 30 to August 16, 2024, to support its review of the LAR. During the regulatory audit, the NRC staff reviewed documents and held discussions with SNC and its representatives.
The regulatory audit was performed consistent with NRC Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-111, Revision 1, Regulatory Audits, dated October 31, 2019 (ML19226A274).
2.0 AUDIT ACTIVITIES AND CONCLUSION The audit was conducted via the use of two online reference portals (CERTEC and Westinghouse SharePoint) set up by the licensee and webinars. An in-person audit session held on July 31, 2024, at Excel Services Corporation in Rockville, MD, which was also supplemented by a simultaneous webinar. Additionally, a virtual audit session was held on August 6, 2024 to further discuss information regarding this LAR. Section 3.0 of this audit summary lists the individuals that took part in or attended the audit. Through the online reference portals, the NRC staff reviewed the licensees documents made available from July 30 through August 16, 2024.
Technical discussions held during the audit were focused on the following major areas:
(a) application specific probabilistic risk assessment (PRA); (b) the proposed change effects on net positive suction head and peak cladding temperature; (c) accident analysis (small and large break loss-of-coolant accidents); and (d) GOTHIC Code inputs and modelling. On August 15, 2024, the NRC staff provided a formal meeting to brief the conclusion of the regulatory audit including audit objectives that were met and details on the path forward. There were no open items from audit discussions and no deviations from the audit plan. Non-docketed information provided by the licensee in response to audit questions raised during the formal audit meetings are listed in section 5.0 of this audit summary.
By letter dated August 16, 2024 (ML24229A245), the licensee provided a supplement that detailed the technical discussion related to the audit questions outlined in section 4.0 of this audit summary. The NRC staff reviewed the licensees supplement to assess if any additional information was needed and determined that no additional information would be needed to complete its review of the subject LARs.
3.0 AUDIT PARTICIPANTS NRC Audit Team Name Email Review Area (Organization)
Zach Turner Zachary.Turner@nrc.gov Plant Licensing Branch LPL2-1 Santosh Bhatt Santosh.Bhatt@nrc.gov Nuclear Systems Performance Branch (SNSB)
Ahsan Sallman Ahsan.Sallman@nrc.gov John Lehning John.Lehning@nrc.gov Nuclear Methods and Fuels Branch (SFNB)
Sunwoo Park Sunwoo.Park@nrc.gov PRA Licensing Branch C (APLC)
Keith Tetter Keith.Tetter@nrc.gov David Nold David.Nold@nrc.gov Containment and Plant Systems Branch (SCPB)
Nicholas Soliz Nicholas.Soliz@nrc.gov Steve Smith Stephen.Smith@nrc.gov Technical Specifications Branch (STSB)
Matthew McConnell Matthew.McConnell@nrc.gov Long Term Operations and Modernization Branch (ELTB)
SNC Audit Team Name Organization Ryan Joyce SNC Wesley Sparkman Thomas Kindred Richard Langford Victoria Sundstrom Jonathan Register Jared Cannon Name Organization Jason Douglass Eulalio Elizondo Eddie Grant Jeremiah Gilbreath Kevin Barber Westinghouse Andrew Bowman Tristan Gibbons Jeffrey Kobelak Brian Ising Tom Elicson Jensen Hughes Frank Hope 4.0 AUDIT QUESTIONS During the course of the regulatory audit, the NRC staff identified the following audit questions to help facilitate an understanding of the licensees LAR:
No.
Audit Question
- 1.
The licensee stated that accumulator temperature sensitivities from similar
[pressurized water reactor (PWR) plant designs (referred as Plant A and Plant B in the LAR) with similar fuel assembly design, power level, and predicted cladding temperature response at FNP were used to estimate the effect of the maximum accumulator temperature limit increase on the peak cladding temperature (PCT) during a large-break LOCA (LBLOCA). The licensee stated in that LAR that the WCOBRA/TRAC code was used to perform sensitivity studies and that the code is essentially the same, allowing for certain documented error corrections and changes, as the one used for the NRC-approved ASTRUM evaluation model applied in the current Analysis of Record (AOR) at FNP.
