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Issuance of Amendment No. 167 Reactor Trip System Instrumentation Technical Specification Changes to Implement WCAP-14333 and WCAP-15376
	ML20293A157
Person / Time
Site:	Seabrook
Issue date:	12/28/2020
From:	Justin Poole; Plant Licensing Branch 1
To:	Moul D; Florida Power & Light Co, NextEra Energy Seabrook
	Poole J, NRR/DORL/LPLI, 415-2048
References
	EPID L-2019-LLA-0237
	Download: ML20293A157 (39)
	v • d • e

December 28, 2020 Mr. Don Moul Executive Vice President Nuclear Division and Chief Nuclear Officer Florida Power & Light Company NextEra Energy Seabrook, LLC Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478

SUBJECT:

SEABROOK STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 167 RE: REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TECHNICAL SPECIFICATION CHANGES TO IMPLEMENT WCAP-14333 AND WCAP-15376 (EPID L-2019-LLA-0237)

Dear Mr. Moul:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 167 to Renewed Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1. This amendment consists of changes to the Technical Specifications in response to your application dated November 1, 2019, as supplemented by letter dated July 13, 2020.

The amendment revises the Technical Specification requirement for the reactor trip system instrumentation and engineered safety features actuation system instrumentation to implement the allowed outage times and bypass test times justified in WCAP-14333-P-A, Probabilistic Risk Analysis of the RPS [Reactor Protection System] and ESFAS [Engineered Safety Features Actuation System] Test Times and Completion Times, and WCAP-15376-P-A, Risk-Informed Assessment of the RTS [Reactor Trip System] and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times. The amendment incorporates changes contained in Technical Specifications Task Force (TSTF) Traveler, TSTF-411, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376),

and TSTF-418, RPS and ESFAS Test Times and Completion Times (WCAP-14333).

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

Sincerely,

/RA/

Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443

Enclosures:

1. Amendment No. 167 to NPF-86

2. Safety Evaluation cc: Listserv

NEXTERA ENERGY SEABROOK, LLC, ET AL.*

DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 167 License No. NPF-86

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

A. The application for amendment filed by NextEra Energy Seabrook, LLC, et al.

(the licensee), dated November 1, 2019, as supplemented by letter dated July 13, 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 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.

NextEra Energy Seabrook, LLC, is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Lighting Plant (collectively, with NextEra Energy Seabrook, LLC, licensees) and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

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 Renewed Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 167, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2020.12.28 Danna 15:53:07 -05'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 28, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 167 SEABROOK STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Renewed Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of 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.

Remove Insert 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23

(3) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) NextEra Energy Seabrook, LLC, 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) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) NextEra Energy Seabrook, LLC, 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 authorized herein.

(7) DELETED C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power).

(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 167, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3*, 4*, 5* 10

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2

b. Low Setpoint 4 2 3 1# #, 2 2

3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate

4. (NOT USED)

5. Intermediate Range, Neutron Flux 2 1 2 1# #, 2 3

6. Source Range, Neutron Flux

a. Startup 2 1 2 2# 4

b. Shutdown 2 0 1 3, 4, 5 5

c. Shutdown 2 1 2 3*, 4*, 5* 10

7. Overtemperature T 4 2 3 1, 2 6A

8. Overpower T 4 2 3 1, 2 6A

9. Pressurizer Pressure--Low 4 2 3 1** 6A

10. Pressurizer Pressure--High 4 2 3 1, 2 6A

11. Pressurizer Water Level--High 3 2 2 1** 6A SEABROOK - UNIT 1 3/4 3-2 Amendment No. 36, 91, 114, 167

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

12. Reactor Coolant FlowLow

a. Single Loop (Above P-8) 3/loop 2/loop in 2/loop in 1 6A any oper- each oper-ating loop ating loop

b. Two Loops (Above P-7 and 3/loop 2/loop in 2/loop 1 6A below P-8) two oper- each oper-ating loops ating loop

13. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 6A Level--Low--Low in any oper- each oper-ating stm. ating stm.

gen. gen.

14. Undervoltage--Reactor Coolant 4-2/bus 2-1/bus 2 on one bus 1** 6A Pumps

15. Underfrequency--Reactor Coolant 4-2/bus 2-1/bus 2 on one bus 1** 6A Pumps

16. Turbine Trip

a. Low Fluid Oil Pressure 3 2 2 1*** 6B

b. Turbine Stop Valve Closure 4 4 4 1*** 11

17. Safety Injection Input from ESF 2 1 2 1, 2 7

18. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6 2 1 2 2# 8 SEABROOK - UNIT 1 3/4 3-3 Amendment No. 36, 114, 167

TABLE 3.3-1 (Continued)

TABLE NOTATIONS

When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.

- Trip function automatically blocked or bypassed below the P-7 (At Power)

Setpoint.

- - Trip function automatically blocked below the P-9 (Reactor Trip/Turbine Trip Interlock) Setpoint.

Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

# Below Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable Channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

SEABROOK - UNIT 1 3/4 3-5 Amendment No. 36, 114, 167

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify that valve RMW-V31 is closed and secured in position within the next hour.

ACTION 6A - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.1.1.

ACTION 6B - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

SEABROOK - UNIT 1 3/4 3-6 Amendment No. 36, 167

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 7 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in a least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

SEABROOK - UNIT 1 3/4 3-7 Amendment No. 167

TABLE 3.3-3 (Continued)

TABLE NOTATIONS

Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

ACTION STATEMENTS ACTION 13 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 14 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirements is met; however, the inoperable channel may be bypased for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels per Specification 4.3.2.1.

ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 16 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

ACTION 17 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

SEABROOK - UNIT 1 3/4 3-22 Amendment No. 36, 114, 167

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 19 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 20 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 21 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 22 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5.

SEABROOK - UNIT 1 3/4 3-23 Amendment No. 36, 167

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 167 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-86 NEXTERA ENERGY SEABROOK, LLC SEABROOK STATION, UNIT NO. 1 DOCKET NO. 50-443

1.0 INTRODUCTION

By letter dated November 1, 2019 (Reference 1), as supplemented by letter dated July 13, 2020 (Reference 2), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted License Amendment Request (LAR) No. 17-06 requesting changes to the Technical Specifications (TSs) for Seabrook Station, Unit No. 1 (Seabrook).

The amendment would revise the Technical Specification (TS) requirement for the reactor trip system instrumentation and engineered safety features actuation system instrumentation to implement the allowed outage times and bypass test times justified in WCAP-14333-P-A, Probabilistic Risk Analysis of the RPS [Reactor Protection System] and ESFAS [Engineered Safety Features Actuation System] Test Times and Completion Times (Reference 3), and WCAP-15376-P-A, Risk-Informed Assessment of the RTS [Reactor Trip System] and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times (Reference 4). The amendment incorporates changes contained in Technical Specifications Task Force (TSTF) Traveler, TSTF-411, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376) (Reference 5), and TSTF-418, RPS and ESFAS Test Times and Completion Times (WCAP-14333) (Reference 6).

The supplement dated July 13, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 28, 2020 (85 FR 5053).

2.0 REGULATORY EVALUATION

2.1 Background The Pressurized Water Reactor Owners Group (PWROG), formerly the Westinghouse Owners Group (WOG), Technical Specifications Optimization Program (TOP) evaluated changes to surveillance test intervals and completion times (CTs), also called allowed outage times, for the Enclosure 2

analog channels, logic cabinets, master and slave relays, and reactor trip breakers (RTBs). The methodology evaluated proposed increases in surveillance intervals, test and maintenance out-of-service times, and the bypassing of portions of the RPS during test and maintenance.

