ML24351A080

Issuance of Amendments Regarding Revision to Technical Specifications to Use Online Monitoring Methodology

ML24351A080
Person / Time
Site:	Hatch, Farley
Issue date:	01/24/2025
From:	John Lamb Plant Licensing Branch II
To:	Coleman J Southern Nuclear Operating Co
Turner, Zachary
References
EPID L-2024-LLA-0058
Download: ML24351A080 (1)

Text

January 24, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, AND EDWIN I.

HATCH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO USE ONLINE MONITORING METHODOLOGY (EPID L-2024-LLA-0058)

Dear Jamie Coleman:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 252 to Renewed Facility Operating License No. NPF-2 and Amendment No. 249 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, respectively; and Amendment No. 325 to Renewed Facility Operating License No. DPR-57 and Amendment No. 270 to Renewed Facility Operating License No. NPF-5 for the Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the License and Technical Specifications (TSs) in response to your application dated May 3, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24124A133).

The amendments revise Technical Specification (TS) 1.1, Use and Application Definitions and add a new TS 5.5.21, Online Monitoring Program, for Farley, Units 1 and 2, and TS 5.5.17, Online Monitoring Program, for Hatch, Units 1 and 2, respectively. The amendments would allow use of an online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed amendments are based on the NRC staff-approved topical report AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (ML21235A493).

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

If you have questions, you can contact me at 301-415-3100 or John.Lamb@nrc.gov.

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, 50-364, 50-321, and 50-366

Enclosures:

1. Amendment No. 252 to NPF-2

2. Amendment No. 249 to NPF-8

3. Amendment No. 325 to DPR-57

4. Amendment No. 270 to NPF-5

5. 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. 252 Renewed License No. NPF-2

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southern Nuclear Operating Company (Southern Nuclear), dated May 3, 2024, 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 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 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 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; 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. 252, 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 90 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 the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.01.24 13:51:03 -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. 249 Renewed License No. NPF-8

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southern Nuclear Operating Company (Southern Nuclear), dated May 3, 2024, 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 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 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 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; 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. 249, 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 90 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 the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.01.24 13:51:45 -05'00'

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 252 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 249 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the License and Appendix A Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Pages Insert Pages License License License No. NPF-2, page 4 License No. NPF-2, page 4 License No. NPF-8, page 3 License No. NPF-8, page 3 TSs TSs 1.1-1 1.1-1 1.1-3 1.1-3 1.1-5 1.1-5 5.5-18 5.5-18 5.5-19

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 252 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 252, 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. 249 (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 2821 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, 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

1.0 USE AND APPLICATION 1.1 Definitions Definitions 1.1

---

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)

CHANNEL CALIBRATION Farley Units 1 and 2 Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, interlock, and trip functions.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

1.1-1 Amendment No. 252 (Unit 1)

Amendment No. 249 (Unit 2)

1.1 Definitions E-AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Farley Units 1 and 2 Definitions 1.1 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 1.1-3 (continued)

Amendment No. 252 (Unit 1)

Amendment No. 249 (Unit 2)

1.1 Definitions PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

QUADRANT POWER TILT RA TIO (QPTR)

RATED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM(RTS)RESPONSE TIME SHUTDOWN MARGIN (SOM)

Farley Units 1 and 2 Definitions 1.1 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2821 MWt.

The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

SOM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SOM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and 1.1-5 (continued)

Amendment No. 252 (Unit 1)

Amendment No. 249 (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.20 5.5.21 Risk Informed Completion Time Program (continued)

If there is a high degree of confidence, based on the evidence collected, that there is no CC failure mechanism that could affect the redundant components, the RICT calculation may use nominal CC factor probability.

If a high degree of confidence cannot be established that there is no CC failure mechanism that could affect the redundant components, the RICT shall account for the increased possibility of CC failure. Accounting for the increased possibility of CC failure shall be accomplished by one of two methods. If one of the two methods listed below is not used, the Technical Specifications Front Stop shall not be exceeded.

1.

