ML25353A551

Issuance of Amendment Nos. 258 and 255 to Revise Technical Specification Table 3.3.3-1, Post Accident Monitoring Instrumentation, by Deleting Function 10, (Reactor Coolant System (Rcs)) Subcooling Margin Monitor

ML25353A551
Person / Time
Site:	Farley
Issue date:	01/08/2026
From:	Turner Z Plant Licensing Branch II
To:	Coleman J Southern Nuclear Operating Co
References
EPID L-2025-LLA-0100
Download: ML25353A551 (0)

Text

January 8, 2026 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company 3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 258 AND 255 TO REVISE TECHNICAL SPECIFICATION TABLE 3.3.3-1, POST ACCIDENT MONITORING INSTRUMENTATION, BY DELETING FUNCTION 10, [REACTOR COOLANT SYSTEM (RCS)]

SUBCOOLING MARGIN MONITOR (EPID L-2025-LLA-0100)

Dear Ms. Coleman:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 258 to Renewed Facility Operating License No. NPF-2 and Amendment No. 255 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. The amendments are in response to your application dated June 30, 2025.

The proposed amendment revises Technical Specification (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation to delete function 10, Reactor Coolant System (RCS) Subcooling Margin Monitor (SMM).

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commissions Federal Register notice.

Sincerely,

/RA - J. Lamb for/

Zachary M. Turner, 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. 258 to NPF-2

2. Amendment No. 255 to NPF-8

3. Safety Evaluation cc: Listserv

SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 258 Renewed License No. NPF-2

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 1 (the facility), Renewed Facility Operating License No. NPF-2 (the license) filed by Southern Nuclear Operating Company (the licensee), dated June 30, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions regulations; D.

The issuance of this 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 Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment. Paragraph 2.C.(2) of the license is hereby amended to read as follows:

2.C.(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 258, are hereby incorporated in the renewed license.

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

3.

This amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: January 8, 2026 SHAWN WILLIAMS Digitally signed by SHAWN WILLIAMS Date: 2026.01.08 09:48:24 -05'00'

SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 255 Renewed License No. NPF-8

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 2 (the facility), Renewed Facility Operating License No. NPF-8 (the license) filed by Southern Nuclear Operating Company (the licensee), dated June 30, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions regulations; D.

The issuance of this 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 Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment. Paragraph 2.C.(2) of the license are hereby amended to read as follows:

2.C.(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 255, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: January 8, 2026 SHAWN WILLIAMS Digitally signed by SHAWN WILLIAMS Date: 2026.01.08 09:49:28 -05'00'

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 255 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 3.3.3-3 3.3.3-3 Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 258 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 258, 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 2 Renewed License No. NPF-8 Amendment No. 255 (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 byproduct, 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 incorporated below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2821 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 255, are hereby incorporated in the renewed license.

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

(3)

Deleted per Amendment 144 (4)

Deleted per Amendment 149 (5)

Deleted per Amendment 144

PAM Instrumentation 3.3.3 Farley Units 1 and 2 3.3.3-3 Amendment No.  (Unit 1)

Amendment No. (Unit 2)

Table 3.3.3-1 (page 1 of 1)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION D.1

1. RCS Hot Leg Temperature (Wide Range) 2 E

2. RCS Cold Leg Temperature (Wide Range) 2 E

3. RCS Pressure (Wide Range) 2 E

4. Steam Generator (SG) Water Level (Wide or Narrow Range) 2/SG E

5. Refueling Water Storage Tank Level 2

E

6. Containment Pressure (Narrow Range) 2 E

7. Pressurizer Water Level 2

E

8. Steam Line Pressure 2/SG E

9. Auxiliary Feedwater Flow Rate 2

E

10. Deleted

11. Containment Water Level (Wide Range) 2 E

12. Core Exit Temperature - Quadrant 1 2(a)

E

13. Core Exit Temperature - Quadrant 2 2(a)

E

14. Core Exit Temperature - Quadrant 3 2(a)

E

15. Core Exit Temperature - Quadrant 4 2(a)

E

16. Reactor Vessel Level Indicating System 2

F

17. Condensate Storage Tank Level 2

E

18. Deleted

19. Containment Area Radiation (High Range) 2 F

(a) A channel consists of two core exit thermocouples.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 255 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By letter dated June 30, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25181A816), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) that contained proposed amendments to the Technical Specifications (TSs) for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2.

