ML25024A017
| ML25024A017 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/23/2025 |
| From: | Perry Buckberg NRC/NRR/DORL/LPL2-2 |
| To: | Baron J Tennessee Valley Authority |
| Buckberg P | |
| References | |
| L-2024-LLA-0156 | |
| Download: ML25024A017 (5) | |
Text
From:
Perry Buckberg Sent:
Thursday, January 23, 2025 3:00 PM To:
Baron, Jesse Shawn Cc:
Wells, Russell Douglas
Subject:
Request for Additional Information - LAR to modify LAR to modify TS SR 3.4.14.1 - L-2024-LLA-0156 Attachments:
RAI - SQN LAR - TS SR 3.4.14.1 - EMIB-STSB L-2024-LLA-0156 1-23-2025.pdf
- Jesse, By letter dated November 26, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24331A178), the Tennessee Valley Authority (TVA) submitted a license amendment request to the U.S. Nuclear Regulatory Commission (NRC). The request is to revise Sequoyah Units 1 and 2 Technical Specification (TS) 3.4.14, Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage, Surveillance Requirement (SR) 3.4.14.1, to only reference the Inservice Testing Program (IST) Program for the Frequency.
The U.S. Nuclear Regulatory Commission staff is reviewing the request and has identified an area where additional information is needed to complete its review. By email dated January 21, 2025, I transmitted a draft request for additional information (RAI) to you. TVA requested a clarification call to discuss the draft RAI and that call was held on January 23, 2025. During the clarification call, each RAI question was discussed and clarified as needed and no changes to the text were deemed necessary. Attached is the final RAI.
Dring the clarification call, you stated that TVA plans to respond to the attached final RAI on for before January 31, 2025. The NRC staff considers that timely responses to RAIs help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the agreed response date, please contact me.
Perry Buckberg Senior Project Manager / Agency 2.206 Petition Coordinator U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation office: (301)415-1383 perry.buckberg@nrc.gov Mail Stop O-8B1a, Washington, DC, 20555-0001
Hearing Identifier:
NRR_DRMA Email Number:
2706 Mail Envelope Properties (SA1PR09MB767912A90261D011B24EE7679AE02)
Subject:
Request for Additional Information - LAR to modify LAR to modify TS SR 3.4.14.1
- L-2024-LLA-0156 Sent Date:
1/23/2025 3:00:28 PM Received Date:
1/23/2025 3:00:00 PM From:
Perry Buckberg Created By:
Perry.Buckberg@nrc.gov Recipients:
"Wells, Russell Douglas" <rdwells0@tva.gov>
Tracking Status: None "Baron, Jesse Shawn" <jsbaron@tva.gov>
Tracking Status: None Post Office:
SA1PR09MB7679.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1803 1/23/2025 3:00:00 PM RAI - SQN LAR - TS SR 3.4.14.1 - EMIB-STSB L-2024-LLA-0156 1-23-2025.pdf 144691 Options Priority:
Normal Return Notification:
No Reply Requested:
No Sensitivity:
Normal Expiration Date:
1 REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION L-2024-LLA-0156 - SEQUOYAH - EXPEDITED REVISION OF TS SR 3.4.14.1 - RCS PIV LEAKAGE TESTING TO IST TENNESSEE VALLEY AUTHORITY SEQUOYAH, UNITS 1, 2 DOCKET NO. 05000327, 05000328 ISSUE DATE: 1/23/2025
=
Background===
By letter dated November 26, 2024 (ML24331A178), Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for changes to the Technical Specifications (TS), for the Sequoyah Nuclear Plant (SQN), Units 1 and 2. The proposed license amendment would revise SQN Units 1 and 2 Technical Specification (TS) 3.4.14, Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage, Surveillance Requirement (SR) 3.4.14.1, to only reference the Inservice Testing Program (IST) Program for the Frequency.
