ML20167A318

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Issuance of Amendment Nos. 174 and 174 Regarding Revision to Technical Specifications to Adopt TSTF-563, Revision 0
ML20167A318
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 07/06/2020
From: Dennis Galvin
Plant Licensing Branch IV
To: Peters K
Vistra Operations Company
Galvin D
References
EPID L-2019-LLA-0260, TSTF-563, Rev 0
Download: ML20167A318 (24)
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Text

July 6, 2020 Mr. Ken J. Peters Senior Vice President and Chief Nuclear Officer Attention: Regulatory Affairs Vistra Operations Company LLC Comanche Peak Nuclear Power Plant 6322 N FM 56 P.O. Box 1002 Glen Rose, TX 76043

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 -

ISSUANCE OF AMENDMENT NOS. 174 AND 174 REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO ADOPT TSTF-563, REVISION 0, REVISE INSTRUMENT TESTING DEFINITIONS TO INCORPORATE THE SURVEILLANCE FREQUENCY CONTROL PROGRAM (EPID L- 2019- LLA- 0260)

Dear Mr. Peters:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 174 to Facility Operating License No. NPF-87 and Amendment No. 174 to Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant (Comanche Peak), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the technical specifications in response to your application dated November 7, 2019, as supplemented by .|letter dated January 16, 2020]].

The amendment adopts Technical Specifications Task Force (TSTF) Traveler TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program. The amendments revise the technical specification definitions of Channel Calibration, Channel Operational Test (COT), and Trip Actuating Device Operational Test (TADOT).

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

Sincerely,

/RA/

Dennis J. Galvin, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446

Enclosures:

1. Amendment No. 174 to NPF-87
2. Amendment No. 174 to NPF-89
3. Safety Evaluation cc: Listserv

COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. NPF-87

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

A. The application for amendment by Vistra Operations Company LLC (Vistra OpCo) dated November 7, 2019, as supplemented by .|letter dated January 16, 2020]], 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.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-87 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 174 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by SAMSON SAMSON S. LEE Date: 2020.07.06 S. LEE 15:47:41 -04'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: July 6, 2020

COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-446 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. NPF-89

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

A. The application for amendment by Vistra Operations Company LLC (Vistra OpCo) dated November 7, 2019, as supplemented by .|letter dated January 16, 2020]], 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.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-89 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 174 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by SAMSON SAMSON S. LEE Date: 2020.07.06 S. LEE 15:54:26 -04'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: July 6, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO. NPF-89 COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-445 AND 50-446 Replace the following pages of the Facility Operating License Nos. NPF-87 and NPF-89, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License No. NPF-87 REMOVE INSERT 3 3 Facility Operating License No. NPF-89 REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 1.1-1 1.1-1 1.1-2 1.1-2 1.1-7 1.1-7

(3) Vistra OpCo, 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, and described in the Final Safety Analysis Report, as supplemented and amended; (4) Vistra OpCo, 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) Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and 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) Vistra OpCo, 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 license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Vistra OpCo is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 13 and 3612 megawatts thermal starting with Cycle 14 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. 174 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Unit 1 Amendment No. 174

(3) Vistra OpCo, 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, and described in the Final Safety Analysis Report, as supplemented and amended; (4) Vistra OpCo, 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) Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and 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) Vistra OpCo, 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 license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Vistra OpCo is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 11 and 3612 megawatts thermal starting with Cycle 12 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. 174 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions DELETED Unit 2 Amendment No. 174

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

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

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

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

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

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals between the (AFD) top and bottom halves of an excore neutron detector.

CHANNEL CALIBRATION 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 all devices in the channel required for channel OPERABILITY. 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.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

COMANCHE PEAK - UNITS 1 AND 2 1.1-1 Amendment No. 150, 174

Definitions 1.1 1.1 Definitions (continued)

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal into TEST (COT) the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY so that the setpoints are within the necessary range and accuracy. The COT 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.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity, or from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

COMANCHE PEAK - UNITS 1 AND 2 1.1-2 Amendment No. 150, 174

Definitions 1.1 1.1 Definitions (continued)

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device and OPERATIONAL TEST (TADOT) verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT 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.

COMANCHE PEAK - UNITS 1 AND 2 1.1-7 Amendment No. 150, 174

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO. NPF-89 COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-445 AND 50-446

1.0 INTRODUCTION

By [[letter::CP-201900393, License Amendment Request (LAR)19-001, Application to Adopt TSTF-513-A, Revision 3, PWR Operability Requirements and Actions for RCS Leakage Instrumentation|letter dated November 7, 2019]] (Reference 1), as supplemented by .|letter dated January 16, 2020]] (Reference 2), Vistra Operations Company LLC (the licensee) submitted a license amendment request (LAR) to revise the technical specifications (TSs) for Comanche Peak Nuclear Power Plant (Comanche Peak), Unit Nos. 1 and 2.

