Text

Issuance of Amendment No. 178 Regarding Revision to Technical Specifications to Adopt TSTF505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b
	ML25338A240
Person / Time
Site:	Seabrook ; (NPF-086)
Issue date:	02/18/2026
From:	V Sreenivas; Plant Licensing Branch 1
To:	Coffey R; Florida Power & Light Co
	Lantigua, R, NRR/DORL/LPLI,
References
	EPID L-2025-LLA-0025
	Download: ML25338A240 (0)
	v • d • e

February 18, 2026 Mr. Robert Coffey Executive Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company Mail Stop: EX/JB 700 Universe Blvd.

Juno Beach, FL 33408

SUBJECT:

SEABROOK STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 178 REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO ADOPT TSTF505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4b (EPID L-2025-LLA-0025)

Dear Mr. Coffey:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 178 to Renewed Facility Operating License No. NPF 86 for the Seabrook Station, Unit No. 1. The amendment consists of changes to the technical specifications (TSs) in response to your application dated February 3, 2025, as supplemented by letter dated September 5, 2025.

The amendment revises the TS requirements to permit the use of Risk Informed completion times for actions to be taken when limiting conditions for operation are not met.

The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b, dated July 2, 2018.

R. Coffey A copy of the related safety evaluation is enclosed. Notice of Issuance will be included in the next Commissions monthly Federal Register notice.

Sincerely,

/RA/

Dr. V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443

Enclosures:

1. Amendment No. 178 to NPF-86

2. Safety Evaluation cc: Listserv NEXTERA ENERGY SEABROOK, LLC DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 178 Renewed License No. NPF-86 1.

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

A.

The application for amendment by NextEra Energy Seabrook, LLC (NextEra, the licensee) dated February 3, 2025, as supplemented by letter dated September 5, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph C.(2) of the 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. 178, 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.

Accordingly, the license is amended by changes as indicated in the attachment to this licensee amendment, and new paragraph 2.L of Renewed Facility Operating License No. NPF-86 will read as follows:

Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times-RITSTF Initiative 4b NextEra Energy Seabrook, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, Risk-Informed CT (RICT). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0, which was approved by the NRC on May 17, 2007.

4.

This amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-86 and the Technical Specifications Date of Issuance: February 18, 2026 UNDINE SHOOP Digitally signed by UNDINE SHOOP Date: 2026.02.18 13:36:47 -05'00'

ATTACHMENT TO SEABROOK STATION, UNIT NO. 1 LICENSE AMENDMENT NO. 178 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 Page 3 Page 3 Page 8 Page 8 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Remove Insert 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 3-19 3/4 3-19 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-23a 3/4 4-11 3/4 4-11 3/4 5-4 3/4 5-4 3/4 6-7 3/4 6-7 3/4 6-12 3/4 6-12 3/4 6-12a 3/4 6-14 3/4 6-14 3/4 6-16 3/4 6-16 3/4 6-16a 3/4 7-3 3/4 7-3 3/4 7-9 3/4 7-9 3/4 7-10 3/4 7-10 3/4 7-12 3/4 7-12 3/4 7-13 3/4 7-13 3/4 7-13A 3/4 7-13A 3/4 7-13B 3/4 7-13B

3/4 8-1a 3/4 8-1a 3/4 8-2 3/4 8-2 3/4 8-2a 3/4 8-2a 3/4 8-12 3/4 8-12 3/4 8-17 3/4 8-17 3/4 8-17a 3/4 8-17a 6-14d 6-14d 6-14e 6-14f Amendment No. 178 (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.

178, 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.

Seabrook, LLC, shall operate the facility in accordance with the Additional Conditions.

K. Inadvertent Actuation of the Emergency Core Cooling System (ECCS)

Prior to startup from refueling outage 11, FPL Energy Seabrook* commits to either upgrade the controls for the pressurizer power operated relief valves (PORV) to safety-grade status and confirm the safety-grade status and water-qualified capability of the PORVs, PORV block valves and associated piping or to provide a reanalysis of the inadvertent safety injection event, using NRC approved methodologies, that concludes that the pressurizer does not become water solid within the minimum allowable time for operators to terminate the event. NextEra Energy Seabrook, LLC submitted an analysis of the inadvertent safety injection event in a letter dated November 7, 2005. In a "L-2006-118, 60-Day Response to NRC Generic Letter 2006-03, Potentially Nonconforming Hemyc and MT [[Topic" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Configurations.|letter dated June 9, 2006]], the NRC concluded the analysis met the requirements of License Condition 2.K.

L. Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times-RITSTF Initiative 4b NextEra Energy Seabrook, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, Risk-Informed CT (RICT). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0, which was approved by the NRC on May 17, 2007.

3.

This renewed license is effective as of the date of issuance and shall expire at midnight on March 15, 2050.

FOR THE NUCLEAR REGULATORY COMMISSION IRAI Ho K. Nieh, Director Office of Nuclear Reactor Regulation Attachments/ Appendices:

1. Appendix A - Technical Specifications (NUREG-1386)

2. Appendix B - Environmental Protection Plan

3. Appendix C - Additional Conditions Date of Issuance: March 12, 2019

On April 16, 2009, the name "FPL Energy Seabrook, LLC" was changed to "NextEra Energy Seabrook, LLC."

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 in accordance with the Risk Informed Completion Time Program, 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 , or in accordance with the Risk Informed Completion Time Program,

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, 178

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 , or in accordance with the Risk Informed Completion Time Program, 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 , or in accordance with the Risk Informed Completion Time Program, 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, 178

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 in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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, 178

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM FUNCTIONAL UNIT TOTAL NO.

CHANNELS OF CHANNELS TO TRIP CHANNELS OPERABLE APPLICABLE MODES ACTION 4.

Steam Line Isolation (continued) b.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2, 3 20 c.

Containment Pressure--

Hi-2 3

2 2

1, 2, 3 18

d.

e.

Steam Line Pressure-Low Steam Generator Pressure - Negative 3/steam line 3/steam line 2/steam line any steam line 2/steam line 2/steam line 2/steam line 1, 2, 3#

3*

18 25 Rate-High any steam line 5.

Turbine Trip a.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2 22 b.

Steam Generator Water Level--

High-High (P-14) 4/stm. gen.

2/stm. gen.

3/stm. gen.

1, 2 18 6.

Feedwater Isolation a.

Steam Generator Water 4/stm. gen.

2/stm. gen.

3/stm. gen. 1, 2 18 Level--High-High (P-14) b.

Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

SEABROOK - UNIT 1 3/4 3-19 Amendment No. 45, 114, 178

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

b.

RWST Level--Low-Low 4

2 3

1, 2, 3, 4 15 Coincident With:

Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

9.

Loss of Power (Start Emergency Feedwater)

a.

4.16 kV Bus E5 and E6-Loss of Voltage 2/bus 2/bus 1/bus 1, 2, 3, 4 14

b.

4.16 kV Bus E5 and E6-Degraded Voltage 2/bus 2/bus 1/bus 1, 2, 3, 4 14 Coincident with SI See Item 1. above for all Safety Injection initiating functions and requirements.

10. Engineered Safety Features Actuation System Interlocks

a.

Pressurizer Pressure, P-11 3

2 2

1, 2, 3 19

b.

Reactor Trip, P-4 2

2 2

1, 2, 3 24

c.

Steam Generator Water Level, P-14 4/stm. gen.

2/stm. gen. 3/stm. gen.

1, 2, 3 25 SEABROOK - UNIT 1 3/4 3-21 Amendment No. 47, 140, 145, 178

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 in accordance with the Risk Informed Completion Time Program, 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 , or in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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, 178

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 , or in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, 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 in accordance with the Risk Informed Completion Time Program, or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5.

ACTION 24 - 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 .

SEABROOK - UNIT 1 3/4 3-23 Amendment No. 36 167, 178

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 25 - 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.

SEABROOK - UNIT 1 3/4 3-23a Amendment No. 178

REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With one or both PORV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s);

otherwise, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to OPERABLE status or close each associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d.

With one or both block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve(s) to OPERABLE status or place its associated PORV(s) control switch to "CLOSE". Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program; otherwise, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SEABROOK - UNIT 1 3/4 4-11 Amendment No. 16, 114, 178

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350ºF LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE centrifugal charging pump,

b.

One OPERABLE Safety Injection pump,

c.

One OPERABLE RHR heat exchanger,

d.

One OPERABLE RHR pump, and

e.

An OPERABLE flow path* capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2, and 3**.

ACTION:

a.

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

During MODE 3, the discharge paths of both Safety Injection pumps may be isolated by closing for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform surveillance testing as required by Specification 4.4.6.2.2.

- The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump and the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.1.2 provided the centrifugal charging pump and the Safety Injection pumps are restored to OPERABLE status within at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375ºF, whichever comes first.

SEABROOK - UNIT 1 3/4 5-4 Amendment No. 114, 178

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT AIR-LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE in accordance with the Containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

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

Enter the ACTION of LCO 3.6.1.2, Containment Leakage, when air lock leakage results in exceeding the overall containment leakage rate acceptance criteria.

a.

