ML21103A039
| ML21103A039 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/23/2021 |
| From: | Dennis Galvin Plant Licensing Branch IV |
| To: | Peters K Vistra Operations Company |
| Galvin D | |
| References | |
| EPID L-2020-LLA-0147 | |
| Download: ML21103A039 (16) | |
Text
April 23, 2021 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. 179 AND 179 REGARDING THE ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-569, REVISION 2, REVISE RESPONSE TIME TESTING DEFINITION (EPID L-2020-LLA-0147)
Dear Mr. Peters:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 179 to Facility Operating License No. NPF-87 and Amendment No. 179 to Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak), respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated July 2, 2020, as supplemented by letter dated August 17, 2020.
The amendments revise the Comanche Peak TSs to adopt Technical Specifications Task Force (TSTF) traveler, TSTF-569, Revision 2, Revise Response Time Testing Definition.
K. Peters A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly 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. 179 to NPF-87
- 2. Amendment No. 179 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. 179 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 July 2, 2020, as supplemented by letter dated August 17, 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. 179 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 Jennifer L. Jennifer L. Dixon-Herrity Date: 2021.04.23 Dixon-Herrity 14:13:27 -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: April 23, 2021
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. 179 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 July 2, 2020, as supplemented by letter dated August 17, 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. 179 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 Jennifer L. Jennifer L. Dixon-Herrity Date: 2021.04.23 Dixon-Herrity 14:13:55 -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: April 23, 2021
ATTACHMENT TO LICENSE AMENDMENT NO. 179 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 179 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 Facility Operating License Nos. NPF-87 and NPF-89, and the 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 1.1-3 1.1-3 1.1-6 1.1-6
(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. 179 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. 179
(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. 179 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. 179
Definitions 1.1 1.1 Definitions (continued)
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conv~rsion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil", or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.
ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from when FEATURE(ESF)RESPONSE the monitored parameter exceeds its ESF actuation setpoint at TIME 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 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.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee program PROGRAM that fulfills the requirements of 10 CFR 50.55a(f).
COMANCHE PEAK - UNITS 1 AND 2 1.1-3 Amendment No. 150, 168, 179
Definitions 1.1 1.1 Definitions (continued)
REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from when (RTS) RESPONSE TIME 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 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.
SHUTDOWN MARGIN (SOM) SOM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping or total steps.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
COMANCHE PEAK - UNITS 1 AND 2 1.1-6 Amendment No. 4§.Q, 179
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 179 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 179 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 application dated July 2, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20184A064), as supplemented by letter dated August 17, 2020 (ADAMS Accession No. ML20230A345), Vistra Operations Company LLC (the licensee) submitted a license amendment request (LAR) for the Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak). The amendments would revise technical specification (TS) definitions for engineered safety feature (ESF) response time and reactor trip system (RTS) response time that are referenced in surveillance requirements (SRs), hereafter, referred to as response time testing (RTT).
The proposed changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-569, Revision 2, Revise Response Time Testing Definition, dated June 25, 2019 (ADAMS Accession No. ML19176A034). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving TSTF-569, Revision 2, on August 14, 2019 (ADAMS Accession No. ML19176A191). The description of the generic changes and their justification are contained in these two documents.
2.0 REGULATORY EVALUATION
2.1 Description of Response Time Testing The RTS for Comanche Peak initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and the reactor coolant system (RCS) pressure boundary during anticipated operational occurrences and to assist the Enclosure 3
engineered safety feature actuation system (ESFAS) in mitigating accidents. The ESFAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary and to mitigate accidents.
The RTT verifies that the individual channel or train actuation response times are less than or equal to the maximum values assumed in the accident analyses. The RTT acceptance criteria are under licensee control. Individual component response times are not modeled in the accident analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (e.g., control and shutdown rods fully inserted in the reactor core).