The RICT calculation shall be adjusted to numerically account for the increased possibility of CC failure, in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG. Specifically, when a component fails, the CC failure probability for the remaining components shall be increased to represent the conditional failure probability due to CC failure of these components, in order to account for the possibility the first failure was caused by a CC mechanism.

OR

2.

Prior to exceeding the front stop, RMAs not already credited in the RICT calculation shall be implemented. These RMAs shall target the success of the redundant and/or diverse SSC of the failed SSC and, if possible, reduce the frequency of initiating events which call upon the function(s) performed by the failed SSCs. Documentation of RMAs shall be available for NRC review.

h.

A RICT entry is not permitted, or a RICT entry made shall be exited, for any condition involving a TS loss of function if a PRA Functionality determination that reflects the plant configuration concludes that the LCO cannot be restored without placing the TS inoperable trains in an alignment which results in a loss of functional level PRA success criteria.

Online Monitoring Program This program provides controls to determine the need for calibration for pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

(continued)

Farley Units 1 and 2 5.5-18 Amendment No. 252 (Unit 1)

Amendment No. 249 (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.21 Online Monitoring Program (continued)

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (proprietary version. The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with NRC approved methodology during the plant operating cycle.

1) Analysis of on line monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration

check,

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the on line monitoring data analysis.

b.

Performance of a calibration checks of any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitter at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

Farley Units 1 and 2 5.5-19 Amendment No. 252 (Unit 1)

Amendment No. 249 (Unit 2)

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 325 Renewed License No. DPR-57

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated May 3, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 Commissions regulations set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and the first part of paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 325, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 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 No. DPR-57 and Technical Specifications Date of Issuance: January 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.01.24 13:53:05 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 325 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the 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 Pages Insert Pages License License License No. DPR-57, page 4 License No. DPR-57, page 4 TSs TSs 1.1-1 1.1-1 1.1-6 1.1-6 5.0-19b Renewed License No. DPR-57 Amendment No. 325 for sample analysis or instrument calibration, or associated with radioactive apparatus or components (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 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 following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal.

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 325, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

(3)

Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. NPF-5

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated May 3, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 Commissions regulations set forth in 10 CFR Chapter I; 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 hereby amended by page 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-5 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 270 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 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 No. NPF-5 and Technical Specifications Date of Issuance: January 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.01.24 13:54:39 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 270 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the 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 Pages Insert Pages License License License No. NPF-5, page 4 License No. NPF-5, page 4 TSs TSs 1.1-1 1.1-1 1.1-4 1.1-4 1.1-5 1.1-5 1.1-6 1.1-6 1.1-7 1.1-7 5.0-19b

Renewed License No. NPF-5 Amendment No. 270 (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 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 following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 270 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(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 license supported by a favorable evaluation by the Commission.

(a)

Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c),

as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020.

Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would 2

The original licensee authorized to possess, use, and operate the facility with Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 AMENDMENT NO. 252 TO RENEWED FACILITY OPERATING LICENSE NPF-2 AMENDMENT NO. 249 TO RENEWED FACILITY OPERATING LICENSE NPF-8 EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 AMENDMENT NO. 325 TO RENEWED FACILITY OPERATING LICENSE DPR-57 AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NOS. 50-348, 50-364, 50-321, AND 50-366

1.0 INTRODUCTION

By application dated May 3, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24124A133), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, and Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2.

The amendments would revise Technical Specification (TS) 1.1, Use and Application Definitions and add a new TS 5.5.21, Online Monitoring Program, for Farley, Units 1 and 2, and TS 5.5.17, Online Monitoring Program, for Hatch, Units 1 and 2, respectively. The amendments would allow use of an online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed amendments are based on the U.S.

Nuclear Regulatory Commission (NRC)-approved topical report AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (ML21235A493).

The NRC staff issued a safety evaluation (SE) approving the dash A version of the AMS-TR-0720R2-A on August 11, 2021, Final Safety Evaluation for AMS [Analysis and Measurement Services] Online Monitoring Topical Report (ADAMS Package ML21179A060).

