ML23054A455

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Issuance of Amendment Nos. 245 and 242, Regarding LAR to Revise Technical Specification 3.4.10, Pressurizer Safety Valves, to Decrease Low Side Setpoint Tolerance LCO Value
ML23054A455
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 03/16/2023
From: John Lamb
Plant Licensing Branch II
To: Gayheart C
Southern Nuclear Operating Co
Devlin-Gill, Stephanie
References
EPID L-2022-LLA-0098
Download: ML23054A455 (21)
v • d • e


Text

March 16, 2023 Mr. R. Keith Brown Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENT NOS. 245 AND 242, REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.4.10, PRESSURIZER SAFETY VALVES, TO DECREASE LOW SIDE SETPOINT TOLERANCE LIMITING CONDITION FOR OPERATION (LCO) VALUE (EPID L-2022-LLA-0098)

Dear Mr. Brown:

The Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. NPF-2 and Amendment No. 242 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated June 30, 2022, as supplemented by letter dated January 17, 2023.

The amendments revise the as found setpoint low side tolerance for the pressurizer safety valves (PSVs) described in TS 3.4.10, Pressurizer Safety Valves.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's monthly Federal Register notice.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 245 to NPF-2
2. Amendment No. 242 to NPF-8
3. Safety Evaluation cc: Listserv

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. NPF-2

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated June 30, 2022, as supplemented by letter dated January 17, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Glenn Glenn E. E. Miller Date: 2023.03.16 Miller 14:07:36 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 16, 2023

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. NPF-8

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated June 30, 2022, as supplemented by letter dated January 17, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Glenn Glenn E. E. Miller Date: 2023.03.16 Miller 14:08:00 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 16, 2023

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 245 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License License NPF-2, page 4 NPF-2, page 4 NPF-8, page 3 NPF-8, page 3 TSs TSs 3.4-10-1 3.4-10-1

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the Issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

a. Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.
b. Deleted per Amendment 13
c. Deleted per Amendment 2
d. Deleted per Amendment 2
e. Deleted per Amendment 152 Deleted per Amendment 2
f. Deleted per Amendment 158
g. Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1) Identification of a sampling schedule for the critical parameters and control points for these parameters;
2) Identification of the procedures used to quantify parameters that are critical to control points;
3) Identification of process sampling points;
4) A procedure for the recording and management of data;
5) Procedures defining corrective actions for off control point chemistry conditions; and Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 245

(2) Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed license.

(3) Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproducts, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporate below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3) Delete per Amendment 144 (4) Delete Per Amendment 149 (5) Delete per Amend 144 Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 242

Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 2423 psig and 2510 psig.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR.

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

The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE------------ A.1 Restore valve to 15 minutes RICT entry is not permitted OPERABLE status.

OR for this loss of function Condition when a In accordance with pressurizer safety valve is the Risk Informed intentionally made Completion Time inoperable. Program One pressurizer safety valve inoperable.

Farley Units 1 and 2 3.4.10-1 Amendment No. 245 (Unit 1)

Amendment No. 242 (Unit 2)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 245 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By application dated June 30, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML22181B145), as supplemented by letter dated January 17, 2023 (ML23017A205), Southern Nuclear Operating Company, Inc. (SNC, the licensee) requested changes to the Technical Specifications (TSs) for Renewed Facility Operating License Nos. NPF-2 and NPF-8 for Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, respectively.

The proposed changes revise the TS 3.4.10, Pressurizer Safety Valves. The proposed changes revise the as found setpoint low side tolerance from -1 percent (> 2460 pounds per square inch gauge (psig)) to -2.5 percent (> 2423 psig).

The supplement letter dated January 17, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on September 6, 2022 (87 FR 54553).

2.0 REGULATORY EVALUATION

2.1 System Description Farley, Units 1 and 2, are three-loop Westinghouse design pressurized water reactors (PWRs) with large dry containments. The containment is prestressed, reinforced concrete cylindrical structure with a shallow domed roof and a reinforced concrete foundation slab. The design includes a 1/4 inch thick welded steel liner to the inside face of the concrete with the floor liner Enclosure 3

installed on top of the foundation and then covered with concrete. The containment encloses the reactor, pressurizer, reactor coolant system, steam generators, and portions of the auxiliary and engineered safeguards systems.

The pressurizer is equipped with the following types of pressure relief valves: 2 power-operated relief valves (PORVs) and 3 pressurizer safety valves (PSVs).