In 1983, the PWROG submitted Westinghouse Topical Report (TR) WCAP-10271-P, Evaluation of Surveillance Frequencies and Out-of-Service Times for the Reactor Protection Instrumentation System (Reference 7), which provided a methodology for justifying revisions to a plants TSs for the RPS. The PWROG stated in WCAP-10271-P that plant staff devoted significant time and effort to perform, review, document, and track surveillance activities that, in many instances, may not be necessary because of the high reliability of the equipment. Part of the justification for the changes was an anticipated small impact on plant risk.

By letter dated February 21, 1985, the NRC staff accepted WCAP-10271-P-A (Reference 8),

including its Supplement 1, with conditions. In 1989, the NRC staff issued a safety evaluation report (SER) for WCAP-10271-P-A, Supplement 2 (Reference 9), which approved similar relaxations for the ESFAS. An additional supplemental SER issued in 1990 (Reference 10) provided consistency between RTS and ESFAS surveillance test intervals and CTs.

The NRC subsequently adopted the TS changes proposed in WCAP-10271 into NUREG-1431, Standard Technical Specifications Westinghouse Plants, Revision 0, issued September 1992 (Reference 11). After the approval of WCAP-10271, and its supplements, the PWROG submitted Westinghouse TR WCAP-14333-P in May 1995. The purpose of this TR was to provide justification for additional TS relaxations beyond those approved in WCAP-10271, including:

Increase the bypass test times and CTs for both the RTS and ESFAS solid-state and relay protection system designs for the analog channels; increase the CT from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass test time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for the logic cabinets, master relays, and slave relays.

When the logic cabinet and RTB both cause their train to be inoperable when in test or maintenance, allow bypassing of the RTB for the period of time equivalent to the bypass test time for the logic cabinets, provided that both are tested at the same time and the plant design is such that both the RTB and the logic cabinet cause their associated electrical trains to be inoperable during test or maintenance.

The NRC staff approved TR WCAP-14333-P by letter dated April 29, 1998 (ADAMS Accession No. ML20013H744). Following the approval of WCAP-14333, the PWROG submitted TR WCAP-15376 to the NRC on November 8, 2000, which the NRC staff subsequently approved by letter dated December 20, 2002. TR WCAP-15376 provided an evaluation of the surveillance test intervals for the Westinghouse RTS and ESFAS analog channels, logic cabinets, master relays, and RTBs. The NRC staffs safety evaluation (SE) supporting the approval of TSTF-411, Revision 1, which references this TR, included the following approvals:

Increase the RTS and ESFAS instrumentation surveillance test interval from 2 or 3 months (WCAP-10271) to 6 months.

Increase the surveillance test interval (from 2 to 4 months), CT (from 1 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ), and bypass test times (from 2 to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) for the RTBs.

2.2 Applicable Regulations Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, paragraph (a)(1), states, Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. Specifically, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether a limiting condition for operation is required to be included in the TSs.

Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC),

includes the following requirements:

GDC 13, Instrumentation and control, requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate, to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

GDC 21, Protection system reliability and testability, requires that the protection system be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed, with redundancy and independence sufficient to assure that loss of the protection function does not result from any single failure and preservation of the required minimum redundancy despite removal from service of any component or channel unless acceptable reliability of operation of the protection system can be otherwise demonstrated.

GDC 22, Protection system independence, requires that the protection system be designed to assure that the effects of natural phenomena, normal operating, maintenance, testing, and postulated accident conditions do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.

Section 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, of 10 CFR requires monitoring the performance or condition of structures, systems, or components against licensee-established goals in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions.

2.3 Regulatory Guides and Staff Review Plans The NRC staff considered the following guidance to facilitate its review of the proposed changes:

NUREG-1431, Revision 4.0, Volume 1, Standard Technical Specifications, Westinghouse Plants (Reference 12)

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 13)

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

Technical Specifications (Reference 14)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 15)

Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, of NUREG 0800 (Reference 16)

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 (Reference 17)

Section 19.2, Revision 0, 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 (Reference 18)

Industry Guidance WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.

WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times.

2.4 Proposed TS Changes The proposed changes revise the Seabrook TS Table 3.3-1, Reactor Trip System Instrumentation, Actions 2, 6, 7, and 9, and TS Table 3.3-3, Engineered Safety Features Actuation System Instrumentation, Actions 13, 15, 18, 20, and 22. In general, the changes include increasing the CTs and bypass test times.

TS 3/4.3.1, Reactor Trip System Instrumentation Specifically, the proposed changes would revise the following functions in TS Table 3.3-1, consistent with the generic evaluations approved in WCAP-14333 or WCAP-15376:

Function System Action Proposed Technical Specification Change 2.a Power Range, Neutron Flux 2 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72

--High Setpoint hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2.b Power Range, Neutron Flux 2 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72

--Low Setpoint hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3 Power Range, Neutron Flux 2 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 High Positive Rate hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 7 Overtemperature T 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

8 Overpower T 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 9 Pressurizer Pressure--Low 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 10 Pressurizer Pressure--High 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 11 Pressurizer Water Level-- 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 High hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 12.a Reactor Coolant Flow--Low, 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Single Loop (Above P-8) hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 12.b Reactor Coolant Flow--Low, 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Two Loops (Above P-7 and hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months below P-8) 13 Steam Generator Water 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Level--Low-Low hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 14 Undervoltage--Reactor 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Coolant Pumps hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 15 Underfrequency--Reactor 6A Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Coolant Pumps hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 16.a Turbine Trip, Low Fluid Oil 6B Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 Pressure hours and bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 17 Safety Injection Input from 7 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 ESF [Engineered Safety hours Features]

19 Reactor Trip Breakers 9 Increase completion time to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and bypass time from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

20 Automatic Trip and Interlock 7 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 Logic hours The following TS Table 3.3-1 actions are revised:

ACTION 2 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable Channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per

Specification 4.3.1.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 6A With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.1.1.

ACTION 6B With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 9 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

TS 3/4.3.2, Engineered Safety Features Actuation System Instrumentation Specifically, the proposed changes would revise the following functions in TS Table 3.3-3, consistent with the generic evaluations approved in WCAP-14333 or WCAP-15376:

Function System Action Proposed Technical Specification Change 1.b Safety Injection - Automatic 13 Increase completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Actuation Logic and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Actuation Relays 2.b Containment Spray - 13 Increase completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Automatic Actuation Logic 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and Actuation Relays 3.a.2 Containment Isolation Phase 13 Increase completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to A - Automatic Actuation 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Logic and Actuation Relays 3.b.2 Containment Isolation Phase 13 Increase completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to B - Automatic Actuation 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Logic and Actuation Relays 8.a Auto Switchover to 13 Increase completion time from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Containment Sump - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Automatic Actuation Logic and Actuation Relays 2.c Containment Spray - 15 Increase one channel bypass time to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Containment Pressure - Hi-3 and additional channel bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.b.3 Phase B Isolation - 15 Increase one channel bypass time to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Containment Pressure - Hi-3 and additional channel bypass time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

8.b Auto Switchover to 15 Increase one channel bypass time to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Containment Sump - RWST and additional channel bypass time from

[Refueling Water Storage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Tank] LevelLow-Low 4.b Steam Line Isolation - 20 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Automatic Actuation Logic 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

and Actuation Relays 7.b Emergency Feedwater - 20 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Automatic Actuation Logic 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

and Actuation Relays 1.c Safety Injection - 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Containment PressureHi-1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1.d Safety Injection - Pressurizer 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Pressure-Low 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1.e Safety Injection - Steam Line 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Pressure-Low 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.c Steam Line Isolation - 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Containment PressureHi-2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.d Steam Line Isolation - 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Steam Line Pressure-Low 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.e Steam Line Isolation - 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Steam Generator (SG) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from Pressure-Negative 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5.b Turbine Trip - SG Water 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to LevelHigh-High (P-14) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

