Text

License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies
	ML23032A162
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Robinson, McGuire
Issue date:	02/01/2023
From:	Gibby S; Duke Energy Carolinas, Duke Energy Progress
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	RA-22-0118
	Download: ML23032A162 (1)
	v • d • e

Shawn Gibby Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 704-519-5136 Shawn.Gibby@duke-energy.com 10 CFR 50.90 February 1, 2023 Serial: RA-22-0118 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 / RENEWED LICENSE NOS. NPF-35 AND NPF-52 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287 / RENEWED LICENSE NOS. DPR-38, DPR-47, AND DPR-55 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63

Subject:

License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, collectively referred to henceforth as Duke Energy, is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station, Units 1 and 2 (CNS), McGuire Nuclear Station, Units 1 and 2 (MNS), Oconee Nuclear Station, Units 1, 2, and 3 (ONS), H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP), and Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment would revise the Surveillance Requirement (SR) Frequency for Reactor Coolant System pressure isolation valve operational leakage testing to reflect being in accordance with the Inservice Testing Program, as governed by 10 CFR 50.55a. Specifically, this change will update TS SR 3.4.14.1 for CNS, MNS, ONS, and RNP and TS SR 4.4.6.2.2 for HNP. An additional revision is proposed to CNS and MNS TS SR 3.3.1.8 to remove restrictive surveillance Frequency content that impedes the full application of the Surveillance Frequency Control Program to establish the Frequency for performance of the Channel Operational Test of select Reactor Trip System instrumentation.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been concluded that the proposed changes involve no

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial: RA-22-0118 significant hazards consideration.

The enclosure to this license amendment request provides Duke Energys evaluation of the proposed changes. In addition, Attachment 1 to the enclosure provides a copy of the existing TS pages marked with the proposed changes. Attachment 2 contains the existing TS Bases pages marked to show the proposed changes for information only. Changes to the TS Bases will be implemented in accordance with the TS Bases Control Program for each site upon implementation of the respective amendment.

Approval of the proposed license amendment is requested within one year of acceptance. The amendment shall be implemented within 120 days from approval.

In accordance with 10 CFR 50.91, a copy of this application, with enclosure, is being provided to the designated North Carolina and South Carolina Officials.

This letter contains no regulatory commitments.

Please refer any questions regarding this submittal to Ryan Treadway, Director - Nuclear Fleet Licensing, at (980) 373-5873.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on February 1, 2023.

Sincerely, Shawn Gibby Vice President - Nuclear Engineering

Enclosure:

Evaluation of the Proposed Change Attachments: 1. Proposed Technical Specification Changes (Mark-up)

2. Proposed Technical Specification Bases Changes (Mark-up) cc: L. Dudes, USNRC Region II - Regional Administrator A. Donley, USNRC Senior Resident Inspector - CNS D. Rivard, USNRC Resident Inspector - CNS G. Hutto, USNRC Senior Resident Inspector - MNS J. Nadel, USNRC Senior Resident Inspector - ONS P. Boguszewski, USNRC Senior Resident Inspector - HNP J. Zeiler, USNRC Senior Resident Inspector - RNP N. Jordan, USNRC NRR Project Manager - Fleet D. Crowley, Radioactive Materials Branch Manager, N.C. DHSR S. Jenkins, Chief, Bureau of Radiological Health (S.C.)

L. Garner, Manager, S.C. DHEC

U.S. Nuclear Regulatory Commission Serial: RA-22-0118 Enclosure ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 16 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 16 Serial: RA-22-0118 Enclosure Evaluation of the Proposed Change License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, collectively referred to henceforth as Duke Energy, is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station, Units 1 and 2 (CNS), McGuire Nuclear Station, Units 1 and 2 (MNS), Oconee Nuclear Station, Units 1, 2, and 3 (ONS), H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP), and Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment would revise the Surveillance Requirement (SR) Frequency for Reactor Coolant System (RCS) pressure isolation valve (PIV) operational leakage testing to reflect being in accordance with the Inservice Testing (IST) Program, as governed by 10 CFR 50.55a. Specifically, this change will update TS SR 3.4.14.1 for CNS, MNS, ONS, and RNP and TS SR 4.4.6.2.2 for HNP. An additional revision is proposed to CNS and MNS TS SR 3.3.1.8 to remove restrictive surveillance Frequency content that impedes the full application of the Surveillance Frequency Control Program (SFCP) to establish the Frequency for performance of the Channel Operational Test (COT) of select Reactor Trip System (RTS) instrumentation.