2.2 Proposed Changes to the Technical Specifications Comanche Peak Limiting Condition for Operation (LCO) 3.3.2 requires the ESFAS instrumentation for each Function in TS Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, to be OPERABLE. To assure the LCO is met, SR 3.3.2.10 requires the licensee to verify that ESF RESPONSE TIMES are within limits. Similarly, Comanche Peak LCO 3.3.1 requires the RTS instrumentation for each Function in TS Table 3.3.1-1, Reactor Trip System Instrumentation, to be OPERABLE, and SR 3.3.1.16 requires the licensee to verify that RTS RESPONSE TIMES are within limits. Section 1.1 of the Comanche Peak TSs define ESF RESPONSE TIME and RTS RESPONSE TIME. The definitions state acceptable means to measure each response time, and provide an alternative that may be used [i]n lieu of measurement.
In its application, the licensee stated that it requests adoption of NRC-approved TSTF-569. The only revision of TSTF-569 that is NRC approved is Revision 2. As described in Section 1, Summary Description, of Revision 2 of TSTF-569:
The proposed change revises the definitions to eliminate the requirement for prior NRC review and approval of the response time verification of similar components, while retaining the requirement for the verification to be performed using the methodology contained in Attachment 1, titled, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing. The proposed change will permit licensees to verify the response time of similar component types using the methodology contained in Attachment 1, without obtaining prior NRC approval for each component.
Accordingly, as shown in the LAR, the request would add an additional in lieu of measurement alternative to measuring ESF RESPONSE TIME and RTS RESPONSE TIME. The additional alternative for ESF RESPONSE TIME would be [i]n lieu of measurement, response time may be verified for selected components provided ... the components have been evaluated in accordance with an NRC approved methodology. Similarly, for RTS RESPONSE TIME, [i]n lieu of measurement, response time may be verified for selected components provided that ...
the components have been evaluated in accordance with an NRC approved methodology.
The application stated that the licensee concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to Comanche Peak and provide the justification for the amendment request. The application identified the requested variations (e.g., identifying the site-specific TS numbers).
2.3 Applicable Regulatory Requirements and Guidance Under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Application for amendment of license, construction permit, or early site permit, whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commission's regulations.
The licensee's request involves adding an option used to satisfy SRs. As described in 10 CFR 50.36(c):
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.
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the request by comparing the licensees proposal against the changes described in TSTF-569, Revision 2. The NRC staff compared Comanche Peaks design and existing TSs with the design and TSs presumed in TSTF-569. As explained below, the NRC staff concluded that the design and licenses (including TSs) were sufficient to justify the licensee's reliance on the staff's SE of TSTF-569 as justification for adopting TSTF-569 in the Comanche Peak licenses.
TSTF-569 is designed to make changes to NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, April 2012, Volume 1, Specifications (ADAMS Accession No. ML12100A222), and Volume 2, Bases (ADAMS Accession No. ML12100A228).
The TSs for Comanche Peak are based on the older Revision 1 of the standard TSs (see ADAMS Package Accession No. ML021820213 (issuing amendments for the conversion to Improved TSs with certain beyond-scope items). The NRC staff compared the TSs assumed in TSTF-569 with the current TSs for Comanche Peak. The NRC staff did not identify any material differences in the relevant TSs.
The licensee is relying on the previous analyses of TSTF-569. For the reasons stated in the NRC staff's SE for TSTF-569, the staff found that the methodology contained in TSTF-569, Revision 2, Attachment 1, provides a consistent, clear, and concise framework for determining that replacement components will operate at a level equivalent to that of the components being replaced. As such, using that methodology will assure that the necessary quality of the components is maintained and that the LCOs will be met. Accordingly, approving the incorporation of that methodology into the licensing basis, and amending the TSs to allow usage of the approved methodology, coupled with approving the aspect of the LAR to use the methodology in TSTF-569, Revision 2, Attachment 1, results in TSs that meet 10 CFR 50.36(c)(3) by assuring that performing SRs 3.3.1.16 and 3.3.2.10 while using the new
[i]n lieu of option, will assure that associated aspects of LCOs 3.3.1 and 3.3.2 will be met.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendments on April 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 September 22, 2020 (85 FR 59563), 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.
Principal Contributor: T. Sweat, NRR Date: April 23, 2021
ML21103A039 *via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DSS/STSB/BC*
NAME DGalvin PBlechman VCusumano DATE 4/23/2021 4/22/2021 03/05/2021 OFFICE NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*
NAME JDixon-Herrity DGalvin DATE 4/23/2021 4/23/2021