The PSVs are totally enclosed spring loaded, self-actuated safety-related valves. In operating Modes 1, 2, and 3, in conjunction with the reactor protection system, the PSVs prevent the reactor coolant pressure boundary (RCPB) pressure from exceeding the American Society of Mechanical Engineers (ASME) pressure safety limit (SL) of 2735 psig, which is 110 percent of the RCPB design pressure. In Modes 4, 5, and while the reactor vessel head is still on in Mode 6, the RCPB overpressure protection is provided by operating procedures and by meeting the requirements in the TSs for the Limiting Condition for Operation (LCO) 3.4.12, Low Temperature Overpressure Protection (LTOP) System. As stated in the Updated Final Safety Analysis Report (UFSAR) Section 15.2.7.1, the PSVs are sized to protect the RCPB against overpressure for all load losses without assuming operation of the steam dump system, pressurizer spray, PORVs, automatic rod cluster control assembly (RCCA) control, or the direct reactor trip on turbine trip.

The PORVs are air-operated valves controlled to open at a specific set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. Both PORVs open at a nominal set pressure of 2335 psig. The functional design of the PORVs is based on maintaining the RCPB pressure below the RTS trip function (Pressurizer Pressure -

High) following a step reduction of 50 percent of full load with steam discharge through the turbine bypass system (also called steam dump system). The PORVs are designed to lift prior to the PSVs.

As stated in UFSAR Section 5.5.13.2, control room position indication is provided for all PSVs.

2.2 Proposed Changes The current PSV lift settings reflect a +/-1 percent tolerance around the nominal lift setpoint of 2485 psig. The licensee stated that since 2015, there have been six instances in which the PSVs tested as-found setpoints were outside the low end (-1 percent) tolerance limit of the nominal setpoint of 2485 psig, which resulted in those PSVs to be declared as inoperable. The test results also indicated that the as-found lift settings were not less than 3 percent of the nominal 2485 psig pressure setting. The proposed PSV lift setting will reflect a

+1 percent, -2.5 percent tolerance around the existing nominal lift setpoint of 2485 psig. The current and proposed TS 3.4.10 are below.

2.2.1 Current TS The Farley, Units 1 and 2, TSs state the current TS 3.4.10, LCO as follows:

Three pressurizer safety valves shall be OPERABLE with lift settings 2460 psig and 2510 psig.

2.2.2 Proposed TS SNC proposes to change the Farley, Units 1 and 2, TS 3.4.10 LCO to the following:

Three pressurizer safety valves shall be OPERABLE with lift settings 2423 psig and 2510 psig.

2.3 Applicable Regulatory Requirements and Guidance The regulation 10 CFR 50.36(c)(1) requires that plant TS will include safety limits, limiting safety system settings and limiting control settings. The regulation 10 CFR 50.36(c)(2)(ii)(C) specifies that a LCO be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure or presents a challenge to the integrity of a fission product barrier. The PSVs are part of the primary success path to mitigate consequences of design-basis events (DBEs), and are credited in the Farley, Units 1 and 2, UFSAR analyses.

General Design Criteria (GDC) 15 (Criterion 15: Reactor coolant system design), as it relates to designing the RCPB and associated auxiliary, control, and protection systems with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences (AOOs).

GDC 31 (Criterion 31 - Fracture prevention of reactor coolant pressure boundary), as it relates to designing the RCPB with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fractures is minimized.

NUREG-0800, Standard Review Plan (SRP), 5.2.2, Overpressure Protection, Revision 3 (ML070540076),Section II states the acceptance criteria are based on meeting the relevant requirements of the Commission regulations in GDC 15 and GDC 31. SRP 5.2.2, in Section II, under heading SRP Acceptance Criteria, in Section 2.B, requires the design of safety valves should have sufficient capacity to limit the pressure to less than 110 percent of the RCPB design pressure during the most severe AOO with reactor scram, as specified by ASME Code Article NB-7000. Sufficient available margin should account for uncertainties in the design and operation of the plant, based on the assumptions stated in SRP 5.2.2,Section II, under heading SRP Acceptance Criteria 2.B.

