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, Unit

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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
ML25041A301
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 02/10/2025
From: Hulvey K
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
CNL-25-027, EPID-L-2024-LLA-0156, SQN-TS-24-04
Download: ML25041A301 (1)
v • d • e


Text

1101 Market Street, Chattanooga, Tennessee 37402 CNL-25-027 February 10, 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:

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. NRC electronic mail to TVA, RE:RE: Draft Request for Additional Information - LAR to modify LAR to modify TS SR 3.4.14.1 -

L-2024-LLA-0156, dated January 28, 2025 (ML25028A375)

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 (IST) Program for the Frequency.

In Reference 2, the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI) and requested that TVA provide a response by January 31, 2025. In Reference 3, NRC granted TVAs request to revise the due date for this RAI response to February 10, 2025. Enclosure 1 to this submittal provides a response to the RAI.

U.S. Nuclear Regulatory Commission CNL-25-027 Page 2 February 10, 2025 In response to NRC RAI STSB-LAR-RAI-1 and RAI STSB-LAR-RAI-2, Enclosure 2 to this submittal provides revised SQN Units 1 and 2 TS Bases pages marked up to show the proposed changes. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program. The TS Bases markups in supersede those provided in Reference 1.

This submittal does not change the no significant hazards consideration or the environmental consideration contained in Reference 1. 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 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 10th day of February 2025.

Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness Enclosures

1.

Response to Nuclear Regulatory Commission Request for Additional Information

2.

Revised TS Bases Page Changes (Mark-Ups) for SQN Units 1 and 2 (For Information Only) 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 Digitally signed by Edmondson, Carla Date: 2025.02.10 16:47:15 -05'00' CNL-25-027 E1-1 of 6 Response to Nuclear Regulatory Commission Request for Additional Information NRC Introduction

=

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 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 CNL-25-027 E1-2 of 6 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).

TVA Response SQN Units 1 and 2 Alternative Request RV-02 (Reference 1 of this RAI response) was approved by the NRC in Reference 2 of this RAI response. Alternative Request RV-02 cites the various requirements of SR 3.4.14.1 in the Reason for Request section including the current requirements for frequency of performance, which include those proposed for deletion in the LAR (Reference 3 of this RAI response). Further, in the Basis for Proposed Alternative section of RV-02, TVA indicated that, for PIVs which are opened by flow during shutdowns, the required seat leakage tests will be performed, and acceptable results obtained prior to entering Mode 2 or the plant cannot startup. Retesting the seat leakage following flow through the valve, as discussed in Alternative Request RV-02, may not be required with approval of the proposed TS changes in Reference 3 of this RAI response.

Therefore, TVA will submit a revision to Alternative Request RV-02 within 30 days following NRC approval of Reference 3 of this RAI response. The revisions to RV-02 will address approved changes to the SR 3.4.14.1 performance frequency requirements and changes in the CNL-25-027 E1-3 of 6 Basis for the Proposed Alternative. TVA will not apply either the existing or proposed revised Alternative Request RV-02 once submitted until the revised request is approved by the NRC.

References for EMIB-LAR-RAI-1

1. TVA letter to NRC, CNL-22-024, Sequoyah Nuclear Plant, Units 1 and 2, American Society of Mechanical Engineers Operation and Maintenance Code, Request for Alternative RV-02, dated March 15, 2022 (ML22074A315)
2. NRC letter to TVA, Sequoyah Nuclear Plant, Units 1 and 2 Revised Authorization of Alternative Request RV-02 for Pressure Isolation Valve Seat Leakage (EPID L-2022-LLR-0072), dated December 1, 2022 (ML22167A167)
3. 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)

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.

TVA Response The IST Program for SQN reflects a frequency of testing for the accumulator discharge PIVs (Unit 1 valves63-561 and 63-562 and Unit 2 valve 63-563) of once every two years; however, SQN performs leakage testing of these valves every 18 months during refueling outages.

Although PIV surveillance leak rate data trends can suggest an exceedance of the acceptance criteria at a future performance, the leak rate trend alone is not an indicator of a future result.

For example, leakage test parameters can vary between tests such as RCS test pressure and temperature. Therefore, the appearance of a leak rate data trend does not affect the proposed changes to SR 3.4.14.1.

SQN does trend PIV leak performance over time, and contingency inspection and repair plans are developed in the event the acceptance criteria is exceeded upon performance of the next surveillance. Additionally, the SQN corrective action program would document and resolve any PIV seat leakage exceeding the SR 3.4.14.1 acceptance criteria and adjust the preventative maintenance activities as necessary.

Event driven PIV leak rate testing is infrequently performed under the current SQN requirements for SR 3.4.14.1. As a result, there are very few data points in the PIV leak test history for SQN from event driven performances. Therefore, the availability of PIV leak rate data points would be essentially the same with the implementation of the proposed TS changes to SR 3.4.14.1.

CNL-25-027 E1-4 of 6 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.

TVA Response TVA has not adopted American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code Case OMN-23 for application in the SQN Units 1 and 2 IST program.

ASME OM Code Case OMN-23 provides adequate controls to ensure component performance with leakage test intervals not to exceed 6 years. OMN-23 is an alternative to the two year test frequency specified in ISTC-3630(a). The TS test frequencies are not addressed in the Code Case OMN-23, therefore there is no impact to the proposed license amendment or future application of OMN-23.

