CNL-25-046, Supplement to Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Testing Frequency
| ML25051A357 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/20/2025 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| CNL-25-046, EPID-L-2024-LLA-0156, SQN-TS-24-04 | |
| Download: ML25051A357 (1) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-25-046 February 20, 2025 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328
Subject:
Supplement to Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04) (EPID-L-2024-LLA-0156)
References:
- 1. TVA letter to NRC, CNL-24-071, Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04), dated November 26, 2024 (ML24331A178)
- 2. NRC electronic mail to TVA, Request for Additional Information - LAR to modify LAR to modify TS SR 3.4.14.1 - L-2024-LLA-0156, dated January 23, 2025 (ML25024A017)
- 3. TVA letter to NRC, CNL-25-027, Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04) (EPID-L-2024-LLA-0156), dated February 10, 2025 (ML25041A301)
In Reference 1, Tennessee Valley Authority (TVA) submitted a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2. The proposed license amendment would revise SQN Units 1 and 2 Technical Specification (TS) Surveillance Requirement (SR) 3.4.14.1, Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage to only reference the Inservice Testing Program for the Frequency.
U.S. Nuclear Regulatory Commission CNL-25-046 Page 2 February 20, 2025 In Reference 2, the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI). In Reference 3, TVA provided a response to each of the RAIs in Reference 2. The enclosure to this letter contains a revised response to STSB-LAR-RAI-1 which replaces the response provided in Reference 3 in its entirety. This revised response is required in order clarify that truncation limits were applied in the SQN Units 1 and 2 Full Power Internal Events Probabilistic Risk Assessment models. Revision bars are included in the enclosure to indicate where changes were made from the response provided in Reference 3.
This supplement does not change the remaining RAI responses and content provided in Reference 3. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.91, Notice for public comment; State consultation, a copy of this supplement is being provided to the Tennessee Department of Environment and Conservation.
There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Amber V. Aboulfaida, Senior Manager, Fleet Licensing, at avaboulfaida@tva.gov.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20th day of February 2024.
Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness
Enclosure:
Revised Response to STSB-LAR-RAI-1 from TVA letter to NRC, CNL-25-027, Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04) (EPID-L-2024-LLA-0156),
(ML25041A301) cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation Respectfully, Digitally signed by Edmondson, Carla Date: 2025.02.20 18:25:10
-05'00'
Enclosure CNL-25-046 E1 of 3 Revised Response to STSB-LAR-RAI-1 from TVA letter to NRC, CNL-25-027, Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04) (EPID-L-2024-LLA-0156), (ML25041A301)
STSB-LAR-RAI-1 As described in the LAR, the licensee proposes to eliminate fixed and event driven pressure isolation valve (PIV) test frequencies from the reactor coolant system (RCS) PIV Leakage SR 3.4.14.1. However, it is not clear to the NRC staff that the licensee verified that there are no licensing basis requirements based on fixed or event driven test frequencies. For example, the Sequoyah TS Bases B 3.4.14, RCS PIV Leakage, References section, contains two references, reference 4: WASH-1400 (NUREG-75/014), Appendix V, October 1975; and reference 5: NUREG-0677, May 1980.The Sequoyah TS Bases state that WASH-1400 identified potential intersystem loss-of-coolant accidents (ISLOCAs) as a significant contributor to plant risk (core melt). NUREG-0677 indicates that testing needs to be done on an event-based frequency and/or more frequent time-based frequency to assure plant risk due to ISLOCA is acceptable. The NRC staff requests the licensee review these references and explain why the proposed frequencies (which eliminate fixed and event driven RCS PIV test frequencies from the TS) are adequate to assure TS are derived from the analyses and evaluation included in the Bases to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. As part of the response the NRC staff requests that the licensee review the Sequoyah licensing basis to ensure that the in-service testing (IST) Program frequencies adequately cover all licensing basis requirements. Additionally, if necessary, the bases should be revised to align with the proposed changes. Furthermore, explain the impact on plant risk of eliminating the fixed and event driven testing frequencies.
