Text

Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System Pressure Isolation Valve Leakage Testing Frequency for the (SQN-TS-24-04)
	ML24331A178
Person / Time
Site:	Sequoyah
Issue date:	11/26/2024
From:	Hulvey K; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	CNL-24-071, SQN-TS-24-04
	Download: ML24331A178 (1)
	v • d • e

1101 Market Street, Chattanooga, Tennessee 37402 CNL-24-071 November 26, 2024 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:

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)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting 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 (IST) Program for the Frequency.

The enclosure provides a description and assessment of the proposed changes. provides the existing TS pages marked to show the proposed changes. provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only. Attachment 3 provides a list of PIVs subject to testing under SR 3.4.14.1. Attachment 4 provides the surveillance test results history for the SQN Units 1 and 2 RCS PIVs.

TVA is requesting an expedited review of this license amendment request as a result of the unanticipated need for a significant repair of the SQN Unit 2 main turbine generator. SQN Unit 2 was originally scheduled to commence the Unit 2 Cycle 26 refueling outage (U2R26) on September 27, 2024. However, recovery from a unit trip that occurred on July 30, 2024, requires a significant unplanned rebuild to the main turbine generator. The projected forced outage duration resulted in a TVA decision to start the U2R26 on August 5, 2024. The U2R26 refueling has been completed and the turbine generator repair efforts are ongoing.

U.S. Nuclear Regulatory Commission CNL-24-071 Page 2 November 26, 2024 The planned evolutions of SQN Unit 2 during the extended forced outage include moving the plant to MODE 2 in order to perform surveillances in January of 2025. Currently, SR 3.4.14.1 requires performance prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. TVA plans to perform SR 3.4.14.1 during the plant evolutions in January of 2025 because SQN Unit 2 will meet this criteria for performance.

Following the planned evolutions, SQN Unit 2 will be returned back to lower MODES where it is planned to be held until the turbine generator is repaired and SQN Unit 2 can return to power operation. The current SR 3.4.14.1 requires at least a partial performance again during ascension to MODE 2 for those systems experiencing flow through the RCS PIVs during the change in MODES. A repetitive performance of SR 3.4.14.1 would be avoided with the TS changes being proposed which will lower current and future outage personnel dose, improve personnel safety, and reduce outage time. SQN Unit 1 is being included in this license amendment request in order to maintain consistent TS requirements for both SQN Units 1 and 2.

The turbine generator repair efforts are targeted for completion by April 1, 2025, and TVA plant operating procedures require revision in order to implement the license amendment request. Accordingly, TVA requests an expedited approval of the proposed license amendment request by February 28, 2025, with implementation within 30 days of issuance of the amendment. TVA will advise the NRC through the Nuclear Reactor Regulation (NRR) project manager of any significant changes to this planned schedule.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosure to the Tennessee State Department of Environment and Conservation.

There are no new regulatory commitments contained in this letter. 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 this 26th day of November 2024.

Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness Enclosure cc: See Page 3 Digitally signed by Edmondson, Carla Date: 2024.11.26 13:39:54

-05'00'

U.S. Nuclear Regulatory Commission CNL-24-071 Page 3 November 26, 2024

Enclosure:

Description and Assessment of the Proposed Change 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 Department of Environment and Conservation

CNL-24-071 Enclosure Description and Assessment of the Proposed Change

Subject:

Expedited Application to Revise Technical Specifications to Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN-TS-24-04)

CONTENTS 1.0

SUMMARY

DESCRIPTION............................................................................................... 1 2.0 DETAILED DESCRIPTION................................................................................................ 1 2.1 Background................................................................................................................... 1 2.2 System Design and Operation...................................................................................... 2 2.3 Current Technical Specification Requirements............................................................. 2 2.4 Reason for the Proposed Change................................................................................. 3 2.5 Description of the Proposed Change............................................................................ 3

3.0 TECHNICAL EVALUATION

............................................................................................... 3

4.0 REGULATORY EVALUATION

........................................................................................... 6 4.1 Applicable Regulatory Requirements and Criteria........................................................ 6 4.2 Precedent...................................................................................................................... 8 4.3 No Significant Hazards Consideration Determination Analysis..................................... 9 4.4 Conclusion................................................................................................................... 10 4.5 References.................................................................................................................. 10