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 SQN TS 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.

CNL-25-027 E1-5 of 6 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.

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 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.

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 CNL-25-027 E1-6 of 6
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 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, USNRC, 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)

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.

10 CFR 50.55a(g).

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.

TVA Response TVA will retain the references noted in the above RAI response and will change the designation of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g) to 10 CFR 50.55a(f). The revised SQN Unit 1 and 2 TS 3.4.14.1 Bases are included in Enclosure 2.

Additional TS Bases Changes The following is a discussion of additional proposed TS Bases changes provided in Enclosure 2 for both SQN Units 1 and 2.

In the Applicable Safety Analyses Section of TS Bases B3.4.14, discussion of References 4 and 5 is removed corresponding to the response to STSB-LAR-RAI-1.

Discussion is added that RCS PIV integrity is not considered in any design basis accident and that this specification provides for monitoring of RCS PIV degradation which could lead to accidents. This additional verbiage clarifies how RCS PIV surveillance testing satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

A sentence is added to the last paragraph of the Surveillance Requirements section identifying the Note that the surveillance is not required to be performed in MODES 3 and 4. This is consistent with the proposed changes to TS SR 3.4.14.1.

CNL-25-027 Revised TS Bases Page Changes (Mark-Ups) for SQN Units 1 and 2 (For Information Only)

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-

Revision 45, B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3), define RCS PIVs as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB), which separate the high pressure RCS from an attached low pressure system.

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.

The PIV leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1).

A known component of the identified LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational LEAKAGE if the other is leaktight.

Although this specification 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 basis for this LCO is the 1975 NRC "Reactor Safety Study" (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt. A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual Heat Removal (RHR) System,

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-2 BASES BACKGROUND (continued) b.

Safety Injection System, and c.

Chemical and Volume Control System.

The PIVs are listed in Table B 3.4.14-1 Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

APPLICABLE Reference 4 identified potential intersystem LOCAs as a significant SAFETY contributor to the risk of core melt. The dominant accident sequence in ANALYSES the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs from the RCS. Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent risk of core melt.

Reference 5 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 leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

RCS PIV integrity is not considered in any design basis accident analyses.

This Specification provides for monitoring the condition of the reactor coolant pressure boundary to detect degradation which could lead to accidents.

Revision 45,

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-3 BASES LCO (continued)

Reference 6 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power.

APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.

A.1 and A.2 The flow path must be isolated by two valves. Required Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB.

Required Action A.1 requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time allows the actions and restricts the operation with leaking isolation valves.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This time frame considers the time required to complete this Action and the low probability of a second valve failing during this period.

Revision 45,

Revision 45, RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 1 B 3.4.14-4 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 LQ

DFFRUGDQFHZLWKWKHUHTXLUHPHQWVRIconsistent with 10 CFR 50.55.a(Ig)

(Ref. 7) as FRQWDLQHGin the Inservice Testing Program, DQGis 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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. This surveillance is modified by a Note that it is not required to be performed in MODES 3 and 4. 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



10 CFR 50.2.



10 CFR 50.55a(c).



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



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



NUREG-0677, May 1980.'HOHWHG



ASME Code for Operation and Maintenance of Nuclear PowerPlants.



10 CFR 50.55a(Ig).

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3), define RCS PIVs as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB), which separate the high pressure RCS from an attached low pressure system.

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.

The PIV leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1).

A known component of the identified LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational LEAKAGE if the other is leaktight.

Although this specification 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 basis for this LCO is the 1975 NRC "Reactor Safety Study" (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt. A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems:

a.

Residual Heat Removal (RHR) System, Revision 45,

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-2 BASES BACKGROUND (continued) b.

Safety Injection System, and c.

Chemical and Volume Control System.

The PIVs are listed in Table B 3.4.14-1 Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

APPLICABLE Reference 4 identified potential intersystem LOCAs as a significant SAFETY contributor to the risk of core melt. The dominant accident sequence in ANALYSES the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs from the RCS. Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent risk of core melt.

Reference 5 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 leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

RCS PIV integrity is not considered in any design basis accident analyses.

This Specification provides for monitoring the condition of the reactor coolant pressure boundary to detect degradation which could lead to accidents.

Revision 45,

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-3 BASES LCO (continued)

Reference 6 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power.

APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.

A.1 and A.2 The flow path must be isolated by two valves. Required Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB.

Required Action A.1 requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time allows the actions and restricts the operation with leaking isolation valves.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This time frame considers the time required to complete this Action and the low probability of a second valve failing during this period.

Revision 45,

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 LQ

DFFRUGDQFHZLWKWKHUHTXLUHPHQWVRIconsistent with 10 CFR 50.55.a(Ig)

(Ref. 7) as FRQWDLQHGin the Inservice Testing Program, DQGis 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the valve has

RCS PIV Leakage B 3.4.14 SEQUOYAH - UNIT 

B 3.4.14-5 Revision 45, BASES SURVEILLANCE REQUIREMENTS (continued) been reseated. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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. This surveillance is modified by a Note that it is not required to be performed in MODES 3 and 4. 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



10 CFR 50.2.



10 CFR 50.55a(c).



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



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



NUREG-0677, May 1980.'HOHWHG



ASME Code for Operation and Maintenance of Nuclear PowerPlants.



10 CFR 50.55a(Ig).

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