ML21015A212

Issuance of Amendment Nos. 178 and 178 Regarding One-Time Revision to Technical Specifications 3.7.8 Station Service Water System (Ssws) and 3.8.1 AC Sources - Operating

ML21015A212
Person / Time
Site:	Comanche Peak
Issue date:	02/12/2021
From:	Dennis Galvin Plant Licensing Branch IV
To:	Peters K Vistra Operations Company
Galvin D
References
EPID L-2020-LLA-0250
Download: ML21015A212 (30)

Text

February 12, 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. 178 AND 178 REGARDING ONE-TIME REVISION TO TECHNICAL SPECIFICATIONS 3.7.8, STATION SERVICE WATER SYSTEM (SSWS), AND 3.8.1, AC SOURCES - OPERATING (EPID L-2020-LLA-0250)

Dear Mr. Peters:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 178 to Facility Operating License No. NPF-87 and Amendment No. 178 to Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant (Comanche Peak), Unit Nos. 1 and 2, respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 19, 2020, as supplemented by System and TS 3.8.1, AC Sources - Operating|letter dated December 16, 2020]].

The amendments revise Comanche Peak TS 3.7.8, Station Service Water System (SSWS),

and TS 3.8.1, AC [Alternating Current] Sources - Operating, to extend the completion times for one station service water train inoperable and for one diesel generator inoperable from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 days on a one-time basis to allow for the replacement of Comanche Peak, Unit No. 2 Station Service Water Pump 2-02 (Train B) during Unit No. 2 Cycle 19. The revised TSs incorporate by reference a regulatory commitment that identifies compensatory measures to be implemented during the extended completion times.

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. 178 to NPF-87

2. Amendment No. 178 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. 178 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 19, 2020, as supplemented by System and TS 3.8.1, AC Sources - Operating|letter dated December 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. 178 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 of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jennifer L. Jennifer L. Dixon-Herrity Date: 2021.02.12 Dixon-Herrity 08:52:22 -05'00' 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: February 12, 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. 178 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 19, 2020, as supplemented by System and TS 3.8.1, AC Sources - Operating|letter dated December 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. 178 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 of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jennifer L. Jennifer L. Dixon-Herrity Date: 2021.02.12 Dixon-Herrity 08:52:56 -05'00' 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: February 12, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 178 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 178 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 3.7-21 3.7-21

--- 3.7-21a 3.8-4 3.8-4

--- 3.8-4a

(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. 178 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. 178

(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. 178 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. 178

SSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One SSWS train --------------------NOTES-------------------

inoperable. 1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources -- Operating," for emergency diesel generator made inoperable by SSWS.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -- MODE 4," for residual heat removal loops made inoperable by SSWS.

B.1 ---------------------NOTE--------------------

Required Action B.1 is not applicable to Unit 2 during replacement of the SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19.

Restore SSWS train to OPERABLE 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> status.

OR COMANCHE PEAK - UNITS 1 AND 2 3.7-21 Amendment No. 150, 156, 178

SSWS 3.7.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 ---------------------NOTE--------------------

Required Action B.2 is applicable on a one-time basis to replace SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19. If Unit 2, Train A SSWS becomes inoperable, immediately enter LCO 3.0.3. Regulatory Commitment 5966825 (Attachment 1 to TXX-20086) will be implemented during the 8 day COMPLETION TIME.

Restore SSWS train to OPERABLE 8 days status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> COMANCHE PEAK - UNITS 1 AND 2 3.7-21a Amendment No. 178

AC Sources -- Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) AND B.4.1 -------------------NOTE-------------------

Required Action B.4.1 is not applicable to Unit 2 during replacement of the SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19.

Restore DG to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR B.4.2 -------------------NOTE-------------------

Required Action B.4.2 is applicable on a one-time basis to replace SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19. If Unit 2, Train A SSWS becomes inoperable, immediately enter LCO 3.0.3.

Regulatory Commitment 5966825 (Attachment 1 to TXX-20086) will be implemented during the 8 day COMPLETION TIME.

Restore DG to OPERABLE status. 8 days COMANCHE PEAK - UNITS 1 AND 2 3.8-4 Amendment No. 150, 178

AC Sources -- Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Two required offsite circuits C.1 --------------------NOTE--------------------

inoperable. In MODES 1, 2 and 3, the TDAFW pump is considered a required redundant feature.

Declare required feature(s) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from discovery inoperable when its redundant of Condition C required feature(s) is inoperable. concurrent with inoperability of redundant required features AND C.2 Restore one required offsite circuit to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

COMANCHE PEAK - UNITS 1 AND 2 3.8-4a Amendment No. 178

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 178 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 178 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 November 19, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20324A627), as supplemented by System and TS 3.8.1, AC Sources - Operating|letter dated December 16, 2020]] (ADAMS Accession No. ML20351A237), Vistra Operations Company LLC (the licensee) submitted a license amendment request (LAR) for Comanche Peak Nuclear Power Plant (Comanche Peak), Unit Nos. 1 and 2.