5.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 11 Attachments

1. Proposed Technical Specification Changes (Markup) for SQN Units 1 and 2

2. Proposed Technical Specification Bases Changes(markup) (For Information Only)

3. SQN Units 1 and 2 Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1

4. SQN Units 1 and 2 Pressure Isolation Valve Leakage Test History

Enclosure CNL-24-071 E1 of 11 Description and Assessment 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting 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 (IST) Program for the Frequency. The IST program complies with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code).

provides the existing TS pages marked to show the proposed changes. provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only. Attachment 3 provides a list of PIVs subject to testing under SR 3.4.14.1. Attachment 4 provides the surveillance test results history for the SQN Units 1 and 2 PIVs. Further discussion on these attachments is provided in Section 2.2 and Section 3.0 of this enclosure.

2.0 DETAILED DESCRIPTION

2.1 Background

TVA is requesting an expedited review of this license amendment request as a result of the unanticipated need for a significant repair of the SQN Unit 2 main turbine generator. SQN Unit 2 was originally scheduled to commence the Unit 2 Cycle 26 refueling outage (U2R26) on September 27, 2024. However, recovery from a unit trip that occurred on July 30, 2024, requires a significant unplanned rebuild to the main turbine generator. The projected forced outage duration resulted in a TVA decision to start the U2R26 on August 5, 2024. The U2R26 refueling has been completed and the turbine generator repair efforts are targeted for completion by April 1, 2025.

The planned evolutions of SQN Unit 2 during the extended forced outage include moving the plant to MODE 2 in order to perform surveillances in January of 2025. Currently, SR 3.4.14.1 requires performance prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. TVA plans to perform SR 3.4.14.1 during the plant evolutions in January of 2025 since SQN Unit 2 will meet this criteria for performance.

Following the planned evolutions, the plant will be returned back to lower MODES where it is planned to be held until the turbine generator is repaired and SQN Unit 2 can return to power operation currently targeted for April 1, 2025. The current SR 3.4.14.1 requires at least a partial performance again during ascension to MODE 2 for those systems experiencing flow through the RCS PIVs during the change in MODES. A repetitive performance of SR 3.4.14.1 would be avoided with the TS changes being proposed, which will lower current and future outage personnel dose, improve personnel safety, and reduce outage time. Performance of SR 3.4.14.1 testing can pose personnel safety risks due to the required system configurations and methods of RCS leakage measurement.

Enclosure CNL-24-071 E2 of 11 2.2 System Design and Operation The RCS PIVs are any two normally closed valves in series which separate the high pressure RCS from an attached low pressure system, such as the residual heat removal (RHR) system, the safety injection (SI) system and the accumulators. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV Leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

Although SR 3.4.14 provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed accident, that could degrade the ability for low pressure injection.

The SQN PIVs are required to be tested for leakage by both the TS and the ASME OM Code.

The ASME OM Code 2004 Edition with Addenda through 2006 is the applicable Code for the current fourth 10-year SQN Units 1 and 2 IST program interval ending June 30, 2026. The ASME OM Code requirement ISTC-3630 applies to seat leakage testing of PIVs subject to SR 3.4.14.1 and requires tests to be conducted at least once every two years. A list of the SQN Unit 1 and 2 PIVs subject to testing under SR 3.4.14.1 are listed in the TS Bases Table B 3.4.14-1 and are provided in Attachment 3.

2.3 Current Technical Specification Requirements The current TS SR 3.4.14.1 for both SQN Units 1 and 2 states:

Notes

1. Not required to be performed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive test loop cannot be avoided.

4. Not required to be performed for RCS PIVs FCV-74-1 and FCV-74-2 following manual or automatic actuation or flow through the valves.

Surveillance Verify leakage from each RCS PIV is equivalent to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2215 psig and 2255 psig.

Frequency In accordance with the Inservice Testing Program, and in accordance with the Surveillance Frequency Control Program.

AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

Enclosure CNL-24-071 E3 of 11 AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

2.4 Reason for the Proposed Change The proposed change will align the SQN Units 1 and 2 TS PIV surveillance testing Frequency with the ASME OM Code. The SQN SR 3.4.14.1 requires testing in accordance with the IST Program. The ASME OM Code requires RCS PIV leakage testing at least every 2 years.

However, SR 3.4.14.1 requires more frequent RCS PIV leakage testing than is required by the ASME OM Code due to additional performance Frequency requirements as listed in the following:

In accordance with the surveillance frequency control program (SFCP), which is a Frequency of every 18 months for SQN Units 1 and 2.

Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

These additional performance Frequency requirements restrict the ability of the IST Program to govern the Frequency in which the RCS PIV leakage testing is performed. The testing history of the SQN PIVs provided in Attachment 4 demonstrates that more frequent testing is not necessary to ensure the PIVs can perform their specified safety functions. Eliminating the additional frequencies would eliminate unnecessary testing, which will reduce occupational radiation exposure, improve personnel safety, and also reduce the length of refueling outages.

This proposed change is requested to be reviewed by the NRC on an expedited basis as discussed in Section 2.1 of this Enclosure.

2.5 Description of the Proposed Change The proposed change will revise the SR 3.4.14.1 performance Frequency for SQN Units 1 and 2 to require performance in accordance with the Inservice Testing Program and remove all other Frequency requirements for SR 3.4.14.1. Notes 2, 3, and 4 of SR 3.4.14.1 are deleted since they are associated with the Frequency requirements that are being removed. The proposed change does not add or remove any RCS PIVs from the TS or ASME OM Code requirements nor does it alter the SR acceptance criteria. provides a marked-up version of the affected TS pages of SQN Units 1 and 2 showing the proposed changes. Attachment 2 provides a marked-up version of the SQN Units 1 and 2 TS Bases. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

3.0 TECHNICAL EVALUATION

The proposed change replaces RCS PIV testing Frequency with a reference to the IST Program. This results in the following changes:

Eliminates the requirement to periodically perform the SR in accordance with the SFCP in addition to the frequency specified in the ASME OM Code.

Enclosure CNL-24-071 E4 of 11 Elimination of a Frequency of 9 months prior to entering MODE 2 if the unit has been in MODE 5 for 7 days or more.

Elimination of a Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve, as well as an SR Note associated with the Frequency.

Elimination of SR Notes that exempt performance for valves located in the RHR flow path when in the shutdown cooling mode, exempt repeat testing of RCS PIVs actuated if a repetitive testing loop cannot be avoided, and exempt testing of specific RCS PIVs in the RHR system.

Each of these changes are evaluated below.

Elimination of the Frequency in accordance with the SFCP The proposed change would remove a surveillance frequency requirement for performing SR 3.4.14.1. The current SR relies on the IST Program and the SFCP to establish the associated frequency for these surveillances. The ASME OM Code establishes a testing frequency for PIV testing of every 2 years. Under the SFCP, changes to the performance frequency requires qualitative considerations to be addressed, such as test intervals specified in applicable industry codes and standards. Because an RCS PIV testing frequency is specified in the ASME OM Code, the frequency cannot be extended beyond every 2 years under the SFCP without prior NRC approval. As a result, the appropriate reference for the SR 3.4.14.1 performance frequency is the IST Program and the ASME OM Code.

Further, the NRC regulations give precedence to the ASME Code when there are conflicts between the TS and the Code. It is stated in 10 CFR 50.55a(f)(5)(ii), "IST program update:

Conflicting IST Code requirements with technical specifications," that "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

The leakage rate testing performance history of each PIV at SQN Units 1 and 2 is shown in. The PIV testing performance history shows that the valves have consistently met the associated leakage criteria when tested. This provides confidence that performing SR 3.4.14.1 at frequencies in accordance with the IST program and eliminating the frequency in accordance with the SFCP from the TS will not degrade the ability of the PIVs to perform their safety function.

The proposed change is acceptable because the performance requirement in accordance with the SFCP is inconsistent with the ASME OM Code requirements, could unnecessarily limit plant operation, and is supported by the RCS PIV testing history. In addition, the proposed change is consistent with NRC regulations.

Testing Every 9 Months The SR 3.4.14.1 Frequency requires the SR to be performed, "Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months." This has the effect of requiring performance of the SR every 9 months, but only if there is an outage of sufficient length to perform the test.

Enclosure CNL-24-071 E5 of 11 Elimination of the more frequent performance of the SR is acceptable because the ASME OM Code, which provides all other aspects of the testing, specifies a longer testing frequency, and the more frequent testing is not warranted given the SQN Units 1 and 2 PIV surveillance performance history provided in Attachment 4. Therefore, leakage testing of the RCS PIVs at the frequency established by the IST Program is satisfactory in demonstrating pressure isolation functional capability and operational readiness.