3.0 TECHNICAL EVALUATION

3.1 Licensee Event Reports (LERs)

In Section 2.2, Reason for the Proposed Change, of its LAR, SNC stated that this change is proposed to reduce an unnecessarily restrictive LCO. The licensee indicated that the Farley PSVs were manufactured by Crosby. SNC reported that there have been six instances since 2015 where the PSVs were tested and found outside the +/-1% tolerance limits. The licensee stated that all of the out-of-tolerance test results were below the low end of the setpoint tolerance (-1%) but did not exceed -3% of the pressure setpoint. SNC submitted licensee event reports (LERs) in accordance with the NRC regulations for each of these instances. The licensee considered that setpoint drift was the cause of the PSVs lifting low out of tolerance, in that the as-found result was lower than the allowed value in the TSs. SNC asserted that the as-found PSV conditions did not have an adverse impact on the overpressurization function of

the PSV in each instance. The licensee stated that the PSV test results were within the safety analysis assumptions that are credited for the PSVs at Farley, and the plant remained bounded by the accident analyses in the FSAR. The licensee considered that generating an LER for a PSV that is performing satisfactorily within the design analysis assumptions becomes an unnecessary burden for licensees and the NRC.

Based on the review of the Farley LAR, the NRC staff requested the licensee to specify whether its evaluations of the six instances of out-of-tolerance test results for the Farley PSVs indicated any material condition concerns with these valves or the performance of their design capabilities. In its supplement to the LAR dated January 17, 2023, SNC stated that the field service reports for each PSV that failed the as-found test had been reviewed. The licensee indicated that there were no material concerns with the valves that would have prevented the valves from performing its design functions, and that all the valves performed within its design capabilities.

In addition to reviewing the licensee submittal, the NRC staff conducted an ADAMS search for LERs related to Farley PSV setpoints since January 1, 2000. The NRC staff located the following LERs from its ADAMS search:

1. Farley, Unit 1 - LER 15-004-00 (ML15282A592)
2. Farley, Unit 1 - LER 16-003-00 (ML16347A116)
3. Farley, Unit 2 - LER 17-003-00 (ML17354A395)
4. Farley, Unit 1 - LER 18-001-00 (ML18163A226)
5. Farley, Unit 1 - LER 19-001-00 (ML19343B093)
6. Farley, Unit 2 - LER 20-002-00 (ML20356A158)

The NRC staff reviewed the listed LERs for the cause of the failure of the PSVs to meet their TS required setpoints.

In 2015, the licensee submitted LER 15-004-00 to report that a Farley, Unit 1, PSV failed its as-found lift test below the TS allowable value because seat leakage reduced the setpoint.

In 2016, SNC submitted LER 16-003-00 to report that a Farley, Unit 1, PSV failed its as-found lift test below the TS allowable value with seat leakage indicated to be the most likely cause of the setpoint drift.

In 2017, the licensee submitted LER 17-003-00 to report that a Farley, Unit 2, PSV failed its as-found lift test below the TS allowable lift setting because of setpoint drift which resulted in seat leakage.

In 2018, SNC submitted LER 18-001-00 to report that a Farley, Unit 1, PSV failed its as-found lift pressure test below the TS allowable lift setting with spring relaxation indicated to be the most likely cause of the failure.

In 2019, the licensee submitted LER 19-001-00 to report that a Farley, Unit 1, PSV failed its as-found lift pressure test below the TS range with setpoint drift indicated to be the cause of the failure.

In 2020, SNC submitted LER 20-002-00 to report that a Farley, Unit 2, PSV failed its as-found lift pressure test below the TS range with setpoint drift due to spring relaxation indicated to be the cause of the failure.

The NRC staff considers these LERs to reveal a variety of possible causes for the PSVs to fail its as-found lift test below the TS allowable range, such as seat leakage, spring relaxation, or general setpoint drift. In that a low lift pressure occurred in each instance for these PSVs, the NRC staff does not consider the overpressure protection function to be impacted by these performance characteristics of the PSVs. In monitoring PSV performance, it is important for the PSVs to operate within a reasonable range above or below their specified lift setpoint. In this case, the operating experience shows that the PSVs tend to drift slightly to the low side of their lift setpoint. The NRC staff does not consider this small amount of drift to the low side of the lift setpoint to reveal a performance concern with the PSVs. Therefore, the NRC staff finds that the operating experience for the PSVs at Farley does not reveal a safety concern with the request by SNC in the LAR to modify the as-found TS setpoint low-side tolerance from -1%

( 2460 psig) to -2.5% ( 2423 psig).