2.0 DETAILED DESCRIPTION

2.1 Background

System Design and Operation RCS Pressure Isolation Valves The RCS PIVs are defined by 10 CFR 50.2, Definitions, 10 CFR 50.55a(c), Reactor coolant pressure boundary, and General Design Criteria (GDC) 55, Reactor coolant pressure boundary penetrating containment, of 10 CFR 50, Appendix A as two normally closed valves in series within the reactor coolant pressure boundary, which separate the high pressure RCS from attached lower pressure systems. Examples of these lower pressure systems include the residual heat removal system (RHR), the safety injection system (SI), and the chemical and volume control system. Failure or excessive PIV leakage could lead to overpressurization of the low-pressure piping or components that could lead to a system rupture, potentially resulting in a loss of coolant accident (LOCA) outside of containment and loss in integrity of a fission product barrier.

Per the requirements of 10 CFR 50.55a, RCS PIVs have to be leak tested in accordance with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code).

RTS Instrumentation The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and RCS pressure boundary during anticipated

U.S. Nuclear Regulatory Commission Page 2 of 16 Serial: RA-22-0118 Enclosure operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents. The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings in terms of parameters directly monitored by the RTS, as well as specifying limiting conditions for operation (LCOs) on other reactor system parameters and equipment performance.

As defined in TS 1.1, Definitions, a COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the operability of required alarm, interlock, and trip functions. The COT includes any necessary adjustments of the required alarm, interlock, and trip setpoints such that the setpoints are within the required range and accuracy.

Surveillance Frequency Control Program With the implementation of Technical Specification Task Force (TSTF)-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b (ADAMS Accession No. ML090850642) by each of the sites, most periodic frequencies of TS surveillances were relocated to a licensee-controlled program, the SFCP, along with the inclusion of the requirements for the new program in the Administrative Controls Section of the TS. The following license amendments addressed the NRCs approval for each respective sites implementation of TSTF-425:

CNS - Amendment Nos. 263 and 259, per letter dated March 29, 2011 (ADAMS Accession No. ML110670536)

HNP - Amendment No. 154, per letter dated November 29, 2016 (ADAMS Accession No. ML16200A285)

MNS - Amendment Nos. 261 and 241, per letter dated March 29, 2011 (ADAMS Accession No. ML110680357)

ONS - Amendment Nos. 372, 374, and 373, per letter dated March 21, 2011 (ADAMS Accession No. ML110470446)

RNP - Amendment No. 265, per letter dated August 15, 2019 (ADAMS Accession No. ML19158A307 The SFCP describes the requirements for the program to control changes to the relocated surveillance frequencies, establishing Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, April 2007 (ADAMS Accession No. ML071360456), as the basis for making any changes to the surveillance frequencies once they are relocated out of the TS. The NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing per letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The technical methodology provided in NEI 04-10 uses a risk-informed, performance-based approach for establishment of surveillance frequencies that is consistent with the philosophy of Regulatory Guide 1.174. Specifically, the use of Probabilistic Risk Assessment (PRA) methods are employed to determine the risk impact of the revised intervals.

Additionally, TSTF-425 established that all surveillance frequencies can be relocated except:

frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

U.S. Nuclear Regulatory Commission Page 3 of 16 Serial: RA-22-0118 Enclosure frequencies that are purely event-driven (e.g., each time the control rod is withdrawn to the full out position);

frequencies that are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching 95% RTP [Rated Thermal Power]); and frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., drywell to suppression chamber differential pressure decrease).

2.2 Current TS Requirements RCS PIVs The existing TS SR 3.4.14.1 for CNS and MNS states:

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

Frequency In accordance with the INSERVICE TESTING PROGRAM, and in accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve The existing TS SR 3.4.14.1 for ONS states:

Surveillance Verify leakage from each require 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 2150 psia and 2190 psia.

Frequency 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, if leakage testing has not been performed in the previous 9 months.

The existing TS SR 3.4.14.1 for RNP states:

Surveillance Verify leakage from each RCS PIV is less than or equal to an equivalent of 5 gpm at an RCS pressure 2235 psig, and verify the margin between the results of the previous

U.S. Nuclear Regulatory Commission Page 4 of 16 Serial: RA-22-0118 Enclosure leak rate test and the 5 gpm limit has not been reduced by 50% for valves with leakage rates > 1.0 gpm.

Frequency In accordance with the INSERVICE TESTING PROGRAM and in accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve The existing TS SR 4.4.6.2.2 for HNP states:

Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At the frequency specified in the Surveillance Frequency Control Program,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and

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

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

The following table provides a simplified view of which Frequencies impact each site.

Duke Energy Site Frequency CNS MNS ONS RNP HNP In accordance with the INSERVICE TESTING X X X PROGRAM In accordance with the Surveillance Frequency X X X X X Control Program Prior to entering MODE 2 whenever the plant has X X X X X been in MODE 5/COLD SHUTDOWN for 7 days or more, if leakage testing has not been performed in the previous 9 months Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to X X X X automatic or manual action or flow through the valve

U.S. Nuclear Regulatory Commission Page 5 of 16 Serial: RA-22-0118 Enclosure Reactor Trip System Instrumentation The existing TS SR 3.3.1.8 for CNS and MNS states:

Surveillance

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

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

Perform COT.

