ML23115A149

Issuance of Amendments No. 218 and 201 Regarding Revision to Technical Specifications to Use Online Monitoring Methodology

ML23115A149
Person / Time
Site:	Vogtle
Issue date:	06/15/2023
From:	John Lamb Plant Licensing Branch II
To:	Brown R Southern Nuclear Operating Co
References
EPID L-2022-LLA-0190
Download: ML23115A149 (26)

Text

June 15, 2023 Mr. R. Keith Brown Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO USE ONLINE MONITORING METHODOLOGY (EPID L-2022-LLA-0190)

Dear Mr. Brown:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 218 to Renewed Facility Operating License NPF-68 and Amendment No. 201 to Renewed Facility Operating License NPF-81 for the Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated December 21, 2022.

The amendments would revise TS 1.1, Use and Application Definitions and add a new TS 5.5.23 Online Monitoring Program. 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-approved topical report AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters.

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

R.

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. 50-424 and 50-425

Enclosures:

1. Amendment No. 218 to NPF-68

2. Amendment No. 201 to NPF-81

3. Safety Evaluation cc: Listserv SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 218 Renewed License No. NPF-68

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the 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 December 21, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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-68 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 218, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this 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 295 days of issuance.

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

Attachment:

Changes to License No. NPF-68 and the Technical Specifications Date of Issuance: June 15, 2023 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2023.06.15 10:45:29 -04'00'

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 201 Renewed License No. NPF-81

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the 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 December 21, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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-81 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 201, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this 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 295 days of issuance.

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

Attachment:

Changes to License No. NPF-81 and the Technical Specifications Date of Issuance: June 15, 2023 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2023.06.15 10:46:20 -04'00'

ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 TO LICENSE AMENDMENT NO. 218 RENEWED FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 201 RENEWED FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of the Licenses 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-68, page 4 License No. NPF-68, page 4 License No. NPF-81, page 3 License No. NPF-81, page 3 TSs TSs 1.1-1 1.1-1 1.1-3 1.1-3 1.1-5 1.1-5 5.5-22 5.5-22 5.5-23 Renewed Operating License NPF-68 Amendment No. 218 (1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 218, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4)

Deleted (5)

Deleted (6)

Deleted (7)

Deleted (8)

Deleted (9)

Deleted (10)

Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training and response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for Implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy Renewed Operating License NPF-81 Amendment No. 201 (2)

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this 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; (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 my be produced by the operation of the facility authorized herein.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 1 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 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 201 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

Vogtle Units 1 and 2 5.5-22 Amendment No. 218 (Unit 1)

Amendment No. 201 (Unit 2) 5.5.22 Risk Informed Completion Time Program

g.

(continued)

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 redundant 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 structures, systems, or components (SSC) of the failed SSC and, if possible, reduce the frequency of initiating events which call upon the function(s) performed by the failed SSC. 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.

5.5.23 Online Monitoring Program 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.

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:

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

Vogtle Units 1 and 2 5.5-23 Amendment No. 218 (Unit 1)

Amendment No. 201 (Unit 2) 5.5.23 Online Monitoring Program (Continued)

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with a 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 check for 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.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 AMENDMENT NO. 218 TO RENEWED FACILITY OPERATING LICENSE NPF-68 AMENDMENT NO. 201 TO RENEWED FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NOS. 50-424, AND 50-425

1.0 INTRODUCTION

By letter dated December 21, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22355A588), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for Vogtle Electric Generating Plant (Vogtle), Units 1 & 2.

The proposed amendments would allow implementation of an on-line instrument monitoring program by revising Technical Specification (TS) 1.1, Use and Application Definitions, and adding a new TS 5.5.23, Online Monitoring Program. SNC proposes to use online monitoring (OLM) methodology as the technical basis to switch from a time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed change is based on the topical report (TR) by Analysis and Measurement Services (AMS), AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (ML21235A494).