Specifically, the proposed changes would revise TS 3.3.3, Post Accident Monitoring (PAM)

Instrumentation, by removing function 10, [Reactor Coolant System] RCS Subcooling Margin Monitor, from Table 3.3.3-1, Post Accident Monitoring Instrumentation.

2.0 REGULATORY EVALUATION

2.1 Background

In the Enclosure to the licensees submittal dated June 30, 2025, the licensee stated:

The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBA).

The instrument channels required to be operable by TS 3.3.3 include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and certain Category I variables. The core subcooling monitor (RCS SMM) is currently listed in the Farley Updated Final Safety Analysis Report (UFSAR) Table 7.5-1, (Sheet 1 of 16) Post Accident Instrumentation as a Type A, Category 2 variable. Type A variables provide primary information required for the control room operator to take specific manually controlled actions for which no automatic control is provided and are required for safety systems to accomplish their safety functions for DBAs. The operability of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97. TS Table 3.3.3-1 lists all Type A and certain Category I variables identified by the unit specific Regulatory Guide 1.97 analysis, as amended by the NRCs SER [Safety Evaluation Report] [(ML20211D523, ML20211D534, and ML20211D508)].

RCS Subcooling Margin is provided to determine safety injection termination and reinitiation and depressurization and cooldown progression. The RCS SMM measures saturation/superheat margin. The function of the RCS SMM is to calculate the subcooled margin which is the difference between the measured temperature of the reactor coolant and the saturation temperature. The saturation temperature is calculated from the minimum primary system pressure input. A maximum or representative temperature input is used for the measured value, which could come from a Resistance Temperature Detector (RTD) loop, or a representative core exit thermocouple (CET).

2.2 Description of Proposed Changes The licensee is proposing to remove the RCS SMM function (function 10) from TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, by deleting function 10 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. The licensee will maintain the RCS SMM indication in Farley UFSAR Table 7.5-1 as a Type B, Category 2 variable.

Specifically, as stated in the Enclosure to the licensees letter dated June 30, 2025, the licensee states:

The proposed Technical Specification change revises the Regulatory Guide 1.97 instrumentation contained in TS 3.3.3 to be consistent with the technical basis for accident monitoring instrumentation identified in WCAP-15981-NP-A [ML103560687],

Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants. This change includes evaluating the current Regulatory Guide 1.97 [Revision 5 (ML18136A762)] categorization of the affected instrumentation with respect to its function as a post accident monitoring instrument based on WCAP-15981-NP-A.

The U.S. Nuclear Regulatory Commission (NRC) staffs evaluation of these changes can be found in Section 3.0 of this safety evaluation.

2.3 Reason for the Proposed Change In Section 2.3 of the Enclosure to its LAR, the licensee stated, in part, that:

The proposed change will make the content of the [Farley] TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation more consistent with the corresponding Standard Technical Specification (STS), NUREG-1431, Revision 5, [TS] 3.3.3, Post Accident Monitoring (PAM) Instrumentation while maintaining the [Farley] response to Regulatory Guide 1.97 unchanged. The proposed TS 3.3.3 change will also eliminate the need for plant shutdown as currently required by TS 3.3.3 Condition E in the event that both RCS Subcooling Margin Monitor channels are inoperable. Note that the proposed activity removes the RCS SMM from the Technical Specifications; it is not proposing to change the safety classification, operation, or maintenance of the RCS SMM System for the [Farley] Units. Any proposed future changes to the RCS SMM following implementation of this proposed change will be controlled by existing design control procedures and regulatory processes.

2.4 Regulatory Requirements and Guidance The NRC staff considered the following NRC regulations and guidance in its review of the proposed LAR:

Under Title 10 of the Code of Federal Regulations (10 CFR) 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired and following as far as applicable, the form prescribed for original applications. Under 10 CFR 50.92(a),

determinations on whether to grant an applied -for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.

The NRCs regulatory requirements related to the content of the TS are set forth in 10 CFR Section 50.36, Technical specifications. This regulation requires that the TSs include items in, among other things, the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) surveillance requirements; (4) design features; and (5) administrative controls.

As stated in 10 CFR 50.36(a)(1), each applicant for an operating license includes in its application proposed technical specifications, and a summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

As stated in 10 CFR 50.36(c)(2), LCO are the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

As stated in 10 CFR 50.47, Emergency plans, Paragraph (b)(9) requires that the onsite and offsite emergency response plans for nuclear power reactors must ensure adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.