With one containment air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed* and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed, 2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days, 3.

Otherwise, 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 , and b.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in accordance with the Risk Informed Completion Time Program, 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 .

Except during entry to repair an inoperable inner door, for a cumulative time not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per year.

SEABROOK - UNIT 1 3/4 6-7 Amendment No. 49, 114, 161, 178

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each 8-inch containment purge supply and exhaust isolation valve shall be OPERABLE and sealed closed except when open for purge system operation for pressure control; for ALARA, respirable, and air quality considerations to facilitate personnel entry; and for surveillance tests that require the valve(s) to be open.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With one or more penetrations with one of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specification 3.6.1.7 above, close the open 8-inch valve(s) or isolate the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or in accordance with the Risk Informed Completion Time Program; otherwise 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 .

b.

With one or more penetrations with one of the 8-inch containment purge supply or exhaust isolation valves having a measured leakage rate in excess of the limits of the Containment Leakage Rate Testing Program, restore the inoperable valve(s) to OPERABLE status or isolate the affected penetration(s) so that the measured leakage rate does not exceed the limits of the Containment Leakage Rate Testing Program, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in accordance with the Risk Informed Completion Time Program, and close the purge supply if the affected penetration is the exhaust penetration, otherwise 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 .

c.

With one or more penetrations with both of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specification 3.6.1.7 above, close at least one of the 8-inch valves within each affected penetration or isolate the penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or otherwise 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 .

SEABROOK - UNIT 1 3/4 6-12 Amendment No. 49, 178

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION

d.

With one or more penetrations with both of the 8-inch containment purge supply or exhaust isolation valves having a measured leakage rate in excess of the limits of the Containment Leakage Rate Testing Program, restore at least one of the valves within each affected penetration(s) to within the allowable leakage limit or isolate the affected penetration(s) so that the measured leakage rate does not exceed the limits of the Containment Leakage Rate Testing Program within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and close the purge supply if the affected penetration is the exhaust penetration, or otherwise 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 .

SEABROOK - UNIT 1 3/4 6-12a Amendment No. 178

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST* and automatically transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 ; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by:

1)

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position**, and 2)

Verifying Containment Spray locations susceptible to gas accumulation are sufficiently filled with water.

b.

By verifying OPERABILITY of each pump when tested in accordance with the INSERVICE TESTING PROGRAM; c.

In accordance with the Surveillance Frequency Control Program by:

1)

Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal, and 2)

Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.

d.

By verifying each spray nozzle is unobstructed following activities that could result in nozzle blockage.

In MODE 4, when the Residual Heat Removal System is in operation, an OPERABLE flow path is one that is capable of taking suction from the refueling water storage tank upon being manually realigned.

- Not required to be met for system vent flow paths opened under administrative control.

SEABROOK - UNIT 1 3/4 6-14 Amendment No. 30, 90, 128, 141, 144, 154, 158, 178

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE*(**).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

NOTES-----------------------------------------------------------

1. Enter applicable ACTIONS for systems made inoperable by containment isolation valves.

2. Enter the ACTION of LCO 3.6.1.2, Containment Leakage, when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

3. Isolation devices in high radiation areas or that are locked, sealed, or otherwise secured may be verified by use of administrative means.

1. With one of the isolation valve(s) inoperable in one or more penetrations, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or in accordance with the Risk Informed Completion Time Program,

a.

Restore the inoperable valve(s) to OPERABLE status, or

b.

Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolation position, or by use of at least one closed manual valve or blind flange or check valve with flow through the valve secured, and following isolation, verify the affected penetration flow path is isolated once per 31 days for isolation devices outside containment and prior to entering MODE 4 from MODE 5 if not performed with the previous 92 days for isolation devices inside containment, or

c.

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 .

2. With two of the isolation valves inoperable in one or more penetrations,

a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , isolate the affected penetration flow path(s) by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange, or

Locked or sealed closed valves may be opened on an intermittent basis under administrative control.

- Refer to TS 3.6.1.7 for the containment purge supply and exhaust isolation valve requirements.

SEABROOK - UNIT 1 3/4 6-16 Amendment No. 120, 141, 158, 161, 178

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 (Continued) b.

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 .

SURVEILLANCE REQUIREMENTS 4.6.3.1 Not used 4.6.3.2 Each containment isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a.

Verifying that on a Phase "A" Isolation test signal, each Phase "A" Isolation valve actuates to its isolation position, b.

Verifying that on a Phase "B" Isolation test signal, each Phase "B" Isolation valve actuates to its isolation position, and SEABROOK - UNIT 1 3/4 6-16a Amendment No. 178

PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a.

One motor-driven emergency feedwater pump, and one startup feedwater pump capable of being powered from an emergency bus and capable of being aligned to the dedicated water volume in the condensate storage tank, and

b.

One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

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

1.

LCO 3.0.4.b is not applicable to the EFW pumps when entering MODE 1.

2.

LCO 3.0.4.b is not applicable to the startup feedwater pump.

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two emergency feedwater pumps inoperable, restore at least one emergency feedwater pump to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and restore both emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 .

c.

With one emergency feedwater pump and the startup feedwater pump inoperable, restore both emergency feedwater pumps to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in accordance with the Risk Informed Completion Time Program, and all three pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SEABROOK - UNIT 1 3/4 7-3 Amendment No. 34, 114, 178

PLANT SYSTEMS TURBINE CYCLE MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3#.

ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or in accordance with the Risk Informed Completion Time Program; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2 and 3:

With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5.0 seconds when tested in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

Entry into this MODE is permitted for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform post-modification or post-maintenance testing to verify OPERABILITY of components. ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 3/4 7-9 Amendment No. 154, 178

PLANT SYSTEMS TURBINE CYCLE ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 At least four atmospheric relief valves and associated manual controls including the safety-related gas supply systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3#, and 4*#.

ACTION:

a.

With one less than the required atmospheric relief valves OPERABLE, restore the required atmospheric relief valves to OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program; or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With two less than the required atmospheric relief valves OPERABLE, restore at least three atmospheric relief valves to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 .

SURVEILLANCE REQUIREMENTS 4.7.1.6 Each atmospheric relief valve and associated manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the nitrogen accumulator tank is at a pressure greater than or equal to 500 psig.

b.

Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, verify that all valves will open and close fully by operation of manual controls.

When steam generators are being used for decay heat removal.

Entry into this MODE is permitted for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform post-modification or post-maintenance testing to verify OPERABILITY of components. ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 3/4 7-10 Amendment No. 141, 178

PLANT SYSTEMS 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent primary component cooling water loops shall be OPERABLE, including one OPERABLE pump in each loop.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

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

Enter applicable ACTIONS of LCO 3.4.1.3, Reactor Coolant Loops and Coolant Circulation, for residual heat removal loops made inoperable by PCCW.

With one primary component cooling water (PCCW) loop inoperable, restore the required primary component cooling water loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 .

SURVEILLANCE REQUIREMENTS 4.7.3 At least two primary component cooling water loops shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and b.

In accordance with the Surveillance Frequency Control Program by verifying that each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Feature actuation signal.

SEABROOK - UNIT 1 3/4 7-12 Amendment No. 32, 141, 158, 161, 178

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM/ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.4 The Service Water System shall be OPERABLE with:

a.

An OPERABLE service water pumphouse and two service water loops with one OPERABLE service water pump in each loop, b.

An OPERABLE mechanical draft cooling tower and two cooling tower service water loops with one OPERABLE cooling tower service water pump in each loop, and c.

A portable cooling tower makeup system stored in its design operational readiness state.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

NOTES--------------------------------------------------

1. Enter applicable ACTIONS of LCO 3.8.1.1, AC Sources-Operating, for diesel generator made inoperable by service water.

2. Enter applicable ACTIONS of LCO 3.4.1.3, Reactor Coolant Loops and Coolant Circulation, for residual heat removal loops made inoperable by service water.

a.

With one service water loop inoperable, return the loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or in accordance with the Risk Informed Completion Time Program, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one cooling tower service water loop or one cooling tower cell inoperable, return the affected loop or cell to OPERABLE status within 21 days, or in accordance with the Risk Informed Completion Time Program, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With two cooling tower service water loops or the mechanical draft cooling tower inoperable, return at least one loop and the mechanical draft cooling tower to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or in accordance with the Risk Informed Completion Time Program, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEABROOK - UNIT 1 3/4 7-13 Amendment No. 32, 116, 161, 172, 178

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM/ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION ACTION: (Continued) d.

With two loops (except as described in c) or the service water pumphouse inoperable, return at least one of the affected loops and the service water pumphouse to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or in accordance with the Risk Informed Completion Time Program, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e.

With the portable tower makeup pump system not stored in its design operational readiness state, restore the portable tower makeup pump system to its required condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or continue operation and notify the NRC within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of actions to ensure an adequate supply of makeup water for the service water cooling tower for a minimum of 30 days.

SURVEILLANCE REQUIREMENTS 4.7.4.1 Each service water loop shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and b.

In accordance with the Surveillance Frequency Control Program by verifying that each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Feature actuation test signal.