The amendments would revise Comanche Peak Technical Specification (TS) 3.7.8, Station Service Water System (SSWS), and TS 3.8.1, AC [Alternating Current] Sources - Operating, to extend the completion times for one station service water (SSW) train inoperable and for one diesel generator (DG) inoperable from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 days on a one-time basis to allow for the replacement of Comanche Peak, Unit No. 2 SSW Pump 2-02 (Train B) during Unit No. 2 Cycle 19. The revised TSs would incorporate by reference a regulatory commitment that identifies compensatory measures to be implemented during the extended completion times, as delineated in Attachment 1 to the LAR.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.91(a)(6), the licensee requested that the proposed amendments be issued under exigent circumstances. As the U.S. Nuclear Regulatory Commission (NRC, the Commission) staff was able to publish a Federal Register notice allowing for the normal 30-day period for public comment on the proposed no significant hazards consideration determination, the NRC staff determined that exigent circumstances do not apply to this LAR.

Enclosure 3

The supplemental System and TS 3.8.1, AC Sources - Operating|letter dated December 16, 2020]], provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 29, 2020 (85 FR 85678).

2.0 REGULATORY EVALUATION

2.1 Description of Systems 2.1.1 Description of Station Service Water System The SSWS removes heat from the component cooling water system (CCWS) heat exchangers and from the emergency DGs. The SSWS also supplies cooling water to the safety injection and centrifugal charging pump lube oil coolers and the containment spray pump bearing oil coolers. In conjunction with the CCWS, the SSWS supplies cooling water to meet the plant cooling requirements during normal operation, shutdown, and during and after a postulated accident of either unit. The required cooling water is taken from the safe shutdown impoundment, which is the ultimate heat sink for Comanche Peak. The SSWS also acts as a backup water supply for the auxiliary feedwater (AFW) system if the condensate storage tank is depleted.

The SSWS consists of two separate and independent full-capacity, safety-related trains. The safety-related trains are redundant in that the components supplied by one train are sufficient to perform the minimum required safety functions. Two full capacity SSWS pumps and two full capacity supply and return headers are provided for each unit. Both pumps normally operate.

This provides a continuous cooling water supply to the two redundant safety-related trains.

Cross connections between the two trains within a unit add operational flexibility to the SSWS.

During a loss of all SSWS on one unit, a SSWS cross-connect between the two units is available to provide additional backup cooling capacity. In addition, fire protection water may be aligned to the jacket water heat exchangers for emergency DGs, and to the lube oil coolers for safety injection pumps and centrifugal charging pumps. During a loss of offsite power (LOOP),

the SSWS pumps are powered by the DGs.

2.1.2 Description of Electrical System The Comanche Peak Updated Final Safety Analysis Report (UFSAR) (ADAMS Package Accession No. ML20315A055) Sections 8.2.2, Analysis, and 8.3.1.1.1, System Structure (Network), state that the offsite power system consists of the 138 and 345 kilovolt (kV) switchyards with each switchyard supplying each unit with an offsite power source that is physically independent and redundant from the other switchyard. The preferred offsite AC source for Comanche Peak, Unit 1, is from the 345 kV switchyard, via startup transformer XST2, and the preferred offsite AC source for Comanche Peak, Unit 2, is from the 138 kV switchyard, via startup transformer XST1. Each startup transformer can safely shutdown both units simultaneously while the stations design basis assumes an accident in just one unit.

Comanche Peak UFSAR Section 8.3.1.1.2, Busing Arrangement, Interconnections, and Load Assignment, states that each of the two (per unit) 6.9 kV, Class 1E buses is supplied by its dedicated DG, assigned to Train A or B, and able to fulfill minimum shutdown safety requirements. Upon loss of voltage on a 6.9 kV, Class 1E bus due to a LOOP with no safety injection signal present, under-voltage relays automatically start a DG and close its output

breaker. Based on UFSAR Tables 8.3-1 and 8.3-2, each DG has sufficient capacity to supply its assigned Class 1E loads and is located in a separate room with walls designed to protect it against a safe shutdown earthquake, tornadoes, missiles, and fire.

2.2 Proposed Technical Specification Changes The licensees proposed changes to TS 3.7.8, Condition B would add one new Required Action B.2 with an associated Completion Time (CT), and two new Notes. The proposed changes are illustrated in bold below:1 1

RCS in the Table below refers to reactor coolant system.