6.a Feedwater Isolation - SG 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Water LevelHigh-High 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from (P-14) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

7.c Emergency Feedwater - SG 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Water LevelLow-Low Start 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from Motor-Driven Pump and Start 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Turbine-Driven Pump 10.c ESFAS Interlocks - SG 18 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Water Level, P-14 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the bypass time increased from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5.a Turbine Trip - Automatic 22 Increase completion time from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Actuation Logic and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Actuation Relays The following TS Table 3.3-3 actions are revised:

ACTION 13 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provide the inoperable channel is placed in the bypassed condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 18 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 20 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 22 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

3.0 TECHNICAL EVALUATION

3.1 Method of Staff Evaluation An acceptable approach for making risk-informed decisions about proposed TS changes, including both permanent and temporary changes, is to show that the proposed changes meet the five key principles stated in RG 1.174, Revision 3, Section C, and RG 1.177, Revision 1, Section B. These key principles are:

Principle 1: The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12).

Principle 2: The proposed LB change is consistent with the defense-in-depth philosophy.

Principle 3: The proposed licensing basis change maintains sufficient safety margins.

Principle 4: When proposed LB changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5: The impact of the proposed LB change should be monitored by using performance measurement strategies.

3.2 Traditional Engineering Evaluation In accordance with Standard Review Plan Sections 16.1, 19.1, and 19.2 of NUREG 0800, the NRC staff reviewed the Seabrook incorporation of WCAP-15376-P-A and WCAP-14333-P-A.

The following sections present the NRC staffs evaluation of the licensees proposed amendment to extend CTs and bypass test times using the five key principles outlined in RG 1.174 and RG 1.177.

The engineering evaluation below addresses the first three key principles of RG 1.174, Revision 3, and IS pertinent to: (1) compliance with current regulations, (2) evaluation of defense in depth, and (3) evaluation of safety margins.

3.2.1 Key Principle 1: The Proposed Change Meets Current Regulations The NRC staff reviewed the licensees proposed TS changes in the LAR and compared them to the description of the proposed changes contained in TSTF-411, Revision 1 (regarding implementation of proposed WCAP-15376) and TSTF-418, Revision 2 (WCAP-14333).

The staff noted that the licensee for Seabrook implemented a Surveillance Frequency Control Program that was approved by the NRC for inclusion in Seabrook License Amendment No. 141, dated July 24, 2014 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML13212A069). As such, the Seabrook RTS and ESFAS TSs do not contain the specific periodic surveillance frequencies for each instrument channel or RTB affected by the changes justified in WCAP-15376-P-A and approved in the staffs SE for that TR. Therefore, the changes proposed in the LAR only include the changes justified within WCAP-15376-P-A that are applicable to the allowed outage time and bypass times for the RTBs.

Evaluation of Changes to RTS The NRC staff reviewed the proposed changes and the associated action statements regarding RTS instrumentation pertaining to Functions 2.a, 2.b, 3, 7 through 11, 12.a, 12.b, 13, 14, 15, 16.a, 17, 19, and 20. The changes were compared and found to be consistent with the intent of NRC-approved TSTF-411 and TSTF-418, and therefore, are found to be acceptable.

Evaluation of Changes to ESFAS The NRC staff also reviewed the proposed changes and the associated action statements regarding ESFAS instrumentation pertaining to Functions 1.b, 1.c/d/e, 2b, 2c, 3.a.2, 3.b.2, 3.b.3, 4.b, 4.c, 4.d, 4.e, 5.a, 5.b, 6.a, 7.b, and 7.c. The changes were compared and found to be consistent with the intent of NRC-approved TSTF-411 and TSTF-418, and therefore, are found to be acceptable.

Evaluations of Deviations from the NRC-Approved Generic TSTF Travelers The following Seabrook ESFAS functions were not included in the generic evaluation (TSTF travelers) and require plant-specific evaluations for acceptance by the NRC staff.

FUNCTION 8.a Automatic Actuation Logic and Actuation Relays 8.b RWST [Refueling Water Storage Tank] Level--Low-Low 10.c Steam Generator Water Level, P-14 Note 1 of Table 2.4-3 submitted in the LAR provides the justification for the extensions requested for Functions 8.a and 8.b. Note 1 states, in part:

The applicability of the changes justified in WCAP-10271-P-A and its supplements to the Seabrook ESP AS Functional Units 8.a, Automatic Actuation Logic and Actuation Relays, and 8.b, RWST Level--Low-Low, was approved by the NRC in License Amendment No. 36 [] issued April 1995. License Amendment 36 included the approval of changes in the Seabrook TS justified in WCAP-10271-PA and its supplements. In Amendment 36, the NRC approved changes from WCAP-10271-P-A and its supplements that included revisions to the Action for Functional Unit 8.a (Action 13) and revisions to both the surveillance test interval (Analog Channel Operational Test) and Action for Functional Unit 8.b (Action 18). The surveillance test interval for Seabrook Functional Unit 8.a, Automatic Actuation Logic and Actuation Relays, was not revised as WCAP-10271-P-A and its supplements did not include changes for the surveillance test interval of Automatic Actuation Logic and Actuation Relays Functional Units. Based on the prior NRC approval of the applicability of WCAP-10271-P-A and its supplements to the Seabrook ESF AS Functional Units 8.a, Automatic Actuation Logic and Actuation Relays, and 8.b, RWST Level--Low-Low, the changes approved in WCAP-14333-P-A and WCAP-15376-P-A are also applicable to these Seabrook Functional Units.

For the reasons stated in Note 1, the NRC staff finds that Functions 8.a and 8.b are reasonable and consistent with the intent of NRC-approved TSTF-411 and TSTF-418, and therefore, are found to be acceptable.

Note 2 of Table 2.4-3 submitted in the LAR provides the justification for the extension requested for Function 10.c. Note 2 states:

The Seabrook ESFAS Functional Unit 10.c, SG Water Level, P-14, is not specifically identified in the TSTF-418 guidance for implementing the changes justified in WCAP-14333-P-A. However, the analog channels that comprise this ESFAS Interlock Function are the same SG Level - High-High analog channels used in the ESFAS Turbine Trip Functional Unit 5.b, SG Water Level -

High-High (P-14) and ESFAS Feedwater Isolation Functional Unit 6.a, SG Water Level-High-High (P-14).

In the NUREG-1431 standard TSs, these two functional units are combined in the turbine trip and feedwater isolation Function 5.b, which is included in TSTF-418. Also, in Seabrook License Amendment No. 36, dated April 10, 1995 (ADAMS Accession No. ML011910374) (discussed

above), the NRC approved changes from WCAP-10271-P-A and its supplements, which included revisions to both the surveillance test interval (analog channel operational test) and action (Action 18) for Functional Unit 10.c, SG Water Level, P-14. Based on the analog channels that comprise Functional Unit 10.c being addressed in TSTF-418 (as the turbine trip and feedwater isolation function), and the prior NRC approval of the applicability of WCAP-10271-P-A and its supplements to this ESFAS functional unit, the changes approved in WCAP-14333-P-A are also applicable to Seabrook Functional Unit 10.c, SG Water Level, P-14.

For the reasons stated in Note 2, the NRC staff finds that Function 10.c is consistent with the intent of NRC-approved TSTF-411 and TSTF-418, and therefore, is acceptable.