4.7.4.2 Each service water cooling tower loop shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and b.

In accordance with the Surveillance Frequency Control Program by verifying that:

1)

Each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Feature actuation test signal, SEABROOK - UNIT 1 3/4 7-13A Amendment No. 32, 116, 141, 158, 161, 178

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM/UTIMATE HEAT SINK SURVEILLANCE REQUIREMENTS 4.7.4.2 (Continued)

2)

Each automatic valve in the flowpath actuates to its correct position on a Tower Actuation (TA) test signal and

3)

Each service water cooling tower pump starts automatically on a TA signal.

4.7.4.3 The service water pumphouse shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the water level to be at or above 25.1' (-15.9' Mean Sea Level).

4.7.4.4 The mechanical draft cooling tower shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying the water in the mechanical draft cooling tower basin to be at a level of greater than or equal to 42.15* feet.

b.

In accordance with the Surveillance Frequency Control Program by verifying that the water in the cooling tower basin to be at a bulk average temperature of less than or equal to 70ºF.

c.

In accordance with the Surveillance Frequency Control Program by:

1)

Starting from the control room each cooling tower fan that is required to be OPERABLE and operating each of these fans for at least 15 minutes, and

2)

Verifying that the portable tower makeup pump system is stored in its design operational readiness state.

d.

In accordance with the Surveillance Frequency Control Program by verifying that the portable tower makeup pump develops a flow greater than or equal to 200 gpm.

With the cooling tower in operation with valves aligned for tunnel heat treatment, the tower basin level shall be maintained at greater than or equal to 40.55 feet.

SEABROOK - UNIT 1 3/4 7-13B Amendment No. 32, 141, 161, 178

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 (Continued)

ACTION:

3. Restore the inoperable offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or in accordance with the Risk Informed Completion Time Program, 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 .

SEABROOK - UNIT 1 3/4 8-1a Amendment No. 175, 178

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 (Continued)

ACTION:

b.

With a diesel generator inoperable:

1) Demonstrate the OPERABILITY of the remaining A.C. sources by performing Specification 4.8.1.1.1a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

Perform ACTION d. Demonstrate the OPERABILITY of the remaining diesel generator by performing Specification 4.8.1.1.2a.5) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*

2) Restore the inoperable diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or in accordance with the Risk Informed Completion Time Program, 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 , unless the following condition exists:

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

ACTIONs b.2(a) and b.2(b) shall not be applied in conjunction with a Risk Informed Completion Time.

(a) The requirement for restoration of the diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months may be extended to 14 days if the Supplemental Emergency Power System (SEPS) is available, as specified in the Bases, and (b)

If at any time the SEPS availability cannot be met, either restore the SEPS to available status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (not to exceed 14 days from the time the diesel generator originally became inoperable), 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 .

The OPERABILITY of the remaining diesel generator need not be verified if it has been successfully operated within the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or if currently operating, or if the diesel generator became inoperable due to:

1.

Preplanned preventive maintenance or testing,

2.

An inoperable support system with no potential common mode failure for the remaining diesel generator, or

3.

An independently testable component with no potential common mode failure for the remaining diesel generator.

SEABROOK - UNIT 1 3/4 8-2 Amendment No. 30, 80, 97, 166, 178

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 (Continued)

ACTION:

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

Enter applicable ACTIONS of LCO 3.8.3.1, Onsite Power Distribution -

Operating, when ACTION c is entered with no AC power to any train.

c.

With one offsite circuit and one diesel generator of the above required A.C.

electrical power sources inoperable:

1)

Demonstrate the OPERABILITY of the remaining A.C. source by performing Specification 4.8.1.1.1a. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Perform ACTION d. Demonstrate the OPERABILITY of the remaining diesel generator by performing Specification 4.8.1.1.2a.5) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .*

2)

Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or in accordance with the Risk Informed Completion Time Program, 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 .

The OPERABILITY of the remaining diesel generator need not be verified if it has been successfully operated within the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or if currently operating, or if the diesel generator became inoperable due to:

1.

Preplanned preventive maintenance or testing,

2.

An inoperable support system with no potential common mode failure for the remaining diesel generator, or

3.

An independently testable component with no potential common mode failure for the remaining diesel generator.

SEABROOK - UNIT 1 3/4 8-2a Amendment No. 30, 80, 97, 161, 178

ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:

a.

Train A

1) 125-volt Battery Bank 1A or 1C,

2)

One full-capacity battery charger on Bus #11A, and

3)

One full-capacity battery charger on Bus #11C.

b.

Train B

1) 125-volt Battery Bank 1B or 1D,

2)

One full-capacity battery charger on Bus #11B, and

3)

One full-capacity battery charger on Bus #11D.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With the required battery bank in one train inoperable, restore the battery bank to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in accordance with the Risk Informed Completion Time Program, 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 .

b.

With one or two full-capacity chargers in one train inoperable, restore the inoperable charger(s) to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in accordance with the Risk Informed Completion Time Program, 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 .

SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that:

1)

The parameters in Table 4.8-2 meet the Category A limits, and

2)

The total battery terminal voltage is greater than or equal to 128 volts on float charge.

b.

In accordance with the Surveillance Frequency Control Program and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

SEABROOK - UNIT 1 3/4 8-12 Amendment No. 141, 157, 178

ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 (Continued)

i.

Train A, 125-volt D.C. Busses consisting of:

1) 125-volt D.C. Bus

11A energized from Battery Bank 1A or 1C, and

2) 125-volt D.C. Bus

11C energized from Battery Bank 1C or 1A.

j.

Train B, 125-volt D.C. Busses consisting of:

1) 125-volt D.C. Bus #11B energized from Battery Bank 1B or 1D, and

2) 125-volt D.C. Bus #11D energized from Battery Bank 1D or 1B.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

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

Enter applicable ACTIONS of LCO 3.8.2.1, DC Sources - Operating, for DC trains made inoperable by inoperable AC power distribution system.

a.

With one of the required trains of A.C. emergency busses (except 480-volt Emergency Bus # E64) not fully energized, reenergize the train within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or in accordance with the Risk Informed Completion Time Program, 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 .

1.

With 480-volt Emergency bus #E64 not fully energized, reenergize the bus within 7 days or in accordance with the Risk Informed Completion Time Program, or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one A.C. vital panel either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. vital panel within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in accordance with the Risk Informed Completion Time Program, 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 ; and (2) reenergize the A.C. vital panel from its associated inverter connected to its associated D.C. bus within 7 days or in accordance with the Risk Informed Completion Time Program, 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 .

c.

With one D.C. bus not energized from an OPERABLE battery bank, reenergize the D.C.

bus from an OPERABLE battery bank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in accordance with the Risk Informed Completion Time Program, 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 .

SEABROOK - UNIT 1 3/4 8-17 Amendment No. 48, 157,161 163, 170, 178

ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 (Continued)

ACTION d.

With two or more A.C. vital panels of the same electrical train either not energized from their associated inverter, or with their inverters not connected to their associated D.C. bus: (1) reenergize or verify energized all A.C. vital panels within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in accordance with the Risk Informed Completion Time Program, 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 ; and (2) reenergize or verify energized at least two A.C. vital panels from their associated inverters connected to their associated D.C. bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or in accordance with the Risk Informed Completion Time Program, 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 .

SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses and panels shall be determined energized in the required manner in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses.

SEABROOK - UNIT 1 3/4 8-17a Amendment No. 48, 141, 157, 170, 178

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

c.

A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate ACTIONS of the LCO in which the loss of safety function exists are required to be entered.

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate ACTIONS to enter are those of the support system.

p.

Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09, Risk-Informed Technical Specifications Initiative 4b: Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A, November 2006. The program shall include the following:

a.

The RICT may not exceed 30 days;

b.

A RICT may only be utilized in MODES 1 and 2;

c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2.

For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the plant configuration change, whichever is less.

3.

Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.

Numerically accounting for the increased possibility of CCF in the RICT calculation, or SEABROOK - UNIT 1 6-14d Amendment No. 138, 141, 161, 178

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) 2.

Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e.

The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

6.8 REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT 6.8.1.1 A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station.

The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

SEABROOK - UNIT 1 6-14e Amendment No. 178

ADMINISTRATIVE CONTROLS 6.8.1.1 (Continued)

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation),

supplementary reports shall be submitted at least every 3 months until all three events have been completed.

SEABROOK - UNIT 1 6-14f Amendment No. 178 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 178 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-86 NEXTERA ENERGY SEABROOK, LLC SEABROOK STATION, UNIT 1 DOCKET NO. 50-443

1.0 INTRODUCTION

By letter dated February 3, 2025 (Reference [1]), as supplemented by letter dated September 5, 2025 (Reference [2]), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted a license amendment request (LAR) for Seabrook Station Unit 1 (Seabrook). The amendment would revise technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) are not met. The proposed changes are based on Technical Specifications Task Force (TSTF)

Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF

[Risk-Informed TSTF] Initiative 4b, dated July 2, 2018 (Reference [3]). The U.S. Nuclear Regulatory Commission (NRC, the Commission) issued a final model safety evaluation (SE) approving TSTF-505, Revision 2, on November 21, 2018 (Reference [4]).

