View source for TSTF-25-03, TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 1, Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency

{{Adams
| number = ML25083A153
| issue date = 03/24/2025
| title = TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 1, Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency
| author name = Jurek S, Richards A, Steinman R, Vaughan J
| author affiliation = Technical Specifications Task Force
| addressee name = 
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 99902042, PROJ0753
| license number = 
| contact person = 
| case reference number = TSTF-25-03
| document type = Letter type:TSTF, Response to Request for Additional Information (RAI)
| page count = 1
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 804-339-7034 Administration by EXCEL Services Corporation TECHNICAL SPECIFICATIONS TASK FORCE A JOINT OWNERS GROUP ACTIVITY TSTF March 24, 2025 TSTF-25-03 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 

==SUBJECT:==
TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 1, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency" On April 22, 2024, the TSTF submitted traveler TSTF-600, Revision 0, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency,"
(Agencywide Documents Access and Management System (ADAMS) Accession No. ML24113A002). On September 20, 2024, the NRC provided a Request for Additional Information (RAI) (ADAMS Accession No. ML24264A023). On October 25, 2024, the TSTF responded to the RAI and provided Revision 1 of TSTF-600 (ADAMS Accession No. ML24299A179).
On January 13, 2025, the NRC provided a second RAI on TSTF-600 by email. The TSTF's response to the NRC's second RAI is attached.
The TSTF and NRC held a teleconference on March 13, 2025, to discuss additional questions.
As a result, TSTF-600 has been revised to address its interaction with approved traveler TSTF-596, Revision 2, "Expand the Applicability of the Surveillance Frequency Control Program (SFCP)."
The responses to the NRC's questions resulted in a revision to the traveler. Revision 2 of TSTF-600 is enclosed. As previously communicated, TSTF travelers will no longer include markups for NUREG-2194, the Westinghouse AP1000 standard technical specifications. As a result, the NUREG-2194 changes have been removed from TSTF-600, Revision 2. 

TSTF 25-03 March 24, 2025 Page 2 Should you have any questions, please do not hesitate to contact us.
Jordan L. Vaughan (PWROG/B&W)
Shane M. Jurek, PE (BWROG)
Andrew M. Richards, Jr. (PWROG/W)
Rebecca L. Steinman, PhD, PE (PWROG/CE)
Attachment Enclosure cc:
Michelle Honcharik, Technical Specifications Branch Shivani Mehta, Technical Specifications Branch 

TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 0, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)
Leakage Testing Frequency" The NRC request is repeated below in italics, followed by the TSTF response in unitalicized text.
By {{letter dated|date=April 22, 2024|text=letter dated April 22, 2024}}, the Technical Specifications Task Force (TSTF) submitted Revision 0 of Traveler TSTF-600, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency" (ML24113A002). You responded to the U.S. Nuclear Regulatory Commission (NRC) staffs initial RAIs on October 25, 2024 (ML24299A179). Upon further review the NRC staff requires the following additional information complete our review of the traveler.
: 1. The traveler 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 (PWRs),
3.4.5 (BWR/4s), and 3.4.6 (BWR/6s). It is not clear to the NRC staff that the traveler justified that there are no licensing basis requirements based on fixed or event driven test frequencies or that the Bases will remain consistent with the proposed changes. For example, the PWR STS 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 STS Bases state that WASH-1400 establishes that intersystem loss-of-coolant accidents (ISLOCAs) are a significant contributor to plant risk. 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 BWR STS Bases for TS 3.4.5 (BWR/4) and TS 3.4.6 (BWR/6) do not refer to WASH-1400, but reference and discuss NUREG-0677.
The NRC staff requests the TSTF 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 STS 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 TSTF propose changes to the STS Bases, as appropriate, to ensure consistency with the proposed changes to the STS. Note that the AP1000 Bases do not refer to either of the references in question in the RCS PIV Integrity specification (3.4.15). Also, as part of the response, explain the impact on plant risk of eliminating the fixed and event driven testing frequencies.
TSTF Response The TSTF evaluated the regulatory history of the RCS PIV requirements. On February 23, 1980, the NRC issued Generic Letter (GL) 80-14, "LWR Primary Coolant System Pressure Isolation Valves." The GL stated that the 1975 "Reactor Safety Study" (WASH-1400 (NUREG-75/014),
Appendix V, October 1975) identified an intersystem Loss of Coolant Accident (LOCA) as a significant contributor to the risk of Pressurized Water Reactor (PWR) core melt accidents. The study examined the configuration of in-series check valves acting as RCS PIVs with and without an in-series Motor Operated Valve (MOV) and concluded that a failure of these valves could lead to overpressurization and rupture of the connected low pressure piping. The GL stated that 

TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 0, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)
Leakage Testing Frequency" the probability of such a failure could be reduced by periodic IST leakage testing on each valve every time the plant is shut down and each time the check valves are moved from the fully closed position, as well as other monitoring methods. The GL required each licensee to respond within 20 days in accordance with 10 CFR 50.54(f) and 1) describe the valve configuration and whether the configuration of concern exists, 2) if the configuration exists, what periodic or surveillance tests are performed to ensure integrity, and 3) if the configuration exists, whether plant procedures should be revised or if plant modifications should be made to increase reliability.
After reviewing the licensee's responses to the GL, the NRC issued orders to those licensees with the RCS PIV configurations of concern. The order immediately incorporated provided Technical Specifications on RCS PIVs into the plant's TS. The provided TS required periodic leakage rate testing prior to entering the operating mode after each time the plant was placed in a cold shutdown condition for 72 hours if testing has not been accomplished in the preceding 9 months and each time the RCS PIV may have been moved from the fully closed position. These testing frequencies are still used in NUREG-1430, NUREG-1431, and NUREG-1432. The RCS PIV configurations of concern did not appear in the Boiling Water Reactor (BWR) plant designs and the event-based testing frequencies were not included in the BWR STS, NUREG-1433 and NUREG-1434.
The state-of-the-art of risk assessment has progressed substantially since the "Reactor Safety Study" in 1975 and NUREG-0677 in 1980. Intersystem LOCAs have been studied extensively by the industry and the NRC. NSAC-154, "ISLOCA Evaluation Guidelines," September 1991, is the industry recognized standard for ISLOCA analysis and it provides detailed instructions on the methodology and documentation. All licensees have internal event Probabilistic Risk Assessment (PRA) models that include consideration of an ISLOCA.
A survey of licensees determined that the risk of an intersystem LOCA is no longer considered a significant contributor to the risk of core melt or large early release. In addition, the 9-month frequency for performing RCS PIV leakage testing is typically not credited, and it is not necessary to reduce the probability of an intersystem LOCA to an acceptable level. While the TSTF believes this to be accurate for all affected plants, a validation of the plant-specific applicability of this information has been added to the model application to be confirmed by each licensee proposing to adopt TSTF-600.
Note that Revision 1 of the TSTF-600 model application included a verification by licensees that the RCS PIV leakage testing Frequencies proposed to be removed from the plant's 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. The NRC verbally expressed a concern at the February 6, 2025, TSTF/NRC public meeting that the TS RCS PIV testing may have been credited by some licensees in justifying an alternative to the American Society of Mechanical Engineers (ASME) Operations and Maintenance Code. Any such reliance would be identified and addressed by affected licensees in response to this required verification. 

TSTF Response to NRC Second Request for Additional Information on TSTF-600, Revision 0, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)
Leakage Testing Frequency" As suggested in the RAI, the Revision 2 of TSTF-600 revises the TS Bases to remove references to WASH-1400 and NUREG-0677 and to discuss the current safety basis of the requirements.
The changes are consistent with the AP1000 standard TS Bases.
: 2. The NRC staff noted on page 13 of 54 of the PDF of TSTF-600, Revision 1, that the NUREG-1434 PIV leakage TS is referred to as 3.4.5 at the top of the page. This should be 3.4.6. The SR is correctly numbered 3.4.6.1.
TSTF Response The error has been corrected in Revision 2 of the traveler. 