SNC has not proposed any deviations from the approved AMS-TR-0720R2-A.

2.0 REGULATORY EVALUATION

2.1 Regulations and Guidance The NRC staff considered the following regulatory requirements and guidance in reviewing the concepts being implemented in the Farley and Hatch OLM programs:

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(1)(ii)(A) states, in part, that limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions. This section requires that, where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded. It also requires that the licensee take appropriate action and notify the NRC if it is determined that an automatic safety system does not function as required. The licensee is then required to review the matter and record the results of the review.

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

The regulation in 10 CFR 50.36(c)(5) states, in part, that Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The regulation in 10 CFR 50.55a(h)(2) states:

o Protection systems. For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements in IEEE [Institute of Electrical and Electronics Engineers] Std 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or the requirements in IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, or the requirements in IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995. For nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.

The Farley, Units 1 and 2, and the Hatch, Unit 2, construction permits were issued on August 16, 1972, August 16, 1972, and December 27, 1972, respectively; therefore, the criteria of IEEE Std 279-1971 are used as a basis for this SE. The Hatch, Unit 1, construction permit was issued on September 30, 1969; therefore, Hatch Unit 1 must be consistent with its licensing basis. Hatch Unit 1 was designed and constructed based on the proposed General Design Criteria (GDC), issued for comment in the Federal

Register (32 FR 10213) by the Atomic Energy Commission (AEC) on July 11, 1967 (hereafter draft GDC). The AEC published the final rule that added Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, in the Federal Register on February 20, 1971 (36 FR 3255, as corrected, 36 FR 12733; July 7, 1971)

(hereafter adopted GDC or GDC). As discussed in the NRC Staff Requirements Memorandum (SRM) for SECY-92-223, Resolution of Deviations Identified During the Systematic Evaluation Program, dated September 18, 1992 (ML18100B279), the Commission decided not to apply the adopted GDC to plants with construction permits issued prior to May 21, 1971. The Commission stated in this SRM that the plants licensed before the GDC were formally adopted were evaluated on a plant-specific basis, determined to be safe, and that current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the intent of the GDC. This SRM also notes that the GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. The Hatch Unit No. 1 Updated Final Safety Analysis Report (UFSAR), Appendix F, Conformance to the Atomic Energy Commission (AEC) Criteria contains an evaluation of the Hatch Unit No. 1 design basis against the adopted GDC.

Clause 4.3, Quality of Components and Modules, of IEEE 279-1971 states the following:

Components and modules shall be of a quality that is consistent with minimum maintenance requirements and low failure rates. Quality levels shall be achieved through the specification of requirements known to promote high quality, such as requirements for design, for the derating of components, for manufacturing, quality control, inspection, calibration, and test.

Clause 4.9 of IEEE 279-1971 states the following:

Means shall be provided for checking, with a high degree of confidence, the operational availability of each system input sensor during reactor operation. This may be accomplished in various ways, for example:

1)

By perturbing the monitored variable, or

2)

Within the constraints of paragraph 4.11, by introducing and varying, as appropriate, a substitute input to the sensor of the same nature as the measured variable, or

3)

By cross-checking between channels that bear a known relationship to each other and that have readouts available.

Appendix A to 10 CFR Part 50, General Design Criterion (GDC) 13, Instrumentation and control, states that instrumentation shall be 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.

Appendix A to 10 CFR Part 50, GDC 20, Protection system functions, states that the protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The following are the specific NRC guidance documents applicable to the NRC staffs evaluation of the Farley and Hatch OLM programs:

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, Branch Technical Position (BTP) 7-12, Guidance on Establishing and Maintaining Instrument Setpoints, Revision 6, (ML16019A200).