As stated in 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, Paragraph (b)(3) identify the post-accident monitoring equipment important to safety covered by this Section.

The NRCs guidance for the format and content of the Farley TS can be found in U.S. Nuclear Regulatory Commission, NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).

Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) for nuclear power plants. Plant-specific design criteria are described in the plants Updated Final Safety Analysis Report (UFSAR). Specifically, Section 3.1 of Farleys UFSAR (ML23319A065) discusses conformance with the following GDCs:

GDC 13, Instrumentation and Control, states:

Instrumentation is provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 19, Control Room, states:

A control room is provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs. Adequate radiation protection is preceded to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room is provided with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

The NRC staff considered the following guidance in its review of the licensees application:

NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980 (ML051400209), provides additional requirements for PAM instrumentation, and includes clarification of the Regulatory Guide 1.97 guidance.

Regulatory Guide 1.97, Revision 2, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, July 1980 (ML060750525), describes methods acceptable to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.

WCAP-15981-NP-A, Revision 0, Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants, September 2008 (ML103560687), provides technical justification for identifying PAM instrumentation that should be included in the TS for Westinghouse Nuclear Steam Supply System (NSSS) plants in addition to providing a methodology to be used by licensees to reassess the PAM instrumentation that should be included in the plant-specific TS.

3.0 TECHNICAL EVALUATION

3.1 Current Partial TS Table 3.3.3-1 (function 10), Post Accident Monitoring Instrumentation The licensee proposed to remove function 10 of TS table 3.3.3-1, which currently states:

Function Required Channels Condition Referenced From Required Action D.1

10. RCS Subcooling Margin Monitor Deleted 2

E The function will be replaced with deleted.

3.2 TS Table 3.3.3 Removal of the Subcooling Margin Monitor (SMM)

SNC proposed to delete the SMM entry from Technical Specification Table 3.3.3-1, Post Accident Monitoring Instrumentation, on the basis that subcooling margin is no longer classified as a Type A variable under Regulatory Guide 1.97, Revision 2.

The NRC staff evaluated SNCs proposal to remove the SMM from TS Table 3.3.3-1. The SMM function calculates the margin between measured reactor coolant system (RCS) temperature and the saturation temperature based on minimum RCS pressure. The staff notes that the underlying inputs to the SMMcore exit temperature, hot leg RTD temperatures, and wide-range RCS pressureremain available on qualified, safety-grade instrumentation, with redundant channels in the control room. These inputs are consistent with the requirements of Regulation Guide (RG) 1.97 and NUREG-0737, Supplement 1. Additionally, the licensee proposes to retain the SMM as a Type B, Category 2 variable in the UFSAR, which reflects its supporting role in post-accident assessment rather than a primary safety decision parameter.

The staff concludes that this approach is consistent with WCAP-15981-NP-A and that removing the SMM from the Farley TS Table 3.3.3-1, Post Accident Monitoring Instrumentation does not impact the availability of information required to assess core cooling. Additionally, RCS SMM information continues to be available to the operators as a secondary process parameter.

3.2.1 Operator Capability to Assess Core Cooling Without SMM The NRC staff determined that operators will continue to be able to assess core cooling and subcooling conditions without the SMM indication. The core exit thermocouples, wide-range RCS pressure indicators, and plant computer-based tools such as the Safety Parameter Display System (SPDS) provide the necessary data for operators to verify that the RCS remains subcooled during accident conditions. The SMM has historically served as a supplementary indicator rather than a primary source of information. Farleys Emergency Operating Procedures (EOPs) provide guidance on interpreting these direct parameters to determine appropriate operator actions. Therefore, the staff concludes that the removal of the SMM from the TS will not impair the operators ability to determine whether core cooling is adequate.

3.2.2 Regulatory Alignment with RG 1.97, NUREG-0737, and GDCs The NRC staff reviewed SNCs basis for reclassifying the SMM from a Type A to a Type B variable under RG 1.97 and found it acceptable. RG 1.97 (Revision 2) considers subcooling margin as a Type B, Category 2 variable, and WCAP-15981-NP-A supports this classification.

NUREG-0737 only requires Type A variables to be included in the TS. The SMM's reclassification and removal from the TS are therefore consistent with NRC guidance.