3.2 Safety Analyses for Events that Model Positive PSV Tolerance SNC stated that the safety analysis of the following events that model the positive PSV (high) tolerance if not terminated prior to overpressurization of the RCPB, assume operation of 3 PSVs to prevent RCPB overpressurization:

Event FSAR Section Uncontrolled rod withdrawal from full power 15.2.2 Complete loss of reactor coolant flow 15.3.4 Loss of external electrical load 15.2.7 Loss of normal feedwater 15.2.8 Loss of all AC power to station auxiliaries 15.2.9 Locked rotor 15.4.4 The NRC staff determined that the ability of the PSVs to perform their safety function of RCPB overpressure protection would not be affected during the above events, because the proposed TS change do not include either the nominal setpoint (2485 psig) change or a change in the setpoint high side (2510 psig) tolerance for the PSVs.

3.3 Safety Analyses for Events that Model Negative PSV Tolerance The licensee stated that the proposed change of the low side tolerance to -2.5 percent of 2485 psig resulting in the low side setpoint pressure change from 2460 psig to 2423 psig could affect the analysis of following non-LOCA events:

Event FSAR Computer Section Code Used Loss of external electrical load - minimum departure from nucleate boiling ratio (DNBR) case and maximum main steam 15.2.7 RETRAN system pressure case.

Feedline break 15.4.2 RETRAN-02 Inadvertent operation of the emergency core cooling system 15.2.14 LOFTRAN Rod cluster control assembly withdrawal at power 15.2.2 LOFTRAN

- DNBR cases

The licensee stated that for the above non-LOCA events analysis, a -3 percent tolerance is applied by both the LOFTRAN and RETRAN base deck calculation notes and is explicitly modeled in each of the safety analyses. The licensee confirmed that for the above events, the analyses of record (AOR) as described in the FSAR use a -3 percent PSV tolerance. Since the -3 percent AOR tolerance bounds the proposed -2.5 percent tolerance, the NRC staff finds it acceptable that the results of the AOR remain valid and are not affected by the proposed TS change.

3.4 Large Break Loss-of-Coolant Accident Analysis SNC stated that the PSVs are not modeled in the postulated large break loss-of-coolant accident (LBLOCA) event AOR because the RCPB does not reach a pressure that would require the PSVs to open.

The NRC staff finds that the LBLOCA AOR will not be affected by the proposed TS change because during this event the RCPB depressurizes rapidly due to double-ended break area size of the recirculation pump pipe.

3.5 Small Break LOCA Analysis The licensee stated that the PSVs are not modeled in the small break LOCA (SBLOCA) event AOR short term hydraulic forces and long-term core cooling because it is a depressurization transient event.

The NRC staff finds that the PSVs are not required to be modeled in the SBLOCA event AOR because it is a depressurization transient event. The RCPB is not expected to re-pressurize to a pressure that would require the PSVs to open. Therefore, the SBLOCA AOR is not affected by the proposed TS change.

3.6 Containment and Radiological Analyses SNC stated the following design basis containment and radiological analysis are not affected by the proposed TS change because the PSVs are either not modeled or credited in their AOR:

Long-term LOCA mass and energy (M&E) releases for containment integrity Short-term LOCA M&E releases for subcompartment analyses Long-term steamline break (SLB) M&E releases inside containment for containment integrity Long-term SLB M&E releases outside containment for compartment response Short-term SLB M&E releases for subcompartment analyses Steam release for dose Steam generator tube rupture The SLB M&E releases inside and outside containment result in depressurization of the RCPB.

The PSV inputs to the analyses of the SLB M&E releases inside or outside the containment have no impact on the transient response since the RCPB depressurizes during this event.

Additionally, PSVs are not modeled in the calculation of steam releases for dose analysis.

Therefore, the PSV setpoint low side tolerance change to -2.5% does not impact the dose analyses for Farley, Units 1 and 2.

The NRC staff finds that the above AOR will not be affected by the proposed TS change, because the RCPB depressurizes during these events.

3.7 Margin Between PORVs and PSVs Opening Setpoints The licensees calculated uncertainty in AOR associated with opening of the PORVs is

+/-48.1 psi. Applying this uncertainty to the nominal PORV setpoint 2335 psig, its actuation band would be between 2285.4 and 2383.1 psig. Table 1 below provides the current and the proposed actuation band of the PSVs and the PORVs and the margin between PSV (low) and PORV (high) pressure end of the tolerance band.