The amendments would revise the current instrumentation testing definitions of channel calibration, channel operational test (COT), and trip actuation device operational test (TADOT) to permit determination of the appropriate frequency to perform the surveillance requirement (SR) based on the devices being tested in each step. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated May 10, 2017 (Reference 3). The U.S. Nuclear Regulatory Commission (NRC, the Commission) issued a final safety evaluation approving TSTF-563, Revision 0, on December 4, 2018 (Reference 4).

A Surveillance Frequency Control Program (SFCP) was incorporated into the Comanche Peak, Units 1 and 2 TSs by Amendment Nos. 156 and 156, respectively, dated June 29, 2012 (Reference 5).

In the LAR dated November 7, 2019, the licensee stated that it is not proposing any variations from the TS changes described in TSTF-563, Revision 0, or the applicable parts of the NRC staffs safety evaluation dated December 4, 2018.

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 Description of Surveillance Frequency Control Program and Instrument Testing The TSs require surveillances for instrumentation channels to be performed within the specified frequency, using any series of sequential, overlapping, or total channel steps. Prior amendments approved by the NRC on June 29, 2012, revised the TSs to relocate all periodic surveillance frequencies to licensee control. Changes to the relocated surveillance frequencies are made in accordance with the TS program referred to as the SFCP. The SFCP allows a new surveillance frequency to be determined for the channel, but that frequency must consider all components in the channel and applies to the entire channel.

A typical instrument channel consists of many different components, such as sensors, rack modules, and indicators. These components have different short-term and long-term performance (drift) characteristics, resulting in the potential for different calibration frequency requirements. Under the current TSs, the most limiting component calibration frequency for the channel must be chosen when a revised frequency is considered under the SFCP. As a result, all components that makeup a channel must be calibrated at a frequency equal to the channel component with the shortest (i.e., most frequent) surveillance frequency.

Some channel components, such as pressure transmitters, are very stable with respect to drift and could support a substantially longer calibration frequency than the other components in the channel. Currently, the SRs in many plants are performed in steps (e.g., a pressure sensor or transmitter is calibrated during a refueling outage and the rack signal conditioning modules are calibrated while operating at power). The proposed change extends this concept to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This will allow each component to be tested at the appropriate frequency based on the components long-term performance characteristics.

Allowing an appropriate surveillance frequency for performing a channel calibration on each component or group of components could reduce radiation dose associated with inplace calibration of sensors, reduce wear on equipment, reduce unnecessary burden on plant staff, and reduce opportunities for calibration errors.

2.2 Proposed Changes to the Technical Specifications Currently, the channel calibration, COT, and TADOT may be performed by any series of sequential, overlapping or total channel steps. The proposed changes to the TSs would revise the definitions of channel calibration, COT, and TADOT to indicate that the step must be performed within the most limiting frequency for the components included in that step by adding the phrase , and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step at the end of the last sentence of each definition.

The following paragraph denotes the changes to the channel calibration definition. Changes are shown in bold italics:

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 all devices in the channel required for channel OPERABILITY. 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 following paragraphs denote the changes to the COT and TADOT definitions. Changes are shown in bold italics:

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY so that the setpoints are within the necessary range and accuracy. The COT 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.

A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT 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 various instrumentation functions in the TSs require surveillances to verify the correct functioning of the instrument channel. The proposed change extends the definitions of instrumentation channel components to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This will allow each component to be tested at the appropriate frequency based on the components long-term performance characteristics.

The proposed changes in the definition for instrument testing would allow the licensee to control the frequency of associated components being tested in each step. The SR for the overall instrumentation channel would remain unchanged. The proposed change would have no effect on the design, fabrication, use, or methods of testing the instrumentation channels and will not affect the ability of the instrumentation to perform the functions assumed in the safety analysis.

These instrumentation testing definitions would state that, The [test type] may be performed by means of any series of sequential, overlapping, or total channel steps. The surveillance frequency of these subsets would be established based on the characteristics of the components in the step rather than the most limiting component characteristics in the entire channel. Each of these steps would be evaluated in accordance with the SFCP.

2.3 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires each applicant for a license authorizing operation of a utilization facility to include the proposed TSs in the application.