Regulatory Basis The NRC regulation at 10 CFR 50.36(c)(3) requires that:
TSs include Surveillance Requirements (SRs). Per 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection 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.
The Commissions Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132, dated July 22, 1993) explained that [e]ach LCO
[limiting condition for operation], Action, and Surveillance Requirement should have supporting Bases. In addition, the Bases in part, for each SR should address the following question, [w]hy is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?
The NRC regulation at 10 CFR 50.54(jj) states:
Structures, systems, and components subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.
The licensee states that these requirements of 10 CFR 50.54(jj) will continue to be met.
The NRC regulations at 10 CFR 50.55a(f) state in part that:
Systems and components of pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel (BPV) Code and ASME OM Code as incorporated by reference into 10 CFR 50.55a.
The licensee states that these requirements of 10 CFR 50.55a will continue to be met.
The NRC regulations in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) establish the necessary design, fabrication, construction, testing, and performance requirements for
2 structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. SQN Units 1 and 2 were designed to meet the intent of the Proposed General Design Criteria (GDC) for Nuclear Power Plant Construction Permits published in July 1967.
The applicable criteria are listed below:
Criterion 1 - Quality standards and records.
Criterion 14 - Reactor coolant pressure boundary Criterion 30 - Quality of reactor coolant pressure boundary Criterion 32 - Inspection of reactor coolant pressure boundary.
Criterion 54 - Piping systems penetrating containment.
Criterion 55 - Reactor coolant pressure boundary penetrating containment.
With the implementation of the proposed change, the licensee states that SQN Units 1 and 2 will continue to meet the identified applicable GDC, regulations, and requirements.
References:
- 1. Sequoyah Nuclear Plants (SQN) - Units 1 and 2, Fourth 10-Year Interval lnservice Testing Program Update, dated January 25, 2017 (ML17026A091).
- 2. Sequoyah Units 1 and 2 Pre-submittal Meeting Summary dated November 26, 2024 (ML24317A001).
- 3. NUREG-1431, Revision 5, Standard Technical Specifications Westinghouse Plants:
Specifications (ML21259A155).
- 4. NUREG-0667, May 1980, The Probability of Intersystem LOCA: impact Due to Leak Testing and Operational Changes (ML19323E667).
Requests for Additional Information EMIB-LAR-RAI-1 The licensee is requested to discuss whether this LAR impacts the authorized SQN Units 1 and 2 Alternative Request RV-02 (ML22304A186) to use mechanical agitation to minimize leakage of pressure isolation valves (PIVs).
EMIB-LAR-RAI-2 LAR Attachment 4, Sequoyah Units 1 and 2 SR 3.4.14.1 Leakage Test History, provides As-Left Measured Value Leakage that in some cases (such as for Unit 1 valves63-561 and 63-562 and Unit 2 valve 63-563) indicates that the next regularly scheduled test will exceed the leakage acceptance criteria. Please discuss the support for this LAR in light of the valve leakage data and the plans to address the leakage data for these valves.
EMIB-LAR-RAI-3 ASME OM Code Case OMN-23, Alternative Rules for Testing Pressure Isolation Valves [PIVs],
allows PIV leakage rate testing to be extended up to 6 years as accepted in NRC Regulatory Guide 1.192 as incorporated by reference in 10 CFR 50.55a. The licensee is requested to describe the impact of its use of Code Case OMN-23 on the changes proposed in the LAR.
3 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 either (a) remove these references or (b) 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.
STSB-LAR-RAI-2 As described in the LAR, the licensee proposes to delete the following references from SR 3.4.14.1 Supporting Bases document.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
It is not clear why these references are being deleted. The NRC staff requests the licensee to explain why it is adequate to remove these references. These references are included in the STS Bases in NUREG-1431. Additionally, the NRC staff believes the reference 10 CFR 50.55a(g) should be 10 CFR 50.55a(f) which is the section for Inservice Testing (IST) for valves and pumps.