Testing Within 24 Hours of Valve Actuation The SR 3.4.14.1 Frequency requires the SR to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve. The purpose of the performance is to verify that the PIVs are closed or seated after being actuated. This test is redundant to other ASME OM Code testing (e.g., exercise testing per ASME OM Code Subsection ISTC-3520 and position indication verification per ASME OM Code Subsection ISTC-3700) and is unnecessary.

In addition, there are other readily available indications that a PIV has failed to close or seat, such as low-pressure system level, temperature, or pressure indications or the lifting of relief valves. Performing the PIV leakage testing for this purpose is unnecessary and will result in higher occupational dose exposure.

Elimination of SR Notes that Exempt Performance SQN Units 1 and 2 SR 3.4.14.1, Note 2 states, "Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation." The Applicability of TS 3.4.14 states:

APPLICABILITY:

MODES 1, 2, and 3, MODE 4, except valves in the residual heat removal (RHR) flow path when in, or during the transition to or from, the RHR mode of operation.

The Applicability already exempts the RCS PIVs associated with the RHR decay heat removal flow path from testing in MODE 4. Further, SR 3.4.14.1, Note 1, states that the testing is not required to be performed in MODES 3 and 4. The RHR decay heat removal system is only used in MODE 4 of the Applicability; therefore, Note 2 is not needed and is eliminated.

SQN Units 1 and 2 SR 3.4.14.1, Note 3 states, "RCS PIVs actuated during performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided." This Note is discussed in the SQN Units 1 and 2 Bases for the 24-hour testing requirement as, "In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided." Therefore, this Note is no longer needed with the elimination of the 24-hour testing Frequency.

SQN Units 1 and 2 SR 3.4.14.1, Note 4 states, "Not required to be performed for RCS PIVs FCV-74-1 and FCV-74-2 following manual or automatic actuation or flow through the valves." This Note is specific to two flow control valves within the RHR system. With the elimination of the 24-hour testing Frequency requirement this Note is no longer needed.

TVA confirms that the RCS PIVs required to be tested by the SQN Units 1 and 2 TS are included in the SQN IST Program. Additionally, SQN confirms that the RCS PIV leakage

Enclosure CNL-24-071 E6 of 11 testing Frequencies proposed to be removed from the SQN Units 1 and 2 TS are not credited for satisfying any other requirements described in the Updated Final Safety Analysis Report or any commitments for reasons other than being a TS requirement.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria 10 CFR 50.54(jj)

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 proposed change does not alter the design, fabrication, erection, construction, or inspection of any components. The proposed change will continue to test the PIVs to the quality standards commensurate with the importance of the safety function being performed. Therefore, the requirements of 10 CFR 50.54(jj) continue to be met.

10 CFR 50.55a In accordance with 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors pursuant to the ASME OM Code as specified in 10 CFR 50.55a(f). Paragraph (f) of 10 CFR 50.55a states 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 proposed change will not affect the current TS requirement to test the PIVs in accordance with the ASME OM Code and, therefore, the requirements of 10 CFR 50.55a continue to be met.

10 CFR Part 50, Appendix A, General Design Criteria (GDC)

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 SQN construction permit was issued in May 1970. The Updated Final Safety Analysis Report (UFSAR), however, addresses the Nuclear Regulatory Commission (NRC) GDC published as Appendix A to 10 CFR 50 in July 1971.

Each applicable criterion listed below is followed by a discussion of the design features and procedures of SQN Units 1 and 2 that meet the intent of the criteria.

Criterion 1 - Quality standards and records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A Quality Assurance Program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

Enclosure CNL-24-071 E7 of 11 Compliance The proposed change does not affect the ability of the Quality Assurance Program to provide adequate assurance that the SQN Units 1 and 2 structures, systems, and components will perform their safety functions. Therefore, the recommendations of GDC 1 continue to be met with the proposed change. Further discussion of compliance with GDC 1 is provided in SQN UFSAR Section 3.1.2.

Criterion 14 - Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Compliance The proposed change does not affect the ability of the reactor coolant pressure boundary to perform its safety function. Therefore, the recommendations of GDC 14 continue to be met with the proposed change. Further discussion of compliance with GDC 14 is provided in SQN UFSAR Section 3.1.2.