Frequency

NOTE------------------------------------------------------

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program or [the] previous 184 days Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program Of note, the use of brackets around the word the above is to illustrate that this word is present in the CNS version of the TS SR, but not in the version for MNS. This is an editorial difference that will be removed with the proposed amendment.

2.3 Reason for the Proposed Change TS SR 3.4.14.1 (TS SR 4.4.6.2.2 for HNP)

The current TS SR 3.4.14.1 for CNS, MNS, ONS and RNP and TS SR 4.4.6.2.2 for HNP contain certain testing requirements that are also required by the ASME OM Code and are implemented in each sites respective IST Program. Performance of leakage rate testing for PIVs is per the requirements of the IST Program, which is governed by 10 CFR 50.55a. In its current state, the respective site TS SR points to the IST Program or SFCP as the basis for the testing Frequency, which is currently on a reload Frequency. Despite the IST Program Frequency being based on the governance of 10 CFR 50.55a, the site TS SR reflect a Frequency requirement of any cold shutdown for 7 days or more, effectively restricting the ability of the IST Program, as well as the SFCP, to govern the Frequency in which the leakage testing is performed. The proposed change to each sites TS SR Frequency would ensure ASME OM

U.S. Nuclear Regulatory Commission Page 6 of 16 Serial: RA-22-0118 Enclosure 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.

Furthermore, in updating the surveillance Frequency for RCS PIV leakage testing, there is the added benefit of potentially lowering outage dose and reducing outage time by eliminating unnecessary testing.

MNS and CNS TS SR 3.3.1.8 The NRC previously approved of the adoption of TSTF-425 and the establishment of a SFCP for use by CNS and MNS. Prior to adopting TSTF-425, TS SR 3.3.1.8 contained a NOTE in the Frequency column that stated, Only required when not performed within previous 184 days.

That note was identical to the revision of the Westinghouse Improved Standard Technical Specifications (ISTS) at the time (i.e., Revision 3 of NUREG-1431, Standard Technical Specifications Westinghouse Plants). The intent of the note was to perform the SR prior to startup unless it had been performed within the normal Frequency, which at the time had been 184 days. The proposed TS markup for SR 3.3.1.8 included in TSTF-425 stated, Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program OR previous 184 days. However, the inclusion of OR previous 184 days restricts the application of the SFCP to establish a new normal Frequency beyond 184 days that applies to the note. As such, MNS and CNS are required to perform a COT more frequently than the Frequency that has been established in accordance with the SFCP. Removing this extraneous content from the TS SR note will allow for the full application of the SFCP and elimination of unnecessary performances of a COT on certain RTS instrumentation.

2.4 Description of the Proposed Change TS SR 3.4.14.1 (TS SR 4.4.6.2.2 for HNP)

The frequencies for the SRs provided above will be revised to reflect the removal of the requirement that the surveillance be performed prior to entering MODE 2 when the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. For those sites with a requirement to perform the surveillance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of valve actuation due to automatic or manual action or flow through the valve, this content is also being removed. Additionally, content related to performing the surveillance at a Frequency in accordance with the SFCP is being removed or replaced to reflect reliance upon the Frequency established by the IST Program for each site.

The CNS and MNS SR 3.4.14.1 will be revised to reflect a Frequency of:

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

U.S. Nuclear Regulatory Commission Page 7 of 16 Serial: RA-22-0118 Enclosure Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve The ONS SR 3.4.14.1 will be revised to reflect a Frequency of:

Frequency In accordance with the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days, if leakage testing has not been performed in the previous 9 months The RNP SR 3.4.14.1 will be revised to reflect a Frequency of:

Frequency In accordance with the INSERVICE TESTING PROGRAM and in accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve The HNP SR 4.4.6.2.2 will be revised as follows:

Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At the frequency specified in the Surveillance Frequency Control Program, In accordance with the INSERVICE TESTING PROGRAM.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, [DELETED]

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and [DELETED]

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

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

Additionally, the proposed deletion of SR 4.4.6.2.2.b and SR 4.4.6.2.2.d will be reflected administratively in HNP TS Table 3.4-1 with the removal of the following note:

Specifications 4.4.6.2.2.b. and d. do not apply to these valves.

U.S. Nuclear Regulatory Commission Page 8 of 16 Serial: RA-22-0118 Enclosure MNS and CNS TS SR 3.3.1.8 The note associated with the Frequency of CNS and MNS SR 3.3.1.8 will be revised to reflect the removal of content related to the previous 184 days as follows:

Frequency

NOTE------------------------------------------------------

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program or [the] previous 184 days

3.0 TECHNICAL EVALUATION

TS SR 3.4.14.1 (TS SR 4.4.6.2.2 for HNP)

Paragraph (a) of 10 CFR 50.55a, "Documents approved for incorporation by reference,"

references the approved documents required to be followed for systems and components of nuclear power reactors. Specifically, paragraph (a)(1)(iv) references the ASME OM Code, which provides requirements for inservice testing of certain components in light-water nuclear power plants. The OM Code was developed and is maintained by the ASME OM Committee, which publishes a new edition of the OM Code periodically. The ASME OM Code identifies the components subject to testing as well as the required methods, intervals, and parameters to be measured and evaluated. It also identifies the criteria for evaluating results, associated corrective actions, personnel qualification requirements, and requirements for record keeping and submittal of reports.