The licensee has proposed variations from the TS changes approved in TSTF-505, Revision 2, which are provided in Section 2.4, Optional Changes and Variations of LAR, as supplemented, and addressed in section 3.2.1 of this SE.

The NRC staff participated in a regulatory audit to ascertain the information needed to support its review of the application and to develop request for additional information (RAIs), as needed (Reference [5]). Following the regulatory audit, the NRC staff issued RAIs (Reference [6]). On December 2, 2025, the NRC staff issued an audit summary.

The supplemental letter dated September 5, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 13, 2025, 90 FR 205218).

2.0 REGULATORY EVALUATION

2.1 Regulatory Review 2.1.1 Applicable Regulations Title 10 of the Code of Federal Regulations (10 CFR) Part 50 provides the general provisions for Domestic Licensing of Production and Utilization Facilities. The general provisions include, but are not limited to, establishing the regulatory requirements that a licensee must adhere to for the submittal of a license application. The NRC staff has identified the following applicable sections within 10 CFR Part 50 for the NRC staffs review of the licensees application to adopt TSTF-505:

10 CFR 50.36, Technical specifications, which specifies the content and information that must be included in a licensee's TS. In accordance with 10 CFR 50.36(c), TSs are required to include (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

10 CFR 50.55a, Codes and standards, of which 10 CFR 50.55a(h), Protection and safety systems, specifies the requirements for protection systems of nuclear power reactors of all types.

10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, (i.e., the Maintenance Rule), which requires that power reactor licensees monitor the performance or condition of structures, systems, and components against licensee-established goals in a manner sufficient to provide reasonable assurance that such structures, systems, and components are capable of fulfilling their intended functions.

2.1.2 Regulatory Guidance NRC Regulatory Guides (RGs) provide one way to ensure that the codified regulations continue to be met. The NRC staff considered the following guidance, along with industry guidance endorsed by the NRC, during its review of the proposed changes:

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009 (Reference [7])

RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (Reference [8]).

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, dated November 2002, Revision 2, dated March 2009, and Revision 3, dated January 2018 (Reference [9], [10], [11]).

RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 0, dated August 1998 and Revision 2, dated January 2021 (Reference [12] and [13]).

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decisionmaking, dated March 2017 (Reference [14]).

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, (SRP) Section 16.1, Risk-Informed Decision Making: Technical Specifications, dated March 2007 and Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, June 2007 (Reference [15] and

[16]).

Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09 Revision 0-A (NEI 06-09-A), Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, dated October 2012 (Reference [17]),

provides guidance for Risk-Informed TSs. The NRC staff issued a final model SE approving NEI 06-09 on May 17, 2007 (Reference [18]).

The licensees submittal cites RG 1.200, Revision 2, for the internal events PRA (which includes internal floods) and fire PRA models. Although RG 1.200 has been updated to Revision 3, the NRC staff finds it acceptable to use Revision 2 to demonstrate the technical acceptability of the Seabrook PRA models because the update does not introduce any technical changes that would impact the determination of PRA acceptability. The licensees submittal also cites various versions of RG 1.174, (i.e., Revisions 1, 2, and 3, and RG 1.177, i.e., Revisions 0 and 2). The updates to RG 1.174 and 1.177 do not include any technical changes that would impact consistency with NEI 06-09-A; therefore, the NRC staff finds the updated revisions to the RGs also applicable for use in the licensees adoption of TSTF-505, Revision 2.

2.2 Description of the RICT Program The TS LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee must shut down the reactor or follow any remedial or required action (e.g., testing, maintenance, or repair activity) permitted by the TSs until the condition can be met. The remedial actions (i.e., ACTIONS) associated with an LCO contain conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s) (CT). The CTs are referred to as the front stops in the context of this SE. For certain conditions, the TSs require exiting the Mode of Applicability of an LCO (i.e., shut down the reactor).

The licensees submittal requested approval to add a RICT program to the Administrative Controls Section of the TS, and modify selected CTs to permit extending the CTs, provided risk is assessed and managed as described in NEI TR 06-09-A. Consistent with Table 1 of TSTF-505, Rev. 2 for Conditions Requiring Additional Technical Justification NUREG-1431, Westinghouse STS (standard technical specifications), in section 2.3 of the SE the licensee provided several plant-specific LCOs and associated Actions for which Seabrook proposed to be included in the RICT Program, along with additional justification. The NRC staffs review of these variations and the justification is provided in section 3.0 of this SE.

The licensee is proposing no changes to the design of the plant or any operating parameter, and no new changes to the design basis in the proposed changes to the TS. The effect of the proposed changes when implemented will allow CTs to vary, based on the risk significance of the given plant configuration (i.e., the equipment out-of-service at any given time), provided that the system(s) retain(s) the capability to perform the applicable safety function(s) without any further failures (e.g., one train of a two-train system is inoperable). These restrictions on inoperability of all required trains of a system ensure that consistency with the defense in depth (DID) philosophy is maintained by following existing guidance when the capability to perform TS safety function(s) is lost.

The proposed RICT program uses plant-specific operating experience for component reliability and availability data. Thus, the allowances permitted by the RICT program are directly reflective of actual component performance in conjunction with component risk significance.

2.3 Other Technical Specification Changes and Modifications In Section 2.4.3 of Attachment 1 to the LAR, the licensee proposed several variations to TSTF-505, Revision 2.

The following changes are proposed administrative variations that would not alter the current requirements:

Delete footnote

permitted 3B Reserve Auxiliary Transformer maintenance using a one-time CT extension from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 30 days. This allowance is currently a footnote to Action a.3 of TS 3.8.1.1, [alternating current] A.C. Sources - Operating, which applies when one offsite circuit is inoperable. The one-time change expired on March 31, 2024.

Delete footnote ** permitted diesel generator maintenance using a one-time CT extension from 14 days to 30 days. This allowance is currently a footnote to Action b.2(a) of TS 3.8.1.1, applicable when one diesel generator is inoperable. The one-time change expired 90 days after approval (February 22, 2021).

Add footnote ** to TS 3.6.3, Containment Isolation Valves, referring to TS 3.6.1.7, Containment Ventilation System to eliminate confusion regarding which LCO applies to the containment purge supply and exhaust isolation valves.

The licensee proposed adding new actions and modifying existing actions so that a RICT may be applied to specific functions or conditions, consistent with the intent of TSTF-505, Revision 2.

The changes were proposed to prevent inadvertent application of a RICT to conditions out of scope of the RICT program, including: 1) a loss of function condition, 2) inapplicable operating modes, and 3) for certain actions, creating an unclear requirement by adding a RICT to the existing TS language. The proposed changes would affect the following Seabrook TS:

TS 3.3.2, Engineered Safety Features Actuation System Instrumentation New Actions 24 and 25 in Table 3.3-3 were proposed for the associated inoperable instrument channels that may not be included in the RICT program.

TS 3.6.1.7, Containment Ventilation System New actions c and d were proposed to apply when the safety function is not maintained for containment purge supply and isolation valves.

TS 3.6.3, Containment Isolation Valves The licensee proposed to add new action 2 to apply when the safety function is not maintained for containment isolation valves. In addition, language similar to NUREG-1431 was proposed for revised action 1.b to add requirements and associated flexibility, in relation to a RICT, of the periodic verification that the affected penetration flow path is isolated.

3.0 TECHNICAL EVALUATION

An acceptable approach for making risk-informed decisions about proposed TS changes, including both permanent and temporary changes, is to demonstrate that the proposed licensing basis (LB) changes meet the five key principles provided in RG 1.174, and the three-tiered approach outlined in RG 1.177.

3.1 Method of Staff Review Each of the key principles and tiers are addressed in NEI TR 06-09-A and approved in the final model SE issued by the NRC for TSTF-505, Revision 2. NEI 06-09-A provides a methodology for extending existing CTs and thereby delay exiting the operational mode of applicability or taking Required Actions if risk is assessed and managed within the limits and programmatic requirements established by a RICT program. The NRC staffs evaluation of the licensees proposed use of RICTs against the key safety principles of RG 1.174 and 1.177 is discussed below.

3.2 Review of Key Principles 3.2.1 Key Principle 1: Evaluation of Compliance with Current Regulations The regulations in 10 CFR 50.36(c)(2) requires that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any required action permitted by the TS until the condition can be met.

The completion times (CTs) in the current TSs were established using experiential data, risk insights, and engineering judgement. The RICT program provides the necessary administrative controls to permit extension of CTs and, thereby, delay reactor shutdown or Required Actions, if risk is assessed and managed appropriately within specified limits and programmatic requirements and the safety margins and DID remains sufficient. The option to determine the extended CT in accordance with the RICT program allows the licensee to perform an integrated evaluation in accordance with the methodology prescribed in NEI TR 06-09-A and proposed TS 6.7.6.p. The RICT is limited to a maximum of 30 days (termed the back stop).

The typical CT is modified by the application of the RICT program as shown in the following example. The changed portion is indicated in italics.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable.

A.1 Restore subsystem to OPERABLE status.