TVA Response The basis for the Sequoyah Nuclear Plant (SQN) technical specification (TS) limiting conditions for operation (LCO) 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage is the 1975 NRC Reactor Safety Study (Reference 1 of this RAI response) that identified potential interfacing system loss of coolant accidents (ISLOCAs) as a significant contributor to the risk of core melt. A subsequent study (Reference 2 of this RAI response) evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic (once per year) leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
NUREG-0677 is a generic evaluation of various combinations of valves that are relied on in nuclear plants to prevent exposure of full RCS pressure to piping designed for lower pressures. This evaluation did not consider the adequacy of the lower pressure piping for withstanding the full power RCS in the event of a failure of the PIVs or whether the lower pressure piping was within containment. It was assumed in NUREG-0677 that an ISLOCA would occur if the series of valves isolating the RCS from the lower pressure piping were to fail.
Enclosure CNL-25-046 E2 of 3 The state-of-the-art of risk assessment has progressed substantially since the "Reactor Safety Study" in 1975 and NUREG-0677 in 1980. TVA has performed a detailed analysis and developed logic models for ISLOCA events as part of the SQN Full Power Internal Events Probabilistic Risk Assessment (FPIE PRA). This analysis uses NSAC-154 (Reference 3 of this RAI response), ISLOCA Evaluation Guidelines, and WCAP-17154-P (Reference 4 of this RAI response), ISLOCA Risk Model, as guidance for developing the ISLOCA event trees, success criteria, failure probabilities, and fault trees. NSAC-154 is the industry recognized standard for ISLOCA analysis and it provides detailed instructions on the methodology and documentation of this analysis. Following NSAC-154 ensures that the ISLOCA analysis is done consistent with industry practice. The objective of WCAP-17154-P is to develop guidance on modeling the risk contribution from ISLOCAs during power operation. WCAP-17154-P addresses the full scope of ISLOCA modeling, including the development of its initiating event frequency, assessing its recovery, estimating the rupture probability of the low pressure interfacing system, and mitigation of its consequences.
The SQN FPIE PRA Model (including internal flooding) was peer reviewed against the requirements of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard (Reference 5 of this RAI response) and any Clarifications and Qualifications provided in the NRC endorsement of the Standard contained in Revision 2 to Regulatory Guide (RG) 1.200 (Reference 6 of this RAI response). A peer review and independent assessment was performed for the SQN Units 1 and 2 Internal Events PRA and are described in Reference 7 of this RAI response.
No ISLOCA Core Damage Frequency (CDF) or Large Early Release Frequency (LERF) cutsets (above the truncation limit of 1E-12/per reactor year for CDF and 1E-13/per reactor year for LERF) are included in the results of the FPIE PRA for SQN Unit 1 or Unit 2. Based on the results of the SQN PRA, the risk of an ISLOCA is considered a negligible contributor to the risk of core melt and large early release. Within the ISLOCA analysis performed for the SQN FPIE, the frequency of the RCS PIV testing credited was once every 18 months, which is equivalent to the IST frequency. As a result, the yearly periodic leakage testing of the PIVs recommended in NUREG-0677 is no longer required to reduce the probability of an intersystem LOCA to an acceptable level. The references to WASH-1400 and NUREG-0677 are removed from the bases to the SQN Technical Specifications as provided in Enclosure 2 to Reference 8 of this RAI response.
TVA has reviewed the SQN IST Program frequencies as related to the license amendment request to verify conformance with licensing basis requirements with the changes being proposed.
References for STSB-LAR-RAI-1
- 1. WASH-1400 (NUREG-75/014), Appendix V, October 1975
- 2. NUREG-0677, May 1980
- 3. NSAC-154, ISLOCA Evaluation Guidelines, September 1991
- 4. WCAP-17154-P, Revision 0, ISLOCA Risk Model, April 2010
- 5. ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear
Enclosure CNL-25-046 E3 of 3 Power Plant Applications, American Society of Mechanical Engineers, New York, February 2009
- 6. Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, US NRC, March 2009
- 7. TVA Letter CNL-21-026, License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times---RITSTF Initiative 4b, (SQN-TS-20-03),
dated August 5, 2021 (ML21217A174)
- 8. TVA Letter CNL-25-027, Response to Request for Additional Information Regarding Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04) (EPID-L-2024-LLA-0156), dated February 10, 2025 (ML25041A301)