TSTF-600, Rev. 2 PWROG-18, Rev. 0 NUREGs Affected:
Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement:
Improvement NRC Fee Status:
Not Exempt Benefit:
Reduces Testing Changes Marked on ISTS Rev 5.0 PWROG RISD & PA (if applicable): PA-LSC-1975 RS-2022-019 See attached.
Revision History OG Revision 0 Revision Status: Closed Original Issue Revision

== Description:==
Revision Proposed by:
PWROG Owners Group Review Information Date Originated by OG:
08-Mar-24 Owners Group Comments (No Comments)
Date: 01-Apr-24 Owners Group Resolution:
Approved TSTF Review Information TSTF Received Date:
01-Apr-24 Date Distributed for Review 01-Apr-24 TSTF Comments:
(No Comments)
Date: 21-Apr-24 TSTF Resolution:
Approved NRC Review Information NRC Received Date:
22-Apr-24 Based on NRC comments on the draft, the justificaiton was revised to focus only on the proposed change and to reflect the NRC's approval of a similar Duke Energy amendment. In addition, two SR Notes were removed from the NUREG-1430, NUREG-1431, and NUREG-1432 SR.
NRC Requests Changes: TSTF Will Revise NRC Comments:
Final Resolution:
24-Mar-25 Copyright(C) 2025, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-600, Rev. 2 PWROG-18, Rev. 0 TSTF Revision 1 Revision Status: Closed On September 20, 2024, the NRC provided an RAI on TSTF-600, Revision 0 (ADAMS Accession No. ML24264A023). The RAI resulted in changes to the traveler justification and model application. The proposed TS and Bases were not affected. The changes are:
: 1) In the justification, the term "valve integrity" was replaced with "pressure isolation functional capability."
: 2) The model application was revised to require licensees to verify that the RCS PIV leakage testing Frequencies proposed to be removed from the 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.
: 3) The justification was revised to clarify that all BWR plant TS that are based on the STS (vice all BWR plant TS) that have an RCS PIV TS have an SR Frequency that only references the IST Program.
: 4) The "Reason for Change" section of the justification was revised to clarify that the ASME Code Case OMN-23 discussion is only an example.
Revision

== Description:==
Revision Proposed by:
TSTF Owners Group Review Information Date Originated by OG:
25-Sep-24 Owners Group Comments (No Comments)
Date: 04-Oct-24 Owners Group Resolution:
Approved TSTF Review Information TSTF Received Date:
11-Oct-24 Date Distributed for Review 11-Oct-24 TSTF Comments:
(No Comments)
Date: 25-Oct-24 TSTF Resolution:
Approved NRC Review Information NRC Received Date:
18-Oct-24 TSTF Revision 2 Revision Status: Active On January 13, 2025, the NRC provided a second RAI on TSTF-600 by email. Revision 2 makes the following changes:
- The TS Bases are revised to no longer reference WASH-1400 and NUREG-0677.
- The model application requires a verfication that an intersystem LOCA isn't risk significant and doesn't rely on the event and frequency based testing for acceptable risk.
- NUREG-2194 is removed from the traveler.
Revision

== Description:==
Revision Proposed by:
TSTF 24-Mar-25 Copyright(C) 2025, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-600, Rev. 2 PWROG-18, Rev. 0 Affected Technical Specifications TSTF Revision 2 Revision Status: Active Added a description of the interaction of TSTF-600 and TSTF-592, "Expand the Applicability of the Surveillance Frequency Control Program (SFCP)."
Owners Group Review Information Date Originated by OG:
17-Feb-25 Owners Group Comments (No Comments)
Date: 03-Mar-25 Owners Group Resolution:
Approved TSTF Review Information TSTF Received Date:
06-Mar-25 Date Distributed for Review 06-Mar-25 TSTF Comments:
(No Comments)
Date:
TSTF Resolution:
SR 3.4.14.1 NUREG(s)- 1430 1431 1432 Only RCS PIV Leakage SR 3.4.14.1 Bases NUREG(s)- 1430 1431 1432 Only RCS PIV Leakage SR 3.4.5.1 NUREG(s)- 1433 Only RCS PIV Leakage SR 3.4.5.1 Bases NUREG(s)- 1433 Only RCS PIV Leakage SR 3.4.6.1 NUREG(s)- 1434 Only RCS PIV Leakage SR 3.4.6.1 Bases NUREG(s)- 1434 Only RCS PIV Leakage 24-Mar-25 Copyright(C) 2025, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-600, Rev. 2 Page 1
: 1. 

==SUMMARY==
DESCRIPTION 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. The proposed change affects the Standard Technical Specifications (STS) in NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and NUREG-14341.
: 2.
DETAILED DESCRIPTION 2.1. System Design and Operation RCS PIVs are two normally closed valves in series on a penetration of the reactor coolant pressure boundary (RCPB) that separate the high pressure RCS from an attached low pressure system, such as the Residual Heat Removal System and the low pressure Emergency Core Cooling System. The number, design, and function of the RCS PIVs is plant-specific. The RCS PIVs are described in the plant's Updated Final Safety Analysis Report (UFSAR) and IST Program documentation.
The RCS PIV Leakage Limiting Condition for Operation (LCO) limits the leakage through the RCS PIVs to amounts that do not compromise safety. The RCS PIV leakage limit applies to each valve in the series and is specified in the Technical Specifications (TS) and the ASME OM Code. Any leakage through both RCS PIVs on a penetration is included in "identified leakage,"
and is limited by the TS LCO on "RCS Operational LEAKAGE." All RCS PIVs required to be tested under the plant's TS are also governed by the plant's IST Program.
1 NUREG-1430 provides the STS for the Babcock & Wilcox plant designs.
NUREG-1431 provides the STS for the Westinghouse plant designs.
NUREG-1432 provides the STS for the Combustion Engineering plant designs.
NUREG-1433 provides the STS for the BWR/4 plant designs, but is also representative of the BWR/2, BWR/3, and, in this case, of the BWR/5 plant design.
NUREG-1434 provides the STS for the BWR/6 plant designs, but is also representative in some cases of the BWR/5 plant design. 

TSTF-600, Rev. 2 Page 2 2.2. Current Technical Specifications Requirements NUREG-1430, Specification 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage,"
SR 3.4.14.1, states:
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 DHR flow path when in the DHR 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.
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] psia and [2255] psia.
[ In accordance with the INSERVICE TESTING PROGRAM OR
[ [18] months]
OR 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

TSTF-600, Rev. 2 Page 3 AND
[ Within 24 hours following valve actuation due to automatic or manual action or flow through the valve ]
NUREG-1431, Specification 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage,"
SR 3.4.14.1, states:
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.
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
[ [18] months OR In accordance with the Surveillance Frequency Control Program ]
AND Prior to entering MODE 2 whenever the unit 

TSTF-600, Rev. 2 Page 4 NUREG-1432, Specification 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage,"
SR 3.4.14.1, states:
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 SDC 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.
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] psia and [2255] psia.
In accordance with the INSERVICE TESTING PROGRAM, and
[ [18] months OR In accordance with the Surveillance has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours following valve actuation due to automatic or manual action or flow through the valve 

TSTF-600, Rev. 2 Page 5 Frequency Control Program ]
AND Prior to entering MODE 2 determine 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 Within 24 hours following valve actuation due to automatic or manual action or flow through the valve NUREG-1433, Specification 3.4.5, "RCS Pressure Isolation Valve (PIV) Leakage," SR 3.4.5.1, states:
SR 3.4.5.1 
-------------------------------NOTE------------------------------
Not required to be performed in MODE 3.
Verify equivalent leakage of each RCS PIV is 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at an RCS pressure [ ] and
[ ] psig.
[ In accordance with the INSERVICE TESTING PROGRAM OR
[ [18] months]
OR In accordance with the Surveillance 