The NRC 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 Online Monitoring Topical Report (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 Vogtle OLM program:

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(1)(ii)(A) states 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 (LSSS) 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 the licensee determines 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 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 10 CFR 50.55a(h), Protection and safety systems, states in Section (h)(2), Protection Systems, in part, that 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 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.

The Vogtle 1 & 2 construction permits were issued in 1974; therefore, the criteria of IEEE 279-1971 are used as a basis for this safety evaluation.

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 Vogtle OLM program:

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 Vogtle OLM program:

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 In its letter dated December 21, 2022, the licensee proposed the following specific changes to the TSs for Vogtle, 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 inputs. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, and trip functions. 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 inputs. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, interlock, and trip functions. 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 TIME 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 5.5.23 - Online Monitoring Program There is currently no TS 5.5.23.

The new TS 5.5.23 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.

The Online Monitoring Program must be implemented in accordance with AMSTR-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 a 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 check for 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 AMS OLM Program The Vogtle OLM program is 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 TS 1.1, Use and Application Definitions, and adding a new TS 5.5.23, 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 Attachment 1 to the LAR dated December 21, 2022.

3.3 TS 1.1, Use and Application Definitions 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 (SFCP).

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.

The TS definition for the terms; ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME and REACTOR TRIP SYSTEM 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 previous exclusion from response time testing had been 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 time response 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 continued exemption of these instruments from time response testing.

3.4 New TS 5.5.23, 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-approved topical report, AMS-TR-0720R2-A and TS 5.5.23 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 Vogtle OLM program is being implemented in accordance with the NRC approved topical report, the NRC staff conducted an audit per audit plan (ML23011A274) and reviewed several Vogtle specific reports that documented program implementation activities. These reports are listed and described in the audit report (ML23080A281). The NRC staff confirmed that all key elements of the OLM program as described in AMS-TR-0720R2-A are being satisfactorily implemented.

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 SE for the AMS OLM TR. These licensee responses are provided in Section 3.4 of the LAR dated December 21, 2022. See Section 3.5 of this SE for additional information on the evaluation of AMS-TR-0720R2-A ASAIs. The NRC determined that all plant specific actions have been performed at an acceptable level and the Vogtle OLM program is being 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 topical report.

The licensee provided responses to each of these ASAIs in Section 3.4 of the LAR dated December 21, 2022. 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 markups that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance. Such TS changes would need to include appropriate markups 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. The criteria of ASAI 1 are met.

3.5.2 MS-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 performed an audit review of the Vogtle 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 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 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 program 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 revised in TS 1.1, which is evaluated in Section 3.2.1 of this SE. The criteria of ASAI 3 are, therefore, 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 groups could be experiencing undetected common mode drift characteristics.

The NRC staff performed an audit review of the backstop calculations performed for the Vogtle 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. The criteria of ASAI 4 are, therefore, 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 SNC proposed OLM program is consistent with the AMS OLM topical report AMS-TR-0720R2-A, and therefore, a different methodology is not being employed. The criteria of ASAI 5 are, therefore, met.

3.6 Technical Summary The NRC staff finds that the licensees proposed implementation of the Vogtle OLM Program is consistent with the approved TR, AMS-TR -720R2-A.

The regulation at 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 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) and 10 CFR 50.36(c)(3) would continue to be met. 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 officials were notified on April 13, 2023, of the proposed issuance of the amendments. On May 12, 2023, the State officials informed the NRC that the State of Georgia has no comments.

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. 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 (88 FR 10558 dated February 21, 2023), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Richard Stattel Joe Ashcraft Date: June 15, 2023

ML23115A149 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NRR/DEX/EICB/BC NAME JLamb KGoldstein VCusumano MWaters DATE 04/13/2023 04/26/2023 04/26/2023 04/13/2023 OFFICE OGC - NLO NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MCarpentier MMarkley JLamb DATE 05/12/2023 06/15/2023 06/15/2023