CONDITION REQUIRED ACTION COMPLETION TIME B. One SSWS train B.1 ------------------NOTES-----------------

inoperable. 1. Enter applicable Conditions and Required Actions of LCO 3.8.1, AC Sources -- Operating, for emergency diesel generator made inoperable by SSWS.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, RCS Loops -- MODE 4, for residual heat removal loops made inoperable by SSWS.

B.1 ------------------NOTE-----------------

Required Action B.1 is not applicable to Unit 2 during replacement of the SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19.

Restore SSWS train to OPERABLE status 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR B.2 ------------------NOTE-----------------

Required Action B.2 is applicable on a one-time basis to replace SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19. If Unit 2, Train A SSWS becomes inoperable, immediately enter LCO 3.0.3. Regulatory Commitment 5966825 (Attachment 1 to TXX-20086) will be implemented during the 8 day COMPLETION TIME.

Restore SSWS train to OPERABLE 8 days status.

The licensees proposed changes to TS 3.8.1, Condition B would renumber Required Action B.4 to B.4.1, add one new Required Action B.4.2 and associated CT, and add two new Notes. The proposed changes are illustrated in bold below:

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) AND B.4.1 ------------------NOTE-----------------

Required Action B.4.1 is not applicable to Unit 2 during replacement of the SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19.

B,4 Restore DG to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR B.4.2 ------------------NOTE-----------------

Required Action B.4.2 is applicable on a one-time basis to replace SSWS Pump 2-02 (Train B) during Unit 2 Cycle 19. If Unit 2, Train A SSWS becomes inoperable, immediately enter LCO 3.0.3.

Regulatory Commitment 5966825 (Attachment 1 to TXX-20086) will be implemented during the 8 day COMPLETION TIME.

Restore DG to OPERABLE 8 days status.

The proposed changes also include compensatory measures through a reference to Regulatory Commitment 5966825, which are described and evaluated in Section 3.4 of this safety evaluation (SE). Inclusion of the regulatory commitment in TSs elevates the status of the referenced commitment to a TS requirement.

2.3 Reason for the Proposed Changes The licensee addressed its reason for the proposed changes in Section 2.3, Reason for Proposed Change, of the enclosure to the LAR. The licensee has been tracking a degradation of performance for Comanche Peak Unit 2 SSW Pump 2-02. While short-term actions have been taken within the time permitted in the action statements of TS 3.7.8 and TS 3.8.1, more extensive corrective actions such as replacement of a pump will take longer than currently allowed. The TS amendments would enable the licensee to replace the SSW Pump 2-02 at power and thus avoid an unnecessary plant transient or shutdown.

The replacement of the Unit 2 SSW Pump 2-02 at power will support the licensees replacement of the Unit 2 SSW Pump 2-01 during the Unit 2 fall 2021 refueling outage. Specifically, the current SSW Pump 2-02 would be refurbished by a vendor and then would be used as the replacement for Unit 2 SSW Pump 2-01. The licensee stated that if these license amendments are approved to extend the two TS CTs on a one-time basis to 8 days, it will replace the Unit 2

SSW Pump 2-02 in February 2021. Without the approval of these license amendments, the licensee may be forced to perform an additional shutdown of Comanche Peak Unit 2.

2.4 Regulatory Requirements and Guidance The regulations in 10 CFR 50.36, Technical specifications, establish the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. The changes proposed in the LAR relate to the LCO category.

The regulations in 10 CFR 50.36(c)(2)(i) state, in part, that:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criteria (GDC) 17, Electric power systems, state, in part, that:

An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. . . .

The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

The regulations in 10 CFR 50.63, Loss of all alternating current power, paragraph (a),

Requirements, state, in part, that Each light-water-cooled nuclear power plant . . . must be able to withstand for a specified duration and recover from a station blackout . . . .

The regulations in 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (the maintenance rule), paragraph (a)(4), state that:

Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety.

RG 1.155, Station Blackout, dated August 1988 (ADAMS Accession No. ML003740034),

provides guidance for complying with 10 CFR 50.63, which requires nuclear power plants to be capable of coping with a station blackout (SBO) event for a specified duration.

Generic Letter (GL) 80-30, Clarification of the Term Operable as it Applies to Single Failure Criterion for Safety Systems Required by TS, dated April 10, 1980 (Legacy Library Accession No. 8401200196), discusses that a plant may temporarily depart from the single-failure design criterion when the plant is operating within a TS action requirement. However, the plant must remain capable to mitigate any postulated accident and safely shutdown.

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) 8-8, Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time Extensions, dated February 2012 (ADAMS Accession No. ML113640138), provides guidance to the NRC staff in reviewing LARs for licensees proposing a one-time or permanent TS change to extend an emergency DG allowed outage time beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees proposed TS changes to determine whether they are consistent with the regulatory requirements and guidance discussed in Section 2.4 of this SE and the licensing and design basis information.