TSTF-600, Rev. 2 Page 6 Frequency Control Program ]
NUREG-1434, Specification 3.4.6 "RCS Pressure Isolation Valve (PIV) Leakage," SR 3.4.6.1, states:
SR 3.4.6.1 
-------------------------------NOTE------------------------------
Not required to be performed in MODE 3.
Verify equivalent leakage of each RCS PIV is 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at an RCS pressure [ ] and
[ ] psig.
[ In accordance with the INSERVICE TESTING PROGRAM OR
[ [18] months]
OR In accordance with the Surveillance Frequency Control Program ]
2.3. Reason for the Proposed Change The RCS PIV leakage testing required by the STS is also required by the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) as implemented in a site's IST Program. However, NUREG-1430, NUREG-1431, and NUREG-1432 require more frequent testing than is required by the ASME OM Code or by the other STS (NUREG-1433 and NUREG-1434). The proposed change would align the STS testing frequency with the requirements of the ASME OM Code and eliminate unnecessary testing. Eliminating unnecessary RCS PIV testing could potentially lower outage dose and reduce outage time.
The proposed change also provides the opportunity for licensees to leverage industry experience.
In addition to publishing a new edition of the OM Code periodically, the ASME OM Committee also publishes new approved or modified Code Cases that provide alternatives developed and approved by ASME or explain the intent of existing Code requirements. Those approved by the NRC may be used voluntarily by licensees as an alternative to compliance with ASME Code provisions that have been incorporated by reference into 10 CFR 50.55a. Revising the Frequency of the RCS PIV leakage verification SR to reference only the IST Program provides 

TSTF-600, Rev. 2 Page 7 licensees with the opportunity to stay aligned with updates in regulation afforded by 10 CFR 50.55a(b). Paragraph (b)(6) of 10 CFR 50.55a allows for the application of the ASME Code Cases listed in Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code," without prior NRC approval subject to the conditions listed in 10 CFR 50.55a(b)(6), paragraphs (i) through (iii). For example, Revision 4 of Regulatory Guide 1.192, endorses ASME OM Code Case OMN-23, "Alternative Rules for Testing Pressure Isolation Valves." This Code Case provides an alternative to the fixed 2-year leakage rate test frequency specified by ASME OM Code Section ISTC-3630(a) and permits the implementation of a condition monitoring program for RCS PIVs. The Code Case does not replace or exclude any test method requirement specified in ISTC-3630, but does allow for the possibility of interval extensions, not to exceed a maximum interval of 6 years, dependent on analysis of tests results and maintenance history of the RCS PIVs. The existing STS testing requirements restrict the ability to implement the condition monitoring program because of the refueling interval and event-driven SR Frequencies. Note that the discussion of OMN-23 is only an example, and the proposed change does not include permission to use Code Case OMN-23, nor is the value of the proposed change dependent on adoption of OMN-23. Licensees will need to follow the appropriate regulatory process should they choose to use the Code Case.
NUREG-1433 and NUREG-1434 (BWR plants) require verification of RCS PIV leakage in accordance with the IST Program or at a fixed Frequency or in accordance with the Surveillance Frequency Control Program (SFCP). There are no testing frequencies related to shutdown conditions or valve actuation. All BWR plants with TS based on the STS that have an RCS PIV TS have a frequency that references the IST Program. The traveler will revise the NUREG-1433 and NUREG-1434 SR Frequency to be in accordance with the IST Program in order to make the BWR STS consistent with the existing BWR plant TS based on the STS.
2.4. Description of the Proposed Change The Frequency for NUREG-1430, NUREG-1431, and NUREG-1432, SR 3.4.14.1, is replaced with, "In accordance with the INSERVICE TESTING PROGRAM." Notes 2 and 3 of SR 3.4.14.1 are deleted.
The Frequency for NUREG-1433, SR 3.4.5.1, and for NUREG-1434, SR 3.4.6.1, is replaced with, "In accordance with the INSERVICE TESTING PROGRAM."
Plants that have adopted TSTF-596, Revision 2, "Expand the Applicability of the Surveillance Frequency Control Program (SFCP)," (ADAMS Accession No. ML24362A054) will use a Frequency of "In accordance with the Surveillance Frequency Control Program." Under TSTF-596, references to the Inservice Testing Program are replaced with references to the SFCP.
The TS Section 5.5, "Surveillance Frequency Control Program," references 10 CFR 50.55a(f)
(the Inservice Testing Program) for the applicable SRs, such as the RCS PIV SR. This is an administrative difference that does not affect the justification for TSTF-600.
Only the SR Frequency is revised. The proposed change does not add or remove any RCS PIVs from the plant's TS or ASME OM Code requirements. The proposed change does not alter the method of testing or the SR acceptance criteria. 

TSTF-600, Rev. 2 Page 8 A model application is attached. The model may be used by licensees desiring to adopt the traveler following NRC approval.
The model application requires licensees proposing to adopt the traveler to confirm that the RCS PIV testing Frequencies being removed from the TS are not credited for satisfying any requirements described in the UFSAR or any commitments for reasons other than being a TS requirement. The model application also requires licensees proposing to adopt the traveler to confirm that an Intersystem Loss-of-Coolant Accident (ISLOCA) is not a risk-significant contributor to core damage or large early release in the plant's probabilistic risk assessment evaluation, and that the event-driven and 9-month Frequency RCS PIV testing is not necessary to ensure the risk associated with an ISLOCA is acceptable.
: 3.
TECHNICAL EVALUATION The traveler replaces RCS PIV testing Frequency with a reference to the IST Program. This results in the following changes:
* Elimination of a fixed Frequency or a Frequency based on the SFCP and retention or addition of a reference to the IST Program;
* Elimination of a Frequency of 24 hours 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 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 an SR Note that exempts performance for valves associated with an in-service decay heat removal system; and
* Change to NUREG-1433 and NUREG-1434 Frequencies.
Each of these changes are evaluated below.
Testing in Accordance with IST Program Instead of the Surveillance Frequency Control Program (SFCP) or on a Fixed Frequency All operating plants except Vogtle 3 & 4 (of the AP1000 design) have adopted a SFCP. If a plant had a fixed Frequency in the RCS PIV testing SR, it would have been replaced with a reference to the SFCP as part of adopting that program. Therefore the [18] month optional Frequency does not appear in any plant's TS and can be removed from the STS.
Under the SFCP, changes to the Frequencies must be made in accordance with NEI 04-10, Revision 1, "Risk-Informed Method for Control of Surveillance Frequencies." NEI 04-10, Step 7, requires identification of qualitative considerations that must be addressed, such as test intervals specified in applicable industry codes and standards, e.g., ASME, IEEE, etc. Because an RCS PIV testing frequency of 24 months is specified in the ASME OM Code, Section ISTC-3630, paragraph (a), the SR Frequency cannot be extended beyond 24 months under the 

TSTF-600, Rev. 2 Page 9 SFCP without NRC approval of a relief request per 10 CFR 50.55a, paragraph (z), "Alternatives to codes and standards requirements." As a result, the appropriate reference is the ASME OM Code and 10 CFR 50.55a instead of the SFCP.
Therefore, changing the SR Frequency to be in accordance with the IST Program instead of a reference to the SFCP or to a fixed Frequency of 24 months has no detrimental effect on testing as the controlling requirement continues to be the ASME OM Code as referenced in 10 CFR 50.55a(f).
Testing Within 24 Hours of Valve Actuation The proposed change would eliminate the requirement to perform the surveillance within 24 hours following valve actuation due to automatic or manual action or flow through the valve.
Although the SR provides a limit on allowable RCS PIV leakage rate, its main purpose is to prevent overpressure failure of the low-pressure portions of connecting systems that would result from gross failure of the associated PIVs. Leakage is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. As such, performing the valve leakage test in conjunction with the other valve tests specified in the ASME OM Code (e.g.,
exercise testing per ASME OM Code Subsection ISTC-3520 and position indication verification per ASME OM Code Subsection ISTC-3700) provides an acceptable method of verifying the RCS PIV's pressure isolation functional capability. The capability of the valves to transition from open to closed provides assurance that the valves can perform their pressure isolation functions as required. Performance of the separate PIV leak rate testing following valve actuation does not contribute any additional assurance of the pressure isolation functional capability.
In removing the Frequency requiring performance of leakage testing within 24 hours following valve actuation due to automatic or manual action or flow through the valves, the TS will instead require a Frequency in accordance with the IST Program, which reinforces the established requirements of the OM Code that are endorsed by the NRC and incorporated by reference in 10 CFR 50.55a(a)(1)(iv). The testing requirements contained within the OM Code are intended to assess the operational readiness of the stated components. Therefore, leakage testing of the RCS PIVs at the frequency established by the IST Program is satisfactory for determining their pressure isolation functional capability. Furthermore, this change in frequency would also align with the corresponding SR 3.4.5.1 in NUREG-1433 and SR 3.4.6.1 in NUREG-1434.
On October 24, 2023 (Reference 3), the NRC approved a license amendment request submitted by Duke Energy that revised the SR Frequency for RCS PIV leakage testing in the same manner that is proposed by this traveler and using a similar justification. See Section 5, "References." In Reference 3, the NRC stated, "the NRC staff has determined that the TS requirement to perform leakage testing within 24 hours following valve actuation is directed to the same end as testing at a frequency in accordance with the IST Program, which meets the established requirements of the OM Code that are incorporated by reference in 10 CFR 50.55a(a)(1)(iv). Accordingly, the NRC staff finds that the requirement for leakage testing within 24 hours following valve actuation is unnecessary and may be deleted from the TS." 