The proposed change would remove overly restrictive and duplicative surveillance Frequency requirements from each sites TS pertaining to verifying that RCS PIV leakage is within its respective limit. Despite the current reliance on the SFCP and/or IST Program to establish the associated Frequency for these surveillances, there remains a requirement that the surveillance be performed prior to entering Mode 2 whenever the plant has been in cold shutdown for 7 days or more and if leakage testing has not been performed in the previous 9 months. This additional Frequency requirement effectively restricts the ability of the SFCP to extend frequencies beyond a singular refueling outage (i.e., 18 or 24 months).

Additionally, even with the ability to establish the surveillance Frequency in accordance with the SFCP, extensions beyond the ASME OM Code Subsection ISTC-3630 paragraph (a) requirement that leakage rate testing for PIVs be performed at least once every two years requires the submittal of a relief request per 10 CFR 50.55a, paragraph (z), Alternatives to codes and standards requirements.

The sites (except ONS) also have a requirement to perform this surveillance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve. 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 that would result from gross failure of the associated PIVs. The leakage limit 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

U.S. Nuclear Regulatory Commission Page 9 of 16 Serial: RA-22-0118 Enclosure per ASME OM Code Subsection ISTC-3700) provides an acceptable method of determining valve integrity. 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 does not contribute any additional assurance of pressure isolation functional capability but rather provides added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA. In removing the Frequency requiring performance of leakage testing within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valves, the sites will instead test 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 valve integrity, particularly as it relates to operational readiness. Furthermore, this change in frequency would also align with the corresponding SR 3.4.5.1 in NUREG-1433, Standard Technical Specifications - General Electric BWR [Boiling Water Reactor]/4 Plants, SR 3.4.6.1 in NUREG-1434, Standard Technical Specifications - General Electric BWR/6 Plants, and SR 3.4.15.1 in NUREG-2194, Standard Technical Specifications - Westinghouse Advanced Passive 1000 (AP1000) Plants.

For HNP only, there is a Frequency requirement that leakage be verified to be within its limit prior to returning the valve to service following maintenance, repair, or replacement work on the valve. This is redundant to the requirement in ASME OM Code Subsection ISTC-3310 that an inservice test be run when a valve or its control system has been replaced, repaired, or undergone maintenance that could affect the valves performance before it can be returned to service.

As documented in the NRC-issued Enforcement Guidance Memorandum (EGM) 12-001, Dispositioning Noncompliance with Administrative Controls Technical Specifications Programmatic Requirements that Extend Test Frequencies and Allow Performance of Missed Tests, dated February 24, 2012 (ADAMS Accession No. ML11258A243), the staff evaluation of inservice testing requirements under 10 CFR 50.55a determined that, if a licensee finds that the requirements of TS conflict with the requirements of 10 CFR 50.55a, then the licensee must amend their TS to comply with 10 CFR 50.55a. While EGM 12-001 was withdrawn per letter dated February 14, 2018 (ADAMS Accession No. ML18016A475) with the expiration of the period of enforcement discretion, the position related to conflicting requirements between TS and 10 CFR 50.55a is further supported by regulation in 10 CFR 50.55a(f)(5)(ii), IST program update: Conflicting IST Code requirements with technical specifications, which states in part that if a revised inservice test program for a facility conflicts 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 proposed change is consistent with NRC staff positions and the regulations with respect to conflicting requirements.

By letter dated August 15, 2018 (ADAMS Accession No. ML18172A172), the NRC issued license amendments to CNS (Amendment Nos. 299 and 295), HNP (Amendment No. 166),

MNS (Amendment Nos. 309 and 288), ONS (Amendment Nos. 409, 411, and 410), and RNP (Amendment No. 259), to adopt TSTF Traveler 545, "TS lnservice Testing Program Removal &

Clarify SR Usage Rule Application to Section 5.5 Testing" (ADAMS Accession No. ML15294A555). Consistent with TSTF-545, these amendments removed the IST detailed

U.S. Nuclear Regulatory Commission Page 10 of 16 Serial: RA-22-0118 Enclosure 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)." In Section 3.0 of the associated safety evaluation for these amendments, it is stated:

In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

In updating the surveillance frequencies for verifying RCS PIV leakage to remove the restrictive and redundant requirements, the frequencies 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).