Regulatory Guide (RG) 1.105, Revision 4, Setpoints for Safety-Related Instrumentation, (ML20330A329), February 2021. This RG describes an approach that is acceptable to the NRC staff to meet regulatory requirements to ensure that:

(a) setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and (b) the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant technical specifications (TS). This RG endorses American National Standards Institute (ANSI)/International Society of Automation (ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. Among other things, the ANSI/ISA 67.04.01 standard provides criteria for assessing the performance of safety related instrument channels to ensure they remain capable of achieving their required safety functions in a reliable manner. This performance monitoring process requires the establishment of acceptable As-Found tolerance limits used to check whether an instrument channel is functioning as required, and the establishment of acceptable As-Left tolerance limits used to establish the maximum allowed deviation from the desired setpoint of the instrument channel and still be considered as within calibration.

The following guidance documents provide information associated with the periodic calibration of safety related instrument channels that was considered by the NRC staff during its evaluation of the Farley and Hatch OLM programs:

Generic Letter 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle, dated April 2, 1991 (ML031140501), provides guidance on acceptable methods for licensees to justify an increase in calibration surveillance intervals using as-found and as-left calibration data from past calibration surveillances.

Regulatory Issue Summary (RIS) 2006-017, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specifications, regarding Limiting Safety System Settings during Periodic Testing and Calibration of Instrument Channels, dated August 24, 2006 (ML051810077), provides regulatory clarification on NRC staff positions in terms of the appropriate determination of TS-related instrument channel operability. The RIS clarifies NRC staff positions about the appropriate establishment of as-found and as-left acceptance tolerances.

2.2 Description of Proposed Changes for Farley, Units 1 and 2 In its letter dated May 3, 2024, the licensee proposed the following specific changes to the TSs for Farley, Units 1 and 2.

TS 1.1 - CHANNEL CALIBRATION The current CHANNEL CALIBRATION definition states:

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include and inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

The revised CHANNEL CALIBRATION definition would state (changes indicated in bold):

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, interlock, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include and inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

TS 1.1 - ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The current ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME definition states:

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series

of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

The revised ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME definition would state (changes indicated in bold):

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

TS 1.1 - REACTOR TRIP SYSTEM (RTS) RESPONSE TIME The current REACTOR TRIP SYSTEM (RTS) RESPONSE TIME definition states:

The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

The revised REACTOR TRIP SYSTEM (RTS) RESPONSE TIME definition would state (changes indicated in bold):

The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

TS 5.5.21 - Online Monitoring Program There is currently no TS 5.5.21.

The new TS 5.5.21 would state:

This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.1 The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version).

The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1.

Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check,

2.

Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3.

Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4.

Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration checks of any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

1 The NRC staff acknowledges that this sentence was not addressed in the licensees LAR section 2.4 Description of the Proposed Changes; however, it was included in Attachments 1 through 6 to the LAR and is, therefore, part of the LAR.

2.3 Description of Proposed Changes for Hatch, Unit 1 In its letter dated May 3, 2024, the licensee proposed the following specific changes to the TSs for Hatch, Unit No. 1.2 TS 1.1 - CHANNEL CALIBRATION The current CHANNEL CALIBRATION definition states:

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

The revised CHANNEL CALIBRATION definition would state (changes indicated in bold):

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

TS 1.1 - REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME The current REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME definition states:

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be 2 The NRC staff notes that Hatch, Unit 1, TSs do not have TS 1.1 definitions for EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME and ISOLATION SYTEM RESPONSE TIME because Hatch Unit 1 is a pre-GDC plant, as explained in Section 2.0 of this safety evaluation. The licensee is incorporating the pressure, level, and flow transmitters in the ECCS and Isolation systems included in the AMS topical report through the adoption of TS 5.5.17.

measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

The revised REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME definition would state (changes indicated in bold):

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

TS 5.5.17 - Online Monitoring Program There is currently no TS 5.5.17.

The new TS 5.5.17 would state:

This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.3 The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version).

The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1.

Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check,

2.

Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3.

Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 3 The NRC staff acknowledges that this sentence was not addressed in the licensees LAR section 2.4 Description of the Proposed Changes; however, it was included in Attachments 1 through 6 to the LAR and is, therefore, part of the LAR.

4.

Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check of any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

2.4 Description of Proposed Changes for Hatch, Unit 2 In its letter dated May 3, 2024, the licensee proposed the following specific changes to the TSs for Hatch, Unit 2.

TS 1.1 - CHANNEL CALIBRATION The current CHANNEL CALIBRATION definition states:

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

The revised CHANNEL CALIBRATION definition would state (changes indicated in bold):

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be

performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

TS 1.1 - EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The current ECCS RESPONSE TIME definition states:

The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

The revised ECCS RESPONSE TIME definition would state (changes indicated in bold):

The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

TS 1.1 - ISOLATION SYSTEM RESPONSE TIME The current ISOLATION SYSTEM RESPONSE TIME definition states:

The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

The revised ISOLATION SYSTEM RESPONSE TIME definition would state (changes indicated in bold):

The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence delays, where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification

has been previously reviewed and approved by the NRC.

TS 1.1 - REACTOR PROECTION SYSTEM (RPS) RESPONSE TIME The current REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME definition states:

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

The revised REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME definition would state (changes indicated in bold):

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

TS 5.5.17 - Online Monitoring Program There is currently no TS 5.5.17.

The new TS 5.5.17 would state:

This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.4 The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version).

The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

4 The NRC staff acknowledges that this sentence was not addressed in the licensees LAR section 2.4 Description of the Proposed Changes; however, it was included in Attachments 1 through 6 to the LAR and is, therefore, part of the LAR.

1.

Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check,

2.

Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3.

Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4.

Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check of any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

3.0 TECHNICAL EVALUATION

3.1 Description of the OLM Program The Farley, Units 1 and 2, and Hatch, Units 1 and 2, OLM programs are based on the AMS OLM topical report, AMS-TR-0720R2-A, which provides a methodology for performing OLM of the output signals of pressure and differential pressure transmitters. This methodology was developed by AMS to be used in nuclear power plants as an analytical tool to measure sensor calibration performance during plant operation between scheduled refueling outages.

3.2 Description and Evaluation of TS Changes The licensees submittal requested approval to implement its OLM program by revising appropriate TS 1.1, Use and Application Definitions, and adding new sections in TS 5.5.17 (Hatch) and 5.5.21 (Farley), Online Monitoring Program. SNC proposes to use the OLM methodology presented in topical report AMS-TR-0720R2-A as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM analysis results. A mark-up of the TS pages was provided in Attachments 1 through 3 to the LAR dated May 3, 2024.

3.3 TS 1.1, Use and Application Definitions For Farley, Units 1 and 2, and Hatch, Unit 1 and 2, the TS definition for the term CHANNEL CALIBRATION is being revised to account for the approved OLM methodologies. The specific change allows transmitters that are included in the licensees OLM program to be excluded from

the scope of instrumentation to be periodically calibrated within the Frequency in the Surveillance Frequency Control Program.

The NRC staff reviewed this change considering the context of the OLM program. This change is acceptable, because the OLM processes include an acceptable method for identifying performance issues as they occur and initiating corrective actions when pre-established OLM limits are exceeded. The corrective actions also include performing instrument calibrations as necessary to restore instrument performance to within acceptable parameters. Data collected during OLM activities is also used to adjust OLM limits such that poorly performing instruments will be calibrated at greater frequencies to address any potential impact on long term plant performance.

For Farley, Units 1 and 2, the TS definition for the terms, ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME and REACTOR TRIP SYSTEM (RTS) RESPONSE TIME, are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The existing exclusion from response time testing is based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

The NRC staff finds this revised definition to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definition for this term allowed exclusion from response time testing based on the fact that instrument failures that affect response time would also affect calibration performance and would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support exclusion of these instruments from response time testing.

For Hatch, Unit 1, the TS definition for the term REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME is being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The existing exclusion from response time testing is based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

The NRC staff finds this revised definition to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definition for this term allowed exclusion from response time testing based on the fact that instrument failures that affect response time would also affect calibration performance and would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support exclusion of these instruments from response time testing.