Criterion 30 - Quality of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Compliance The proposed change does not affect the quality standards applied to the reactor coolant pressure boundary or the means for identifying reactor coolant leakage. Therefore, the recommendations of GDC 30 continue to be met with the proposed change. Further discussion of compliance with GDC 30 is provided in SQN UFSAR Section 3.1.2.

Criterion 32 - Inspection of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Compliance The proposed change does not affect the design of the reactor coolant pressure boundary to permit inspection and testing or the material surveillance program for the reactor pressure vessel. Therefore, the recommendations of GDC 32 continue to be met with the proposed change. Further discussion of compliance with GDC 32 is provided in SQN UFSAR Section 3.1.2.

Criterion 54 - Piping systems penetrating containment. Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

Compliance The proposed change does not affect the piping systems penetrating primary reactor containment. It also does not affect the capability to test periodically the operability of the

Enclosure CNL-24-071 E8 of 11 isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. Therefore, the recommendations of GDC 54 continue to be met with the proposed change. Further discussion of compliance with GDC 54 is provided in SQN UFSAR Section 3.1.2.

Criterion 55 - Reactor coolant pressure boundary penetrating containment. Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

Compliance The proposed change does not affect the safety function or features of the containment isolation valves for the reactor coolant pressure boundary. Therefore, the recommendations of GDC 55 continue to be met with the proposed change. Further discussion of compliance with GDC 55 is provided in SQN UFSAR Section 3.1.2.

The proposed change does not alter the design of any plant components. Compliance with the Appendix A GDCs, as described in the SQN Updated Final Safety Analysis Report, is not affected by the proposed change. Therefore, with the implementation of the proposed change, SQN Units 1 and 2 continue to meet the identified applicable GDC, regulations and requirements.

4.2 Precedent On February 1, 2023, Duke Energy submitted a license amendment request (Reference 1) for Catawba Nuclear Station, McGuire Nuclear Station, Oconee Nuclear Station, H. B. Robinson Steam Electric Plant, and Shearon Harris Nuclear Power Plant, that proposed to revise the SR Frequency for RCS PIV leakage testing in the same manner that is proposed by this license amendment request using a similar justification. A supplement was submitted on July 7, 2023 (Reference 2), which confirmed that all RCS PIVs required to be tested by the TS are

Enclosure CNL-24-071 E9 of 11 included in the respective site's IST Program. The license amendment was approved on October 24, 2023 (Reference 3).

On May 1, 2023, Southern Nuclear Company submitted a license amendment request (Reference 4) for Vogtle Units 1 and 2 that proposed a revision to the SR frequency for RCS PIV leakage testing in the same manner that is proposed by this license amendment request using a similar justification. The Vogtle license amendment request included an IST alternative request to test the RCS PIVs on a performance based frequency. The license amendment was approved on April 10, 2024 (Reference 5). The NRC's technical evaluation of the Vogtle Units 1 and 2 change is also applicable to this license amendment request. The Vogtle IST alternative request was also approved on April 10, 2024 (Reference 6). SQN is not submitting an IST alternative request as part of this license amendment request.

4.3 No Significant Hazards Consideration Determination Analysis In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SQN),

Units 1 and 2. The proposed change revises the Surveillance Requirement (SR) to perform reactor coolant system (RCS) pressure isolation valve (PIV) leakage testing to only reference the Inservice Testing Program (IST Program) for the Frequency.

SQN has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). RCS PIV testing is performed during a plant shutdown and is not an initiator to any accident previously evaluated.

The RCS PIV operational testing acceptance criteria are not affected by the proposed change. The RCS PIVs will continue to be tested to ensure leakage is within the technical specification allowable leakage limits. As a result, the consequences of any accident previously evaluated are unchanged.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the ASME OM Code.

Enclosure CNL-24-071 E10 of 11 RCS PIV operational testing is only performed during a plant shutdown. The testing methodology and acceptance criteria remain unchanged. The proposed change does not involve a physical change to the plant or the manner in which the plant is operated or controlled.

The proposed change does not alter the design function or operation of the RCS PIVs. The proposed change does not alter the ability of the RCS PIVs to perform their design function. Since pressure boundary leakage is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the ASME OM Code.

The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of allowed RCS PIV leakage is not increased. The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, SQN concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.5 References

1. Duke Energy letter to the NRC, "License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies," dated February 1, 2023 (ML23032A162).