3.1 Proposed One-Time Change to Station Service Water System CT In the LAR, the licensee proposed a one-time extension to the CT for the Comanche Peak TS Required Action to restore the Unit 2 SSWS Train B to operable status from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 days in order to replace the SSW Pump 2-02 in Unit 2 Train B. Since Comanche Peak Unit 2 has two 100-percent capacity SSWS trains, as previously described, its safety function would remain unchanged and Unit 2 would remain capable of mitigating any postulated accident and safely shutting down with SSWS Train B unavailable during the extended CT.

The licensees proposed changes would not allow plant operation in a configuration outside the design basis. The licensees proposed changes would incorporate by reference in the TS change the regulatory commitment in Attachment 1 to the LAR to implement the compensatory measures listed in Section 3.5.3, Regulatory Commitment 5966825, of the enclosure to the LAR. These compensatory measures are intended to reduce the risk of any potentially risk-significant configurations during the proposed CT and are further evaluated in Section 3.4 of this SE.

The NRC staff reviewed the licensees use of the maintenance rule to manage risk for multiple systems inoperable. In Section 3.7, Risk Informed Configuration Management, of the enclosure to the LAR, the licensee stated that:

Risk will also be managed during the extended COMPLETION TIME via the Maintenance Rule 10 CFR 50.65(a)(4) Configuration Risk Management Program (CRMP), which has been reviewed in prior risk informed Technical Specification change requests. The process governing CRMP is described in Technical Specifications, Section 5.5.18 . . . .

In addition to the management of risk by the maintenance rule, the TSs play an important part in the risk management of maintenance activities by limiting the duration of safety equipment outages. Because a failure to meet a systems TS LCO usually involves a loss of redundancy, the affected system cannot withstand a single-failure and still perform its intended safety function. TSs provide a time by which the design-basis must be restored or the plant shut down

(see GL 80-30). This is acceptable because, by limiting the duration of plant operation with inoperable safety equipment, the TSs manage the associated increase in risk to an acceptable level. For plant operation with only one or two TS-required components inoperable, probabilistic risk assessment has indicated that the TS strategy for managing risk is usually conservative.

For operation with multiple TS-required components inoperable, TSs may actually allow operation that is non-conservative compared to the established risk guidelines. One reason for this is that the NRC never intended TSs to fully address inoperable components in several systems at the same time. In addition to observing the limitations of TSs, licensees must also assess and manage risk associated with maintenance activities in accordance with paragraph (a)(4) of the maintenance rule. More information can be found in the NRC SE, Application of Generic Letter 80-30 Guidance to an Inoperable Non-Technical Specification Support Subsystem, dated April 5, 2002 (ADAMS Accession No. ML020950074). Therefore, based on the requirement to asses and manage risk associated with maintenance activities in accordance with paragraph (a)(4) of the maintenance rule and the licensees description of its application of the maintenance rule, the NRC staff finds that the licensee can manage the risk of multiple inoperable systems. The management of risk beyond the use of the maintenance rule, including the use of risk insights and compensatory measures, is evaluated in Section 3.4 of this SE.

Based on the above, the NRC staff concludes that with the proposed extension of the CT, the SSWS and all supported safety systems will remain capable of performing their safety functions to safely shut down the plant and mitigate the effects of a design-basis accident.

3.2 Proposed One-Time Change to Diesel Generator CT In the LAR, the licensee proposed the changes to TS 3.8.1 as identified in Section 2.2 of this SE. The changes allow the licensee to replace the Unit 2 SSW Pump 2-02 while the unit is operating at power with both offsite AC sources available to supply at least one of the two redundant Unit 2 Class 1E buses (2EA1-Train A or 2EA2-Train B) and DG 2EG1 (Train A) in standby mode. The grid and transmission network supplying the two units are stable with the 138 and 345 kV switchyards having consistent availability and high reliability. Therefore, the loss of one offsite AC source during the planned maintenance is not probable with the loss of both being highly unlikely. The loss of the preferred offsite AC source to a 6.9 kV Class 1E bus would cause the alternate offsite AC source to connect automatically to that bus. If both offsite AC sources fail, then the operational DG 2EG1 would supply the Class 1E loads of its bus.

The NRC staff determined that multiple redundant AC sources are available to Unit 2 during the proposed plant maintenance despite the inoperability of Unit 2 DG 2EG2. DG 2EG2s inoperability does not defeat Unit 2s safe shutdown capability since either offsite AC source to either bus 2EA1 or 2EA2 or DG 2EG1 to bus 2EA1 or the uncredited alternate power diesel generators (APDGs) can perform the designated safety functions. Per the LAR, the licensee will verify the availability of both offsite AC sources and the Train A DG 2EG1 at the start of (and as expected during) the Unit 2 maintenance and will also take other compensatory measures. A total LOOP for Unit 2 is not likely during the expanded outage window of 8 days. In addition, in accordance with GL 80-30, the single failure criterion for a system is temporarily relaxed when in an LCO statement for that system, so a second failure in the system (i.e., the loss of DG 2EG1) is not required to be postulated. Therefore, since the loss of both offsite AC circuits to both Unit 2 Class 1E buses is unlikely and the concurrent loss of DG 2EG1 being highly improbable, Unit 2 is still capable of performing its required minimum safety functions.