For Hatch, Unit 2, the TS definition for the terms, EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME, ISOLATION SYSTEM RESPONSE TIME, and REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME, are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The exclusion from response time testing is based on the periodic channel calibration

program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

The NRC staff finds this revised definition to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definition for this term allowed exclusion from response time testing based on the fact that instrument failures that affect response time would also affect calibration performance and would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support exclusion of these instruments from response time testing.

3.4 New TS 5.5.17 (Hatch, Unit Nos. 1 and 2) and 5.5.21 (Farley, Units 1 and 2), Online Monitoring Program This new TS section provides a description of the AMS based OLM program. The new TS stipulates that the OLM program must be implemented in accordance with the NRC staff-approved topical report, AMS-TR-0720R2-A, TS 5.5.17 (Hatch, Units 1 and 2) and 5.5.21 (Farley, Units 1 and 2) lists the key elements of the OLM program. The NRC staff reviewed the TS description of the OLM program and found that it is consistent with the program descriptions provided in the approved topical report AMS-TR-0720R2-A. To verify that the Farley and Hatch OLM programs are being implemented in accordance with the NRC-approved topical report, the NRC staff conducted an audit per audit plan (ML24149A049) and reviewed several Farley and Hatch specific reports that documented program implementation activities. These reports are listed and described in the staffs audit report (ML24302A297). The NRC staff confirmed that all key elements of the OLM program as described in AMS-TR-0720R2-A would be implemented satisfactorily.

The NRC staff also reviewed SNCs responses to each of the Application Specific Action Items (ASAIs) that were provided in Section 4.0 of the NRC staffs SE for the AMS OLM TR. The licensees responses to these ASAIs are provided in Section 3.4 of the LAR dated May 3, 2024.

Section 3.5 of this SE provides the staffs evaluation of the licensees responses to the AMS-TR-0720R2-A ASAIs. The NRC determined that all plant specific actions have been performed at an acceptable level and the Farley and Hatch OLM programs would be implemented in conformance with the approved TR AMS-TR-0720R2-A.

3.5 AMS TR-0720R2-A - ASAIs The NRC identified five ASAIs in the safety evaluation of the AMS OLM program TR. The licensee provided responses to each of these ASAIs in Section 3.4 of the LAR dated May 3, 2024. The NRC staff evaluation of these ASAIs is provided below:

3.5.1 AMS-TR-0720R2-A ASAI 1 - Evaluation and Proposed Mark-up of Existing Plant Technical Specifications Evaluation and Proposed Mark-up of Existing Plant Technical Specifications - When preparing a license amendment request to adopt OLM methods for establishing calibration frequency, licensees should consider mark-ups that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance. Such TS changes

would need to include appropriate mark-ups of the TS tables describing limiting conditions for operation and surveillance requirements, the technical basis for the changes, and the administrative programs section.

This evaluation is provided in Section 3.2 of this SE. The licensee provided mark-ups of the applicable TSs that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance for transmitters that are included in the OLM program. Mark-ups of the TS BASES were also provided which describe the technical basis for the online monitoring program. Therefore, the staff finds the criteria of ASAI 1 are met.

3.5.2 AMS-TR-0720R2-A ASAI 2 - Identification of Calibration Error Source When determining whether an instrument can be included in the plant OLM program, the licensee shall evaluate calibration error source in order to account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system. Calibration errors identified through OLM should be attributed to the transmitter until testing can be performed on other support devices to correctly determine the source of calibration error and reallocate errors to these other loop components.

The NRC staff conducted an audit of the Farley and Hatch OLM program reports to verify that calibration error sources were being factored into account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system.

The NRC staff found that the OLM program attributes calibration errors to the transmitter unless testing is subsequently performed to determine and reallocate calibration error to other instrument loop components. Therefore, the staff finds the criteria of ASAI 2 are met.