TSTF-600, Rev. 2 Page 10 NUREG-1430, NUREG-1431, and NUREG-1432, SR 3.4.14.1, Note 3, states, "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." This Note is discussed in the 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." With the elimination of the 24-hour testing Frequency, the Note is no longer needed. The Note does not appear in the STS NUREGs that do not have the 24-hour testing Frequency.
Testing Every 9 Months NUREG-1430, NUREG-1431, and NUREG-1432 require RCS PIV testing every 9 months prior to entering Mode 2 if the plant has been in Mode 5 for seven days or more. The proposed change would remove this SR Frequency. In removing the 9-month Frequency, the TS will instead require leakage testing at a Frequency in accordance with the IST Program, which reinforces the established requirements of the OM Code that are endorsed by the NRC and incorporated by reference in 10 CFR 50.55a(a)(1)(iv). The testing requirements contained within the OM Code are intended to assess the operational readiness of the stated components.
Therefore, leakage testing of the RCS PIVs at the frequency established by the IST Program is satisfactory for determining their pressure isolation functional capability.
In Reference 3, the NRC stated, "Despite the IST Program Frequency being based on the requirements of the ASME OM Code incorporated by reference into 10 CFR 50.55a, the site TS SR reflect a requirement to test after any cold shutdown for 7 days or more, effectively restricting the ability of the IST Program and the SFCP to govern the frequency of leakage testing. The proposed change to each sites TS SR Frequency would ensure ASME OM Code testing requirements for the RCS PIVs will be retained in the IST Program and duplicative testing requirements in the TS SR will be removed."
Elimination of an SR Note that Exempts Performance for Valves Associated with an In-Service Decay Heat Removal System NUREG-1430, NUREG-1431, and NUREG-1432, SR 3.4.14.1, Note 2 states, "Not required to be performed on the RCS PIVs located in the [DHR/ RHR/SDC] flow path when in the shutdown cooling mode of operation." The NUREG-1430 Note refers to the Decay Heat Removal (DHR) system, NUREG-1431 refers to the Residual Heat Removal (RHR) system, and NUREG-1432 refers to the Shutdown Cooling (SDC) system. This Note is not needed and is eliminated.
The Applicability of TS 3.4.14 states:
APPLICABILITY:
MODES 1, 2, and 3, MODE 4, except valves in the [decay heat removal (DHR) /
residual heat removal (RHR) / shutdown cooling (SDC)]
flow path when in, or during the transition to or from, the
[DHR/RHR/SDC] mode of operation. 

TSTF-600, Rev. 2 Page 11 The Applicability already exempts the RCS PIVs associated with the 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 decay heat removal system is only used in Modes 3 and 4 of the Applicability. Therefore, Note 2 is not required and is eliminated. There is no similar note in NUREG-1433 or NUREG-1434.
Change to NUREG-1433 and NUREG-1434 The proposed change updates the NUREG-1433 and NUREG-1434 RCS PIV leakage testing frequency as an administrative change. The BWR STS require verification of RCS PIV leakage in accordance with the IST Program and in accordance with the SFCP. However, all BWR plants with TS based on the STS that have an RCS PIV TS have an SR Frequency that references the IST Program and does not reference the SFCP. Therefore, the STS frequency is revised to be consistent with the TS of the STS BWR plants and to be consistent with the proposed change to the PWR plants.
Additional Considerations In Reference 3, the NRC stated, "The NRC staff finds that the proposed changes to each sites TS SR Frequency would ensure ASME OM Code testing requirements for the RCS PIVs will be retained in the IST Program and duplicative testing requirements in the TS SR will be removed.
Accordingly, the proposed TS SR Frequency change to 'INSERVICE TESTING PROGRAM,' is consistent with the previously approved TSTF-545, 'TS Inservice Testing Program Removal &
Clarify SR Usage Rule Application to Section 5.5 Testing'."
On December 11, 2015, the NRC approved (ADAMS Accession No. ML15314A365) TSTF-545 (ADAMS Accession No. ML15294A555). TSTF-545 removed detailed IST Program requirements from the Administrative Controls section of TS and added a new TS definition for INSERVICE TESTING PROGRAM, which is defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The Safety Evaluation states (emphasis added):
Furthermore, the NRC staff notes that the licensee's IST, which is required by 10 CFR 50.54 and 50.55a(f), and which is outside the scope of this amendment request, already contains requirements and considerations similar to those of [the Administrative Controls requirement]. First, the NRC staff finds that the proposed deletion of [the Administrative Controls requirement] is acceptable on the basis that each licensee has its self-contained IST program developed and implemented in accordance with the requirements of the ASME OM Code which adequately defines frequencies. Including testing frequencies and [the Administrative Controls requirement] is duplicative of the requirements in the ASME OM Code as required by 10 CFR 50.55a(f). Therefore, the NRC staff finds the proposed deletion of [the Administrative Controls requirement]
acceptable on the basis that the changes eliminate an inconsistency between TS and IST program with regard to test delays and are in conformance with 10 CFR 50.55a(f)(5)(ii).
Last, concerning any conflict between IST Code requirements and TS, 10 CFR 50.55a(f)(5)(ii) already specifies that, if a revised inservice test program for a facility conflicts with the TS, then the licensee must apply for an appropriate amendment.
Therefore, deleting the contrary statement from the TS is acceptable. 

TSTF-600, Rev. 2 Page 12 The same logic is applicable to removing the event-driven and fixed Frequencies from the RCS PIV SR.
In updating the Frequency for verifying RCS PIV leakage, the Frequency would instead reflect being in accordance with the IST Program, which is required by regulation per both 10 CFR 50.54, Condition of licenses, and 10 CFR 50.55a(f). Specifically, paragraph (jj) of 10 CFR 50.54 states that structures, systems, and components (SSCs) 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. Paragraph 10 CFR 50.55a(f), "lnservice Testing Requirements," requires, in part, that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda incorporated by reference in the regulations. Exceptions are allowed where alternatives have been authorized by the NRC pursuant to paragraphs 10 CFR 50.55a(z)(1) and 10 CFR 50.55a(z)(2).
The SR 3.0.2 Bases discuss testing performed in accordance with the IST Program. It states, "Examples of where SR 3.0.2 does not apply are the Containment Leakage Rate Testing Program required by 10 CFR 50, Appendix J, and the inservice testing of pumps and valves in accordance with applicable American Society of Mechanical Engineers Operation and Maintenance Code, as required by 10 CFR 50.55a. These programs establish testing requirements and Frequencies in accordance with the requirements of regulations. The TS cannot in and of themselves extend a test interval specified in the regulations directly or by reference."
Therefore, the proposed change is consistent with the NRC staff's approval of TSTF-545, the NRC staff's approval of a similar change as documented in Reference 3, 10 CFR 50.54 and 10 CFR 50.55a(f), and the treatment of similar Frequencies in the TS.
TS Bases Changes The SR Bases are revised to reflect the changes to the SR Frequency. In addition, a typographical error in the SR Bases, which refers to 10 CFR 50.55a(g) instead of 10 CFR 50.55a(f), is corrected.
The Bases are updated to reflect the current safety basis of the requirements. References to outdated studies, such as WASH-1400 (1975) and NUREG-0677 (1980), are removed.
On December 30, 2024, the NRC approved TSTF-596, Revision 2, "Expand the Applicability of the Surveillance Frequency Control Program (SFCP)." This traveler, among other changes, replaced references to the Inservice Testing Program in the TS Bases with references to 10 CFR 50.55a(f). For consistency, and to assist licensees adopting both travelers, a similar change has been made to the RCS PIV Bases in this traveler.
: 4.
REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria The RCS PIVs are referred to in 10 CFR 50.2, "Definitions," under "Reactor coolant pressure boundary," and 10 CFR 50.55a(c), "Reactor coolant pressure boundary," as two normally closed 