7 days OR In accordance with the Risk Informed Completion Time Program In LAR Attachment 1, Section 2.4, Optional Changes and Variations, and LAR Attachment 2, Proposed Technical Specification Changes (Mark-Up), as supplemented, the licensee provided a list of the TSs, associated LCOs, and Required Actions for the CTs that included modifications and variations from the approved TSTF-505. The modifications and variations consisted of proposed changes to the Required Actions and CTs. Furthermore, consistent with Table 1 of TSTF-505, Revision 2, for Seabrook TSs 3.3.1, 3.5.2, 3.6.1.3, 3.6.2, 3.7.1.5, and 3.7.1.6 in LAR Enclosure 1, Section 4.0, Additional Justification for Specific Actions, the licensee included additional technical justification to demonstrate the acceptability for including these TS in the RICT program. The NRC staff reviewed the proposed changes to the TS, associated LCOs, Required Actions and CTs provided by the licensee for the scope of the RICT program and concluded, with the incorporation of the RICT program, that the required performance levels of equipment specified in LCOs are not changed, only the required CT for the Required Actions are modified, such that 10 CFR 50.36(c)(2) will remain met. Based on the discussion provided above, the NRC staff finds that the TS program provided in section 2.0 of this SE, LCOs, Required Actions, and CTs meet the first key principle of RG 1.174 and RG 1.177.

3.2.2 Key Principle 2: Evaluation of Defense in Depth (DID)

In RG 1.174, the NRC identified the following considerations used for evaluation of how the LB change is maintained for the DID philosophy:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential common cause failures (CCFs).

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

The licensee requested to use the RICT program to extend the existing CTs for the respective TS LCOs prescribed in Attachment 2 of the LAR, as supplemented. For the TS LCOs, in of the LAR, as supplemented, the licensee provided a description and an assessment of the redundancy and diversity for the proposed changes. The NRC staffs evaluation of the proposed changes for these LCOs assessed the Seabrooks redundant or diverse means to mitigate accidents to ensure consistency with the plant LB requirements using the guidance prescribed in RG 1.174, RG 1.177, and TSTF-505 Revision 2, to ensure adequate DID (for each of the functions) to operate the facility in the proposed manner (i.e., that the changes are consistent with the DID criteria).

of the LAR, Tables E1-3(1), Information to Support RPS Instrumentation Redundancy and Diversity, and E1-3(2), Information to Support ESFAS Instrumentation Redundancy and Diversity, provided information supporting the Seabrook evaluation of the redundancy, diversity, and DID for each TS LCO and TS Required Action as it relate to I&C, electrical, and power systems. The NRC confirmed that for the following TS LCOs, the above DID criteria were applicable except for the criteria for maintaining multiple fission product barriers.

The licensee proposed adding RICTs to the following LCOs related to I&C:

TS 3.3.1, Reactor Trip System Instrumentation TS 3.3.2, Engineered Safety Features Actuation System Instrumentation For the TS LCOs specific to I&C, the NRC staff reviewed the specific trip logic arrangements, redundancy, backup systems, manual actions, and diverse trips specified for each of the protective safety functions and associated instrumentation as described in the associated Updated Final Safety Analysis Report (UFSAR) (Reference [19]) sections, and as reflected in of the LAR, as supplemented, for each I&C LCO above. The NRC staff verified that in accordance with the Seabrook UFSAR and equipment and actions credited in Enclosure 1 of the LAR, as supplemented, in all applicable operating modes, the affected protective feature would perform its intended function by ensuring the ability to detect and mitigate the associated event or accident when the CT of a channel is extended. Furthermore, the NRC staff concludes that there is sufficient redundancy, diversity, and DID, to protect against CCFs and potential single failure for the Seabrook instrumentation systems evaluated in LAR Enclosure 1 during a RICT. There is at least one diverse means specified by the licensee for initiating mitigating action for each accident event, thus providing DID against a failure of instrumentation during the RICT for each TS LCO. The DID specified by the licensee does not overly rely on manual actions as the diverse means; therefore, there is not overreliance of programmatic activities as compensatory measures. Therefore, the NRC staff finds that the intent of the plants design criteria (e.g., safety functions) for the above TS LCOs related to I&C are maintained.

The licensee proposed adding RICTs to the following TS actions related to electrical power systems:

3/4.8.1 A.C. SOURCES - OPERATING o TS LCO 3.8.1.1, Action a.3 - One offsite AC circuit inoperable o TS LCO 3.8.1.1, Action b.2 - One diesel generator inoperable o TS LCO 3.8.1.1, Action c.2 - One offsite AC circuit and one diesel generator inoperable 3/4.8.2 D.C. SOURCES - OPERATING o TS LCO 3.8.2.1, Action a - One required battery bank in one train inoperable o TS LCO 3.8.2.1, Action b - One or two full capacity chargers in one train inoperable 3/4 8.3 ONSITE POWER DISTRIBUTION - OPERATING o TS LCO 3.8.3.1, Action a - One of the required trains of AC emergency buses (except 480-volt Emergency Bus #E64) not fully energized o TS LCO 3.8.3.1, Action a.1 - 480-volt Emergency Bus #E64 not fully energized o TS LCO 3.8.3.1, Action b - One AC vital panel either not energized from its associated inverter, or with the inverter not connected to its associated DC bus o TS LCO 3.8.3.1, Action c - One DC bus not energized from an operable battery bank o TS LCO 3.8.3.1, Action d - Two or more AC vital panels of the same electrical train either not energized from their associated inverter, or with their inverter not connected to their associated DC bus.

In addition to the changes listed above, the proposed changes would revise certain TS 3/4.8 LCO Actions to align with the STS in TSTF-505, Revision 2, as follows:

TS LCO 3.8.1.1, ACTION a.3: Restore at least two offsite circuits is replaced with Restore the inoperable offsite circuit TS LCO 3.8.1.1, ACTION b.2: Restore at least two diesel generators is replaced with Restore the inoperable diesel generator TS LCO 3.8.1.1, ACTION c.3: Action and subparts deemed redundant to TS 3.8.1.1, action a (one inoperable offsite circuit) and action b (one inoperable diesel generator) will be deleted.

TS LCO 3.8.2.1, ACTION b: With one of the full-capacity chargers, restore the inoperable charger is replaced with With one or two full-capacity chargers in one train, restore the inoperable charger(s)

For the TS LCOs specific to electrical and power systems, the NRC staff reviewed the information provided by the licensee in the LAR, as supplemented, for the proposed TS LCOs, TS Bases, and the UFSAR to verify the capacity and capability of the affected electrical power systems to perform their safety functions (assuming no additional failures) is maintained. The NRC staff verified that the design success criteria in enclosure 1, Table E1-1, List of Revised Required ACTIONs to Corresponding PRA Functions, the Seabrook UFSAR states that the plant is designed such that the safety functions are maintained assuming a single failure within the electrical power system. Single-failure requirements are typically suspended for the time that a plant is not meeting an LCO (i.e., in an ACTION statement).

The NRC staff reviewed the licensees proposed electrical TS LCO changes and supporting documentation. The staff finds that given each LCOs reduced redundancy, the CT extensions, as allowed by the RICT Program, are acceptable because (a) the capacity and capability of the remaining operable electrical systems to perform their safety functions (assuming no additional failures) is maintained, and (b) the licensees identification and implementation of Risk Management Actions (RMAs) as compensatory measures, in accordance with the RICT Program will be effective.

The NRC staff notes that while in a TS LCO condition, the redundancy of the function will be temporarily relaxed and, consequently, the system reliability will be degraded accordingly. The NRC staff examined the design information from the Seabrook UFSAR and the risk informed TS LCO conditions for the affected safety functions. Based on this information, the NRC staff confirmed that under any given design-basis accident (DBA) evaluated in the Seabrook UFSAR, the affected protective features maintain adequate DID.

Considering that the CT extensions will be implemented in accordance with the NEI TR 06-09-A guidance, that also considers RMAs, and the redundancy of the offsite and onsite power system, the staff finds that the plant will maintain adequate DID. Therefore, the staff finds the TS LCOs proposed by the licensee in Attachment 2 of the LAR, as supplemented, are acceptable for the RICT program.

The NRC staff reviewed all TS LCOs proposed by the licensee in Attachment 2 of the LAR, as supplemented, and concludes that the proposed changes do not alter the ways in which the Seabrook systems fail, do not introduce new CCF modes, and the system independence is maintained. The NRC staff finds that extending the selected CTs with the RICT program following loss of redundancy, but maintaining the capability of the system to perform its safety function, is an acceptable reduction in DID during the proposed RICT period provided that the licensee identifies and implements compensatory measures in accordance with the RICT program during the extended CT.

Based on the above, the NRC staff finds that the licensees proposed changes are consistent with the NRC-endorsed guidance prescribed in NEI 06-09-A and satisfy the second key principle in RG 1.177. Additionally, the NRC staff concludes that the changes are consistent with the DID philosophy as described in RG 1.174.

3.2.3 Key Principle 3: Evaluation of Safety Margins Paragraph 50.55a(h) of 10 CFR (Codes and Standards) requires in part, that protection systems of nuclear power reactors of all types must meet the requirements specified in this paragraph. Section 2.2.2 of RG 1.177 states, in part, that sufficient safety margins are maintained when:

Codes and standards or alternatives approved for use by the NRC are met.