2. Duke Energy letter to NRC, "Response to Request for Additional Information Regarding the License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies,"

dated July 7, 2023 (ML23188A176).

Enclosure CNL-24-071 E11 of 11

3. NRC letter to Duke Energy, "Catawba Nuclear Station, Unit Nos. 1 and 2; Shearon Harris Nuclear Power Plant, Unit No. 1; McGuire Nuclear Station, Unit Nos. 1 and 2; Oconee Nuclear Station, Unit Nos. 1, 2, and 3; and H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendments to Revise Restrictive Technical Specification Surveillance Requirement Frequencies," dated October 24, 2023 (ML23241A987).

4. Southern Nuclear Company letter to the NRC, Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02, dated May 1, 2023 (ML23121A267).

5. NRC letter to Southern Nuclear Company, Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 224 and 207, Regarding Revision to Technical Specification 3.4.14 - Reactor Coolant System Pressure Isolation Valve Leakage, Surveillance Requirement 3.4.14.1, dated April 10, 2024 (ML24030A909).

6. NRC letter to Southern Nuclear Company, Vogtle Electric Generating Plant, Units 1 and 2 - Proposed Alternative to the Requirements of American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants for Pressure Isolation Valve Testing Frequency, dated April 10, 2024 (ML24079A006).

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure CNL-24-071 Proposed Technical Specification Changes (Mark-Up) for SQN Units 1 and 2

RCS PIV Leakage 3.4.14 SEQUOYAH - UNIT 1 3.4.14-3 Amendment 334, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1

NOTES-----------------------------

1. Not required to be performed in MODES 3 and

4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

4. Not required to be performed for RCS PIVs FCV-74-1 and FCV-74-2 following manual or automatic actuation or flow through the valves.

Verify leakage from each RCS PIV is equivalent to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2215 psig and 2255 psig.

In accordance with the Inservice Testing Program, and In accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND

RCS PIV Leakage 3.4.14 SEQUOYAH - UNIT 1 3.4.14-4 Amendment 334, SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve

RCS PIV Leakage 3.4.14 SEQUOYAH - UNIT 2 3.4.14-3 Amendment 327, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1

NOTES-----------------------------

1. Not required to be performed in MODES 3 and

4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

4. Not required to be performed for RCS PIVs FCV-74-1 and FCV-74-2 following manual or automatic actuation or flow through the valves.

Verify leakage from each RCS PIV is equivalent to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2215 psig and 2255 psig.

In accordance with the Inservice Testing Program, and In accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND

RCS PIV Leakage 3.4.14 SEQUOYAH - UNIT 2 3.4.14-4 Amendment 327, SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve

Enclosure CNL-24-071 Proposed Technical Specification Bases Changes (Mark-Up) for SQN Units 1 and 2 (For Information Only)

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 1 B 3.4.14-4 Revision 45, BASES ACTIONS (continued)

B.1 and B.2 If Required Actions and associated Completion Times of Condition A are not met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 9 months, but may be extended, if the plant does not go into MODE 5 for at least 7 days. The Frequency is consistent with 10 CFR 50.55.a(g) (Ref. 7) as containedin accordance with in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref.6), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential for an unplanned transient if the surveillance were performed with the reactor at power.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 1 B 3.4.14-5 Revision 45, BASES SURVEILLANCE REQUIREMENTS (continued) been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complementary to the Frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

REFERENCES

1.

10 CFR 50.2.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME Code for Operation and Maintenance of Nuclear Power Plants.

7.

10 CFR 50.55a(g).

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 2 B 3.4.14-4 Revision 45, BASES ACTIONS (continued)

B.1 and B.2 If Required Actions and associated Completion Times of Condition A are not met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 9 months, but may be extended, if the plant does not go into MODE 5 for at least 7 days. The Frequency is consistent with 10 CFR 50.55.a(g) (Ref. 7) as containedin accordance with in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref.6), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential for an unplanned transient if the surveillance were performed with the reactor at power.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 2 B 3.4.14-5 Revision 45, BASES SURVEILLANCE REQUIREMENTS (continued) been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complementary to the Frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

REFERENCES

1.

10 CFR 50.2.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, Section V, GDC 55.

4.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5.

NUREG-0677, May 1980.

6.

ASME Code for Operation and Maintenance of Nuclear Power Plants.

7.