In addition, in the LAR, Section 3.1.5, Conformance with NUREG 0800 BTP 8-8 Recommendations, the licensee postulates two scenarios for concurrent failures of both offsite AC circuits and the existing shutdown train in Unit 2 that demonstrates additional shutdown capability for Unit 2 using Unit 1 as a SSWS supply or uncredited APDGs as an electrical source. The first considers a Unit 2 LOOP with a concurrent failure of the Unit 2 Train A SSWS pump, which causes DG 2EG1 to be non-functional until other cooling can be established. For this scenario, the licensee could utilize the SSWS cross-tie between the two units within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to establish a flow path to DG 2EG1 to safely shutdown Unit 2 using the full complement of Train A equipment. The second scenario considers a Unit 2 LOOP and a concurrent non-recoverable loss of DG 2EG1. In this scenario, the Unit 2 APDGs could be aligned to bus 2EA1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to provide for an orderly safe shutdown of Unit 2. Even after additional failures are postulated in Unit 2, as demonstrated in those two scenarios, Unit 2 has enough safe shutdown margin to carry out its required safety functions.

Regarding the SBO requirements in 10 CFR 50.63 and the guidelines in RG 1.155, Comanche Peak was evaluated and determined to meet that regulation and to be consistent with the regulatory guidance. As indicated in BTP 8-8, if Unit 2 experienced SBO conditions during the scheduled maintenance, a backup AC source must be available. For Comanche Peak, that source is the APDGs. As discussed in UFSAR Section 8B, Station Blackout, the reactor coolant system with associated support systems can ensure proper core cooling and containment integrity during a 4-hour postulated SBO event. If no DG is available in the non-SBO Unit, no credit is taken for common cooling systems for the SBO unit, which can cope without them. The NRC staff agrees because, although the APDGs are not credited AC sources, they are available and sufficient to allow the SBO unit to be safely shut down.

As stated in BTP 8-8, there are seven conditions that a licensee should meet during an extended CT other than an allowed outage time of less than 14 days. The licensee adheres to this guidance since the allowed outage time is 8 days including any unforeseen contingencies, and the licensee has included applicable compensatory measures proposed as a regulatory commitment incorporated by reference in the one-time TS change. Those seven conditions are presented below, along with how the licensee addresses them in the LAR.

The first condition is addressed in Sections 3.1.2, Station Service Water System Pump 2-02 Replacement, and 3.1.5, Conformance with NUREG-0800 BTP 8-8 Recommendations, of the enclosure to the LAR. The licensee does not have a substitute DG for Unit 2 DG 2EG2, which will be declared inoperable for the entire planned maintenance. However, the existing APDGs assigned to Unit 2 provide safe shutdown capability if other preferred AC sources are unavailable. The NRC staff notes that the APDGs do not depend on the SSWS for their function. Thus, the Unit 2 minimum shutdown safety requirements are not compromised due to DG 2EG2s unavailability. The NRC staff notes that this provision is consistent with the BTP 8-8 position that a supplemental power source be provided for the duration of the CT extension requested by the licensee.

The licensee addressed the remaining six conditions of BTP 8-8 in Sections 3.1.5 and 3.5.3 of the enclosure to the LAR.

The second condition is addressed in that this is only a one-time deviation from TS 3.8.1 and will not be repeated during the next 24 months.

The third condition is met by not scheduling the subject work to begin if severe weather is predicted and by the control room personnel checking weather conditions during the maintenance with briefings to operators to reinforce LOOP planned actions.

The fourth condition is met by the licensee contacting the transmission grid controller prior to entry into the LCO to verify that the grid is stable with no foreseen challenges being found and both offsite AC sources available.

The fifth and sixth conditions are addressed by the operations shift manager controlling access to, and work on, the 138 and 345 kV switchyards with the suspension of discretionary testing and maintenance for the necessary components for the entire duration of the extended CT and assurance that the grid is stable.

The seventh condition is addressed by suspending all discretionary testing and maintenance for the turbine-driven AFW Pump 2-01 for the entire duration of the extended CT.

The NRC staff reviewed the licensees proposal to incorporate by reference in the TS change the regulatory commitment in Attachment 1 of the LAR. The regulatory commitment is to implement the compensatory measures listed in Section 3.5.3 of the enclosure to the LAR.

