ML18247A538

Issuance of Amendment No. 158 Removing Requirement to Perform Certain Surveillance Requirements During Shutdown and Changes to Administrative Technical Specifications

ML18247A538
Person / Time
Site:	Seabrook
Issue date:	11/27/2018
From:	Justin Poole Plant Licensing Branch 1
To:	Nazar M NextEra Energy Seabrook
Poole J, NRR/DORL/LPLI, 415-2048
References
EPID L-2017-LLA-0407
Download: ML18247A538 (33)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 27, 2018 Mr. Mano Nazar President and Chief Nuclear Officer Nuclear Division NextEra Energy Seabrook, LLC Mail Stop: EX/JB 700 Universe Blvd.

Juno Beach, FL 33408

SUBJECT:

SEABROOK STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 158 RE: REMOVING REQUIREMENT TO PERFORM CERTAIN SURVEILLANCE REQUIREMENTS DURING SHUTDOWN AND CHANGES TO ADMINISTRATIVE TECHNICAL SPECIFICATIONS (EPID L-2017-LLA-0407)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 158 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 1, 2017.

The amendment revises certain 18-month TS surveillance requirements to eliminate the condition that testing be conducted "during shutdown" and revises the administrative portion of the TSs regarding plant staff and responsibilities.

A copy of our related safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

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

Enclosures:

1. Amendment No. 158 to NPF-86

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY SEABROOK. LLC. ET AL.*

DOCKET NO. 50-443 SEABROOK STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 158 License No. NPF-86

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

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

(the licensee), dated December 1, 2017, 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.

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

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 158, and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook. LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION J<] G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: November 27, 2018

ATTACHMENT TO LICENSE AMENDMENT NO. 158 SEABROOK STATION, UNIT NO. 1 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages, as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Remove Insert 3/4 5-6 3/4 5-6 3/4 6-14 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 7-4 3/4 7-4 3/4 7-12 3/4 7-12 3/4 7-13A 3/4 7-13A 3/4 8-8 3/4 8-8 6-1 6-1 6-2 6-2 6-3 6-3 6-4 6-4 6-23 6-23 6-24 6-24 6-25 6-25

(4) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 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 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; (6) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 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; and (7) DELETED C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's 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; 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, as revised through Amendment No. 158, and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) License Transfer to FPL Energy Seabrook, LLC**

a. On the closing date(s) of the transfer of any ownership interests in Seabrook Station covered by the Order approving the transfer, FPL Energy Seabrook, LLC**, shall obtain from each respective transferring owner all of the accumulated decommissioning trust funds for the facility, and ensure the deposit of such funds and additional funds, if necessary, into a decommissioning trust or trusts for Seabrook Station established by FPL Energy Seabrook, LLC**, such that the amount of such funds deposited meets or exceeds the amount required under 10 CFR 50. 75 with respect to the interest in Seabrook Station FPL Energy Seabrook, LLC**,

acquires on such dates(s).

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

AMENDMENT NO. 158

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS-Tava GREATER THAN OR EQUAL TO 350°F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued)

d. In accordance with the Surveillance Frequeocy Control Program by:

1) Verifying automatic interlock action of the RHR system from the Reactor Coolant System to ensure that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 440 psig, the interlocks prevent the valves from being opened.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. 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 (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.

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

1) Centrifugal charging pump;

2) Safety Injection pump; and

3) RHR pump.

SEABROOK - UNIT 1 3/4 5-6 Amendment No. aa, 74, 83, 141, 164, 158

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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, 1<11, 1<1<1, 154, 158

CONTAINMENT SYSTEMS DEPRESSURIZATION AND COOLING SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:

a. A spray additive tank containing a volume of between 9420 and 9650 gallons of between 19 and 21 % by weight NaOH solution, and

b. Two gravity feed paths each capable of adding NaOH solution from the chemical additive tank to the Refueling Water Storage Tank.

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

ACTION:

With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by 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;

b. In accordance with the Surveillance Frequency Control Program by:

1) Verifying the contained solution volume in the tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis.

c. In accordance with the Surveillance Frequency Control Program by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal.