Additionally, the instrumentation supporting the subcooling assessment remains compliant with the requirements of GDC 13 and GDC 19 for control room indication and system-level monitoring. Therefore, the NRC staff concludes the proposed change continues to satisfy applicable regulatory criteria.

3.2.3 Post-Accident Human-System Interface and Data Integrity The NRC staff considered human-system interface (HSI) and data integrity impacts associated with removing the SMM from the TS. The current SMM display will be retained in the control room, and no hardware changes are being made as part of this amendment. The inputs to the SMM, including temperature and pressure signals, are already validated and used for other safety functions. Therefore, the NRC staff finds that data integrity and HSI aspects are not adversely affected by the proposed amendment.

3.2.4 Systems-Level Justification for Reclassification The NRC staff reviewed the licensees justification for reclassifying the SMM function as Type B and determined it is consistent with a systems-level approach to post-accident monitoring. The subcooling margin calculation is derived from parameters in the Farley, Units 1 and 2, TS Table 3.3.3-1 that are individually monitored by redundant and qualified sensors. These parameters are already used in the operator decision-making process. The removal of the SMM from the TS does not eliminate the availability or integrity of this information but reflects that the SMM function is a secondary calculation rather than a unique indicator. As such, the staff agrees that the function is appropriately categorized as Type B.

3.3 Design Basis Accidents The instrumentation that provides the primary information that is essential for the direct accomplishment of the specified manual actions (including long-term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish its safety functions for design basis accidents or transients were identified from the design basis accident analyses documented in Farley, Units 1 and 2, UFSAR, Chapter 15. The manual actions specifically credited in the DBA analyses for which no automatic control is provided are described below.

Inadvertent Operation of the Emergency Core Cooling System (ECCS) During Power Operation (UFSAR Section 15.2.14):

This event is classified as a Condition II event, Incidents of Moderate Frequency, commonly known as an Anticipated Operational Occurrence (AOO), meaning the plant's design and operational procedures are intended to safely manage such occurrences without major consequences. The safety concern addressed by the NRC is ensuring that an AOO does not escalate into a more severe event, such as a large break loss of coolant accident (LOCA), or resulting in a breach of the reactor coolant pressure boundary.

In the event of an operation of the ECCS during power operation, the DBA analyses assume that the operators will terminate Safety Injection (SI) according to the plant EOPs. One of the reasons for the SI termination in the EOPs is based on the RCS subcooling indications. The inputs to the RCS SMM are RCS hot leg and cold leg temperatures from loop RTDs, CET temperature, RCS wide range pressure, and pressurizer pressure. Since these indications are independently displayed in the control room, the RCS SMM provides redundant calculation and display functions as a backup of these input indications.

Main Steam Line Break (UFSAR Section 15.4.2.1)

A steam line break accident involves the sudden rupture of a high-pressure steam pipe, causing rapid, dangerous release of superheated steam, potentially leading to severe burns, fatalities, structural damage, and asbestos contamination. At Farley, Units 1 and 2, it is a main steam line break (MSLB) that rapidly cools the reactor (primary loop), challenging safety systems. One of the reasons for termination of SI is based on the RCS subcooling.

Steam Generator Tube Rupture (UFSAR Section 15.4.3)

A steam generator tube rupture (SGTR) in Farley, Units 1 and 2, is a DBA where a tube breaks, creating a direct leak path from the radioactive primary system to the non-radioactive secondary system, potentially releasing radionuclides via safety valves. This causes primary coolant to flow into the steam generator, leading to rapid water level increase, pressure changes, and safety system activation to prevent core damage and control radioactive release, demanding quick operator action. One of the reasons for operator action is to cooldown and depressurize is the RCS subcooling indications to assure that the pressurizer is not overfilled.

DBA Conclusion The three DBA events described above make use of the SMM for certain operator actions. As stated by the licensee in the LAR, if the SMM indication is not available, operators are trained to use alternate indications to calculate subcooling. As described above in Section 3.2 of this SE, the underlying inputs to the SMM remain available on qualified, safety-grade instrumentation, with redundant channels in the control room. Therefore, NRC staff finds that the DBA analysis in Chapter 15 of the Farley, Units 1 and 2, UFSAR will not be impacted and the proposed change to remove the SMM from TS acceptable.