TSTF-600, Rev. 2 Page 13 valves in series within the reactor coolant pressure boundary, which separate the high pressure RCS from attached lower pressure systems.
Section 50.55a, Codes and standards, in 10 CFR in 50.55a(f)(4), Inservice testing standards requirement for operating plants, states, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code, Division 1 must meet the inservice testing (IST) requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in 10 CFR 50.55a(f)(2) and (3) and that are incorporated by reference in 10 CFR 50.55a(a)(1)(iv), to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Section 50.55a(f)(5)(ii), IST program update: Conflicting IST Code requirements with technical specifications, states, in part, that if a revised inservice test program for a facility conflict with the TS for the facility the licensee must apply for an amendment of the TS to conform the TS to the revised program.
The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(b) requires:
Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"]. The Commission may include such additional technical specifications as the Commission finds appropriate.
Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for an amendment must be filed with the Commission, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.
Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate.
Section IV, "The Commission Policy," of the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58FR39132), dated July 22, 1993, states in part that improved STS have been developed and will be maintained for each NSSS owners group. The Commission Policy encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications. The industry's proposal of travelers and the NRC's approval of travelers is the method used to maintain the improved STS as described in the Commission's Policy. Following NRC approval, licensees adopt travelers into their plant-specific technical specifications following the requirements of 10 CFR 50.90. Therefore, the traveler process facilitates the Commission's policy while satisfying the requirements of the applicable regulations. 

TSTF-600, Rev. 2 Page 14 The regulation at 10 CFR 50.36(a)(1) also requires the application to include a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls." The proposed traveler revises the Bases to be consistent with the Technical Specifications, and therefore, is in compliance with 10 CFR 50.36(a)(1).
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 Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.
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 traveler and 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 included in the respective site's IST Program. The model application of this traveler requires each licensee adopting the traveler to make a similar confirmation. The license amendment was approved on October 24, 2023 (Reference 3). The NRC's technical evaluation of the Duke Energy change is also applicable to the proposed traveler.
: 5.
REFERENCES
: 1. Letter from Shawn Gibby (Duke Energy) to the Document Control Desk (NRC), 
"License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies," dated February 1, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23032A162).
: 2. Letter from Shawn Gibby (Duke Energy) to Document Control Desk (NRC), 
"Response to Request for Additional Information Regarding the License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies," dated July 7, 2023 (ADAMS Accession No. ML23188A176).
: 3. Letter from Shawn A. Williams (NRC) to Shawn Gibby (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 (ADAMS Accession No. ML23241A987). 

TSTF-600, Rev. 2 Model Application 

TSTF-600, Rev. 2 Page 1
[DATE]
10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 DOCKET NO. PLANT NAME
[5X]-[xxx] 

==SUBJECT:==
Application to Revise Technical Specifications to Adopt TSTF-600, "Revise the Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage Testing Frequency" Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].
[LICENSEE] requests adoption of TSTF-600, "Revise the Reactor Coolant System (RCS)
Pressure Isolation Valve (PIV) Leakage Testing Frequency," which is an approved change to the Standard Technical Specifications (STS), into the [PLANT NAME, UNIT NOS] TS. TSTF-600 revises the Surveillance Requirement (SR) to perform RCS PIV leakage testing to only reference the Inservice Testing Program (IST Program) for the Frequency.
The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. [Attachment 2 provides revised (clean) TS pages.] Attachment 3 provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only.
[LICENSEE] requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within 6 months of completion of the NRCs acceptance review. Once approved, the amendment shall be implemented within [90] days.
There are no regulatory commitments in this letter.
[In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.]
[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.
Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.] 

TSTF-600, Rev. 2 Page 2 If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].
Sincerely,
[Name, Title] 

==Enclosure:==
Description and Assessment Attachments: 1.
Proposed Technical Specification Changes (Mark-Up)
[2. Revised Technical Specification Pages]
: 3.
Proposed Technical Specification Bases Changes (Mark-Up) - For Information Only
[The attachments are to be provided by the licensee and are not included in the model application.]
cc:
NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact 

TSTF-600, Rev. 2 Page 3 ENCLOSURE DESCRIPTION AND ASSESSMENT 

==1.0 DESCRIPTION==
[LICENSEE] requests to adoption of TSTF-600, "Revise the Reactor Coolant System (RCS)
Pressure Isolation Valve (PIV) Leakage Testing Frequency," which is an approved change to the Standard Technical Specifications (STS), into the [PLANT NAME, UNIT NOS] TS. TSTF-600 revises the Surveillance Requirement (SR) to perform RCS PIV leakage testing to only reference the Inservice Testing Program (IST Program) for the Frequency. 

==2.0 ASSESSMENT==
2.1 Applicability of Safety Evaluation
[LICENSEE] has reviewed the safety evaluation for TSTF-600 provided to the Technical Specifications Task Force in a letter dated [DATE]. This review included the NRC staffs evaluation, as well as the information provided in TSTF-600. [LICENSEE] has concluded that the justifications presented in TSTF-600 and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.
((LICENSEE] has adopted TSTF-596, Revision 2, "Expand the Applicability of the Surveillance Frequency Control Program (SFCP)," into the [PLANT] TS. Therefore, the Frequency for SR [3.4.14.1] is revised to state, "In accordance with the Surveillance Frequency Control Program." ]
[LICENSEE] confirms that the RCS PIVs required to be tested by the [PLANT] TS are included in the [PLANT] IST Program.
[LICENSEE] confirms that an Intersystem Loss-of-Coolant Accident (ISLOCA) is not a risk-significant contributor to core damage or large early release in the plant's probabilistic risk assessment evaluation, and that the event-driven and 9-month Frequency RCS PIV testing is not necessary to ensure the risk associated with an ISLOCA is acceptable.
[LICENSEE] confirms that the RCS PIV leakage testing Frequencies proposed to be removed from the [PLANT] 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.
2.2 Variations
[LICENSEE is not proposing any variations from the TS changes described in TSTF-600 or the applicable parts of the NRC staffs safety evaluation.] [LICENSEE is proposing the following variations from the TS changes described in TSTF-600 or the applicable parts of the NRC staffs safety evaluation:] 

TSTF-600, Rev. 2 Page 4
[The [PLANT] TS utilize different [numbering][and][titles] than the STS on which TSTF-600 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-600 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-600 to the [PLANT] TS.]
[The [PLANT] TS contain requirements that differ from the STS on which TSTF-600 was based but are encompassed by the TSTF-600 justification. (Describe the differences)]
[The [PLANT] TS do not include the changes in TSTF-545, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing." As a result, the
[PLANT] TS include a TS Section [5.5], "Inservice Testing Program," and do not include a defined term, "INSERVICE TESTING PROGRAM." Therefore, the proposed SR Frequency does not use a capitalized defined term.]
[The [PLANT] TS are not based on the standard TS on which TSTF-600 is based. (Describe the differences. For example, changes to SR 4.0.5, SRs that reference the IST or 4.0.5, and SRs that contain requirements that are in STS programs.)] 