Table 1 Nominal Lift Setpoint Current Actuation Band Proposed Actuation Band Valve Pressure (psig) (psig)

(psig)

Low High Low High PSV 2485 2460 2510 2423* 2510 High Low High Low PORV 2335 2383.1 2285.4 2383.1 2285.4 Minimum Margin = PSV Not 76.9 psi 39.9 psi Not applicable (Low) minus PORV (High) applicable

  • Proposed change The proposed change results in a reduction in the minimum margin [(PSV (Low) - PORV (High)]

from 76.9 psi to 39.9 psi, and a reduction in the maximum margin [PSV (low) - PORV (low)]

from 174.6 psi to 137.6 psi. Based on the sufficient maximum and minimum margins of 137.6 psi and 39.9 psi respectively, the NRC staff finds it acceptable that the PORVs will actuate prior to the PSVs during a RCPB overpressure transient.

3.8 High Pressure Reactor Trip System Setting and Uncertainty The pressurizer pressure RTS function (Pressurizer Pressure - High) in conjunction with PORVs and PSVs ensures that RCPB overpressure protection is provided. The limiting safety system setting of the RTS function is selected below the PSV actuation pressure and above the PORV setting to minimize PSV actuation as the RTS function will trip the reactor and therefore mitigate most overpressure transients before the RCPB pressure increases to the PSV setting.

The nominal high pressurizer pressure RTS trip setting is 2385 psig with an associated uncertainty of +28.8 psi. The licensee calculated this uncertainty using Farley-specific setpoint methodology based on a statistical square root of the sum of the squares (SRSS) technique consistent with the NRC-approved methodology in WCAP-8567-P-A, Improved Thermal Design Procedure, February 1989 (ML081770447), which is the current licensing basis. The licensee briefly described the uncertainty calculation method in letter dated January 17, 2023 (ML23017A205) with the following key points:

The method identifies the distinct uncertainty components and combines the components statistically to determine the total channel uncertainty.

The total channel uncertainty is compared to the total allowance (i.e., the difference between the safety analysis limit (SAL) and the TS trip setpoint) to demonstrate margin.

The method is used in several PWRs and is based on Westinghouse setpoint uncertainty calculation methodology and conforms to the industry practices such as Integrated Safety Analysis ( ISA) Standard S67.04, 1987.

Table 2 below provides the current and the proposed actuation pressure of the PSVs and the RTS trip and the margin between PSV (low) and RTS trip (high) pressure end tolerance band.

Table 2 Current Proposed Nominal Set Actuation Actuation Item Pressure Pressure Pressure (psig)

(psig) (psig)

Low Low PSV 2485 2460 2423*

High High RTS Trip 2385 2413.8 2413.8 Margin = PSV (Low) minus 46.2 psi 9.2 psi RTS Trip (High)

  • Proposed change The proposed change results in a reduction in the margin (PSV (Low) pressure minus RTS trip (High) pressure) from 46.2 psi to 9.2 psi. Although the margin is reduced, the NRC staff finds the resulting 9.2 psi margin, which includes uncertainty, is sufficient to ensure that the RTS trip will actuate prior to the PSVs.

The worst scenario of reactor overpressure protection would be PORV (high) actuation at 2383.1 psig, RTS trip (high) at 2413.8 psig, and PSV (high) actuation at 2510 psig.

3.9 TS 3.4.10 Change SNC proposes to change the Farley, Units 1 and 2, TS 3.4.10 LCO to the following:

Three pressurizer safety valves shall be OPERABLE with lift settings > 2423 psig and

> 2510 psig.

3.9.1 Commission Policy Statement on TSs Questions The Federal Register Notice for the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132) had the following questions:

1. What is the justification for the Technical Specification, i.e., which Policy Statement criterion requires it to be in the Technical Specifications?
2. What are the Bases for each LCO, i.e., why was it determined to be the lowest functional capability or performance level for the system or component in question necessary for safe operation of the facility and, what are the reasons for the Applicability of the LCO?
3. What are the Bases for each Action, i.e., why should this remedial action be taken if the associated LCO cannot be met; how does this Action relate to other Actions associated with the LCO; and what justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?
4. What are the Bases for each Safety Limit?
5. What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?

3.9.2 NRC Staff Responses to the Commission Policy Statement on TSs Questions

1. What is the justification for the Technical Specification, i.e., which Policy Statement criterion requires it to be in the Technical Specifications?