Severe Accident Management Guidelines (SAMGs)

RCS subcooling margin is not used in the SAMGs as a primary indication and is not expected to be a reliable indication after significant core damage has occurred. The Farley SAMG technical support guideline for instrumentation states that subcooling margin monitor is not accurate after significant core damage. It does, however, list the RCS SMM as a secondary instrument that could be available for core exit thermocouples.

This program includes training and a procedure to manually calculate subcooling margin, using control room pressure and temperature instruments.

Additional Plant-Specific Information Section 4.1 of the NRC SER of WCAP-15981-NP-A requires that amendment requests based on WCAP-15981-NP-A submit additional plant-specific information. SNC provided the plant-specific information in section 3.6 of the letter dated June 30, 2025.

SNC has confirmed this LAR contains only a single PAM function change and not the indications for accident management such as DBA analysis, EOPs, SAMGs similar to Tables 7 of WCAP-15981-NP-A. The licensee states that Requirements 7 and 8 are unnecessary since this amendment request contains a single PAM function change.

3.4 Review of Risk Information Provided in the LAR TR WCAP-15981-NP, Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants, (ML103560687) was not submitted as a risk-informed application pursuant to RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, (ML17317A256), but uses Probabilistic Risk Assessment (PRA) information as one element of the overall method to determine the instrumentation to be included in the Post Accident Monitoring (PAM) Technical Specifications (TSs).

TR WCAP-15981-NP states that the licensee should ensure that the internal events PRA is technically adequate for the application, but that only a limited assessment is required since this it is not a risk-informed application.

TR WCAP-15981-NP states that the assessment of PRA technical adequacy needs to consider the areas of the accident sequence analysis and the Human Reliability Analysis (HRA) to assure that the treatment of operator actions based on plant instrumentation is appropriate. It further states that the licensee should confirm that all operator actions potentially impacted by the subject instruments have been identified, that the treatment of these operator actions in the PRA is appropriate (including the human error probability values and dependencies), and that there are no peer review comments that can affect the conclusions regarding instrument importance.

TR WCAP-15981-NP, Table 14, indicates that the licensee should confirm that the PRA reflects the as-built, as operated design, and that any plant modifications and operational changes not reflected in the PRA do not impact the plant-specific PAM instrumentation application.

TR WCAP-15981-NP states that the assessment of technical adequacy is limited to the internal events PRA.

TR WCAP-15981-NP states that the plant-specific Risk Achievement Worth (RAW) and Fussel-Vessey (FV) importances are to be used to identify the risk important operator actions for both Corde Damage Frequency (CDF) and Large Early Release Frequency (LERF).

Licensees that submit an LAR based on TR WCAP-15981-NP need to submit the following plant-specific information regarding its PRA:

A general description of the PRA, including the scope of the analyses, PRA update history (including version peer reviewed, version(s) in which peer review comments were addressed, and version used for PAM application), and the licensees PRA updating and quality assurance process.

A description of the most relevant peer reviews, a characterization of the peer review findings, a summary of the status of resolution of the peer review comments, and a listing of all unresolved facts and observations that potentially impact the application of TR WCAP-15981-NP.

A conclusion regarding PRA quality assessment for the PAM TS application, and verification that the quality is acceptable for the application. This should include confirmation that the PRA reflects the as-built, as-operated design, and that any recent plant modifications and operational changes not reflected in the PRA do not impact the plant-specific PAM application; all peer review comments have been resolved or do not impact plant-specific PAM application; the PRA and HRA is sufficiently complete and applicable for evaluating the risk associated with the PAM application.

Listings of the important operator actions identified based on RAW and FV importance values for CDF and for LERF, along with these values.

Additions to the list of important operator actions based on review of results from the plant-specific external event assessments, or verification that the plant specific risk assessments do not result in identification of additional risk significant operator actions or variables/instruments.

A listing of variables/instruments related to the important operator actions. This should indicate how each variable/instrument considered in the methodology application was related to or mapped to a PRA model element or operator action.

In LAR Section 3.3 the licensee provided a general description of its PRA including scope, update history, peer reviews, version used for this application, and its PRA update and quality assurance process.

On August 23, 2019, the NRC issued amendment nos. 225 and 222 for Farley that allowed the use of Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, (TSTF-505) (ML19175A243). In the accompanying safety evaluation, the NRC staff found the licensees PRA models to be technically adequate to support the Risk Informed Completion Time (RICT) program. Because the NRC staff found the licensees PRA models to be technically adequate for the TSTF-505 application and determined that they can be used in regulatory decision making, the NRC staff finds the licensees PRA models are technically adequate for this application. As indicated in TR WCAP-1 5981-NP the assessment of PRA technical adequacy is less than what is required for a risk-informed application because the instrumentation importance in the PRA is just one of several considerations in the WCAP methodology.