These compensatory measures are intended to reduce the risk of any potentially risk-significant configurations during the proposed CT and are further evaluated in Section 3.4 of this SE.

Based on the above, the NRC staff concludes that during the proposed extension of the TS 3.8.1 DG CT, Unit 2 will remain capable of safely shutting down and mitigating the effects of a design-basis accident. This is further demonstrated by its capability to address additional failures of either the Unit 2 Train A SSWS pump or DG 2EG1, as postulated by the licensee in the LAR, during the planned maintenance. In addition, the licensee will undertake compensatory measures to ensure the viability of the redundant Class 1E equipment and electrical AC sources in Unit 2.

3.3 Defense-in-Depth and Safety Margin Although one train of the SSWS and its associated DG will be out of service longer than the current TSs allow, the NRC staff finds that the capability to fulfill the function of that system will be retained since one train will remain to fulfill the function. The licensee stated that the potential for common-cause failures would not increase because there is no change in failure mechanisms associated with the SSWS or DG CT changes. The licensee also stated that there is no change to the likelihood for an initiating event, successful mitigative action, or required operator actions, and that the plant design will not be modified with the proposed extension of the CTs.

For Unit 2 Train B components affected by the Unit 2 Train B SSWS being inoperable, the licensee identified the redundant Train A components: the Emergency Core Cooling System Train A, Containment Spray System Train A, AFW System Train A, CCWS Train A, and specified Train A electrical power sources as protected in accordance with station procedures.

The turbine-driven AFW is also identified as a protected system. The NRC staff notes that the turbine-driven AFW does not depend on the SSWS for its function. In Sections 3.1.4, Defense in Depth, and 3.1.5 of the enclosure to the LAR , the licensee stated that the SSWS cross-connect between Unit 1 and Unit 2 provides additional backup cooling capability, available to respond in the event of a total loss of SSW. Comanche Peak is also equipped with APDGs assigned to each unit, which would supply adequate power for safe shutdown in the event that

the Train A DG is lost concurrent with a LOOP. Therefore, the cross-tie and APDGs provide mitigation capability for two different scenarios. The decrease in redundancy of the SSWS is addressed by existing mitigation and safe shutdown capability as well as the compensatory measures discussed in Section 3.4 of this SE. Additional aspects of the electrical power system defense-in-depth is discussed in Section 3.2 of this SE. After a review of the information in the LAR, combined with the conclusions in Section 3.2 and 3.4 of this SE, the NRC staff finds that defense-in-depth is preserved commensurate with the expected frequency and consequences of challenges from the proposed changes.

The safety analysis acceptance criteria stated in the Comanche Peak UFSAR are not impacted by these changes. The proposed changes will not allow plant operation in a configuration outside the design basis. The requirements regarding the SSWS and DG credited in the accident analysis will remain the same. The design and operation of the SSWS and DG are not modified by this LAR. No codes or standards approved for use by the NRC relevant to the SSWS, DG, and associated systems are modified or affected. Based on its review, the NRC staff concludes that safety margins will continue to be maintained during the proposed CT extension.

3.4 Risk Insights In the LAR, as supplemented, the licensee described the risk insights that form the basis for its proposed compensatory measures and support the deterministic evaluation of the proposed one-time change to the SSWS and DG TS CT extensions. In Section 3.4, Supplemental Risk Information, of the enclosure to the LAR, the licensee stated that the risk information and insights were considered in the overall decisionmaking process and were useful in the development of effective risk management strategies. In Section 3.4.1, Probabilistic Risk Assessment Capability and Insights, of the enclosure to the LAR, the licensee described that the scope of the risk assessments used to determine risk insights are quantitative models for internal events, internal flooding, and internal fire, and qualitative assessments for external hazards. The risk insights were used to determine the compensatory measures listed in Section 3.5.3 of the enclosure to the LAR, which are included in the proposed TS change by means of the regulatory commitment in Attachment 1 to the LAR. As mentioned by the licensee in Section 3.4 of the enclosure to the LAR, this is not a risk-informed LAR (i.e., not formally submitted using the guidance in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (ADAMS Accession No. ML17317A256)) and, therefore, the NRC staff did not review the licensees probabilistic risk assessment models to determine their technical acceptability.

Therefore, the NRC staff did not rely on the quantitative risk information provided by the licensee in Attachments 5 and 6 to the LAR. However, the NRC staff considered the licensee-provided qualitative risk insights and associated compensatory measures in its decision on the proposed changes.

The NRC staff determined that special circumstances, as discussed in NUREG-0800, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, dated June 2007 (ADAMS Accession No. ML071700658), did not exist for the proposed changes. As such additional risk information was not required to be provided.