Within the scope of its review, the staff agrees that the licensee has demonstrated the proposed applicability of WCAP-14333 and WCAP-15376 to Seabrook and has met the limitations and conditions as outlined in the NRC staffs SERs approving TSTF-411, Revision 1 (regarding implementation of proposed WCAP-15376), and TSTF-418, Revision 2 (WCAP-14333), and therefore, the LAR meets GDC 13, 21, and 22.

3.2.2 Key Principle 2: The Proposed Change is Consistent with Defense-in-Depth Philosophy Defense in depth is an approach to designing and operating nuclear facilities involving multiple independent and redundant layers of defense to compensate for human and system failures.

Regulatory Position C.2.1.1 in RG 1.174, Revision 3, states that defense in depth consists of seven elements, and consistency with the defense-in-depth philosophy is maintained if the following occurs:

1. A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.

2. Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

3. System redundancy, independence, and diversity are maintained commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers).

4. Defenses against potential common cause failures are maintained, and the potential for introduction of new common cause failure mechanisms is assessed.

5. Independence of physical barriers is not degraded.

6. Defense against human errors is maintained.

7. The intent of the plants design criteria is maintained.

Condition and Limitation No. 1 of the SEs for WCAP-15376-P and WCAP-14333-P requested the licensee to confirm applicability of the TRs to the plant for which it is being applied and perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposed changes.

In LAR Section 3.2.3, the licensee assessed the applicability of WCAP-15376 for Seabrook.

The licensee determined that the Seabrook RPS and ESFAS are similar in design to the reference plant in the TR. Additionally, the licensee confirmed that plant-specific evaluations of current and future unavailability (e.g., increased structures, systems, and components unavailability due to the CT extension) were considered.

Furthermore, to determine that WCAP-14333 is applicable to Seabrook, the licensee addressed the implementation guidance developed by the PWROG in LAR Section 3.2.1, Tables 3.2.1-1, 3.2.1-2, and 3.2.1-3. These tables compare plant-specific data to the generic analysis assumptions. The evaluation provided by the licensee compared the analysis assumptions in WCAP-14333-P-A to plant-specific parameters, including surveillance and maintenance intervals, operator actions, transient and anticipated transient without scram frequencies, actuation signals, safety functions, and certain component failure probabilities.

In addition, the licensee considered the probability of failure of the master relays and safeguards driver card developed for WCAP-15376-P to confirm applicability to Seabrook. The licensee assessed the reliability data of the components over a 5-year period and identified zero failures recorded. The licensee performed an analysis to estimate the expected number of failures for the given component failure probabilities and actuations and concluded zero failures would be expected. The NRC staff finds that the licensee appropriately assessed the data to confirm that the failure probabilities used in WCAP-15376-P are applicable to Seabrook.

Based on the above discussion and the NRC staffs review of the licensees evaluation for applicability of WCAP-15376-P-A and WCAP-14333-P-A, the staff concludes that the licensee confirmed that the generic evaluation assumptions used in the WCAPs are applicable to Seabrook. Therefore, the NRC staff finds this first part of Condition and Limitation No. 1 for both WCAPs is satisfied. Refer to Section 3.2.4 of this SE for the staffs review of the licensees evaluation for the Tier 1 analysis.

The second part of Condition and Limitation No. 1 of the SE for WCAP-15376-P-A requires the licensee to perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposed changes. WCAP-15376 was based on a large dry containment and assumed that the only contributions to large early release frequency (LERF) would come from containment bypass events and core damage events with the containment not isolated. The licensee stated in LAR Section 3.3.2 that a plant-specific assessment was performed to confirm the applicability of the WCAP analysis to Seabrook. The licensee provided the results of this assessment in Section 3.3.2 and Table 3.3.2-1 of the LAR, which included the appropriate containment failures and other failure modes that could result in a release. The NRC staff finds the second part of Condition and Limitation No. 1 for WCAP-13576-P-A is satisfied because the licensee performed a plant-specific assessment using a model that included the specific design and performance of the Seabrook containment.

Refer to Section 3.2.4 of this SE for the staffs review of the licensees evaluation for the Tier 1 analysis, which includes staff review of the risk metrics that include contribution to LERF.

3.2.3 Key Principle 3: The Proposed Change Maintains Sufficient Safety Margins As previously discussed, the NRC staff had approved the use of WCAP-14333 and WCAP-15376, provided the limitations and conditions were met. As previously discussed, the NRC staff approved the use of TSTF-411 and TSTF-418 to implement the allowed outage times and bypass test times justified in WCAP-15376 and TSTF-418, respectively. Section 3.2.1 of this SE describes how the NRC staff found that the licensee had adequately demonstrated the

proposed applicability of WCAP-14333 and WCAP-15376 to Seabrook and has met the limitations and conditions as outlined in the NRC staffs SERs approving TSTF-411 and TSTF-418. By meeting the limitations and conditions previously outlined in the NRC staffs SEs, the staff finds that sufficient safety margins exist.

3.2.4 Key Principle 4: Change in Risk Is Consistent with the Safety Goal Policy Statement The NRC staff evaluated Key Principle 4 using the three-tiered approach described in Standard Review Plan Section 16.1 and RG 1.177.

The three-tiered approach ensures that adequate programs and procedures are in place to identify risk-significant plant configurations resulting from maintenance or other operational activities and to take appropriate compensatory measures to avoid such configurations. In order to determine whether the probabilistic risk assessment (PRA) used in support of the proposed CT extension is of sufficient quality, scope, and level of detail, the NRC staff evaluated the relevant information in the LAR submittal and considered the results of the PRA reviews. Consistent with RG 1.177, Revision 1, the staffs review of the licensees submittal focused on the capability of the licensees PRA model to analyze the risks stemming from the proposed CT extensions.

3.2.4.1 PRA Capability and Insights The licensees Tier 1 risk evaluations as delineated in RG 1.177, Revision 1, assess the impact of the proposed CTs and bypass time changes on core damage frequency (CDF), LERF, incremental conditional core damage probability (ICCDP), and incremental conditional large early release probability (ICLERP). Furthermore, the PRA acceptability review performed by the NRC staff assesses whether the PRA model used to evaluate the proposed TS changes is of sufficient scope and detail for this application. WCAP-15376-P-A and WCAP-14333-P-A provided generic PRA models for the evaluation of extensions to surveillance test intervals, RTB CTs, bypass test times, and RPS and ESFAS test and CTs. The NRC staff found these generic models and both WCAP evaluations to be acceptable on a generic basis in the SEs for TRs WCAP-15376-P-A and WCAP-14333-P-A. Although the SEs accepted the use of a representative model as generally reasonable, the application of the representative models and the associated results to a specific plant introduce a degree of uncertainty because of modeling, design, and operational differences. Therefore, the staff determined that each licensee adopting these WCAPs would need to confirm that the TR analyses and results are applicable.

As discussed above for Conditions and Limitations No. 1, the licensee confirmed the applicability of both WCAP-15376-P-A and WCAP-14333-P-A to Seabrook. A detailed NRC staff review of the remaining conditions and limitations is provided throughout the applicable portions of this SE.

The licensee performed an assessment using its PRA, Level 1/LERF model, which includes assessment of internal events and internal floods events. A qualitative assessment was provided for internal fire events and seismic events. Assessment of high winds, external floods, and other external events used NUREG-1407, NUREG/CR-2300, and ASME/ANS RA-Sa-2009 (Reference 19, Reference 20, and Reference 21, respectively) to screen out the applicable hazards. A more detailed review of the licensees qualitative and screening assessments is provided below in this SE.