SNC stated that its PRA provides an accurate representation of the design and operation of Farley, Units 1 and 2, and that all its models including Internal Events and Internal Flooding, Fire Hazards, Seismic, and Other External Hazards, have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. In addition, SNC stated that its PRA Configuration Control processes ensure that the PRA models used in this application acceptably reflect the as-built and as-operated plant.

In Section 3.3 of the LAR, the licensee stated that its PRA models have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (ML090410014) and Revision 3 (ML20238B871).

SNC stated that the full scope internal events (including internal flooding) PRA peer review was subject to a self-assessment and a full scope peer review conducted in March 2010 against ASME/ANS RASa-2009, Addenda A to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, December 2008 (ML092870592), and any clarifications and qualifications provided in the NRC endorsement of the Standard contained in Revision 2 to RG 1.200.

SNC stated that a finding closure review was conducted on the internal events (including internal flooding) in October 2018 and that closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) as accepted by the NRC in the letter dated May 3, 2017 (ML17079A427). The licensee stated that all Finding level F&Os were closed out.

The licensee stated that its fire PRA (FPRA) was subject to a self-assessment and a full scope peer review conducted in October 2011 by the Pressurized Water Reactors Owners Group (PWROG) against all technical elements in Section 4 of ASME/ANS RA-Sa-2009 and any clarifications and qualifications provided in RG 1.200, Revision 2. SNC stated that after the full scope peer review, two focused-scope peer reviews (February 2018 and July 2018) and one Appendix X closure review (September 2018) were held and that in the September 2018 closure review, all F&Os from the October 2011 full scope peer review, all F&Os from the February 2018 focused-scope peer review, and all F&Os from the July 2018 focused scope peer review were closed out.

SNC stated that in November-December 2019, a focused-scope peer review of the internal events, internal flooding, fire, and seismic PRAs against applicable requirements of the ASME/ANS PRA standard was conducted of the following PRA model upgrades:

Reactor Coolant Pump Shutdown Seal Model (applicable to both internal events and internal flooding PRA models);

Main Control Room Abandonment (applicable to fire PRA model only); and, Diverse and Flexible Coping (FLEX) modeling with FLEX HRA (applicable to all PRA models).

SNC stated that the FLEX HRA portion of the review was performed using the Integrated Human Event Analysis System (IDHEAS) Method based on Electric Power Research Institute (EPRI) FLEX HRA Report 3002013018 and that as a result of this focused-scope peer review, a total of seven F&Os were generated, all of which were Suggestion type F&Os. The licensee stated that a Focused Scope Peer Review (FSPR) was performed in January 2023, against RG 1.200, Revision 3, to review some significant changes to the Farley PRA and that these included revisions to the FPRA to incorporate the updated methods provided in NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants (NUREG-2230, EPRI 3002016051) (ML20157A148), and NUREG-2178, Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) (NUREG-2178), (ML16110A140), SNC stated that in the May 2023 Appendix X closure review, all F&Os from the January 2023 FSPR were closed out.

SNC stated that RCS subcooling margin is provided to determine SI termination and reinitiation and depressurization and cooldown progression. The RCS SMM measures saturation/superheat margin. The function of the RCS SMM is to calculate the subcooled margin which is the difference between the measured temperature of the reactor coolant and the saturation temperature. The saturation temperature is calculated from the minimum primary system pressure input. A maximum or representative temperature input is used for the measured value, which could come from a resistance temperature detector (RTD) loop, or a representative core exit thermocouple (CET). Since the RCS SMM is not relied upon to support any PRA modeled system to mitigate core damage or large early release, the RCS SMM has no contribution to the total CDF and LERF. Due to no contribution to CDF and LERF, the RCS SMM and any components associated with the RCS SMM are not High Safety Significant (HSS) or Low Safety Significant (LSS) with respect to Internal Events, Internal Flooding, Fire, or Seismic risk hazards.

SNC stated that the proposed change does not add important operator actions and is not risk-based in nature.