Of note, the proposed change to each sites respective TS SR Frequency for RCS PIV leakage testing would not impact the TS PIV allowable leakage limits of each site (i.e., the Surveillance Requirement itself). With the proposed change, testing of the RCS PIVs would still be performed in accordance with the IST Program in order to provide a means of assessing whether the PIVs between the RCS and the connecting systems are degraded or degrading.

Furthermore, 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 applicants or licensees as an alternative to compliance with ASME Code provisions that have been incorporated by reference into 10 CFR 50.55a. In aligning the Frequency of the RCS PIV leakage verification SR with the IST Program, it allows for the potential 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, Regulatory Guide (RG) 1.192, Revision 4, Operation and Maintenance Code Case Acceptability, ASME OM Code (ADAMS Accession No. ML21181A223), became effective on April 4, 2022, as documented in the final rule, Approval of American Society of Mechanical Engineers Code Cases, issued in the Federal Register (FR) on March 3, 2022 (87 FR 11934). This RG includes information reviewed by the NRC on OM Code Cases listed in the 2020 Edition of the OM Code and on the ASME Codes & Standards (C&S) Connect website.

One code case in particular, Code Case OMN-23, Alternative Requirements for Testing

U.S. Nuclear Regulatory Commission Page 11 of 16 Serial: RA-22-0118 Enclosure Pressure Isolation Valves, was approved by the NRC for use without conditions and establishes requirements for implementing and maintaining a condition monitoring program for RCS PIVs that may be implemented in lieu of the 2-year leakage rate test Frequency specified in Section ISTC-3630(a). It does not replace or exclude any test method requirement specified in ISTC-3630, but does allow for the possibility of 24-month interval extensions, not to exceed a maximum interval of 6 years, dependent on analysis of tests results and maintenance history of the RCS PIVs.

Given that none of the sites IST Programs are currently based on the 2020 Edition of the OM Code, relief would need to be requested in accordance with 10 CFR 50.55a(z) to take full advantage of NRC-approved Code Case OMN-23. However, updates to the IST Programs for the sites are outside the scope of this license amendment request and will be addressed once the leakage verification tests are no longer restricted by the redundant testing requirements and the Frequency to perform leakage testing prior to entering Mode 2 whenever the unit has been in cold shutdown for 7 days is eliminated.

MNS and CNS TS SR 3.3.1.8 As discussed previously, the NRC issued license amendments to MNS and CNS allowing for the relocation of specific surveillance frequencies to the SFCP, a licensee-controlled program.

In relocating these surveillance frequencies, changes could be made in accordance with NEI 04-10, Revision 1, as consistent with the NRC-approved TSTF-425, Revision 3. In terms of which surveillance frequencies could be relocated to the SFCP, TSTF-425 established that all surveillance frequencies could be relocated except those that: 1) reference other approved programs for the specific interval; 2) are purely event-driven; 3) are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs; and 4) are related to specific conditions or conditions for the performance of a surveillance requirement. It was also established that the surveillance Frequency associated with the performance of a COT in NUREG-1431 ISTS SR 3.3.1.8 could be relocated to the SFCP.

While TSTF-425 relocated the 184-day surveillance Frequency for performance of the COT from ISTS SR 3.3.1.8 to the SFCP, it inadvertently left it in the corresponding note. As such, it needlessly restricts the ability to utilize the SFCP to control the surveillance Frequency in accordance with NEI 04-10, Revision 1, beyond the stated 184 days. In its current state, a COT is required to be performed prior to any reactor startup that falls outside the 184 days from the previous performance of TS SR 3.3.1.8. The proposed change to MNS and CNS TS SR 3.3.1.8 would remove the extra content from the surveillance Frequency note and allow the SFCP to establish the Frequency in accordance with the program requirements previously approved by the NRC.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Guidance 10 CFR 50.36 The NRC's regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TS include items in the following five specific categories: (1) safety

U.S. Nuclear Regulatory Commission Page 12 of 16 Serial: RA-22-0118 Enclosure limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) SRs, (4) design features, and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TS.

Per 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

10 CFR 50.55a In accordance with 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include IST of pumps and valves at nuclear power reactors pursuant to the ASME OM Code as specified in 10 CFR 50.55a(f). Paragraph (f) of 10 CFR 50.55a states in-part that systems and components of pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel (BPV) Code and ASME OM Code. The ASME OM Code, a consensus standard incorporated by reference into 10 CFR 50.55a, was reviewed by NRC staff during the incorporation process to ensure the OM Code requirements were technically sufficient. The NRC staffs review for technical sufficiency found that the ASME OM Code IST program requirements were suitable for incorporation into the NRC's rules.

Paragraph 50.55(a)(f)(5)(ii) of 10 CFR states in part that if a revised IST Program for a facility conflicts with the TS for the facility, the licensee must apply for an amendment of the TS to conform the TS to the revised program.