The regulation at 10 CFR 50.36(b) states:

Each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to

§ 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). One such category is SRs, which are defined in 10 CFR 50.36(c)(3) as 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 at 10 CFR 50.36(c)(5) requires TSs to include administrative controls, which 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.

Prior amendments revised and relocated most periodic surveillance frequencies to licensee control. Changes to the relocated surveillance frequencies are made in accordance with the SFCP. The SFCP requires that changes to the relocated frequencies be made in accordance with NRC staff-approved Topical Report (TR) Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, dated April 2007 (Reference 6).

TR NEI 04-10, Revision 1, describes an evaluation process and a multi-disciplinary plant decisionmaking panel that considers the detailed evaluation of proposed surveillance frequency revisions. The evaluations are based on operating experience, test history, manufacturers recommendations, codes and standards, and other deterministic factors, in conjunction with risk insights. The evaluation considers all components being tested by the SR. Process elements are included for determining the cumulative risk impact of the changes, updating the licensees probabilistic risk assessment (PRA) models, and for imposing corrective actions, if necessary, following implementation of a revised frequency.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical Specifications, dated March 2010 (Reference 7) of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition (SRP). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications (STS) for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the applicable reference STS (i.e., the current STS), as modified by NRC-approved travelers. In addition, the guidance states that comparing the change to previous STS can help clarify the intent of the TSs.

Regulatory Guide (RG) 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated May 2011 (Reference 8), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications, dated May 2011 (Reference 9), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009 (Reference 10),

describes an acceptable approach for determining the technical adequacy of PRAs.

The NRC staffs guidance for evaluating the technical basis for proposed risk-informed changes is provided in SRP, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, dated June 2007 (Reference 11). The NRC staffs guidance on evaluating PRA technical adequacy is provided in SRP, Section 19.1, Revision 3, Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load, dated September 2012 (Reference 12). More specific guidance related to risk-informed TS changes is provided in SRP, Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, dated March 2007 (Reference 13), which includes changes to surveillance test intervals (STIs) (i.e., surveillance frequencies) as part of risk-informed decisionmaking. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.174, Revision 2, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:

The proposed change meets the current regulations, unless it explicitly relates to a requested exemption or rule change.

The proposed change is consistent with the defense-in-depth philosophy.

The proposed change maintains sufficient safety margins.

When proposed changes result in an increase in risk associated with core damage frequency or large early release frequency, the increase(s) should be small and consistent with the intent of the Commissions Safety Goal Policy Statement.

The impact of the proposed change should be monitored using performance measurement strategies.

The STS applicable to the proposed changes are provided in NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, dated April 2012 (References 14 and 15, respectively).

3.0 TECHNICAL EVALUATION

Revising the frequency of a channel calibration, COT, and TADOT instrument channel under the SFCP requires assurance that component performance characteristics, such as drift between each test, will not result in undetected instrument errors that exceed the assumptions of the safety analysis and supporting instrument loop uncertainty calculations. These requirements are consistent with the methodology described in TR NEI 04-10, which is required by the SFCP.

The SFCP does not permit changes to the TS allowable values or nominal trip setpoints; but allows only the surveillance frequency to be changed when determined permissible by TR NEI 04-10. Therefore, prior to extending the test intervals for an instrument channel component or components associated with a given calibration step, the component performance characteristics must be evaluated to verify the allowable value or nominal trip setpoint will still be valid and to establish a firm technical basis supporting the extension. In addition, each change must be reviewed by the licensee to ensure the applicable uncertainty allowances are conservative (bounding) (e.g., sensor drift, rack drift, indicator drift). Documentation to support the changes shall be retained per the guidance in TR NEI 04-10.

Five key safety principles that must be evaluated before changing any surveillance frequency are identified in Section 3.0 of TR NEI 04-10. Principle 3, The proposed change maintains sufficient safety margins, requires confirmation of the maintenance of safety margins, which, in this case, includes performance of deterministic evaluations to verify preservation of instrumentation trip setpoint and indication safety margins.

The evaluation methodology specified in NEI 04-10 also requires consideration of common-cause failure effects and monitoring of the instrument channel component performance following the frequency change, to ensure channel performance is consistent with the analysis to support an extended frequency.

The method of evaluating a proposed surveillance frequency change is not dependent on the number of components in the channel. Each step needs to be evaluated to determine the acceptable surveillance frequency for that step. The proposed change to permit changing the surveillance frequency of channel component(s) does not affect the test method or evaluation method. The requirement to perform a channel calibration, channel functional test, COT, or TADOT on the entire channel is not changed.