Safety analysis acceptance criteria in the final safety analysis report (FSAR) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainties.

The licensee is not proposing to change any quality standard, material, or operating specification in this application. In the LAR, as supplemented, the licensee proposed to add a new program, Risk Informed Completion Time Program, in Section 6.7.6 p. under the section title Procedures and Programs, of the TSs, which would require adherence to NEI 06-09-A.

The NRC staff evaluated the effect on safety margins when the RICT is applied to extend the CT up to a backstop of 30 days in a TS condition with sufficient trains remaining operable to fulfill the TS safety function. Although the licensee will be able to have design-basis equipment out of service longer than the current TS allow any increase in unavailability is expected to be insignificant and is addressed by the consideration of the single failure criterion in the design-basis analyses. Acceptance criteria for operability of equipment are not changed and, if sufficient trains remain operable to fulfill the TS safety function, the operability of the remaining train(s) ensures that the current safety margins are maintained. The NRC staff finds that if the specified TS safety function remains operable, sufficient safety margins would be maintained during the extended CT of the RICT program.

Safety margins are also maintained if PRA functionality is determined for the inoperable train which would result in an increased CT. Credit for PRA functionality, as described in NEI 06-09-A, is limited to the inoperable train, LOOP, or component.

Based on the above, the NRC staff finds that the design-basis analyses for Seabrook remains applicable and unchanged, sufficient safety margins would be maintained during the extended CT and the proposed changes to the TSs do not include any change in the standards applied or the safety analysis acceptance criteria. The NRC staff concludes that the proposed changes meet 10 CFR 50.55a(h), and therefore the third key principle of RG 1.174 and RG 1.177.

3.2.4 Key Principle 4: Change in Risk Consistent with the Safety Goal Policy Statement NEI 06-09-A provides a methodology for a licensee to evaluate and manage the risk impact of extensions to TS CTs. Permanent changes to the fixed TS CTs are typically evaluated by using the three-tiered approach described in Chapter 16.1 of the SRP, RG 1.177 and RG 1.174. This approach addresses the calculated change in risk as measured by the change in core damage frequency (CDF) and large early release frequency (LERF), as well as the incremental conditional core damage probability and incremental conditional large early release probability; the use of compensatory measures to reduce risk; and the implementation of a configuration risk management program (CRMP) to identify risk-significant plant configurations.

The NRC staff evaluated the licensees processes and methodologies for determining that the change in risk from implementation of RICTs will be small and consistent with the intent of the Commissions Safety Goal Policy Statement. In addition, the NRC staff evaluated the licensees proposed changes against the three-tiered approach in RG 1.177 for the licensees evaluation of the risk associated with a proposed TS CT change. The results of the staffs review are discussed below.

Tier 1: PRA Capability and Insights The first tier evaluates the impact of the proposed changes on plant operational risk. The Tier 1 review involves two aspects: (1) scope and acceptability of the PRA models and their application to the proposed changes, and (2) a review of the PRA results and insights described in the licensees application., Information Supporting PRA Consistency with Regulatory Guide 1.200, Revision 2, and Enclosure 4, Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models, of the LAR identified the following modeled hazards and alternate methodologies that the licensee proposed to be used in the Seabrook RICT program to assess the risk contribution for extending the CT of a TS LCO.

Internal Events PRA model (includes internal floods)

Internal Fire Events PRA model Seismic Hazard: CDF penalty of 2.4 x 10--5 per year and a LERF penalty of 6.2 x 10--6 per year Other External Hazards: screened out from RICT program based on Appendix 6-A of the ASME/ANS RA-Sa-2009 PRA Standard (Reference [20])

Evaluation of Modeled PRAs For the modeled PRAs, which includes the internal events PRA (IEPRA), internal flooding PRA, and fire PRA (FPRA), as mentioned in Enclosure 2 of the LAR, the licensee confirmed that the PRA models had been peer reviewed, consistent with RG 1.200, Revision 2. For the open F&Os resulting from these peer reviews the licensee stated that closure of the F&Os was performed using an independent assessment process. The NRC staff confirmed that the licensee performed closure of the F&Os consistent with NEI 17-07 (Reference [21]). The NRC evaluated the remaining open F&Os, along with the dispositions for impact on this application, as listed in Table E2-1 of Enclosure 2 of the LAR. In response to RAI-08, the licensee confirmed that the two open F&Os associated with the IEPRA and FPRA had been closed using an independent closure assessment. Therefore, the implementation item proposed in Attachment 6 of the LAR associated with the finding level F&Os, LE-D6-01 and 03-002 (SR HR-E1) for the PRA models is no longer needed. With regards to the remaining open F&Os listed in Table E2-1, the NRC finds that the licensees disposition for impact on this application is appropriate. of the LAR provided a list of the key assumptions and sources of uncertainty, along with treatment for the application of TSTF-505. In response to RAI-03, the licensee provided a description for how the Seabrook PRA sources of uncertainty were evaluated as potential key sources of uncertainty for the TSTF-505 application and also provided the results of sensitivity studies that determined the impact for each associated source of uncertainty.

The NRC staff reviewed the PRA models peer review history provided by the licensee in of the LAR, as supplemented. The NRC staff notes that there were self-assessments performed to specifically address the gaps from use of the historical standard to the use of ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2 for both the internal events and internal flood PRA models, as confirmed by the licensees response to RAI-7, part vi and the NRCs prior approval of their surveillance frequency control program (Reference [22]). Thus, the licensee adequately applied the guidance for establishing PRA acceptability for the aforementioned models. The NRC staff further considered the key assumptions and key sources of uncertainty identified by the licensee and the licensees proposed use of surrogates in the PRA models for specific TS functions. Furthermore, in the LAR, Section 1.0, NextEra confirmed that there were no portable FLEX mitigating strategies incorporated into the Seabrook PRA models for use in performing a RICT calculation. The NRC staff finds the Seabrook scope, and acceptability of the modeled internal events, internal flooding events, and fire events to be commensurate with the RICT application for use in the integrated decision-making process and consistent with NEI 06-09 as endorsed and RG 1.200.

Evaluation of Seismic Hazard The licensees approach for including the seismic risk contribution in the RICT calculation is to add a penalty seismic CDF and a penalty seismic LERF to each RICT calculation. The proposed seismic CDF estimate is based on using the plant-specific seismic hazard curves developed in response to the Near-Term Task Force (NTTF) recommendation 2.1 (Reference [23]), and a plant-level high confidence of low probability of failure (HCLPF) capacity of 0.27g referenced to peak ground acceleration (PGA). The uncertainty parameter for seismic capacity was represented by a composite beta factor (c) of 0.52. The plant-level HCLPF value of 0.27g was obtained from the median seismic capacity (Am) of 0.90g and c of 0.52 provided in the LAR. The calculated seismic CDF penalty is 2.4 x 10-5 per year. The NRC staff finds that the method to determine the baseline seismic CDF is acceptable because it is consistent with the approach used in NRC Generic Issue (GI)-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants (Reference [24]). The staff performed an independent convolution using the input parameters identified by the licensee to confirm the proposed seismic CDF penalty.

Concerning the proposed seismic LERF estimate, the licensee explained in the LAR that an estimate of the seismic LERF was obtained by convolving the estimated seismic CDF (as described above) with a limiting fragility for containment function, expressed in terms of a HCLPF of 0.59g PGA and c of 0.36. The containment function HCLPF value of 0.59g was obtained from the median seismic capacity (Am) of 1.37g and c of 0.36 provided in the LAR.

The calculated seismic LERF is 6.2 x 10-6 per year. The NRC staff finds the licensees approach to estimating the baseline seismic LERF acceptable because the use of a 0.59g PGA HCLPF as the limiting fragility for containment function is based on consideration of the seismic fragilities of the containment and associated components.

The licensee addressed the incremental risk associated with seismic-induced LOOP in its supplement. The seismic LOOP frequency across the entire hazard interval is 3.9 x 10-5 per year. The seismic LOOP frequency across the entire hazard interval is about 5.9 percent of the total internal events 24-hour non-recovered LOOP frequency of 6.6 x 10-4 per year. The NRC staff evaluated the licensees analysis and finds that it adequately addresses the impact of a seismically induced LOOP on risk and that the exclusion of the impact of a seismically induced LOOP on risk from the non-recovered LOOP frequency has an insignificant impact on the RICT calculations.

In its supplement (response to RAI-11), the licensee stated that it conducted seismic walkdowns in accordance with the post-Fukushima NTTF Recommendation 2.3 (Reference [25]). The licensee further noted that these walkdowns did not identify any adverse findings related to the containment function or seismic-induced spatial interactions that could affect the TSTF-505 application.