10 CFR Part 50, Appendix A, General Design Criteria (GDC) 1, 14, 32, 54 Appendix A, GDC 1, Quality standards and records, states in part:

Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions.

Appendix A, GDC 14, Reactor coolant pressure boundary, states:

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

Appendix A, GDC 32, Inspection of reactor coolant pressure boundary, states:

Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess

U.S. Nuclear Regulatory Commission Page 13 of 16 Serial: RA-22-0118 Enclosure their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Appendix A, GDC 54, Piping systems penetrating containment, states:

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

While ONS and RNP were not licensed to 10 CFR 50, Appendix A, GDC, conformance to the applicable GDC is discussed in Section 3.1 of each sites respective Updated Final Safety Analysis Report.

Conclusion Duke Energy has evaluated the proposed changes against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance that the health and safety of the public will remain unaffected following the approval of the proposed changes.

4.2 Precedents TS SR 3.4.14.1 (TS SR 4.4.6.2.2 for HNP)

While a precedent was not identified for this proposed change, it should be noted that Beaver Valley Power Station, Unit 1, was issued Amendment No. 124 to Facility Operating License No.

DPR-66 per letter dated April 25, 1988 (ADAMS Accession No. ML003767331). This amendment imposed additional and revised requirements on RCS PIVs in accordance with Generic Letter 87-06. Noticeably missing from these requirements was the 24-hour Frequency for leak testing following valve actuation due to automatic or manual action or flow through the valves. Beaver Valley Power Station, Unit 1, is a Westinghouse PWR similar to CNS, MNS, HNP, and RNP.

MNS and CNS TS SR 3.3.1.8 Many sites that were issued license amendments to implement TSTF-425, Revision 3, proposed and obtained NRC approval of a Note in the Frequency column for SR 3.3.1.8 (COT) that stated the Surveillance was only required when not performed within the Frequency specified in the SFCP. The part of the Note stating or the previous 184 days was removed upon issuance of those license amendments. Recent examples include:

H. B. Robinson Steam Electric Plant, Unit No. 2, Amendment No. 265 to Renewed Facility Operating License No. DPR-23 per letter dated August 15, 2019 (ADAMS Accession No. ML19158A307)

U.S. Nuclear Regulatory Commission Page 14 of 16 Serial: RA-22-0118 Enclosure Wolf Creek Generating Station, Unit 1, Amendment No. 227 to Renewed Facility Operating License No. NPF-42 per letter dated April 8, 2021 (ADAMS Accession No. ML21053A117).

4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, collectively referred to henceforth as Duke Energy, hereby requests a revision to the Technical Specifications (TS) for Catawba Nuclear Station, Units 1 and 2 (CNS), McGuire Nuclear Station, Units 1 and 2 (MNS), Oconee Nuclear Station, Units 1, 2, and 3 (ONS), H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP), and Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment would revise the Surveillance Requirement (SR) Frequency for Reactor Coolant System (RCS) pressure isolation valve (PIV) operational leakage testing to reflect being in accordance with the Inservice Testing (IST) Program, as governed by 10 CFR 50.55a.

Specifically, this change will update TS SR 3.4.14.1 for CNS, MNS, ONS, and RNP and TS SR 4.4.6.2.2 for HNP. An additional revision is proposed to CNS and MNS TS SR 3.3.1.8 to remove restrictive surveillance Frequency content that impedes the full application of the Surveillance Frequency Control Program (SFCP) to establish the Frequency for performance of the Channel Operational Test (COT) of select Reactor Trip System (RTS) instrumentation.

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed amendment addresses revising TS SR wording related to surveillance frequencies that are overly restrictive and redundant. The proposed change to the RCS PIV TS SR addresses revising the surveillance testing interval for leakage testing of the RCS PIVs to align it with the IST Program, which is governed by Title 10, Code of Federal Regulations (CFR), Section 50.55a, "Codes and standards." The RCS PIVs will continue to be tested to ensure they are within the TS allowable leakage limits, which remain unchanged as part of this amendment. The proposed change to MNS and CNS TS SR 3.3.1.8 removes extraneous content from the Note associated with the Frequency of performing the COT for select RTS instrumentation. Neither change involves a physical change to the plant or a change in the manner in which the plant is operated or controlled.

The TS SRs associated with the proposed changes are not initiators of any previously evaluated accidents. As such, the probability of each of these previously evaluated accidents is not affected by the proposed changes.

The proposed changes do not alter the design, function, or operation of any plant structure, system, or component (SSC). The proposed changes do not alter or prevent the ability of any TS-required SSC from performing its intended function to mitigate the consequences on an initiating event with the assumed acceptance limits. The proposed

U.S. Nuclear Regulatory Commission Page 15 of 16 Serial: RA-22-0118 Enclosure changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure. As a result, the outcomes of accidents previously evaluated are unaffected.

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

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

Response: No.