Consistent with the 10 CFR 50.46 requirements which state that emergency core cooling system (ECCS) performance must be calculated in accordance with an acceptable evaluation model and that the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident, provide the following:
a) A detailed discussion on why the evaluation was performed using similar PWR plant designs and not the actual FNP model from the existing AOR.
b) A comparison of changes made to the WCOBRA/TRAC code used for the sensitivity study performed for Plant A and Plant B to the WCOBRA/TRAC code version from the NRC-approved ASTRUM methodology used for FNP.
c) A comparison of key plant input parameters used for Plants A and B to the input parameters for the existing FNP LBLOCA evaluation to support that the analytical technique used to perform the sensitivity studies realistically describes the behavior of the reactor system for FNP during the LBLOCA event.
No.
Audit Question
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
- 2.
Provide results from the plant accumulator temperature sensitivity performed for the plant designs similar to FNP (i.e., Plants A and B). Include details on development of the trendline for the sensitivity study based on the available data.
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
- 3.
Provide results for the eighteen accumulator temperature sensitivity studies mentioned in the LAR that were performed as a confirmation of the estimated magnitude of impact of the change to the accumulator temperature limit. Further provide key information concerning the plants considered in the sensitivity studies to confirm similarity to FNP.
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
- 4.
Provide details on how the impacts of changes and errors in the evaluation model were considered when evaluating their impact on accumulator temperature changes.
The FNP analysis of record was performed in 2004 using the ASTRUM code, and changes and error corrections affecting this analysis over the past 20 years have resulted in an estimated PCT increase of approximately 200 F. In particular, the method used to estimate the impact of the change in accumulator temperature does not appear to consider the impacts of thermal conductivity degradation and other errors from Final Safety Analysis Report (FSAR) PCT rack-up, which would impact fuel stored energy and could influence the calculated PCT impact associated with the increased accumulator temperature limit to a greater degree than the estimate provided in the LAR.
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
- 5.
Based on the revised mass and energy release corresponding to an accumulator temperature of 124°F using the NRC-approved WCAP-10325-P-A methodology, for the limiting double-ended pump suction break with the most limiting single failure, using a bounding containment initial temperature of 127°F, provide the following:
(a). Changes in the GOTHIC model used for containment pressure, temperature, and sump temperature responses.
(b). Results of peak pressure, peak temperature, peak sump temperature, and maximum recirculation sump temperature including comparison of these parameters with those in the analysis of record.
(c). Result graphs of containment pressure, temperature, and sump temperature responses.
(d). Results showing available net positive suction head (NPSH), required NPSH, and NPSH margins for the residual heat removal (RHR) and containment spray (CS) pumps.
No.
Audit Question
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
6 In the LAR enclosure, refer to the following statement under the heading, Net Positive Suction Head (NPSH) Evaluation, The strainer head losses are also shown to decrease as the sump temperature increases above 140°F. Based on the competing effects between vapor pressure of the sump inventory and strainer head losses, the pump NPSH margin would be expected to increase or stay the same as the sump temperature increases above 212°F.
Provide a basis and results of the strainer head losses decrease as the temperature increases above 140°F.
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
7 Provide details on the impact of accumulator temperature change on small break loss-of-coolant accident (SBLOCA) analysis for various break-sizes analyzed. Include discussion on PCT time vs accumulator injection time for each of the break-sizes and how those will be impacted by the change, especially for cases where the accumulator injection times are relatively close to the PCT time.
==
Conclusion:==
The licensee agreed to voluntarily supplement the LAR and provide information pertaining to this audit question on the docket.
5.0 DOCUMENTS REVIEWED The licensee provided the following supporting documents (e.g., analyses, calculations, reports, drawings, and procedures) available on the Farley and Westinghouse document portals during the audit period. Below lists the documents available on the portals that the NRC audit team reviewed during the audit.
SNC Report: SNC Report No. NMP-ES0050-F01, RER No. SNC1823599-01 Sequence No. 1, Evaluation of Increasing FNP Containment TS Temperature to 124F, Version 5, June 11, 2024.
SNC Report: SNC Report No. F-RIE-FIREPRA-U00-018-002, 2022 Focused Scope Peer Review for FIRE PRA Model, January 31, 2023.
SNC Report: SNC Report No. F-RIE-FIREPRA-U00-018-003, 2023 F&O Close Out for Fire PRA Modell, June 15, 2023.
SNC Report: SNC Report No. F-RIE-IEIF-U00-012-001, 2019 Focused Scope Peer Review for Internal Events (including Internal Flooding) PRA Model, Revision 1.0, December 16, 2019.