3.2.4.2 Internal Events PRA (includes internal floods)

In the LAR, the licensee provided a summary of the peer review history for the internal events (includes internal floods) PRA (IEPRA) for Seabrook. The Seabrook IEPRA (includes internal floods) has been subjected to four peer reviews (1999, 2005, 2009, and 2012) against internal events and internal floods technical elements using the industry probabilistic safety analysis certification process or the ASME/ANS PRA standard. In addition, in the amendment request dated October 3, 2019, to support the Seabrook one-time TS change for onsite power distribution requirements (Reference 22), the licensee noted four self-assessments against the internal event supporting requirements from the ASME/ANS RA-Sa-2009 PRA standard, as endorsed by RG 1.200, Revision 2. In the one-time TS change, the licensee confirmed that the self-assessment performed in 2010 evaluated the 2009 PRA against each of the technical elements for internal events using the ASME/ANS RA-Sa-2009 PRA standard as endorsed by the NRC, in addition to reviewing the results from previous peer reviews. The NRC staff finds that the peer review processes and self-assessments that were performed assessed the appropriate technical elements and gaps to RG 1.200, Revision 2, using the endorsed guidance.

Table 3.2.1-1-f of the LAR summarized the significant PRA model changes and CDF impacts for the Seabrook IEPRA (includes internal floods). In request for additional information (RAI) 01, the NRC staff requested the licensee provide justification to ascertain whether each of the PRA model changes constituted a PRA maintenance or upgrade, consistent with the endorsed PRA standard (Reference 23). In its response to the RAI for the PRA changes made in 2011 and 2014, the licensee discussed the nature of the PRA changes that included data updates, human reliability analysis (HRA) success criteria, and plant-specific procedure changes, and identified maintenance changes or upgrades. The NRC staff finds that the licensee appropriately identified these PRA updates and incorporated them into the Seabrook model, along with performing a peer review that addressed the upgrades, consistent with the ASME/ANS RA-Sa-2009 PRA standard, as endorsed by RG 1.200.

In RAI 02, the NRC staff requested the licensee to justify that there was no impact for the requested TS change or that the licensee has resolved fact and observation HR-E3-1 to meet Capability Category II for the associated supporting requirement provided in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2. In response to the RAI, the licensee provided a summary of the comprehensive review performed for the modeled operator actions and included the roles and responsibilities that were performed by the retired shift manager, training instructor, and current operations. The licensee further confirmed that the retired shift supervisor was the individual leading the interviews with operations staff and that active operators were involved and current procedures and images were used. Based upon the above discussion, the NRC staff finds that the licensees disposition for the fact and observation associated with supporting requirement HR-E3 is appropriate for the risk-informed application.

In Table 3.2.1-1-f of the LAR, NextEra provided a summary of significant PRA model changes and CDF impacts that include an update of the reactor coolant pump seal loss-of-coolant accident (LOCA) model that is credited in the IEPRA. In RAI 03, the NRC staff requested the licensee to validate and confirm any credit for an update to the reactor coolant pump seal LOCA model that was used to support this risk-informed application. In response to the RAI, the licensee confirmed that no credit was taken for reactor coolant pump seal modifications for the PRA evaluation that was used to support this risk-informed application.

WCAP-15376 SE Condition and Limitation No. 4 To further assess the licensees process used to identify the key assumptions and key sources of uncertainty for this risk-informed application, in RAI 04, the staff requested the licensee to provide a description of the process. In response to the RAI, the licensee provided a summary of the process and stated that WCAP-16432, Process for Identifying Assumptions Within a PRA; WCAP-16282, WOG Guidelines for PRA Key Assumptions; and Electric Power Research Institute TR-1009652, Guideline for the Treatment of Uncertainty in Risk-informed Applications, Technical Basis Document, were used to develop a comprehensive list of uncertainties and assumptions for the Seabrook IEPRA and identify those that were determined to be key for the risk-informed application. The licensee provided five steps that were used to facilitate the process. The NRC staff finds that the process described by the licensee in response to the RAI is inclusive of guidance provided in NUREG 1855 (Reference 24),

RG 1.200, and the ASME/ANS PRA standard, and therefore, is sufficient to assess key assumptions and sources of uncertainty for these risk-informed applications. Furthermore, in review of Tables 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.3-1 provided in the LAR, the NRC staff finds that the licensee appropriately applied the WCAPs and assessed them for applicable assumptions.

Condition and Limitation No. 4 of the SE for WCAP-15376-P-A requested the licensee to confirm that the plant-specific model assumptions for the HRA are consistent with the assumptions delineated in NRC-approved WCAP-15376. In LAR Section 3.3.2, Table 3.3.2-3, the licensee listed the operator actions credited in the WCAP-15376 analysis and identified the corresponding human failure events used in the Seabrook IEPRA (includes internal floods) for the HRA. The licensee confirmed that the Seabrook plant-specific HRA analysis is either consistent with or more conservative than the assumptions used in WCAP-15376-P-A.

Accordingly, the NRC staff concludes that Condition and Limitation No. 4 is satisfied.

Closure of Peer Review Results In the LAR dated October 3, 2019, to support the one-time TS change, the licensee confirmed that an independent assessment was performed for closure of the IEPRA (includes internal floods) findings and that All findings were reviewed to Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, as accepted by the NRC in the staff memorandum dated May 3, 2017

[Reference 25]. In the SE issued for the approval of the one-time TS change (Reference 26),

the NRC staff performed a review to confirm that the independent assessment was performed consistent with the NRC-accepted Appendix X guidance. In addition, the licensee provided a response to RAI 02 for the staff review of the one-time TS change LAR (Reference 27). In response to the RAI, the licensee provided a table that included all the findings, along with the dispositions and acceptability evaluations performed by the independent assessment team.

The NRC staff reviewed the table provided in the licensees response to RAI 02 for the one-time TS change, in addition to Tables 3.4.1 and 3.4.2 provided in this LAR, and concludes that the open findings did not impact this application. Therefore, the NRC staff finds the IEPRA (includes internal floods) is appropriate for this risk-informed application.

PRA Capability Conclusions RG 1.177 states that The licensee should provide the rationale that supports the acceptability of the proposed changes by integrating probabilistic insights with traditional considerations to arrive at a final determination of risk. In summary, the licensee has evaluated the Seabrook

IEPRA against RG 1.200, Revision 2; evaluated the findings identified from the peer reviews; and addressed the findings impact on this risk-informed application.

The NRC staff concludes that the Seabrook IEPRA (includes internal floods) was subjected to a peer review process using the guidance in RG 1.200, Revision 2; and that findings from those reviews have either been closed in accordance with Appendix X of NEI 05-04, 07-12, 12-13, as accepted with conditions by the NRC staff, or determined to have no adverse impact on this application. Therefore, the NRC staff finds that the Seabrook IEPRA is sufficient to assess the risk impact for this risk-informed application.

3.2.4.3 PRA Results and Insights Satisfaction of the fourth key principle of risk-informed decisionmaking may be demonstrated with reasonable assurance by comparing risk metrics that reflect the proposed TS changes to the numerical risk acceptance guidelines in RG 1.174, Revision 3, and RG 1.177, Revision 1.

Furthermore, Condition and Limitation Nos. 2 and 3 provided in Section 5.0 of the NRC SE for WCAP-15376 delineate information the NRC staff requested the licensee to provide in the LAR submittal to assess the applicability of the plant-specific TS change to the approved TR.