3.5.3 AMS-TR-0720R2-A ASAI 3 - Response Time Test Elimination Basis If the plant has eliminated requirements for performing periodic response time (RT) testing of transmitters to be included in the OLM program, then the licensee shall perform an assessment of the basis for RT test elimination to determine if this basis will remain valid upon implementation of the OLM program and to determine if the RT test elimination will need to be changed to credit the OLM program rather than the periodic calibration test program.

The transmitters that are being incorporated into the OLM programs were exempt from response time testing. The licensee, therefore, performed an assessment of the basis for response time testing exemptions and determined that the OLM program will continue to support exemption from response time testing, because the OLM methods will detect transmitter failures that would affect response time performance. The basis for this exclusion is extended to include transmitters in the OLM program by the TS changes incorporated in TS 1.1, which is evaluated in Section 3.3 of this SE. Therefore, the staff finds criteria of ASAI 3 are met.

3.5.4 AMS-TR-0720R2-A ASAI 4 - Use of Calibration Surveillance Interval Backstop In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe how they intend to apply backstop intervals as a means for mitigating the potential that a process group could be experiencing undetected common mode drift characteristics.

The NRC staff performed an audit review of the backstop calculations performed for the FNP and HNP transmitters being incorporated into the OLM program and confirmed that these calculations were performed in a manner consistent with the processes outlined in the approved AMS OLM topical report for determining maximum calibration intervals. Therefore, the staff finds criteria of ASAI 4 are met.

3.5.5 AMS-TR-0720R2-A ASAI 5 - Use of Criteria other than in AMS OLM TR for Establishing Transmitter Drift Flagging Limit In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe whether they intend to adopt the criteria within the AMS OLM TR for flagging transmitter drift or whether they plan to use a different methodology for determining this limit.

The NRC staff determined that the Farley and Hatch proposed OLM programs are consistent with the AMS OLM topical report AMS-TR-0720R2-A, and therefore, a different methodology is not being employed. Therefore, the staff finds criteria of ASAI 5 are met.

3.6 Technical Summary The NRC staff finds that the licensees proposed implementation of the Farley and Hatch OLM Programs are consistent with the approved TR, AMS-TR -720R2-A.

The regulation in 10 CFR 50.36(a)(1) states, in part, [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.

Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that correspond to the proposed TS changes, to provide the reasons for those TSs.

The TS bases changes were determined to be consistent with the approved AMS OLM topical report methods, and are therefore, acceptable.

The NRC staff determined that implementation of the OLM program for Farley, Units 1 and 2, and Hatch, Units 1 and 2, will continue to support establishment of limiting safety system settings associated with the transmitters that are included in the program. These settings will continue to ensure that associated automatic protective actions will correct abnormal situations before safety limits are exceeded. The surveillance requirements relating to test, calibration, and inspection of these transmitters will also continue to ensure that the adequate quality of systems and components is maintained. Therefore, the NRC staff finds that the requirements of 10 CFR 50.36(c)(1)(ii)(A), 10 CFR 50.36(c)(3), and 10 CFR 50.36(c)(5) would continue to be met. Additionally, the NRC staff finds that the licensees implementation of the OLM Program in accordance with approved TR, AMS-TR-720R2-A will continue to meet the requirements of Appendix A to 10 CFR Part 50, GDC 13 and 20. The licensee will still be required to notify the NRC if an associated automatic safety system does not function as required.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State and Alabama State officials were notified on December 13, 2024, of the proposed issuance of the amendments. The State

of Georgia officials had no comments on December 20, 2024. The State of Alabama officials had no comments on December 23, 2024.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. 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 (89 FR 49241 dated June 11, 2024), 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 Contributor: G. Blas Rodriguez, NRR Joseph Ashcraft, NRR Date: January 24, 2025

ML24351A080 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NAME JLamb KZeleznock SMehta DATE 12/13/2024 12/18/2024 12/17/2024 OFFICE NRR/DEX/EICB/BC OGC - NLO NRR/DORL/LPL2-1/BC NAME FSacko JEzell MMarkley DATE 12/17/2024 01/17/2025 01/24/2025 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/PM NAME JMinzer Bryant JLamb DATE 01/24/2025 01/24/2025