The NRC staff finds that, during RICTs for SSCs credited in the design basis to mitigate seismic events, the licensees proposed methodology captures the risk associated with seismically induced failures of redundant SSCs because such SSCs are assumed to be fully correlated. By assuming full correlation, the seismic risk for those RICTs will not increase if one of the redundant SSCs is unavailable because simultaneous failure of all redundant trains would be assumed in a seismic PRA. During RICTs for SSCs that are not credited in seismic events, the proposed methodology for considering seismic risk contributions is conservative because the seismically induced failure of such SSCs would not result in a risk increase associated with the plant configuration during the RICT but the seismic penalty is still included in the calculation.

During RICTs for SSCs that are credited in seismic events, the proposed methodology is acceptable for this application because the plant-level HCLPF value used for the RICT calculations provides a conservative estimate of HCLPF values for all the credited SSCs.

In summary, the NRC staff finds the licensees proposal to use a seismic CDF penalty of 2.4 x 10-6 per year and a seismic LERF penalty of 6.2 x 10-6 per year to be acceptable for the licensees RICT Program for Seabrook because (1) the licensee used the most current site-specific seismic hazard information; (2) the licensee determined a seismic CDF penalty based on a plant-level HCLPF value of 0.27g and a combined beta factor of 0.52, consistent with the information for Seabrook in the GI-199 evaluation; (3) the licensee determined a seismic LERF penalty based on a plant-specific limiting containment HCLPF value of 0.59g PGA and a combined beta factor of 0.36; and (4) adding baseline seismic risk to RICT calculations, which assumes fully correlated failures, is acceptable for this application.

Evaluation of Other External Hazards Besides the seismic hazard discussed above, the licensee confirmed that other external hazards for Seabrook have an insignificant contribution to the configuration risk and proposed that these hazards be screened out from the RICT Program. The licensee provided its assessment of other external hazard risk for the RICT Program in Table E4-8, Evaluation of Risks from External Hazards, of Enclosure 4 to the LAR. The hazards are screened per a plant-specific evaluation in accordance with GL 88-20, updated to use the criteria in ASME PRA Standard RA-Sa-2009.

The staff identified a discrepancy in the notation of the screening criteria for external hazards between Section 2.0, Technical Approach, and Table E4-9, Progressive Screening Approach for Addressing External Hazards, in Enclosure 4 of the LAR. In its supplement (response to RAI-12), the licensee acknowledged this discrepancy and revised the text in Section 2.0 to ensure consistency with the screening criteria presented in table E4-9.

The NRC staff reviewed the information in the LAR and finds that the risk from other external hazards has an insignificant contribution to the configuration risk, and the other external hazards can be screened from RICT calculations because they either do not challenge the plant or they are bounded by the external hazards analyzed for the plant. The NRC staff notes that the preliminary screening criteria and progressive screening criteria used and presented in table E4-9 of Enclosure 4 to the LAR are the same criteria that are presented in supporting requirements for screening external hazards EXT-B1, EXT-B2, and EXT-C1 of the ASME/ANS RA-Sa-2009 PRA Standard.

Application of PRA Models, Results and Insights in the RICT Program The Seabrook base PRA models that have been determined to be acceptable in this SE will be modified as an application-specific PRA model (i.e., Configuration Risk Assessment and Management Program-CRMP tool), that will be used to analyze the risk for an extended CT.

The CRMP model produces results (i.e., risk metrics) that are consistent with the NEI 06-09-A guidance. In the LAR, the licensee provided all information needed to support the requested LCO actions proposed for the Seabrook RICT program consistent with all the Limitations and Conditions prescribed in Section 4.0 of NEI 06-09-A.

The NRC staff did not identify any insufficiencies in the licensees information or the CRMP tool as described in Enclosure 8, Attributes of the Real-Time Risk Model, to the LAR. Furthermore, as stated in Attachment 1 to the LAR, the proposed changes do not change the design, configuration, or method of operation of the plant. The proposed changes do not involve a physical alteration of the plant (no new or different kinds of equipment will be installed). The NRC staff finds that the Seabrook PRA models and CRMP tool used reflects the as-built, as-operated plant consistent with RG 1.200, Revision 2, for the Integrated Evaluation PRA-IEPRA (includes internal floods) and FPRA for ensuring PRA acceptability is maintained.

Therefore, the NRC staff concludes that the proposed application of Seabrook RICT Program is appropriate for use in the adoption of TSTF-505 for performing RICT calculations.

The licensee provided in Enclosure 5, Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), to the LAR, as supplemented, the estimated mean total CDF and LERF of the base PRA models to demonstrate that Seabrook meets the 1E-4/year CDF and 1E-5/year LERF criteria of RG 1.174 consistent with the guidance in NEI 06-09-A, and that these guidelines are satisfied for implementation of a RICT.

The licensee has incorporated NEI 06-09-A into TS 6.7.6.p. The estimated current mean total CDF and LERF for Seabrook PRAs meet the RG 1.174, Revision 3 guidelines, therefore, the NRC staff concludes the PRA results and insights to be used by the licensee in the RICT program will continue to be consistent with NEI 06-09-A.

Based on the above conclusions the NRC staff finds that the licensee has satisfied the intent of Tier 1 in RG 1.177 and RG 1.174 for determining the PRA acceptable, and that the scope of the PRA models (i.e., internal events, internal flooding, and fire), seismic methodology and other external hazards is appropriate for this application.

Tier 2: Avoidance of Risk-Significant Plant Configurations As described in RG 1.177, 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. The limits established for entry into a RICT and for RMA implementation are coordinated with the NEI guidance of Nuclear Management and Resources Council (NUMARC) 93-01, Revision 4F (Reference [26]), endorsed by RG 1.160, Revision 4 (Reference [27]), as applicable to plant maintenance activities. In response to RAI-04, the licensee confirmed that the Seabrook Maintenance Rule monitoring program incorporates the guidance of NUMARC 93-01, as endorsed by RG 1.160, specifically the quantitative performance measures.

The NRC staff concludes that the Tier 2 attributes of the proposed RICT Program, including limits established for entry into a RICT and implementation of RMAs, are consistent with NEI 06-09-A. Therefore, the proposed changes are consistent with the intent of Tier 2 in RG 1.177.

Tier 3: Risk-Informed Configuration Risk Management The third tier stipulates that a licensee should develop a program that ensures that the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. The proposed RICT program establishes a CRMP based on the underlying PRA models. The CRMP is then used to evaluate configuration-specific risk for planned activities associated with the RMTS extended CT, as well as emergent conditions which may arise during an extended CT This required assessment of configuration risk, along with the implementation of compensatory measures and RMAs, is consistent with the principle of Tier 3 for assessing and managing the risk impact of out-of-service equipment.

Paragraph 50.36(c)(5) of 10 CFR identifies administrative controls as the provisions relating to organization and management, procedures, [thereby] assuring operation of the facility in a safe manner. In Enclosure 8 of the submittal, Attributes of the Real-Time Risk Model, the licensee confirmed that future changes made to the baseline PRA models and changes made to the online model (i.e., CRMP) are controlled and documented by plant procedures. 0 of the LAR, provided the attributes that the licensees RICT program procedures will address, which are consistent with NEI TR 06-09-A. The NRC staff finds that the licensee has identified appropriate administrative controls consistent with NEI 06-09-A and 10 CFR 50.36(c)(5).

Based on the licensees incorporation of NEI 06-09-A in the TS, as discussed in LAR and its use of RMAs as discussed in LAR Enclosure 12, and because the proposed changes are consistent with the Tier 3 guidance of RG 1.177, the NRC staff finds the licensees Tier 3 program is acceptable and supports the proposed implementation of the RICT program.

Key Principle 4: Conclusions The licensee has demonstrated the technical acceptability and scope of its PRA models and alternative methods, including consideration of the impact of seismic events, extreme winds and tornado hazards, and other external hazards, and that the models can support implementation of the RICT program for determining extensions to CTs. The licensee has made proper consideration of key assumptions and sources of uncertainty. The risk metrics are consistent with the approved methodology of NEI 06-09-A and the acceptance guidance in RG 1.177 and RG 1.174. The RICT program will be controlled administratively through plant procedures and training and follows the NRC-approved methodology in NEI 06-09-A. The NRC staff concludes that the RICT program satisfies the fourth key principle of RG 1.177 and is, therefore, acceptable.

Key Principle 5: Performance Measurement Strategies - Implementation and Monitoring RG 1.177 and RG 1.174 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 degradation occurs due to unanticipated degradation or common cause mechanisms.

In their response, the licensee clarifies how they plan on handling potential common cause failures as a result of an emergent condition. In Enclosure 11 to the LAR, the licensee states, the SSCs in the scope of the RICT program are also in the scope of 10 CFR 50.65 for the Maintenance Rule. The Maintenance Rule monitoring programs will provide for evaluation and disposition of unavailability impacts which may be incurred from implementation of the RICT program. Furthermore, in Enclosure 11 of the LAR, the licensee confirmed that the cumulative risk is calculated at least every refueling cycle, but the recalculation period does not exceed 24 months, which is consistent with NEI 06-09-A.

In response to RAI-05, the licensee provided the criteria that would result in an unscheduled PRA model update. The criteria is related to an overall change in risk (CDF or LERF),

exceedance of overall CDF or LERF risk value, changes in calculated RICT, and significant changes of SSC risk metrics. The NRC staff determined that the Seabrook criteria are consistent with previously approved criteria and are acceptable for this application.