==3.0 REGULATORY ANALYSIS==
3.1 No Significant Hazards Consideration Analysis
[LICENSEE] requests adoption of TSTF-600, "Revise the Reactor Coolant System (RCS)
Pressure Isolation Valve (PIV) Leakage Testing Frequency," which is an approved change to the Standard Technical Specifications (STS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). TSTF-600 revises the Surveillance Requirement (SR) to perform RCS PIV leakage testing to only reference the Inservice Testing Program (IST Program) for the Frequency.
[LICENSEE] 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 change revises the SR to perform RCS PIV leakage testing to only reference the IST Program for the Frequency. Performance of the RCS PIV leakage testing is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The RCS PIVs for which the Surveillance Frequencies are revised are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigative function assumed in the accident analysis. The proposed test frequency has been endorsed in the American Society of Mechanical Engineers (ASME) Operations and Maintenance Code for Nuclear Power Plants, and in NRC regulation 10 CFR 50.55a. As a result, the consequences of any accident previously evaluated are not significantly increased. 

TSTF-600, Rev. 2 Page 5 Therefore, the proposed change 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 change revises the SR to perform RCS PIV leakage testing to only reference the IST Program for the Frequency. No new or different accidents result from utilizing the proposed change. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements on the affected components. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.
Therefore, the proposed change 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 change revises the SR to perform RCS PIV leakage testing to only reference the IST Program for the Frequency. The proposed change does not change the SR acceptance criteria or adversely affect existing plant safety margins. The reliability of the equipment assumed to operate in the safety analysis is not significantly affected. As such, there are no changes being made to safety analysis assumptions, safety limits or limiting safety system settings that would significantly affect plant safety as a result of the proposed change.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, [LICENSEE] 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.
3.2 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 Commissions 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. 

TSTF-600, Rev. 2 Page 6 

==4.0 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 a 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. 

TSTF-600, Rev. 2 Technical Specifications Changes 

RCS PIV Leakage 3.4.14 Babcock & Wilcox STS 3.4.14-3 Rev. 5.0 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 DHR flow path when in the DHR 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.
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] psia and [2255] psia.
[ In accordance with the INSERVICE TESTING PROGRAM OR
[ [18] months]
OR In accordance with the Surveillance Frequency Control Program ]
AND TSTF-600, Rev. 2

RCS PIV Leakage 3.4.14 Babcock & Wilcox STS 3.4.14-4 Rev. 5.0 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 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
[ Within 24 hours following valve actuation due to automatic or manual action or flow through the valve ]
SR 3.4.14.2 
-------------------------------NOTE------------------------------
[ Not required to be met when the DHR System autoclosure interlock is disabled in accordance with LCO 3.4.12.
Verify DHR System autoclosure interlock prevents the valves from being opened with a simulated or actual RCS pressure signal [425] psig.
[ [18] months OR In accordance with the Surveillance Frequency Control Program ] ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-1 Rev. 5.0 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 RCS pressure boundary that 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 leakage 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.
Decay Heat Removal (DHR) System,
: b.
Emergency Core Cooling System (ECCS), and TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-2 Rev. 5.0 BASES BACKGROUND (continued)
: c.
Makeup and Purification System.
The PIVs are listed in [FSAR section] Reference 46.
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 DHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the reactor coolant pressure boundary (RCPB), and the subsequent pressurization of the DHR System downstream of the PIVs from the RCS. Because the low pressure portion of the DHR System is typically designed for 600 psig, overpressurization failure of the DHR 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 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.
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.
Reference 57 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-3 Rev. 5.0 of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-4 Rev. 5.0 BASES LCO (continued) 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 DHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the DHR 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 is not sufficient to overpressurize the connected low pressure systemsfor a LOCA outside the containment.
ACTIONS The ACTIONS are modified by two Notes. Note 1 is added to provide 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 on the RCS pressure boundary [or the high pressure portion of the system].
Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. 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 hours allows the actions and restricts the operation with leaking isolation valves.
[ Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing some other valve qualified for isolation or restoring one leaking PIV. [The 72 hour time after exceeding the limit considers the time required to complete the Action and the low probability of a second valve failing during this time period.
or TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-5 Rev. 5.0 BASES ACTIONS (continued)
The 72 hour time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This timeframe considers the time required to complete this Action and the low probability of a second valve failing during this period. ] 
-----------------------------------REVIEWERS NOTE-----------------------------------
Two options are provided for Required Action A.2. The second option (72 hour restoration) is appropriate if isolation of a second valve would place the unit in an unanalyzed condition.
B.1 and B.2 If leakage cannot be reduced, [the system isolated,] or other Required Actions accomplished, 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 hours and to MODE 5 within 36 hours. This Required 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.
C.1 The inoperability of the DHR autoclosure interlock renders the DHR suction isolation valves incapable of isolating in response to a high pressure condition and preventing inadvertent opening of the valves at RCS pressures in excess of the DHR systems design pressure. If the DHR autoclosure interlock is inoperable, operation may continue as long as the DHR suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours. This action accomplishes the purpose of the autoclosure function.
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 or 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.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-6 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
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 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. 
-----------------------------------REVIEWERS NOTE-----------------------------------
If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of [18] months or the reference to the Surveillance Frequency Control Program should be used.
[ The [18 month] Frequency is in accordance with the requirements of consistent with 10 CFR 50.55a(fg) (Ref. 58) as contained in the INSERVICE TESTING PROGRAM, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref. 7), and is based on the need to perform such surveillances under conditions that apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the plant at power.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
[ 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 hours after the valve has been reseated. Within 24 hours is a reasonable and practical time limit for performing this test after opening or reseating a valve. ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-7 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
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 complimentary 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 DHR System when the DHR System is aligned to the RCS in the decay heat removal mode of operation. PIVs contained in the DHR flow path must be leakage rate tested after DHR is secured and stable unit conditions and the necessary differential pressures are established. 
-----------------------------------REVIEWERS NOTE-----------------------------------
The "24 hour..." Frequency of performance for Surveillance Requirement 3.4.14.1 is not required for B&W Owner's Group plants licensed prior to 1980. These plants were licensed prior to the NRC establishing formal Technical Specification controls for pressure isolation valves. Subsequently, these earlier plants had their licenses modified by NRC Order to require certain PIV testing Frequencies (excluding the "24 hour..." Frequency) be included in that plant's Technical Specifications. Based upon the information available to the Staff at the time, the content of those Orders was considered acceptable. Since 1980, the NRC Staff has determined an additional PIV leakage rate determination is required within 24 hours following actuation of the valve and flow through the valve. This is necessary in order to ensure the PIV's ability to support the integrity of the reactor coolant pressure boundary.
The Revised Standard Technical Specifications include the "24 hours..."
Frequency to reflect current NRC Staff position on the need to include this test requirement within Technical Specifications.
[ SR 3.4.14.2 and SR 3.4.14.3 Verifying that the DHR autoclosure interlocks are OPERABLE ensures that RCS pressure will not pressurize the DHR system beyond 125% of its design pressure of [600] psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be 
< [425] psig to open the valves. This setpoint ensures the DHR design pressure will not be exceeded and the DHR relief valves will not lift. [ The 18 month Frequency is based on the need to perform this Surveillance TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Babcock & Wilcox STS B 3.4.14-8 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued) under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance was performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
These SRs are modified by Notes allowing the DHR autoclosure function to be disabled when using the DHR System suction relief valve for cold overpressure protection in accordance with LCO 3.4.12. ]
REFERENCES
: 1.
10 CFR 50.2.
: 2.
10 CFR 55a(c).
: 3.
10 CFR 50, Appendix A, Section V, GDC 55.
: 4.
NUREG-75/014, Appendix V, October 1975.
: 5.
NUREG-0677, NRC, May 1980.
: 64. [Document containing list of PIVs.]
: 7.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
: 58. 10 CFR 50.55a(fg).
TSTF-600, Rev. 2

RCS PIV Leakage 3.4.14 Westinghouse STS 3.4.14-3 Rev. 5.0 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.
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
[ [18] months OR 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 TSTF-600, Rev. 2