SEABROOK - UNIT 1 3/4 6-15 Amendment No. 4Mi- 158

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:

With one or more of the isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve(s) to OPERABl-E status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange; or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

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

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

PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position;

2) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and

3) Verifying that valves FW-156 and FW-163 are OPERABLE for alignment of the startup feedwater pump to the emergency feedwater header.

b. *1n accordance with the Surveillance Frequency Control Program by verifying the following pumps. develop the required discharge pressure and flow as specified in the Technical Requirements Manual:

1) The motor-driven emergency feedwater pump;

2) The steam turbine-driven emergency feedwater pump when the secondary steam supply pressure is greater than 500 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) The startup feedwater pump.

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 upon receipt of an Emergency Feedwater System Actuation test signal;

2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal; SEABROOK - UNIT 1 3/4 7-4 Amendment No. 30, 90,114,141, 158

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:

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM/ULTIMATE HEAT SINK SURVEILLANCE REQUIREMEN'1"S

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,

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 70°F.

• 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 7-13A Amendment No. 32, 118, 141, 158

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued}

14) Simulating a Tower Actuation (TA} signal while the diesel generator is loaded with the permanently connected loads and auto-connected emergency (accident) loads, and verifying that the service water pump automatically trips, and that the cooling tower pump automatically starts. After energization the steady state voltage and frequency of the emergency buses shall be maintained at 4160 +/- 420 volts and 60 +/- 1.2 Hz;and

15) While diesel generator 1A is loaded with the permanently connected loads and auto-connected emergency (accident} loads, manually connect the 1500 hp startup feedwater pump to 4160-volt bus ES. After energization the steady-state voltage and frequency of the emergency bus shall be maintained at 4160 +/- 420 volts and 60 +/- 1.2 Hz.

g. In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously from standby condition and verifying that both diesel generators achieve:

1) A generator voltage and frequency greater than or equal to 3740 volts and 58.8 Hz within 10 seconds after the start signal, and

2) A steady-state generator voltage and frequency of 4160 +/- 420 volts and 60 +/- 1.2 Hz.

SEABROOK - UNIT 1 3/4 8-8 Amendment No. 13, 38, 54, 8Q, 141, 158

6.0 . ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall station operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Manager (SM) shall be responsible for the control room command function. During any absence of the SM from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the SM from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions for departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the FSAR and updated in accordance with the requirements of 10 CFR 50.71.

b. The plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence ftom operating pressures.

SEABROOK - UNIT 1 6-1 Amendment No. 55, 88, 104, 158

6.0 ADMINISTRATIVE CONTROLS 6.2.2 STATION STAFF

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4.

b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 6.2.2.a and 6.2.2.d for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

c. A radiation protection technician shall be on site when fuel is in the reactor.

The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

d. An individual (Shift Technical Advisor (STA)) shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. The STA position shall be manned in MODES 1, 2, 3, and 4 unless the SM or the individual with a Senior Operator license meets the qualifications for the STA.

e. While the unit is in MODE 1, 2, 3 or 4, a licensed senior operator, either the SM or SRO, shall be on shift having had at least 6 months of hot operating experience.

f. The Operations Manager shall meet one of the following:

1. Hold a senior operator license,

2. Have held a senior operator license on a similar unit (PWR), or

3. Have been certified for equivalent senior operator knowledge.

g. The Assist_ant Operations Manager shall hold a senior reactor operator license.

SEABROOK - UNIT 1 6-2 Amendment No. ;;w, 121 124, 158

TABLE 6.2-1 DELETED SEABROOK - UNIT 1 6-3 Amendment No. 4.Q4, 158

ADMINISTRATIVE CONTROLS 6.2.3 ITHIS SPECIFICATION NUMBER IS NOT USED) 6.2.4 (THIS SPECIFICATION NUMBER IS NOT USED) 6.3 (THIS SPECIFICATION NUMBER IS NOT USED}

6.4 (THIS SPECIFICATION NUMBER IS NOT USED}

6.5 ITHIS SPECIFICATION NUMBER IS NOT USED) 6.6 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT 1 6-4 Amendment No. 55, 104, 113, 118, 158

ADMINISTRATIVE CONTROLS HIGH RADIATION AREA 6.11.2 (Continued)

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located ~ithin large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained as required by the Operr;1tional Quality Assurance Program (OQAP). This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b. Shall become effective after review and acceptance by the Onsite Review Group and approval of the plant manager.