The NRC staff concludes that the RICT program satisfies the fifth key principle of RG 1.177 and RG 1.174 because: (1) the RICT program will monitor the average annual cumulative risk increase as described in NEI 06-09-A, thereby ensuring the program, as implemented, continues to meet RG 1.174 guidance for small risk increases; and (2) all affected SSCs are within the Maintenance Rule program, which is used to monitor changes to the reliability and availability of these SSCs.

4.0 CONCLUSION

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

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the NRC staff notified the Commonwealth of Massachusetts and State of New Hampshire officials on December 3, 2025, of the proposed issuance of the amendment. The State officials had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 amendments involve no significant hazards consideration, as published in Federal Register on May 13, 2025 (90 FR 205218), and there has been no public comment on such findings. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

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

8.0 References

[1]

Mack, K. A., NextEra Energy Seabrook, LLC to U.S. Nuclear Regulatory Commission, "License Amendment Request to Revise Technical Specifications to Adopt Risk-Inofrmed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF-Initiatie 4b," ADAMS Accession No. ML25034A143, February 3, 2025.

[2]

Mack, K. A., NextEra Energy Seabrook, LLC to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding RITSTF Initiative 4b, and 10 CFR 50.69,," ADAMS Accession No. ML25251A063, September 5, 2025.

[3]

U.S. Nuclear Regulatory Commission, "TSTF-505, Revision 2, TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, Provide Risk-Informed Extended Completion Times and Submittal of TSTF-505, Revision 2," July 2, 2018 (ADAMS Package Accession No. ML18183A493), dated July 2, 2018 (ADAMS Package Accession No. ML18183A493).

[4]

U.S. Nuclear Regulatory Commission, "Final Revised Model Safety Evaluation of Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B," dated November 21, 2018 (ADAMS Package Accession No. ML18269A041).

[5]

Sreenivas, V., U.S. Nuclear Regulatory Commission to Florida Power & Light Company, "Seabrook Station, Uni 1 - Regulatory Audit Plan in Support of LAR to Revise Technical Specifications to Adopt Risk-Informed Completion Times (EPID L-2025-LLA-0025),"

ADAMS Accession No. ML25219A763, March 28, 2025.

[6]

EMAIL CAPTURE, FROM: U.S. Nuclear Regulaory Commission, "

SUBJECT:

Seabrook Station, Unit 1 - Request for Additional Information re: License Amendement Request to Adopt TSTF-505 and Provisions of 10 CFR 50.69 (EPID L-2025-LLA-0025, L-2025-LLA-0058)," ADAMS Accession No. ML25219A763, August 7, 2025.

[7]

U.S. Nuclear Regulatory Commission, RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (ADAMS Accession No. ML090410014), March 2009.

[8]

U.S. Nuclear Regulatory Commission, RG 1.200, Revision 3, "Acceptability of Probabilistic Risk Asssessment Results for Risk-Informed Activities," ADAMS Accession No. ML20238B871, dated December 2020.

[9]

U.S. Nuclear Regulatory Commission, RG 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," ML023240437, November 2002.

[10] U.S. Nuclear Regulatory Commission, RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," ML100910006, May 2011.

[11] U.S. Nuclear Regulatory Commission, RG 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," ML17317A256, January 2018.

[12] U.S. Nuclear Regulatory Commission, RG 1.177, Revision 0, "Plant Specific, Risk-Informed Decisionmaking: Tehcnical Specifications," ADAMS Accession No. ML003740176, August 1998.

[13] U.S. Nuclear Regulatory Commission, RG 1.177, Revision 2, "Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," ADAMS Accesssion No.

ML20164A034, Janauary 2021.

[14] U.S. Nuclear Regulatory Commission, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking Final Report,," March 2017 (ADAMS Accession No. ML17062A466).

[15] NUREG-0800, Chapter 16, Section 16,1, Risk-informed Decision Making: Technical Specifications, Issued March 2007, ADAMS Accession No. ML07038228.

[16] NUREG-0800, Chapter 19, Section 19.1, DETERMINING THE TECHNICAL ADEQUACY OF PROBABILISTIC RISK ASSESSMENT FOR RISK-INFORMED LICENSE AMENDMENT REQUESTS, Issued September 2012, ADAMS Accession No. ML12193A107.

[17] Bradley, B., NEI, letter to S. D. Stuchell, NRC, "NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b; Risk Managed Technical Specifications (RMTS) Guidelines, Revision 0-A," ML122860402, October 2012.

[18] BGolder, J. M., NRC, letter to B. Bradley, NEI, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specification 4b, Risk-Managed Technical Specifications (RMTS) Guidlines'," ML071200238, May 17, 2007.

[19] U.S. Nuclear Regulatory Commission, "Updated Final safety Analysis Report (UFSAR),

Revision 23," ADAMS Accession No. ML25132A063, dated May 12, 2025.

[20] American Society of Mechnical Engineers and American Nuclear Society, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RA-S-2008, PRA Standard ASME/ANS RA-Sa-2009, dated February 2009, New York, NY (Copyright).

[21] Nuclear Energ Institute, NEI 17-07, Revision 2, "Performance of PRA Peer Reviews using the ASME/ANS PRA Standard," ADAMS Accession No. ML19231A182, August 2019.

[22] Lamb, J.G., U.S. Nuclear Regulatory Commission, letter to Site Vice President, NextEra Energy Seabrook, LLC, "Seabrook Station, Unit No. 1 - Issuance of Amendment Regarding the Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program (TAC No. MF1958),"

ML13212A069, July 24, 2014.

[23] NextEra, SBK-L-14052, Seismic Hazard and Screening Report (CEUS Sites), "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," ADAMS Accession No. ML14092A413, March 2014.

[24] Hiland, P., U.S. Nuclear Regulatory Commission letter to Sheron, B.W., "Safety/Risk Assessment Results for GI 199, Implications of Updated Probabilistic Seismic Hazard Estiates in CEUS on Existing Plants," ADAMS Accession No. ML100270582, September 2010.

[25] NextEra (Seabrook) Letter SBK-L-12242, "Response to 10 CFR 50.54(f) Request for Information Regarding Near-Term Task Force Recommendation 2.3, Seismic,"

ML12340A487, November 26, 2012.

[26] NEI, NUMARC 93-01, Revision F, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," ADAMS Accession No. ML18120A069, April 2018.

[27] U.S. Nuclear Regulatory Commission, RG 1.160, Revision 4, "Monitoring the Effectiveness of Maintenance aat Nuclear Power Plants," August 2018.

[28] Mack, K. A., NextEra Energy Seabrook, LLC to U.S. Nuclear Regulatory Commission, "Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of SSCs for Nuclear Power Reactors," March 21, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25080A172).

[29] Mack, K. A., NextEra Energy Seabrook, LLC to U.S. Nuclear Regulatory Commission, "Response to Requests for Additional Information Regarding Seabrook LAR to Adopt TSTF-505, Revision 2 and 10 CFR 50.69," September 5, 2025 (ML25251A063).

Principal Contributors: A Schwab, NRR A Brown, NRR A Foli, NRR C Moulton, NRR D Scully, NRR E Kleeh, NRR E Coffman, NRR H Wagage, NRR J English, NRR J Robinson, NRR J Ambrosini, NRR K West, NRR M Li, NRR S Alferink, NRR S Park, NRR K Hsu, NRR Date: February 18, 2026

ML25338A240 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/APLB/BC NRR/DRA/APLC/BC(A)

NAME VSreenivas KEntz EDavidson RPascarelli DATE 12/02/2025 12/30/2025 1/15/2026 1/13/2026 OFFICE NRR/DRA/APLA/BC(A)

NRR/DEX/EICB/BC(A)

NRR/DEX/EMIB/BC NRR/DSS/SCPB/BC NAME MGonzalez SDarbali SBailey MValentin DATE 1/16/2026 1/15/2026 1/15/2026 1/15/2026 OFFICE NRR/DEX/EEEB/BC NRR/DSS/SNSB/BC NRR/DSS/STSB/BC(A)

NRR/DORL/LPL1/BC(A)

NAME WMorton ITsang SMehta UShoop DATE 1/20/2026 1/15/2026 1/15/2026 2/18/2026 OFFICE NRR/DORL/LPL1/PM NAME VSreenivas DATE 2/18/2026

v t e Letter Sequence Approval
EPID:L-2025-LLA-0025, TSTF505 (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions 7 August 2025 Answers 5 September 2025 Results Approval Other: ML25057A434, ML25114A113, ML25171A122, ML25322A260
MONTHYEAR2025-03-28 [Table View]

Person / Time
ML25338A240
Site:	Seabrook (NPF-086)
Issue date:	02/18/2026
From:	V Sreenivas Plant Licensing Branch 1
To:	Coffey R Florida Power & Light Co
Lantigua, R, NRR/DORL/LPLI,
References
EPID L-2025-LLA-0025
Download: ML25338A240 (0)
v • d • e

ML25338A240

Text

SUBJECT:

Dear Mr. Coffey:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

SUBJECT:

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