RCS PIV Leakage 3.4.14 Westinghouse STS 3.4.14-4 Rev. 5.0 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY Within 24 hours following valve actuation due to automatic or manual action or flow through the valve SR 3.4.14.2 
-------------------------------NOTE------------------------------
[ Not required to be met when the RHR System autoclosure interlock is disabled in accordance with SR 3.4.12.7.
Verify RHR System autoclosure interlock prevents the valves from being opened with a simulated or actual RCS pressure signal [425] psig.
[ [18] months OR In accordance with the Surveillance Frequency Control Program ] ]
SR 3.4.14.3 
-------------------------------NOTE------------------------------
[ Not required to be met when the RHR System autoclosure interlock is disabled in accordance with SR 3.4.12.7.
Verify RHR System autoclosure interlock causes the valves to close automatically with a simulated or actual RCS pressure signal [600] psig.
[ [18] months OR In accordance with the Surveillance Frequency Control Program ] ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-1 Rev. 5.0 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, TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-2 Rev. 5.0 BASES BACKGROUND (continued)
: b.
Safety Injection System, and
: c.
Chemical and Volume Control System.
The PIVs are listed in the FSAR, Section [ ] (Ref. 46).
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 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.
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.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-3 Rev. 5.0 BASES LCO (continued)
Reference 68 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 is not sufficient to overpressurize the connected low pressure systemsfor 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 [or the high pressure portion of the system].
Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. 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 hour Completion Time allows the actions and restricts the operation with leaking isolation valves.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-4 Rev. 5.0 BASES ACTIONS (continued)
[ Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing some other valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time after exceeding the limit considers the time required to complete the Action and the low probability of a second valve failing during this time period.
[or]
The 72 hour Completion Time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This timeframe considers the time required to complete this Action and the low probability of a second valve failing during this period. ] 
-----------------------------------REVIEWERS NOTE-----------------------------------
Two options are provided for Required Action A.2. The second option (72 hour restoration) is appropriate if isolation of a second valve would place the unit in an unanalyzed condition.
B.1 and B.2 If leakage cannot be reduced, [the system can not be isolated,] or the other Required Actions accomplished, the plant must be brought to a MODE in which overall plant risk is reduced. To achieve this status, the plant must be brought to MODE 3 within 6 hours and MODE 4 within 12 hours. This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment.
Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 75). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 75, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-5 Rev. 5.0 BASES ACTIONS (continued)
Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.
However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
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.
C.1 The inoperability of the RHR autoclosure interlock renders the RHR suction isolation valves incapable of isolating in response to a high pressure condition and preventing inadvertent opening of the valves at RCS pressures in excess of the RHR systems design pressure. If the RHR autoclosure interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours. This Action accomplishes the purpose of the autoclosure function.
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.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-6 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
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 [18 month]
Frequency is in accordance with the requirements of consistent with 10 CFR 50.55a(fg) (Ref. 69) and the INSERVICE TESTING PROGRAM, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref. 8), 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
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 hours after the valve has been reseated. Within 24 hours 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.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-7 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
[ SR 3.4.14.2 and SR 3.4.14.3 Verifying that the RHR autoclosure interlocks are OPERABLE ensures that RCS pressure will not pressurize the RHR system beyond 125% of its design pressure of [600] psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be 
< [425] psig to open the valves. This setpoint ensures the RHR design pressure will not be exceeded and the RHR relief valves will not lift. [ The
[18] month Frequency is based on the need to perform the Surveillance under conditions that apply during a plant outage. The [18] month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
These SRs are modified by Notes allowing the RHR autoclosure function to be disabled when using the RHR System suction relief valves for cold overpressure protection in accordance with SR 3.4.12.7. ]
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.
[ 46. Document containing list of PIVs. ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Westinghouse STS B 3.4.14-8 Rev. 5.0 BASES REFERENCES (continued)
: 57. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.
: 8.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
: 69. 10 CFR 50.55a(fg).
TSTF-600, Rev. 2

RCS PIV Leakage 3.4.14 Combustion Engineering STS 3.4.14-3 Rev. 5.0 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 SDC 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.
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] psia and [2255] psia.
In accordance with the INSERVICE TESTING PROGRAM, and
[ [18] months OR In accordance with the Surveillance Frequency Control Program ]
AND Prior to entering MODE 2 determine 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 TSTF-600, Rev. 2

RCS PIV Leakage 3.4.14 Combustion Engineering STS 3.4.14-4 Rev. 5.0 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY Within 24 hours following valve actuation due to automatic or manual action or flow through the valve SR 3.4.14.2 
-------------------------------NOTE------------------------------
[ Not required to be met when the SDC System autoclosure interlock is disabled in accordance with SR 3.4.12.7.
Verify SDC System autoclosure interlock prevents the valves from being opened with a simulated or actual RCS pressure signal [425] psig.
[ [18] months OR In accordance with the Surveillance Frequency Control Program ] ]
SR 3.4.14.3 
-------------------------------NOTE------------------------------
[ Not required to be met when the SDC System autoclosure interlock is disabled in accordance with SR 3.4.12.7.
Verify SDC System autoclosure interlock causes the valves to close automatically with a simulated or actual RCS pressure signal [600] psig.
[ [18] months OR In accordance with the Surveillance Frequency Control Program ] ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-1 Rev. 5.0 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 RCS pressure boundary that 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 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 PIVs in series 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 valves in series 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 leakage 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 condition 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:
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-2 Rev. 5.0 BASES BACKGROUND (continued)
: a.
Shutdown Cooling (SDC) System,
: b.
Safety Injection System, and
: c.
Chemical and Volume Control System.
The PIVs are listed in FSAR section (Ref. 46).
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 SDC System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the reactor coolant pressure boundary (RCPB), and the subsequent pressurization of the SDC System downstream of the PIVs from the RCS. Because the low pressure portion of the SDC System is typically designed for [600] psig, overpressurization failure of the SDC 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 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.
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.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-3 Rev. 5.0 BASES LCO (continued)
Reference 57 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 SDC flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the SDC 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 is not sufficient to overpressurize the connected low pressure systemsfor a LOCA outside the containment.
ACTIONS The Actions are modified by two Notes. Note 1 is added to provide clarification that each flow path allows separate entry into a Condition.
This is allowed based on 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 stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be in the RCPB [or the high pressure portion of the system].
Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate if leakage cannot be reduced.
The 4 hours allows the actions and restricts the operation with leaking isolation valves.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-4 Rev. 5.0 BASES ACTIONS (continued)
[ Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing some other valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time after exceeding the limit considers the time required to complete the action and the low probability of a second valve failing during this time period.
or The 72 hour Completion Time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This timeframe considers the time required to complete this Action and the low probability of a second valve failing during this period.] 
-----------------------------------REVIEWERS NOTE-----------------------------------
Two options are provided for Required Action A.2. The second option (72 hour restoration) is appropriate if isolation of a second valve would place the unit in an unanalyzed condition.
B.1 and B.2 If leakage cannot be reduced [the system isolated] or other Required Actions accomplished, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and to MODE 5 within 36 hours. This Action reduces 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.
C.1 The inoperability of the SDC autoclosure interlock renders the SDC suction isolation valves incapable of: isolating in response to a high pressure condition and preventing inadvertent opening of the valves at RCS pressures in excess of the SDC systems design pressure. If the SDC autoclosure interlock is inoperable, operation may continue as long as the affected SDC suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours. This Action accomplishes the purpose of the autoclosure function.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-5 Rev. 5.0 BASES 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 or 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 [18] month Frequency is in accordance with the requirements of consistent with 10 CFR 50.55a(fg) (Ref. 58) and the INSERVICE TESTING PROGRAM and is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref. 7), and is based on the need to perform the Surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
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 hours after the valve has been reseated. Within 24 hours is a reasonable and practical time limit for performing this test after opening or reseating a valve.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-6 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
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 complimentary 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 SDC System when the SDC System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in the SDC shutdown cooling flow path must be leakage rate tested after SDC is secured and stable unit conditions and the necessary differential pressures are established.
SR 3.4.14.2 and SR 3.4.14.3 Verifying that the SDC autoclosure interlocks are OPERABLE ensures that RCS pressure will not pressurize the SDC system beyond 125% of its design pressure of [600] psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be 
< [425] psig to open the valves. This setpoint ensures the SDC design pressure will not be exceeded and the SDC relief valves will not lift. [ The 18 month Frequency is based on the need to perform these Surveillances under conditions that apply during a plant outage. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.14 Combustion Engineering STS B 3.4.14-7 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
The SRs are modified by Notes allowing the SDC autoclosure function to be disabled when using the SDC System suction relief valves for cold overpressure protection in accordance with SR 3.4.12.7.
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.
: 46. [ Document containing list of PIVs. ]
: 7.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
: 58. 10 CFR 50.55a(fg).
TSTF-600, Rev. 2