10 CFR 50.55a(g).

Enclosure CNL-24-071 SQN Units 1 and 2 Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1 CNL-24-071 A3-1 of 1 RCS Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1 for SQN Units 1 and 2 Valve Number1 Function System2 ASME Code Class OM Code Category3 63-560 Accumulator Discharge SI 1

A/C 63-561 Accumulator Discharge SI 1

A/C 63-562 Accumulator Discharge SI 1

A/C 63-563 Accumulator Discharge SI 1

A/C 63-622 Accumulator Discharge SI 1

A/C 63-623 Accumulator Discharge SI 1

A/C 63-624 Accumulator Discharge SI 1

A/C 63-625 Accumulator Discharge SI 1

A/C 63-551 Safety Injection (Cold Leg)

SI 1

A/C 63-553 Safety Injection (Cold Leg)

SI 1

A/C 63-557 Safety Injection (Cold Leg)

SI 1

A/C 63-555 Safety Injection (Cold Leg)

SI 1

A/C 63-632 Residual Heat Removal (Cold Leg)

SI 1

A/C 63-633 Residual Heat Removal (Cold Leg)

SI 1

A/C 63-634 Residual Heat Removal (Cold Leg)

SI 1

A/C 63-635 Residual Heat Removal (Cold Leg)

SI 1

A/C 63-641 Residual Heat Removal/Safety Injection (Hot Leg)

SI 1

A/C 63-644 Residual Heat Removal/Safety Injection (Hot Leg)

SI 1

A/C 63-558 Safety Injection (Hot Leg)

SI 1

A/C 63-559 Safety Injection (Hot Leg)

SI 1

A/C 63-543 Safety Injection (Hot Leg)

SI 1

A/C 63-545 Safety Injection (Hot Leg)

SI 1

A/C 63-547 Safety Injection (Hot Leg)

SI 1

A/C 63-549 Safety Injection (Hot Leg)

SI 1

A/C 63-640 Residual Heat Removal (Hot Leg)

SI 1

A/C 63-643 Residual Heat Removal (Hot Leg)

SI 1

A/C FCV-74-1 Residual Heat Removal RHR 1

A FCV-74-2 Residual Heat Removal RHR 1

A Notes:

1) All listed valves are included in the scope of the SQN Units 1 and 2, 4th Interval Inservice Testing Plan

2) SI - Safety Injection, RHR - Residual Heat Removal

3) A - Valves for which seat leakage is limited to a specific maximum amount in the closed position, A/C - Check valves for which seat leakage is limited to a specific maximum amount in the closed position

Enclosure CNL-24-071 SQN Units 1 and 2 Pressure Isolation Valve Leakage Test History Valve Number Date of Test As-Left Measured Value (gpm)

Maximum Allowable Leak Rate (gpm) 12/25/2016 0.10 10/8/2017 3.10 5/3/2018 0.12 11/3/2019 0.00 11/24/2019a 0.00 5/7/2021 0.10 11/20/2022 0.14 4/27/2024 0.31 12/25/2016 0.06 10/8/2017 1.41 5/3/2018 0.08 11/3/2019 0.54 11/24/2019a 1.40 5/7/2021 1.55 11/20/2022 2.96 4/28/2024 4.66 12/25/2016 0.29 10/8/2017 1.55 5/3/2018 0.07 11/3/2019 0.00 11/24/2019a 0.00 5/7/2021 0.00 11/20/2022 2.50 4/28/2024 3.27 12/25/2016 0.38 10/8/2017 1.72 5/3/2018 0.23 11/3/2019 0.89 11/24/2019a 0.50 5/7/2021 0.03 11/20/2022 0.00 4/27/2024 0.46 12/21/2016 0.00 5/2/2018 0.00 11/2/2019 0.00 5/6/2021 0.00 11/18/2022 0.00 4/29/2024 0.00 12/21/2016 0.00 5/2/2018 0.00 11/2/2019 0.00 5/6/2021 0.00 11/18/2022 0.00 4/21/2024 2.12 Note 5.0 63-563 5.0 63-562 63-623

a. Second performance during same refueling outage due to re-entering Mode 5 with RHR in service.

5.0 63-622 5.0 Sequoyah Unit 1 PIV SR 3.4.14.1 Leakage Test History 5.0 63-561 5.0 63-560 Accumulator Discharge CNL-24-071 A4-1 of 10 Valve Number Date of Test As-Left Measured Value (gpm)