The important operator actions related to the RCS subcooling margin are discussed in Sections 3.1 and 3.2 of the LAR. The licensee stated that the manual actions specifically credited in the design basis accident analyses for which no automatic control is provided include: Inadvertent Operation of the Emergency Core Cooling System During Power Operation (UFSAR Section 15.2.14); Steam Line Break (UFSAR Section 15.4.2.1); and, Steam Generator Tube Rupture (UFSAR Section 15.4.3).

SNC stated that it reviewed its EOPs to identify key operator actions that would be implemented during an optimal recovery from Design Basis Accidents. The licensee further stated that the RCS SMM indication is used in multiple Farley EOPs, but if the RCS SMM indication is not available, Farley Operators are trained to use alternate indications to calculate subcooling. The licensee concluded that the proposed change to remove the RCS SMM function from the PAM TS 3.3.3 would retain the core subcooling monitor (RCS SMM) function in Farley UFSAR Table 7.5-1 as a Type B, Category 2 variable and would not require procedural step changes to the Farley EOPs.

Based on the information provided in the LAR, the NRC staff determined that the LAR included the information described in TR WCAP-15981-NP, as modified by the safety evaluation, including a general description of the PRA, a description of the most relevant peer reviews, a conclusion regarding PRA quality assessment for this LAR, discussions of important operator actions, and, an evaluation of variables/instruments related to important operator actions.

Because SNC submitted the proposed change in accordance with TR WCAP-15981-NP and the risk information provided in the LAR meets the criteria of TR WCAP-15981-NP, as modified by the safety evaluation, the NRC staff concludes that the proposed change is acceptable. In addition, the licensee has confirmed the applicability of TR WCAP-15981-NP to their plant and completed all parts of the stated methodology and provided the information required for the NRC to approve the proposed change. In addition, the proposed change is not risk-informed and the system described in the proposed change is not relied upon to support any PRA modeled system to mitigate core damage or large early release.

The NRC staff notes that, although the original application requested a 60-day implementation period, the licensee subsequently made a verbal request to extend the implementation period to 120 days. The licensee indicated that this extension was requested to accommodate the possibility that NRC approval may be granted earlier than originally estimated.

3.5 Technical Conclusion TR WCAP-15981-NP recommends that RCS subcooling be reclassified as Type B Category 3 and be relocated from the TS. The justification provided in TR WCAP-1 5981-NP is that core exit temperature and RCS pressure are inputs to RCS subcooling and provide the most direct indication of the accomplishment of the core cooling function. RCS subcooling provides information to indicate whether the core cooling function is being accomplished. Therefore, it is a Type B variable and is a backup to core exit temperature and RCS pressure and should be reclassified as a Category 3 variable. Further, TR WCAP-15981-NP concluded that RCS subcooling does not satisfy the criteria in 10 CFR 50.36(c)(2)(ii) that list the content to be included in the TS. The NRC staff agrees that RCS subcooling is not a key variable for the core cooling function and may be relocated from the TS.

The NRC staff determined that the regulatory requirements of 10 CFR 50.36(c)(2) will continue to be met because the LCO will continue to describe the lowest functional capability or performance level of equipment required for safe operation of the facility. Based on these determinations, the NRC staff concludes the proposed changes to TS Table 3.3.3-1 are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Alabama State official was notified of the proposed issuance of the amendments on December 15, 2025, and the State official confirmed that the State of Alabama had no comments (ML25350A011).

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change 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, and there has been no public comment on such finding published in the Federal Register on September 2, 2025 (90 FR 42456). 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 Commissions 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: R. Hardin, NRR G. Vasudevamurthy, NRR F. Forsaty, NRR J. Robinson, NRR A. Russell, NRR Date of Issuance: January 8, 2026

ML25353A551 NRR-058 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DRA/APLB/BC NRR/DSS/SNSB/BC NAME ZTurner (JLamb for)

KZeleznock EDavidson NDifrancesco DATE 12/19/2025 12/29/2025 01/06/2026 01/07/2026 OFFICE NRR/DEX/EEEB/BC NRR/DEX/EICB/BC NRR/DSS/STSB/BC (A)

NRR/DORL/LPL2-1/BC NAME WMorton (DMurdock for) SDarbali SMehta (KWest for)

MMarkley (SWilliams for)

DATE 01/06/2026 01/05/2026 01/06/2026 01/08/2026 OFFICE NRR/DORL/LPL2-1/PM NAME ZTurner (JLamb for)

DATE 01/08/2026