ML25209A497
| ML25209A497 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 08/08/2025 |
| From: | Tyree C NRC/NRR/DNRL/NLRP |
| To: | Wilson C Constellation Energy Generation |
| References | |
| Download: ML25209A497 (1) | |
Text
Christopher D. Wilson Director, License Renewal Constellation Energy Generation, LLC 200 Energy Way Kennett Square, PA 19348
SUBJECT:
CLINTON POWER STATION, UNIT NO. 1 - REPORT FOR THE AGING MANAGEMENT AUDIT REGARDING THE LICENSE RENEWAL APPLICATION REVIEW SUPPLEMENT
Dear Christopher Wilson:
By letter dated February 14, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24045A024), as supplemented by letter dated November 27, 2024 (ML24332A050), December 20, 2024 (ML24355A050), January 30, 2025 (ML25030A182),
March 25, 2025 (ML25084A044), and April 10, 2025 (ML25100A083). Constellation Energy Generation, LLC (Constellation or CEG) applied for license renewal (LR) of Facility Operating License No. NPF-62 for Clinton Power Station, Unit No. 1, to the U.S. Nuclear Regulatory Commission (NRC). Constellation submitted its application pursuant to Title 10 of the Code of Federal Regulations part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for LR.
The NRC staff completed its aging management audit from August 12, 2024 - February 11, 2025, in accordance with the audit plan (ML24163A325), and published the audit report (ML25090A201) on May 6, 2025. A supplement to the audit report is enclosed.
If you have any questions regarding this matter, I may be reached at 301-415-3754, or by email at Christopher.Tyree@nrc.gov.
Sincerely, Chris S. Tyree, Project Manager License Renewal Project Branch Division of Materials and License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-461
Enclosure:
Audit Report Supplement cc w/encl: Listserv August 8, 2025 Signed by Tyree, Christopher on 08/08/25
- via e-concurrence NRR-106 OFFICE NLRP/DNRL/PM NLRP/DNRL/LA BC:NVIB:DNRL NAME CTyree KBratcher ABuford DATE 08/06/2025 07/30/2025 2/6/2025 OFFICE NLRP/DNRL/BC (A)
NLRP/DNRL/PM NAME ANeuhausen CTyree DATE 08/07/2025
Enclosure LRA TLAA Section 4.2.7, Reactor Pressure Vessel Reflood Thermal Shock Analysis Summary of Information in the Application. License renewal application (LRA) Section TLAA 4.2.7, Reactor Pressure Vessel Reflood Thermal Shock Analysis, discusses the thermal shock analysis of the reactor vessel as a result of a Loss of Coolant Accident (LOCA) event. The applicant dispositioned the Time-Limited Aging Analysis (TLAA) in accordance with 10 CFR 54.21(c)(1)(ii). To verify that the applicant provided an acceptable technical basis to support its disposition of the TLAA, the staff audited the TLAA.
Audit Activities. A search of the applicants operating experience database was conducted using keywords: flaw evaluation, LOCA, reactor vessel, design cycle, neutron fluence, and reference temperature, thermal shock, and heat transfer, No significant plant-specific operating experience associated with the LRA TLAA 4.2.7 was noted by the staff during its review.
During its audit, the staff interviewed the applicants staff and reviewed documentation provided by the applicant. The staff reviewed the following relevant documents.
Relevant Documents Reviewed Document Title Revision /
Date LRA Section 4.2.7 Reactor Pressure Vessel Reflood Thermal Shock
- Analysis, Revision 0 2/14/2024 LRA Reference 4.8.35 Ranganath, S., Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Postulated Loss of Coolant Accident, Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany.
08/1979 LRA Section 4.2.3 Reactor Pressure Vessel Adjusted Reference Temperature (ART) Analysis Revision 0 2/14/2024 GE Hitachi ePortal Reference 4.9.6 Constellation Energy Generation, LLC Clinton Power Station Licensing Renewal Fluence & RPV Internal Analysis Revision 0 12/2022 During the audit of the TLAA, the staff verified that the applicant has provided its technical basis that supports its disposition of 10 CFR 54.21(c)(1)(ii).
During the audit, the staff reviewed the generic fracture mechanics analysis by the author, Ranganath, in 1979 (LRA Reference 4.8.35) that evaluates the effects of the limiting LOCA event and Constellation Energy Generation, LLC, Clinton Power Station Licensing Renewal Fluence & RPV Internal Analysis. The Ranganath paper provides the technical basis for the TLAA of the reactor vessel reflood thermal shock analysis. The Clinton Power Station licensing Renewal Fluence & RPV Internal Analysis provides information on the impact of neutron fluence on the fracture toughness of RPV materials.
The staff made the following observations:
The staff noted that the LOCA analysis in the Ranganath paper postulated a steam line break and a recirculation line break. The evaluation of heat transfer conditions and temperature gradients showed that the steam line break is more severe than the recirculation line break based on thermal stresses and brittle fracture. Therefore, the steam line break is the controlling design basis accident for the purpose of the LOCA evaluation.
The staff observed that during the entire LOCA event, the water level in the vessel stays well above the top of the active fuel zone. Therefore, the Extended Beltline region that is being evaluated is surrounded by water on the inside surface.
As shown in the Ranganaths paper, the LOCA analysis shows that at 300 seconds into the event, the applied stress intensity factor, KI-applied, reaches a peak value of approximately 100 ksiin, then slowly decreases.
The staff confirmed that by this 300-second point in the event, the vessel wall temperature at the 1/4T depth (T = the wall thickness of the reactor vessel) from the inside surface is reduced from 550ºF to approximately 400ºF. As a result, the heat transfer regime is forced two-phase convection with boiling for which the heat transfer coefficient is potentially as high as 10,000 BTU/hr-ft2 ºF.
The staff confirmed that after 300 seconds, the lower heat transfer coefficient was assumed for the remainder of the LOCA event. Therefore, the maximum applied thermal stress intensity factor, KI, was determined to be 100 ksiin at 300 seconds after the LOCA.
The staff noted that because the material fracture toughness exceeds the maximum applied stress intensity factor at that point in the event, an existing flaw in the vessel would not propagate due to brittle fracture during a LOCA. This was done by determining the temperature required to achieve a fracture toughness of 200 ksiin when using the equation for fracture toughness stress intensity for crack initiation (KIC) presented in Appendix A of the ASME Code,Section XI.
The staff confirmed that by setting KIC = 200 ksiin, and using the limiting 52 EFPY ART value for the CPS Extended Beltline, the temperature at which KIC reaches 200 ksiin, was determined to be 176.25ºF as shown in the Clinton Power Station licensing Renewal Fluence & RPV Internal Analysis.
The staff found that sufficient information was available to complete its review of TLAA Section 4.2.7.
During the audit of the operating experience associated with the TLAA, the staff independently searched the plant-specific database to identify any previously unknown or recurring aging effects. The staff did not identify any additional aging effects that would have an impact on the evaluation of the TLAA.
The staff also audited the summary description of the TLAA for the reactor vessel reflood thermal shock analysis in Section A.4.2.7 of the UFSAR supplement. The staff verified this summary description is consistent with the generic description provided in the SRP-LR.