For example, an evaluation in accordance with NEI 04-10 may determine that a field sensor (e.g., a transmitter) should be calibrated every 48 months, the rack modules should be calibrated every 30 months, and the indicators should be calibrated every 24 months. Under the current TS requirements, all devices in the channel must be calibrated every 24 months.

However, under the proposed change, sensors, rack modules, and indicators would be calibrated at the appropriate frequency for the tested devices. As required by the channel calibration definition, the test would still encompass all devices in the channel required for channel operability.

TR NEI 04-10 methodology is used to evaluate surveillance frequency changes to determine if such SR extensions could be applied. Process elements are used to determine the cumulative risk impact of changes, update the PRA, and impose corrective actions, if needed, following implementation. Several steps are required by TR NEI 04-10, Section 4.0, Step 7, to be evaluated prior to determining the acceptability of changes. These steps include history of surveillance tests, industry and plant specific history, impact on defense-in-depth, vendor recommendations, required test frequencies for the applicable codes and standards, ensuring

that plant licensing basis would not be invalidated and other factors. The NRC staff finds these measures acceptable in determining the SR extensions.

In addition, TR NEI 04-10, Section 4.0, Step 16 requires an independent decisionmaking panel to review the cumulative impact of all STI changes over a period of time. This is also required by RGs 1.174 and 1.177. The independent decisionmaking panel is composed of the site Maintenance Rule expert panel, surveillance test coordinator, and subject matter expert who is a cognizant system manager or component engineer. Based on the above information, the NRC staff finds that the setpoint changes will be tracked in an acceptable manner.

Licensees with an SFCP may currently revise the surveillance frequency of instrumentation channels. The testing of these channels may be performed by means of any series, sequential, overlapping, or total channel steps. However, all required components in the instrumentation channel must be tested in order for the entire channel to be considered operable.

The NRC staff notes that industry practice is to perform instrument channel surveillances, such as channel calibrations and channel functional tests, using separate procedures based on the location of the components. Each of these procedures may be considered a step. The results of all these procedures are used to satisfy the SR using the existing allowance to perform it by means of any series of sequential, overlapping, or total channel steps. The proposed changes would allow for determining an acceptable surveillance frequency for each step.

The NRC staff notes that the TR NEI 04-10 methodology includes the determination of whether the structure, system, and components (SSCs) affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy, consistent with the guidance contained in RG 1.200, and by sensitivity studies identified in TR NEI 04-10. The licensee is not proposing to change the methodology, or the acceptance criteria for extending STIs, and licensees will need to evaluate changes in the frequency for performing each of the steps in the instrumentation surveillance test per the methodology in TR NEI 04-10.

Therefore, the NRC staff concludes that the proposed change to determine an acceptable test frequency for individual steps within instrumentation channel surveillance tests is acceptable because any extended STIs will be developed within the established constraints of the SFCP and TR NEI 04-10.

The regulatory requirements in 10 CFR 50.36 are not specific regarding the frequency of performing surveillance tests. The proposed change only affects the frequency of performance and does not affect the surveillance testing method or acceptance criteria. Therefore, the proposed change is consistent with the surveillance testing requirements of 10 CFR 50.36.

PRA Acceptability The guidance in RG 1.200 states that the quality of a licensees PRA should be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the greater the change in risk or the greater the uncertainty in that risk as a

result of the requested TS change, or both, the more rigor that should go into ensuring the quality of the PRA.

The NRC staff will have performed an assessment of the PRA models used to support the approved SFCP that uses TR NEI 04-10, using the guidance of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category II of the NRC-endorsed PRA standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies where appropriate, in accordance with TR NEI 04-10.

The SFCP permits revising of the surveillance frequency for instrumentation channels. The NRC staff evaluated whether TR NEI 04-10 can be applied to subsets in an instrument channel when the SFCP currently specifies a surveillance interval that is applied to the entire channel.

The NRC staff notes that the current channel surveillance may be performed by means of any series of sequential, overlapping, or total channel steps. In practice, this means that a channel is divided into subsets and each subset is tested separately. Therefore, the current instrument channel testing is already composed of a sequence of individual tests.

The instrument function may be modeled in the PRA differently depending on the site and the function (e.g., channel may be modeled individually, subsets may be modeled, or the channel function may be modeled as a single entity). There are different steps through the evaluation methodology in TR NEI 04-10 that could be used based on the different PRA modeling approaches. The appropriate modeling of these different approaches is included in the NRC staffs review of the PRA modeling during the review of the application to implement an SFCP that uses TR NEI 04-10.