The proposed amendment addresses revising TS SR wording related to surveillance frequencies that are overly restrictive and redundant. The proposed change to the RCS PIV TS SR addresses revising the surveillance testing interval for leakage testing of the RCS PIVs to align it with the IST Program, which is governed by 10 CFR 50.55a. The proposed change to MNS and CNS TS SR 3.3.1.8 removes extraneous content from the Note associated with the Frequency of performing the COT for select RTS instrumentation. The technical testing methodology and associated acceptance criteria remain unchanged. The testing requirements involved to periodically demonstrate the integrity of the SSCs exist to ensure the plant's capability to mitigate the consequences of an accident. There are not any accident initiators or precursors affected by these changes. The proposed TS changes do not involve a physical change to the plant or the manner in which the plant is operated or controlled.

The proposed changes neither install or remove any plant equipment, nor alter the design, physical configuration, or mode of operation of any plant SSC. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the Updated Final Safety Analysis Report. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Specifically, no new hardware is being added to the plant as part of the proposed changes, no existing equipment design or function is being modified, and no significant changes in operations are being introduced. No new equipment performance burdens are imposed.

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

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

Response: No.

The proposed amendment addresses revising TS SR wording related to surveillance frequencies that are overly restrictive and redundant. The proposed change to the RCS PIV TS SR addresses revising the surveillance testing interval for leakage testing of the RCS PIVs to align it with the IST Program, which is governed by 10 CFR 50.55a. The

U.S. Nuclear Regulatory Commission Page 16 of 16 Serial: RA-22-0118 Enclosure proposed change to MNS and CNS TS SR 3.3.1.8 removes extraneous content from the Note associated with the Frequency of performing the COT for select RTS instrumentation. The technical testing methodology and associated TS allowable leakage limits/acceptance criteria remain unchanged. The proposed changes in this license amendment request do not alter the design, configuration, operation, or function of any plant SSC. The ability of any operable SSC to perform its designated safety function is unaffected by these changes. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. They do not alter any safety analysis assumptions, initial conditions, or results of any accident analyses. The design, operation, testing methods, and acceptance criteria for the SSCs will continue to be met. The proposed changes will not result in plant operation in a configuration outside the design basis.

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

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

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

5.0 ENVIRONMENTAL CONSIDERATION

S Duke Energy 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 by 10 CFR 20, or it would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent 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 needs be prepared in connection with the proposed amendment.

U.S. Nuclear Regulatory Commission Serial: RA-22-0118 Enclosure ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) 9 PAGES PLUS THE COVER

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.8 ----------------------------------NOTE--------------------------------

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

---------NOTE-------

Only required Perform COT. when not performed within the Frequency specified in the Surveillance Frequency Control Program or the previous 184 days Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.1-11 Amendment Nos. 263/259

RCS PIV Leakage 3.4.14 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. In accordance with

the INSERVICE TESTING Verify leakage from each RCS PIV is equivalent to < 0.5 PROGRAM, and gpm per nominal inch of valve size up to a maximum of 5 in accordance with gpm at an RCS pressure > 2215 psig and < 2255 psig. 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 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve (continued)

Catawba Units 1 and 2 3.4.14-3 Amendment Nos. 299/295

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated, at the frequency specified in the Surveillance Frequency Control Program, to be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or Particulate Radioactivity Monitor;

b. Monitoring the containment sump inventory and Flow Monitoring System;

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 +/- 20 psig with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance*; and

e. Monitoring the Reactor Head Flange Leakoff System.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

B At the frequency specified in the Surveillance Frequency Control Program, C Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for

7 days or more and if leakage testing has not been performed in the previous 9

months, (DELETED)

D Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and (DELETED)

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

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.6.2.3 Primary-to-secondary leakage shall be verified to be 150 gallons per day through any one steam generator at the frequency specified in the Surveillance Frequency Control Program **.

InaccordancewiththeINSERVICETESTINGPROGRAM

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady-state operation. Not applicable to primary-to-secondary leakage.

- Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady-state operation.

SHEARON HARRIS - UNIT 1 3/4 4-24 Amendment No. 154

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.8 ------------------------------NOTES----------------------------------

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

Perform COT. ---------NOTE-------

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program or previous 184 days Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program (continued)

McGuire Units 1 and 2 3.3.1-11 Amendment Nos. 261/241

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ------------------------------NOTE------------------------------------

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 In accordance with gpm per nominal inch of valve size up to a maximum of 5 the INSERVICE gpm at an RCS pressure > 2215 psig and < 2255 psig. TESTING PROGRAM, and in accordance with the Surveillance Frequency Control Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve (continued)

McGuire Units 1 and 2 3.4.14-3 Amendment Nos. 309/288

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RCS PIVs 3.4.14 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 less than or In accordance with equal to an equivalent of 5 gpm at an RCS pressure the INSERVICE 2235 psig, and verify the margin between the TESTING results of the previous leak rate test and the 5 gpm PROGRAM and In limit has not beHQUHGXFHGE\ 50% for valves with accordance with leakage rates > 1.0 gpm. 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 (continued)