The Farley, Units 1 and 2, TS 3.4.10 satisfies Criterion 3. Criterion 3 states, A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The PSVs are a component that is part of the primary success path and which functions to mitigate a design basis accident. In operating Modes 1, 2, and 3, in conjunction with the reactor protection system, the PSVs prevent the RCPB pressure from exceeding the ASME pressure safety limit (SL) of 2735 psig, which is 110 percent of the RCPB design pressure. Therefore, the PSVs meets Criterion 3.

2. What are the Bases for each LCO, i.e., why was it determined to be the lowest functional capability or performance level for the system or component in question necessary for safe operation of the facility and, what are the reasons for the Applicability of the LCO?

The LCO for Farley, Units 1 and 2, TS 3.4.10 is discussed in Sections 3.1 through 3.5 above.

3. What are the Bases for each Action, i.e., why should this remedial action be taken if the associated LCO cannot be met; how does this Action relate to other Actions associated with the LCO; and what justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?

The Actions for Farley, Units 1 and 2, TS 3.4.10 are not being changed.

4. What are the Bases for each Safety Limit?

Farley, Units 1 and 2, TS 3.4.10 does not change a Safety Limit.

5. What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?

Farley, Units 1 and 2, TS 3.4.10 does not change a Surveillance Requirement.

3.10 Technical Conclusions The NRC staff concludes that the proposed TS 3.4.10 change in the PSVs setpoint as-found low side tolerance from -1 percent (> 2460 psig) to -2.5 percent (> 2423 psig) is acceptable based on the following:

The ability of the PSVs to perform their safety function of RCPB overpressure protection would not be affected during the non-LOCA events that model the PSV (high) side of the tolerance band.

The ability of the PSVs to perform their safety function of RCPB overpressure protection would not be affected during the non-LOCA events that model the PSV (low) side of the tolerance band.

The LBLOCA and SBLOCA AOR are not affected by the proposed TS change.

The containment and dose AORs are not affected by the proposed TS change.

The proposed margin 39.9 psi between the PSV (low) pressure and PORV (high) pressure is sufficient to ensure that the PORVs will actuate prior to the PSVs during RCPB overpressure events.

The proposed margin 9.2 psi between PSV (low) pressure and RTS trip (high) pressure is sufficient to ensure that the RTS trip will actuate prior to the PSVs.

Actuation of the PSVs during RCPB overpressure events would be minimized because the RTS trip function will trip the reactor and therefore mitigate most overpressure transients before pressure increases to the PSV (low) pressure setting.

The reliability of the PSVs to perform their safety function is not affected by the proposed TS change.

3.11 Regulatory Conclusions Based on the technical evaluations, the NRC staff conclusions regarding the applicable regulatory requirements given in Section 2.1 above are as follows:

The requirements of 10 CFR 50.36(c)(1) and (2)(ii)(C) are satisfied because the proposed TS 3.4.10 includes safety limits, limiting safety system settings, and the LCO is established for structure, system, or component that is part of the primary success path and which

functions or actuates to mitigate a design basis accident or transient that either assumes the failure or presents a challenge to the integrity of a fission product barrier.

GDC 15 continues to be satisfied because the RCPBs associated protection systems ensures integrity of the RCPB with adequate margins during normal operation and during transient events.

GDC 31 continues to be satisfied because RCPB has sufficient margin to assure that when stressed under a postulated overpressure transient event, the boundary behaves in a nonbrittle manner, and the probability of rapidly propagating fracture is minimized.

Additionally, the NUREG-0800, Standard Review Plan 5.2.2 acceptance criteria is met because the PSVs capacity remains unchanged and is sufficient to limit the RCPB pressure less than 110% of its design pressure during the most severe operational transient events, and therefore will continue to provide overpressure protection.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amendments on February 12, 2023. On February 13, 2023, the State official confirmed that the State of Alabama had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on February 22, 2022 (87 FR 9652), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: Ahsan Sallman Tom Scarbrough John G. Lamb Date: March 16, 2023

ML23054A455 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DEX/EICB/BC NAME JLamb KGoldstein MWaters DATE 2/22/2023 03/01/2023 02/24/2023 OFFICE NRR/DSS/SNSB/BC (A) NRR/DEX/EMIB/BC NRR/DSS/STSB/BC NAME DWoodyatt SBailey VCusumano DATE 2/22/2023 02/23/2023 03/01/2023 OFFICE OGC - NLO NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME KDowling MMarkley (EMiller for) JLamb DATE 03/14/2023 03/16/2023 03/16/2023

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