In Section 3.4.3, Avoidance of Risk Significant Plant Configurations, of the enclosure to the LAR, the licensee stated that the dominant risk scenarios associated with the loss of the SSWS include the loss of the remaining SSW pump with the loss of turbine-driven AFW, transient

events leading to a plant trip with failure of the turbine-driven AFW pump, internal flood events leading to induced seal loss-of-coolant accident with loss of mitigating functions, and fires in the unaffected SSW related areas, safety-related switchgear rooms, DG rooms, cable spreading rooms, and the main control room leading to abandonment and/or induced LOOP. The main impact of the above scenarios on critical safety functions is the loss of heat removal from the steam generators due to failure of all the AFW pumps (random or induced) or loss of the remaining SSW pump and loss of turbine-driven AFW.

In order to manage the dominant risk scenarios, the licensee proposed to incorporate by reference in the TS change the regulatory commitment in Attachment 1 to the LAR to implement the compensatory measures listed in Section 3.5.3 of the enclosure to the LAR for the duration of the proposed one-time CT extension, in addition to the site procedures for those scenarios.

The proposed compensatory measures include control of access as well as suspension of all planned maintenance in both switchyards, posting of protected equipment, ensuring grid stability prior to entering the required action with the proposed extended CT, and not scheduling entry into the required action with the proposed extended CT if severe weather conditions are anticipated. In Sections 3.4.1, and 3.3.2, FLEX Equipment, of the enclosure to the LAR, the licensee stated that the probabilistic risk assessment models do not include quantitative credit for any flexible coping strategies (FLEX) or portable equipment, which further defined the scope of the probabilistic risk assessment model used by the licensee, and that FLEX was not used in developing risk insights. However, in Section 3.3.2, the licensee explained that diverse and FLEX equipment remains available for use in accordance with the licensees FLEX procedures.

The licensee proposed the suspension of all discretionary testing and maintenance activities on protected equipment, which includes equipment supported by Train B of the SSWS. The protected equipment includes the APDGs, 1E switchgear, motor-driven AFW pumps, residual heat removal pumps, centrifugal charging pumps, safety chillers, safety chilled water recirculation pumps, and turbine-driven AFW pump for Unit 2; the component cooling water pumps, DGs, and startup transformers for both units; and the SSW pumps for both units except for the one being replaced.

Further, the licensee evaluated the impact of the proposed changes on fire protection and proposed compensatory measures to address those impacts in Sections 3.4.3 and 3.4.5, Fire Protection Program Considerations, of the enclosure to the LAR. The fire safe shutdown analysis assumes that there is a LOOP in conjunction with a fire and any equipment requiring power can be fed from onsite power sources. In Section 3.4.5, the licensee proposed restrictions on combustible storage, suspension of hot work, and an hourly roving fire watch for the areas that credit the unavailable SSW pump in the fire safe shutdown analysis. These areas are listed in the Section 3.4.3 of the enclosure to the LAR. The restrictions on the storage of combustibles during the extended CT includes both Units 1 and 2 transient combustible safe zones identified in the fire assessment as well as in the main control room, the cable spreading rooms, and the cable routing paths for the inservice startup transformers. The hourly roving fire watch protects areas credited by the fire assessment, specifically, the main control room and cable spreading rooms, and areas containing power and control cabling of the inservice startup transformers.

In addition to reviewing the licensee-provided risk insights, the NRC staff also reviewed the NRCs Standardized Plant Analysis Risk (SPAR) model for Comanche Peak. The NRC staffs review of the SPAR model was used to assess the proposed changes, including identifying the dominant risk contributors for the proposed changes, and the adequacy of the compensatory measures proposed by the licensee to manage the risk from the proposed changes. This

review increases the NRC staffs confidence in the appropriateness of the licensee-provided risk insights. The NRC staffs review of the SPAR model did not identify the need for any additional compensatory measures. The NRC staffs review also noted that the risk due to external hazards such as high winds and tornadoes, seismic events, and external flooding does not significantly change the risk from the proposed changes because the licensees compensatory measures include consideration of severe weather; the licensee reviewed Individual Plant Evaluation of External Events insights and conclusions; the licensee has addressed LOOP, which is the dominant impact of the external hazards; and any additional impacts from an external hazard are independent of the CT.

The NRC staffs review of the licensees risk insights and proposed compensatory measures supplemented by its review of the Comanche Peak SPAR model finds that (1) the licensee appropriately identified the dominant risk scenarios for the proposed changes, and (2) the licensees compensatory measures, proposed as a regulatory commitment incorporated by reference in the one-time TS change (which elevates the status of the referenced commitment to a TS requirement), appropriately manage the risk from the dominant risk scenarios.

Therefore, the NRC staff concludes that the licensees risk insights support the proposed changes and that the compensatory measures are acceptable for the proposed changes.