The licensee is using a PRA that was used to support its application that implemented an SFCP at Comanche Peak that uses TR NEI 04-10. The proposed amendments would change the capability of the licensee to change the surveillance frequency of an entire channel to now change the frequency of each subset of the channel. The NRC staff finds that changes to the surveillance frequency caused by defining and using individual, testable component subsets can be appropriately evaluated with the current SFCP and PRAs. The NRC staff finds that the risk-informed methodology review and the PRA acceptability review that was performed during the review of the licensees application to implement an SFCP at Comanche Peak that uses TR NEI 04-10 are adequate.

The NRC staff determined that the proposed changes to the TS meet the standards for TS in 10 CFR 50.36(b). The regulations at 10 CFR 50.36 require that TS include items in specified categories, including SRs. The proposed changes modify the definitions applicable to instrumentation channel components but do not alter the technical approach that was approved by the NRC in TR NEI 04-10, and the TS, as revised, continue to specify the appropriate SRs for tests and inspections to ensure the necessary quality of affected SSCs is maintained.

Additionally, the changes to the TS were reviewed and found to be technically clear and consistent with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff reviewed the proposed changes against the regulations and concludes that the changes continue to meet the requirements of Sections 50.36(b), 50.36(c)(3), and 50.36(c)(5),

of 10 CFR, for the reasons discussed above, and thus provide reasonable assurance that

adoption of these TS will have the requisite requirements and controls to operate safely.

Therefore, the NRC staff concludes that the proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendments on June 12, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change SRs.

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, as published in the Federal Register on February 25, 2020 (85 FR 10735), 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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. McCool, T. P, Vistra Operations Company LLC, letter to U.S. Nuclear Regulatory Commission, Comanche Peak Nuclear Power Plant Docket Nos. 50-445 and 50-446 License Amendment Request (LAR)19-002 Application to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated November 7, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19325C595).
2. McCool, T. P, Vistra Operations Company LLC, letter to U.S. Nuclear Regulatory Commission, Comanche Peak Nuclear Power Plant Docket Nos. 50-445 and 50-446 Editorial Correction to License Amendment Request (LAR)19-002 Application to Adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated January 16, 2020 (ADAMS Accession No. ML20028D385).
3. Technical Specifications Task Force, letter to U.S. Nuclear Regulatory Commission, Transmittal of TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated May 10, 2017 (ADAMS Accession No. ML17130A819).
4. Cusumano, V. G., U.S. Nuclear Regulatory Commission, letter to Technical Specifications Task Force, Final Safety Evaluations of Technical Specifications Task Force Traveler TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program (EPID L-2017-PMP-0006),

dated December 4, 2018 (ADAMS Package Accession No. ML18333A152).

5. Singal, B. K., U.S. Nuclear Regulatory Commission, letter to Rafael Flores, Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Units 1 and 2 -

Issuance of Amendments Re: Relocation of Surveillance Frequencies from the Technical Specifications to a Licensee-Controlled Document (TAC Nos. ME6789 and ME6790), dated June 29, 2012 (ADAMS Accession No. ML12067A244).

6. Nuclear Energy Institute, (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, dated April 2007 (ADAMS Accession No. ML071360456).
7. U.S. Nuclear Regulatory Commission, Chapter 16.0, Revision 3 Technical Specifications, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2010 (ADAMS Accession No. ML100351425)
8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated May 2011 (ADAMS Accession No. ML100910006).
9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 1, dated May 2011 (ADAMS Accession No. ML100910008).
10. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009 (ADAMS Accession No. ML090410014).
11. U.S. Nuclear Regulatory Commission, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated June 2007 (ADAMS Accession No. ML071700658).
12. U.S. Nuclear Regulatory Commission, Section 19.1, Revision 3, Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated September 2012 (ADAMS Accession No. ML12193A107).
13. U.S. Nuclear Regulatory Commission, Section 16.1, Revision 1, Risk Informed Decision Making: Technical Specifications, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2007 (ADAMS Accession No. ML070380228).
14. U.S. Nuclear Regulatory Commission, NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, dated April 2012 (ADAMS Accession No. ML12100A222).
15. U.S. Nuclear Regulatory Commission, NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, Volume 2, Bases, dated April 2012 (ADAMS Accession No. ML12100A228).

Principal Contributor: K. Bucholtz, NRR Date: July 6, 2020

ML20167A318 *Via e-mail OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DSS/STSB/BC*

NAME DGalvin PBlechman VCusumano DATE 06/18/2020 06/18/2020 12/04/2019 OFFICE OGC* NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*

NAME AGhosh Naber JDixon-Herrity (SLee for) DGalvin DATE 07/02/2020 07/06/2020 07/06/2020

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