HBRSEP Unit No. 2 3.4-39 Amendment No. 265

RCS PIVs 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 (continued) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve SR 3.4.14.2 Verify RHR System interlock prevents the valves In accordance with from being opened with a simulated or actual RCS the Surveillance pressure signal > 474 psig. Frequency Control Program HBRSEP Unit No. 2 3.4-40 Amendment No. 265

U.S. Nuclear Regulatory Commission Serial: RA-22-0118 Enclosure ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP) 10 PAGES PLUS THE COVER

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) channel shall be declared inoperable. The second Note also requires that the methodologies for calculating the as-left and the as-found tolerances be in the UFSAR. The NOMINAL TRIP SETPOINT definition includes a provision that would allow the as-left setting for the channel to be outside the tolerance band, provided the setting is conservative with respect to the NTSP. This provision is not applicable to Functions for which the second Note applies.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition. The verification is performed by visual observation of the permissive status light in the unit control room. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within the Frequency specified in the Surveillance Frequency Control Program or 184 days of the Frequencies prior to reactor startup and four hours after reducing power below P-10 and P-6.

The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels.

Catawba Units 1 and 2 B 3.3.1-47 Revision No. 8

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RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This provision is not applicable to Functions for which the second NOTE applies.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition. The verification is performed by visual observation of the permissive status light in the unit control room. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within the frequency specified in the Surveillance Frequency Control Program or 184 days of the Frequencies prior to reactor startup and four hours after reducing power below P-10 and P-6. The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then the testing required by this surveillance must be performed prior to the expiration of the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-10 or < P-6) for periods > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

For Functions for which TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions (Reference 12) has been implemented, this SR is modified by two Notes as identified in Table 3.3.1-1. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is outside its as-found tolerance but McGuire Units 1 and 2 B 3.3.1-42 Revision No. 146

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RCS PIV Leakage B 3.4.14 BASES ACTIONS A.1 and A.2 (continued)

Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing some other valve qualified for isolation. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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.

B.1 and B.2 If Required Actions and associated Completion Times are not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each required 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 detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program..

INSERVICE TESTING PROGRAM OCONEE UNITS 1, 2, & 3 B 3.4.14-4 Rev. 001

RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS 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.

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

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 LPI System when the LPI 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. For the purposes of meeting this SR, test activities including contingencies may be performed prior to declaring a PIV inoperable. A PIV will be considered in testing until the test procedure is complete, or the test coordinator determines that further test contingencies would not be expected to produce an acceptable result.

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. NRC letter to DPC, "Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves," dated April 20, 1981.

4. NUREG-75/014, Appendix V, October 1975.

5. NUREG-0677, NRC, May 1980.

6. 10 CFR 50.36.

7. ASME Code for Operation and Maintenance of Nuclear Power Plants.

OCONEE UNITS 1, 2, & 3 B 3.4.14-5 Rev. 001

RCS PIVs B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria. Leakage rates > 1.0 gpm and 5.0 gpm are considered unacceptable if the latest measured rate exceeds the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the 5.0 gpm limit by 50%.

Leakage rates > 5.0 gpm are considered to be unacceptable.

More than one valve may be tested in parallel. The combined leakage must be within the limits of this SR. In addition, the minimum differential pressure when performing the SR shall not be < 150 psid. For 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 must be performed once prior to entering MODE 2 whenever the unit has been in MODE 5 for at least 7 days if leakage testing has not been performed in the previous 9 months. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

INSERVICE TESTING 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 it has been established per Note 3 that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated if in MODES 1 or 2, or prior to entry into MODE 2 if not in MODES 1 or 2 at the end of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

(continued)

HBRSEP Unit No. 2 B 3.4-87 Revision No. 83

RCS PIVs B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS 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.

SR 3.4.14.2 Verifying that the RHR interlock is OPERABLE ensures that RCS pressure will not pressurize the RHR system beyond 125% of its design pressure of 600 psig. The interlock setpoint prevents the valves from being opened and is set so the actual RCS pressure must be < 474 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. UFSAR, Section 3.1.

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

(continued)

HBRSEP Unit No. 2 B 3.4-88 Revision No. 83

v t e Letter Sequence Request
EPID:L-2023-LLA-0015, TSTF-493, TSTF-425, TSTF-545 (Open)
Initiation Request Acceptance Administration Questions 2 June 2023 20 June 2023 Answers Results Approval... Other:
MONTHYEAR2023-02-27 [Table View]

RA-22-0118, License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies

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5.0 ENVIRONMENTAL CONSIDERATION

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RA-22-0118, License Amendment Request to Revise Restrictive Technical Specification Surveillance Requirement Frequencies

Text

Subject:

Enclosure:

SUMMARY

2.1 Background

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

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