3.5 Technical Evaluation Conclusion

Based on its review, the NRC staff concludes that the Comanche Peak Unit 2 SSWS and all supported safety systems, including the electrical power system, will remain capable of performing their safety functions to safely shut down the plant and mitigate the effects of a design-basis accident. The NRC staff further concludes that the compensatory measures, proposed as a regulatory commitment incorporated into the TS by reference for the one-time TS change, acceptably manage the risk from the proposed changes. Accordingly, the NRC staff concludes that there is reasonable assurance that the proposed TS changes will have minimal impact on the licensees ability to continue to comply with the requirements of 10 CFR 50.36; 10 CFR Part 50, Appendix A, GDCs 17 and 18; 10 CFR 50.63; and 10 CFR 50.65(a)(4) and, therefore, they are acceptable.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRC staff proposed to find that the requested amendments involve no significant hazards consideration in its Federal Register notice on December 29, 2020 (85 FR 85678). The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91(a), in its application dated November 19, 2020, the licensee provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes add provisions to increase the COMPLETION TIME (CT) from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 days, on a one-time basis for Comanche Peak Nuclear Power Plant Station Service Water System Train B (replacement of Station Service Water System Pump 2-02). This one-time increase will be used once at power during Unit 2 Cycle 19. An additional REQUIRED ACTION, new Note, and associated COMPLETION TIME is specified for Technical Specification (TS) 3.7.8, Station Service Water System, and TS 3.8.1, AC Sources - Operating, when Station Service Water System Train B is declared inoperable to replace the Station Service Water Pump 2-02. The proposed changes do not physically alter any plant structures, systems, or components, and are not an accident initiator: therefore, there is no effect on the probability of accidents previously evaluated. As part of the single failure design feature, loss of one Station Service Water System train or one Emergency Diesel Generator does not prevent the minimum safety function from being performed. Also, the proposed changes do not affect the type or amounts of radionuclides released following an accident, or affect the initiation and duration of their release. Therefore, the consequences of accidents previously evaluated, which rely on Station Service Water System or Emergency Diesel Generator to mitigate, are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not involve a change in design, configuration, or method of operation of the plant. The proposed changes will not alter the manner in which equipment is operated, nor will the functional demands on credited equipment be changed. The proposed changes do not impact the interaction of any systems whose failure or malfunction can initiate an accident. There are no identified redundant components affected by these changes and thus there are no new common cause failures or any existing common cause failures that are affected by extending the CT. The proposed changes do not create any new failure modes.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes are based upon a deterministic evaluation. This evaluation is supplemented by risk insights.

The deterministic evaluation concluded with one inoperable Station Service Water System train and one inoperable Emergency Diesel Generator, the redundant OPERABLE Station Service Water System train and Emergency Diesel Generator will be able to perform the safety function as described in the accident analysis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Supplemental risk information supporting this license amendment request concluded that the additional REQUIRED ACTIONs, new NOTEs, and associated COMPLETION TIMEs have a small impact on overall plant risk and is consistent with the NRC Safety Goal Policy statement and the thresholds in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, and RG 1 .177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications.

The deterministic evaluation and the supplemental risk information provide assurance that the Station Service Water System and AC Sources Operating will be able to perform their design function with a longer COMPLETION TIME for inoperable Station Service Water System Train B (replacement of Station Service Water System Pump 2-02) and inoperable Emergency Diesel Generator Train B during Unit 2 Cycle 19, and risk is not significantly impacted by the change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff reviewed the licensees no significant hazards consideration analysis. Based on this review and on the NRC staffs evaluation of the underlying LAR as discussed above, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendments on January 15, 2021. The State official had no comments.

6.0 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. 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, published in Federal Register on December 29, 2020 (85 FR 85678), and there has been no public comment on such finding. Additionally, the Commission has made a final determination that no significant hazards consideration is involved for the proposed amendments as discussed above in Section 4.0 of this SE. 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.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: N. Chien, NRR E. Kleeh, NRR A. Neuhausen, NRR S. Vasavada, NRR Date: February 12, 2021

ML21015A212 *by e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DSS/SCPB/BC*

NAME DGalvin PBlechman BWittick DATE 1/19/21 1/19/21 1/11/21 OFFICE NRR/DRA/APLC/BC* NRR/DSS/SNSB/BC* NRR/DSS/STSB/BC*

NAME SRosenberg SKrepel VCusumano DATE 1/11/21 1/11/21 1/14/21 OFFICE NRR/DEX/EEEB/BC* OGC - NLO* NRR/DORL/LPL4/BC*

NAME BTitus Wachutka JDixon-Herrity DATE 1/13/21 2/3/21 2/12/21 OFFICE NRR/DORL/LPL4/PM*

NAME DGalvin DATE 2/12/21