WCAP-14333-P-A Tier 1 Evaluation In Section 3.2.1 of the LAR, the licensee discussed the evaluation performed to demonstrate the WCAP-14333-P-A Tier 1 analysis and results for Seabrook. The licensee confirmed that (1) the signals available at Seabrook to actuate reactor trip and safeguards equipment for various events are consistent with those credited in the WCAP analysis, and (2) the analog channel, logic cabinet, master and slave relay, and RTB maintenance intervals at Seabrook are consistent with those assumed in the WCAP-14333 analysis.

The licensee provided the baseline CDF and LERF values for Seabrook as 1.20E-05/year and 1.55E-07/year, respectively. The licensee also provided the risk metrics for the proposed change from the unavailability of RTS and ESFAS instrumentation as a result of implementing WCAP-14333-P-A and WCAP-15376-P-A (Reference 13 and Reference 3). In Section 3.1 of the LAR, the licensee confirmed that Seabrooks current licensing basis is that of a WCAP-10271-P-A (or TOP taken from the Technical Specification Optimization Program) plant, and the system instrumentation design is predominately 2-of-4 logic. The NRC staff confirmed that both WCAP base case CDF and LERF values bounded the baseline Seabrook CDF and LERF, thus meeting the RG 1.174 risk acceptance guidelines. Therefore, the increase in CDF and LERF, and the incremental core damage and incremental large early release probabilities (ICCDPs and ICLERPs) provided in the WCAP analyses meet the risk metrics provided in RG 1.177. The NRC staff finds that the licensees Tier 1 evaluation confirms the applicability of WCAP-14333-P-A to Seabrook to the WCAP; therefore, the Tier 1 analysis and risk metrics provided in the WCAP are acceptable for this risk-informed application.

WCAP-15376-P-A Tier 1 Evaluation In Section 3.2.3 of the LAR, the licensee discussed the evaluation performed to demonstrate the WCAP-15376-P-A Tier 1 analysis and results for Seabrook. The licensee confirmed that (1) the signals available at Seabrook to actuate reactor trip and safeguards equipment for various events are consistent with those credited in the WCAP analysis, and (2) the applicable analog

channel, logic cabinet, RTB test intervals, bypass test times, and CTs are consistent with the WCAP analysis, and plant procedures are in place for the relevant operator actions credited.

The NRC staff reviewed Tables 8.29 and 8.32 provided in the WCAP and confirmed that both WCAP base case CDF and LERF values bounded the baseline Seabrook CDF and LERF values, thus meeting the RG 1.174 risk acceptance guidelines. Therefore, the increase in CDF and LERF, and the incremental core damage and incremental large early release probabilities (ICCDPs and ICLERPs) provided in the WCAP analyses meet the risk metrics provided in RG 1.177. The NRC staff finds that the licensees Tier 1 evaluation confirms the applicability of WCAP-15376-P-A to Seabrook; therefore, the Tier 1 analysis and risk metrics provided in the WCAP are acceptable for this risk-informed application.

WCAP-15376 SE Condition and Limitation No. 3 Condition and Limitation No. 3 of the SE for WCAP-15376 requested the licensee to perform a plant-specific evaluation of the risk impact of concurrent testing of one logic cabinet and associated RTB to ensure conformance with WCAP-15376, RG 1.174, and RG 1.177.

In Section 3.2.3.1 of the LAR, the licensee discussed that for Seabrook, the RTB under test can be in the open or closed position during testing; therefore, it is necessary to have the RTB closed to verify that the RTB will open when testing the RTB trip actuating devices. The licensee stated that the time duration when the RTB is closed when the other RTB is in service is small compared to the allowed bypass test time and is estimated to be less than 30 minutes per RTB surveillance test. The licensee evaluated the impact of this alternate test configuration and determined that in either test configuration, the RTB unavailability associated with the RTB bypass test time remains consistent with the WCAP. The NRC staff finds that the licensee appropriately evaluated the Seabrook RTB configuration and confirmed that the proposed RTB test unavailability is equal to or less than the RTB test unavailability assumed in the WCAP; therefore, Condition and Limitation No. 3 is satisfied.

3.2.4.4 Tier 2: Avoidance of Risk-Significant Plant Configurations RG 1.177, Revision 1, states that the licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service, consistent with the proposed TS change. The second tier evaluates the capability of the licensee to recognize and avoid risk-significant plant configurations that could result if equipment, in addition to that associated with the proposed change, is taken out of service simultaneously, or if other risk-significant operational factors such as concurrent system or equipment testing are also involved.

Tier 2 Evaluation for WCAP-14333-P-A and WCAP-15376-P-A In LAR Sections 3.2.2 and 3.2.4, the licensee provided Tier 2 restrictions for implementation of WCAP-14333 and WCAP-15376, respectively. Furthermore, the licensee discussed that entry into these conditions is not typical and is entered due to equipment failure. In the event of an emergent condition during the extended CT, the licensee will use its Tier 3 configuration risk management program (CRMP) to assess the emergent condition and direct activities (e.g.,

restore the logic train and exit the action statement or fully implement the Tier 2 restrictions).

From the given restrictions, and using the analyses provided in the WCAPs, the licensee evaluated concurrent component outage configurations and provided the applicable Tier 2 restrictions for Seabrook. The NRC staff finds these restrictions are consistent with the Tier 2 restrictions identified in the NRC staffs SEs on WCAP-14333 and WCAP-15376, and therefore, are acceptable for the risk-informed applications (i.e., TSTF-411 and TSTF-418).

3.2.4.5 Tier 3: Risk-Informed Plant Configuration and Control Management Section 2.3 of RG 1.177 discusses Tier 3 of the three-tiered approach for evaluating risk associated with proposed changes to TS CTs. Tier 3 is the establishment of a risk-informed plant configuration control program (i.e., a CRMP) to ensure that other potentially lower probability, but nonetheless risk-significant configurations resulting from maintenance and other operational activities are identified and compensated for. Because the Maintenance Rule, as codified in 10 CFR 50.65(a)(4), requires licensees to assess and manage the potential increase in risk that may result from activities such as surveillance testing and corrective and preventive maintenance, a licensee may use its existing Maintenance Rule program to satisfy Tier 3.

As described in Section 3.2.5 of the LAR, the licensee confirmed that the risk associated with the unavailability of plant equipment (i.e., solid-state protection system) is modeled in the Seabrook CRMP and specifically states that failure of the logic train (cabinet) results in a failure of the associated train of the RTS/RTB and ESFAS. Section 50.65(a)(4) of 10 CFR requires, in part, Before performing maintenance activitiesthe licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities, and that their removal from service is monitored, analyzed, and maintained. The licensee further confirms that Seabrooks CRMP is consistent with the guidance provided in NUMARC 93-01, Revision 4F, endorsed by the NRC in RG 1.160, Revision 4, for meeting 10 CFR 50.65(a)(4)

(Reference 28 and Reference 29, respectively).

WCAP-14333-P-A SE Condition No. 2 and Condition and Limitation No. 2 of the SE for WCAP-15376 requested that the licensee perform Tier 2 and Tier 3 analyses, including a determination of risk-significant configuration insights and confirmation that these insights are incorporated into the plant-specific CRMP.

The NRC staff finds the licensees Tier 3 CRMP is in accordance with the regulatory position specified in RG 1.177 and is acceptable to the extent needed to support this application.

Furthermore, based on the NRC staffs review of the licensees Tier 2 and Tier 3 analyses provided in this section of the SE, the NRC staff concludes that Condition and Limitation No. 2 is fully satisfied.

3.2.4.6 Non-PRA Methods: Internal Fires, Seismic, and Other External Hazards Section 3.1 of RG 1.200, Revision 2, states that missing hazard groups may be evaluated using bounding arguments to cover the risk contributions not address by the PRA model.