RCS PIV Leakage 3.4.5 General Electric BWR/4 STS 3.4.5-2 Rev. 5.0 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME A.2 Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, de-activated automatic, or check valve.
72 hours B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
12 hours 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 
-------------------------------NOTE------------------------------
Not required to be performed in MODE 3.
Verify equivalent leakage of each RCS PIV is 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at an RCS pressure [ ] and
[ ] psig.
[ In accordance with the INSERVICE TESTING PROGRAM OR
[ [18] months]
OR In accordance with the Surveillance Frequency Control Program ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.5 General Electric BWR/4 STS B 3.4.5-1 Rev. 5.0 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3). RCS PIVs are defined as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB). PIVs are designed to meet the requirements of Reference 64. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration.
The RCS PIV 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 these valves is not included in any allowable LEAKAGE specified in LCO 3.4.4, "RCS Operational LEAKAGE."
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 event that could degrade the ability for low pressure injection.
A study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.
PIVs are provided to isolate the RCS from the following typically connected systems:
: a.
Residual Heat Removal (RHR) System,
: b.
Core Spray System,
: c.
High Pressure Coolant Injection System, and
: d.
Reactor Core Isolation Cooling System.
The PIVs are listed in Reference 46.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.5 General Electric BWR/4 STS B 3.4.5-2 Rev. 5.0 BASES APPLICABLE Reference 5 evaluated various PIV configurations, leakage testing of the SAFETY valves, and operational changes to determine the effect on the ANALYSES probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
PIV leakage is not considered in any Design Basis Accident analyses.
This Specification provides for monitoring the condition of the RCPB to detect PIV degradation that has the potential to cause a LOCA outside of containment.
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.
RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO RCS PIV leakage is 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. 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.
The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 46).
Reference 67 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). 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, and 3, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 3, valves in the RHR shutdown cooling flow path are not required to meet the requirements of this LCO when in, or during transition to or from, the RHR shutdown cooling mode of operation.
In MODES 4 and 5, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and is not sufficient to overpressurize the connected low pressure systemsfor a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.5 General Electric BWR/4 STS B 3.4.5-3 Rev. 5.0 BASES ACTIONS The ACTIONS are modified by two Notes. Note 1 has been provided to modify the ACTIONS related to RCS PIV flow paths. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV 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. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems.
A.1 and A.2 If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated automatic, or check valve within 4 hours.
Required Action A.1 and Required Action A.2 are modified by a Note stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB [or the high pressure portion of the system].
Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours allows time for these actions and restricts the time of operation with leaking valves.
Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing another valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time considers the time required to complete the action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the low pressure ECCS piping when overpressurized to reactor pressure (Ref. 57).
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.5 General Electric BWR/4 STS B 3.4.5-4 Rev. 5.0 BASES ACTIONS (continued)
B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and MODE 4 within 36 hours. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS Performance of leakage testing on each RCS PIV 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. 
-----------------------------------REVIEWERS NOTE-----------------------------------
If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.
[ The 18 month Frequency is in accordance with the requirements of required by 10 CFR 50.55a(f) (Ref. 6) the INSERVICE TESTING PROGRAM is within the ASME Code Frequency requirement and is based on the need to perform this Surveillance during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
OR The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.5 General Electric BWR/4 STS B 3.4.5-5 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued) 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
This SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3. Entry into MODE 3 is permitted for leakage testing at high differential pressures with stable conditions not possible in the lower MODES.
REFERENCES
: 1.
10 CFR 50.2.
: 2.
10 CFR 50.55a(c).
: 3.
10 CFR 50, Appendix A, GDC 55.
: 4.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
: 5.
NUREG-0677, May 1980.
: 46. FSAR, Section [ ].
: 57. NEDC-31339, November 1986.
: 6.
10 CFR 50.55a(f).
TSTF-600, Rev. 2

RCS PIV Leakage 3.4.6 General Electric BWR/6 STS 3.4.6-2 Rev. 5.0 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME A.2 Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.
72 hours B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
12 hours 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 
-------------------------------NOTE------------------------------
Not required to be performed in MODE 3.
Verify equivalent leakage of each RCS PIV is 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at an RCS pressure
[1040] psig and [1060] psig.
[ In accordance with INSERVICE TESTING PROGRAM OR
((18] months OR In accordance with the Surveillance Frequency Control Program ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-1 Rev. 5.0 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND RCS PIVs are defined as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB). The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3). PIVs are designed to meet the requirements of Reference 64. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration.
The RCS PIV 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 these valves is not included in any allowable LEAKAGE specified in LCO 3.4.5, "RCS Operational LEAKAGE."
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 which could degrade the ability for low pressure injection.
A study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.
PIVs are provided to isolate the RCS from the following typically connected systems:
: a.
Residual Heat Removal (RHR) System,
: b.
Low Pressure Core Spray System,
: c.
High Pressure Core Spray System, and
: d.
Reactor Core Isolation Cooling System.
The PIVs are listed in Reference 46.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-2 Rev. 5.0 BASES APPLICABLE Reference 5 evaluated various PIV configurations, leakage testing of the SAFETY valves, and operational changes to determine the effect on the ANALYSES probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
PIV leakage is not considered in any Design Basis Accident analyses.
This Specification provides for monitoring the condition of the RCPB to detect PIV degradation that has the potential to cause a LOCA outside of containment.
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.
RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO RCS PIV leakage is 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. 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.
The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 64).
Reference 67 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). 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, and 3, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 3, valves in the RHR flowpath are not required to meet the requirements of this LCO when in, or during transition to or from, the RHR shutdown cooling mode of operation.
In MODES 4 and 5, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and is not sufficient to overpressurize the connected low pressure systemsfor a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-3 Rev. 5.0 ACTIONS The ACTIONS are modified by two Notes. Note 1 has been provided to modify the ACTIONS related to RCS PIV flow paths. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-4 Rev. 5.0 BASES ACTIONS (continued) result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV 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. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems.
A.1 and A.2 If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated, automatic, or check valve within 4 hours. Required Action A.1 and Required Action A.2 are modified by a Note stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB [or the high pressure portion of the system.]
Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours allows time for these actions and restricts the time of operation with leaking valves.
Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing another valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time after exceeding the limit considers the time required to complete the Required Action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the low pressure ECCS piping when overpressurized to reactor pressure (Ref. 57).
B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and to MODE 4 within 36 hours. This action may reduce the leakage and also TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-5 Rev. 5.0 BASES ACTIONS (continued) reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.6.1 REQUIREMENTS Performance of leakage testing on each RCS PIV 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. 
-----------------------------------REVIEWERS NOTE-----------------------------------
If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.
[ The 18 month Frequency is in accordance with the requirements of required by 10 CFR 50.55a(f) (Ref. 6)the INSERVICE TESTING PROGRAM is within the ASME Code Frequency requirement and is based on the need to perform this Surveillance 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.
OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. 
-----------------------------------REVIEWERS NOTE-----------------------------------
Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement. 
------------------------------------------------------------------------------------------------ ]
TSTF-600, Rev. 2

RCS PIV Leakage B 3.4.6 General Electric BWR/6 STS B 3.4.6-6 Rev. 5.0 BASES SURVEILLANCE REQUIREMENTS (continued)
Therefore, thisThis SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3. Entry into MODE 3 is permitted for leakage testing at high differential pressures with stable conditions not possible in the lower MODES.
REFERENCES
: 1.
10 CFR 50.2.
: 2.
10 CFR 50.55a(c).
: 3.
10 CFR 50, Appendix A, GDC 55.
: 4.
ASME Code for Operation and Maintenance of Nuclear Power Plants.
: 5.
NUREG-0677, May 1980.
: 46. FSAR, Section [ ].
: 57. NEDC-31339, November 1986.
: 6.
10 CFR 50.55a(f).
TSTF-600, Rev. 2}}