SNC Report: SNC Report No. F-RIE-IEIF-U00-012-002, 2022 F&O Closeout by Independent Assessment, Revision 1.0, January 31, 2023.
SNC Report: SNC Report No. PRA-BC-F-24-001, Farley Containment MAAP Analysis, July 30, 2024.
SNC Report: SNC Report No. RBA-23-004-F, Farley Containment Temperature Increase Evaluation, August 22, 2023.
SNC Report: File Name, [request for additional information] (RAI) 6-1, August 12, 2024.
SNC Report: File Name, RAI 5, August 15, 2024.
SNC Report: File Name, RAI 5d RHR and CTMT Spray NPSH margin, August 15, 2024.
Westinghouse Report: Transmittal of WEC Report No. ALAM-CONT-AA-000003, Results of LOCA Mass & Energy Release Analyses for Farley Units 1 and 2 (ALA/APR) for an Increase in the Containment Temperature to 124F, Revision 1, April 4, 2024.
Westinghouse Report: WEC Report No. ALAM-LOCA-TM-AA-000008, LBLOCA Input to J.M. Farley Units 1 and 2 Engineering Report and/or License Amendment Request for a Change to the Maximum Containment Temperature Technical Specification from 120F to 124F, Revision 0, November 30, 2023.
Westinghouse Report: WEC Report No. ALAM-LOCA-TM-AA-000011, Small Break LOCA Evaluation of the J.M. Farley Units 1 and 2 Containment Temperature Increase to 124F, March 12, 2024.
Westinghouse Report: WEC Report No. ALAM-LOCA-TM-AA-000014, Suggested Large Break LOCA (LBLOCA) LAR Input for J.M. Farley Units 1 and 2 for the Containment Temperature Increase to 122F, Revision 0.
Westinghouse Report: WEC Report No. ALA-24-41, Transmittal of Responses to Draft RAIs 1 through 4 on the Farley Units 1 and 2 Technical Specification 3.6.5 (Containment Air Temperature) Change, Revision 0, August 13, 2024.
Westinghouse Report: WEC Report No. ALA-24-44, Transmittal of Response to Draft RAI 7 on the Farley Units 1 and 2 Technical Specification 3.6.5 (Containment Air Temperature) Change, Revision 0, August 14, 2024.
Westinghouse Report: WEC Report No. WCAP-16272-P, Best-Estimate Large-Break LOCA Analysis for J.M. Farley Units 1&2 Using the ASTRUM Methodology, Revision 2, January 1, 2006.
Westinghouse Report: WEC Report No. LTR-NRC-03-5, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2002, Revision 0, March 7, 2003.
Westinghouse Report: WEC Report No. LTR-NRC-04-17, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2003, Revision 0, March 25, 2004.
Westinghouse Report: WEC Report No. LTR-NRC-05-20, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2004, Revision 0, April 11, 2005.
Westinghouse Report: WEC Report No. LTR-NRC-06-8, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2005, Revision 0, March 16, 2006.
Westinghouse Report: File Name, Best Estimate of the Large Break Loss of Coolant Accident for Plant A, June 1, 2003.
Westinghouse Report: File Name, Best Estimate of the Large Break Loss of Coolant Accident for Plant B, December 1, 2002.
Westinghouse Report: File Name, Background Information for Code Versions.
Westinghouse Report: File Name, SI Flow Comparison for FNP, Plant A and Plant B.
Westinghouse Report: File Name, Accumulator Temperature Study Results.
- concurrence via email NRR-106 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL1/LAiT*
NRR/DORL/LPL2-1/LA*
NRR/DRA/APLC/ABC*
NAME ZTurner CAdams KGoldstein ANeuhausen DATE 8/19/2024 8/15/2024 8/21/2024 8/19/2024 OFFICE NRR/DSS/STSB/BC*
NRR/DSS/SNSB/BC*
NRR/DSS/SCPB/BC*
NRR/DEX/ELTB/BC*
NAME SMehta (RElliott for)
PSahd MValentin JPaige (KMiller for)
DATE 8/19/2024 8/19/2024 8/20/2024 8/20/2024 OFFICE NRR/DORL/LPL2-1/BC*
NRR/DORL/LPL2-1/PM*
NAME MMarkley (SWilliams for)
ZTurner DATE 8/22/2024 8/22/2024