The licensee provided qualitative assessments for the internal fires, seismic, and other external hazards in Section 3.2.6 of the LAR. For seismic, internal fires, and other external hazards, the NRC staff reviewed the licensees qualitative assessments as follows.

Seismic Hazard In Section 3.2.6.1 of the LAR, the licensee stated that the seismic risk insights from NUREG-1742 (Reference 30) and risk insights from the Seabrook IEPRA were used in assessing the impact of seismic events on the WCAPs risk assessments as they related to the signal unavailability change for the proposed TS changes to CTs and bypass times. The licensee considered only the loss-of-offsite power and small-break LOCA because for larger plant damage states, the small increase in the signal unavailability does not affect the seismic plant risk for the higher-level seismic events (e.g., large LOCA). In summary, for the LOOP and small-break LOCA, the licensee concluded that due to the fail-safe design and reliability of the normal reactor trip function and operator actions to shut down the reactor following lower-level seismic events, the risk contribution from seismic events associated with the proposed CT extension is assessed to be negligible. The NRC staff finds the licensee provided reasonable assurance that the contribution from seismic hazards for these risk-informed TS changes is not a significant contributor to risk because the Seabrook evaluation was consistent with the WCAP-15376-P-A and WCAP-14333-P-A methodology, the licensee considered defense in depth, and the licensee confirmed that the seismic contribution to risk was small for the requested TS changes.

Internal Fires To assess the contribution from internal fires, the licensee used NUREG-1742 and risk insights from the Seabrook IEPRA. The approach for the assessment performed included (1) use of a plant-level fire initiating event frequency, (2) an assumption that total fire event frequency results in a plant transient event with the need for steam generator decay heat removal, (3) all transient events credit ESFAS and are subject to the WCAP-assessed change in signal unavailability, and (4) credit for alternate signals and manual capability. In Section 3.2.6.2 of the LAR, the licensee provided the quantitative results for one ESFAS and emergency feedwater logic train available and two ESFAS and emergency feedwater logic trains available, along with a qualitative discussion to support consideration of fire-transient events and reactor trip signal.

The licensees evaluations demonstrated that the fail-safe design and reliability of the normal reactor trip function, the availability of the anticipated transient without scram mitigation system actuation circuitry system to trip the reactor, and the reliability of operator actions to manually trip the reactor, all contribute to the negligible internal fire risk contribution associated with the proposed CT extension. The NRC staff finds the licensee provided reasonable assurance that the contribution from internal fires for this risk-informed TS change is not a significant contributor because the licensee performed an evaluation using the risk insights provided in NUREG-1742 and the IEPRA to assess the TS changes, considered defense in depth, and confirmed that the fire contribution to risk was small for these TS changes.

High Winds Hazard, External Flood, and Other External Events In Section 3.2.6.3 of the LAR, the licensee stated that high winds, external flood, and other external hazards were reviewed and screened using the guidance provided in NUREG-1407; NUREG/CR-2300, Volume 2, A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants; and the ASME/ANS RA-Sa-2009 PRA standard. The licensee stated that the assessment demonstrated that actuation signal unavailability is a very small contributor to the CDF for these external events. The NRC staff finds the licensees conclusion that these external events are not significant risk contributors to this application is acceptable because the

licensee screened these external events consistent with NRC-endorsed guidance and the risk results provided in Tables 1.3, 1.4, and 1.5 of TSTF-418 for WCAP-14333-P and Tables 8.29 and 8.32 of WCAP-15376-P-A demonstrated margin to the risk acceptance guidelines provided in RG 1.174 and RG 1.177.

3.2.4.7 Conclusions of Key Principle 4 The NRC staff finds that the licensee has demonstrated the applicability of WCAP-14333-P and WCAP-15376 to Seabrook and has met the limitations and conditions of WCAP-14333-P and WCAP-15376. The Tier 1 analyses for both WCAPs are acceptable given that the evaluation for the CDF and LERF values and the estimates provided by the licensee for CDF, LERF, ICCDP, and ICLERP are within the acceptance guidelines of RG 1.174 and RG 1.177 for the proposed TS changes. The Tier 2 analyses evaluated the risk of concurrent outage configurations to identify potential risk-significant configurations. For the Tier 3 analyses, the licensees CRMP at Seabrook was found to be consistent with the guidance in RG 1.177 for the implementation of WCAP-14333 and WCAP-15376. The NRC staff finds the licensee has followed the three-tiered approach outlined in RG 1.177 to evaluate the risk associated with the proposed TS CT changes, and therefore, the proposed changes satisfy Key Principle 4 of RG 1.177.

3.2.5 Key Principle 5: Performance Monitoring RG 1.174 and RG 1.177 also establish the need for an implementation and monitoring program to ensure that extensions to TS CTs do not degrade operational safety over time and that no adverse effects occur from unanticipated degradation or common cause mechanisms. The purpose of an implementation and monitoring program is to ensure that the impact of the proposed TS change continues to reflect the reliability and availability of structures, systems, and components impacted by the change. In addition, the application of the three-tiered approach in evaluating the extensions to CTs provides additional assurance that the changes will not significantly impact the key principle of defense in depth.

The licensee monitors the reliability and availability of the RTS and ESFAS instrumentation under the Maintenance Rule (10 CFR 50.65), which requires a licensee to monitor the performance or condition of structures, systems, and components against licensee-established goals. The NRC staff finds that Seabrook satisfies the RG 1.174 and RG 1.177 guidelines for an implementation and monitoring program for the proposed change, and therefore, Key Principle 5 is met.

3.2.6 Summary of WCAP-14333 and WCAP-15376 Conditions and Limitations Condition and Limitation No. 1 of the SEs for WCAP-14333-P-A and WCAP-15376-P-A requested the licensee to confirm applicability of the TRs to the plant proposing the TS change.

In Sections 3.2.2 and 3.2.4 of this SE, the NRC staff found that Condition and Limitation No. 1 for WCAP-14333-P-A and WCAP-15376-P-A has been met.

Condition and Limitation No. 2 of the SEs for WCAP-14333-P-A and WCAP-15376-P-A requested the licensee to perform Tier 2 and Tier 3 analyses, including determination of risk-significant configuration insights and confirmation that these insights are incorporated into the plant-specific CRMP. In Section 3.2.4.5 of this SE, the NRC staff found that Condition and Limitation No. 2 for WCAP-14333-P-A and WCAP-15376-P-A has been met.

Condition and Limitation No. 3 of the SE for WCAP-15376 requested the licensee to perform a plant-specific evaluation of the risk impact of concurrent testing of one logic cabinet and associated RTB to ensure conformance with WCAP-15376, RG 1.174, and RG 1.177. In Section 3.2.4.3 of this SE, the NRC staff found that Condition and Limitation No. 3 for WCAP-15376-P-A has been met.

Condition and Limitation No. 4 of the SE for WCAP-15376-P-A requested the licensee to confirm that the plant-specific model assumptions for the HRA is consistent with the assumptions delineated in the NRC-approved WCAP-15376. In Section 3.2.4.2 of this SE, the NRC staff found that Condition and Limitation No. 4 for WCAP-15376-P-A has been met.

Condition and Limitation No. 5 of the SE for WCAP-15376-P-A stated for future digital upgrades with increased scope, integration, and architectural differences beyond that of Eagle 21, the staff finds that generic applicability of WCAP-15376-P, Revision 0, to future digital systems is not clear and should be considered on a plant-specific basis. The current RPS and ESFAS systems at Seabrook are analog. In the LAR, the licensee stated that there are presently no plans to implement upgrades to either system. Therefore, this condition does not apply to the current Seabrook LAR and is not approved for future digital upgrades.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of New Hampshire and Commonwealth of Massachusetts officials were notified of the proposed issuance of the amendment on December 15, 2020. The officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, published in the Federal Register on January 28, 2020 (85 FR 5053), and there has been no public comment on such finding.

Accordingly, the amendment meets 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 amendment.

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

7.0 REFERENCES

1. McCartney, E., NextEra Energy Seabrook, LLC, letter to U.S. Nuclear Regulatory Commission, License Amendment Request 17-06, Change to the Technical Specification Requirements for Reactor Trip System Instrumentation and Engineered Safety Features Actuation System Instrumentation to Implement WCAP-14333 and WCAP-15376, dated November 1, 2019 (ADAMS Accession No. ML19310D804).

2. McCartney, E., NextEra Seabrook, LLC, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-411 and TSTF-418, dated July 13, 2020 (ADAMS Accession No. ML20195B098).

3. Newton, R. A. Westinghouse Owners Group, letter to U.S. Nuclear Regulatory Commission, Transmittal of Reports: WCAP-14333-P [Proprietary] and WCAP-14334-NP [Non-proprietary] Entitled Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, dated June 20, 1995 (ADAMS Accession No. ML17263B245).

4. Bryan, R. H. Westinghouse Owners Group, letter to U.S. Nuclear Regulatory Commission, Transmittal of Approved Topical Reports: WCAP-15376-P-A, Rev. 1 (Proprietary) and WCAP-15377-NP-A, Rev. 1 (Non-Proprietary), Entitled Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, (MUHP-3046), dated March 19, 2003 (ADAMS Accession No. ML030870033).

5. Technical Specification Task Force (TSTF) Traveler TSTF-411, Revision 1, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P), dated August 7, 2002 (ADAMS Accession No. ML022470164).

6. Technical Specification Task Force (TSTF) Traveler TSTF-418, RPS and ESFAS Test Times and Completion Times (WCAP-14333), dated August 26, 2001 (ADAMS Accession No. ML012530049).

7. Sheppard, J.J., Westinghouse Owners Group, letter to Denton, H. R., U.S. Nuclear Regulatory Commission, Westinghouse Owners Group Submittal of WCAP-10271, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System, dated February 3, 1983 (ADAMS Accession No. ML20028G404).

8. WCAP-10271-P-A, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System, dated February 21, 1985 (ADAMS Accession No. ML18029A333 (proprietary, non-public)).

9. Newton, R. A., Westinghouse Owners Group, letter to Hodges, M. W., U.S. Nuclear Regulatory Commission, Westinghouse Owners Group Transmittal of Approved Versions of WCAP-10271, Supplement 2, Revision 1-P-A, and WCAP-10272, Supplement 2, Revision 1-A, Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System, dated May 12, 1989 (ADAMS Accession No. ML20246J878).

10. Ross, C., U.S. Nuclear Regulatory Commission, to Russell, W., Waiver of CRGR

[Committee to Review Generic Requirements] Review of WCAP-10271, Supplement 2, Revision 1, Safety Evaluation Re: Evaluation of Surveillance Frequency and Out of Service Times for ESFAS, dated April 19, 1990 (ADAMS Accession No. ML19332H492 (non-public).

11. U.S. Nuclear Regulatory Commission, NUREG-1431, Volume 1, Standard Technical Specifications Westinghouse Plants, dated September 1992 (ADAMS Accession No. ML13196A330).

12. U.S. Nuclear Regulatory Commission, NUREG-1431, Volume 1, Standard Technical Specifications Westinghouse Plants, Revision 4, dated April 2012 (ADAMS Accession No. ML12100A222).

13. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ADAMS Accession No. ML17317A256).

14. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, dated May 2011 (ADAMS Accession No. ML100910008).

15. 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).

16. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Chapter 16, Section 16.1, Revision 1, Risk-informed Decision Making: Technical Specifications, dated March 2007 (ADAMS Accession No. ML070380228).

17. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan (SRP),

Chapter 19, 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 (ADAMS Accession No. ML12193A107).

18. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan (SRP),

Chapter 19, 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).

19. U.S. Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, Final Report, dated June 1991 (ADAMS Accession No. ML063550238).

20. U.S. Nuclear Regulatory Commission, NUREG/CR-2300, Volume 2, PRA Procedures Guide, A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, Final Report, Chapters 9-13 and Appendices A-G, dated January 1983 (ADAMS Accession No. ML063560440).

21. American Society of Mechanical Engineers and American Nuclear Society Standard, ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated February 2009, New York, NY.

22. McCartney, E., NextEra Seabrook, LLC, letter to U.S. Nuclear Regulatory Commission, License Amendment Request 19-02, One-Time Change to the Seabrook Technical Specifications Onsite Power Distribution Requirements, dated October 3, 2019 (ADAMS Accession No. ML19276G055).

23. Poole, J., U.S. Nuclear Regulatory Commission, e-mail to Brown, K. and Thomas, C.,

Request for Additional Information Related to Seabrook License Amendment Request Regarding TSTF-411 and TSTF-418 (L-2019-LLA-0237), dated June 11, 2020 (ADAMS Accession No. ML20167A184).

24. U.S. Nuclear Regulatory Commission, NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Final Report, dated March 2017 (ADAMS Accession No. ML17062A466).

25. Giitter, J. and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, letter to Krueger, G.,

Nuclear Energy Institute, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ADAMS Accession No. ML17079A427).

26. Poole, J., U.S. Nuclear Regulatory Commission, letter to Moul, D., NextEra Seabrook, LLC, Seabrook Station, Unit No. 1 - Issuance of Amendment No. 163 Re: One-Time Change to the Onsite Power Distribution Requirements (EPID L-2019-LLA-0216), dated December 5, 2019 (ADAMS Accession No. ML19326C480).

27. McCartney, E., NextEra Seabrook, LLC, letter to U.S. Nuclear Regulatory Commission, Seabrook Station, Response to Request for Additional Information Related to Seabrook Inverter Amendment, dated October 30, 2019 (ADAMS Accession No. ML19305A301).

28. Nuclear Energy Institute, NUMARC 93-01, Revision 4F, Industry Guideline for Monitoring The Effectiveness of Maintenance At Nuclear Power Plants, dated April 2018 (ADAMS Accession No. ML18120A069).

29. U.S. Nuclear Regulatory Commission, RG 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4, dated August 2018 (ADAMS ML No. 18220B281).

30. U.S. Nuclear Regulatory Commission, NUREG-1742, Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program, Final Report, Volume 1, dated April 2002 (ADAMS Accession No. ML021270070).

Principal Contributors: A. Brown J. Ashcraft Date: December 28, 2020

ML20293A157 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EICB/BC NRR/DRA/APLA/BC NAME JPoole LRonewicz MWaters RPascarelli DATE 10/22/2020 10/21/2020 9/23/2020 9/25/2020 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME VCusumano STurk JDanna JPoole DATE 10/02/2020 12/17/2020 12/28/2020 12/28/2020

v t e Letter Sequence Approval
EPID:L-2019-LLA-0237, TSTF-411, TSTF-418 (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions 11 June 2020 Answers 13 July 2020 Results Approval Other:
MONTHYEAR2020-12-28 [Table View]

ML20293A157: Difference between revisions

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SUBJECT:

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1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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ML20293A157: Difference between revisions

Revision as of 01:58, 8 March 2021

Text

SUBJECT:

Dear Mr. Moul:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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