6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Control Program (OQAP). This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after review and acceptance by the Onsite Review Group and the approval of the plant manager.

SEABROOK - UNIT I 6-23 Amendment No. 22, ee, 104, 107, 115, 158

ADMINISTRATIVE CONTROLS OFFSITE DOSE CALCULATION MANUAL (ODCM)

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and each affected page shalt indicate the revision number the change was implemented.

6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS*

6.14.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid): *

a. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Onsite Review Group. The discussion of each change shall contain:

1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;

5) An evaluation of the change, which shows the expected maximum.

exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

• Licensees may choose to submit the information called for in this Specification as part of the FSAR update, pursuant to 10 CFR 50. 71.

SEABROOK - UNIT 1 6-24 Amendment No. 44&,

158

ADMINISTRATIVE CONTROLS 6.14.1 (Continued)

6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; and *

8) Documentation of the fact that the change was reviewed and found acceptable by the Onsite Review Group.

b. Shall become effective upon review and acceptance by the Onsite Review Group.

6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option 8, as modified by approved exemptions. This program shall be in accordance with Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," and conditions and limitations specified in NEI 94-01, Revision 2-A.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.15% of primary containment air weight per day.

The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

Containment leakage rate acceptance criterion is ~ 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are~ 0.60 La for the Type Band Type C tests and~ 0.75 La for Type A tests.

SEABROOK - UNIT 1 6-25 Amendment No. 115, 153, 158

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 158 TO FACILITY OPERATING LICENSE NO. NPF-86 NEXTERA ENERGY SEABROOK, LLC SEABROOK STATION, UNIT NO. 1 DOCKET NO. 50-443

1.0 INTRODUCTION

By letter dated December 1, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17339A428), NextEra Energy Seabrook, LLC (NextEra or the licensee) submitted License Amendment Request (LAR) 17-04, requesting changes to the Technical Specifications (TSs) for Seabrook Station, Unit No. 1 (Seabrook). Specifically, the licensee proposes to revise certain 18-month TS surveillance requirements (SRs) to eliminate the condition that testing be conducted "during shutdown" and revise the administrative TSs regarding plant staff and responsibilities.

2.0 REGULATORY EVALUATION

2.1 System Descriptions The systems and components associated with the effected SRs are described in Section 2.1, "System Design and Operation," of the LAR, as follows:

Emergency Core Cooling System (ECCS)

The ECCS consists of the centrifugal charging pumps, safety injection pumps, a refueling water storage tank (RWST), the residual heat removal pumps, the residual heat removal heat exchangers, the safety injection accumulators, and the associated valves and piping. The ECCS is comprised of two identical trains, each train independent of the other and fully redundant. The primary function of the ECCS following an accident is to remove the stored and fission product decay heat from the reactor core so that fuel rod damage, to the extent that it would impair effective cooling of the core, is prevented. The ECCS is designed to cool the reactor core as well as to provide additional shutdown capability following a loss of coolant accident, control rod ejection accident, steam or feedwater system break, or a steam generator tube rupture.

Enclosure 2

Containment Spray System (CBS)

The CBS system is designed to remove the energy discharged to the containment following a loss-of-coolant accident or main steam line break to prevent the containment pressure from exceeding design pressure and to reduce and maintain containment temperature and pressure within acceptable limits.

The CBS system is actuated by high pressure in the containment. The CBS system is comprised of two identical trains, each train independent of the other and fully redundant.

Spray Additive Tank (SAT)

The spray additive tank (SAT) is mounted adjacent to the RWST, and drains by gravity into the RWST mixing chamber. The SAT provides the correct amount of sodium hydroxide solution to insure that the final containment recirculation sump pH after injection will be between 8.5 and 11.0 units for the various reactor coolant conditions.

Containment Isolation Valves The Containment Isolation System is comprised of the valves, piping and actuators required to isolate the containment following a LOCA or steam line rupture. Containment isolation valve closure speeds and leak tightness will prevent radiological effects from exceeding the guidelines established by 10 CFR [Title 1O of the Code of Federal Regulations Part] 100.

Emergency Feedwater (EFW) System The EFW system provides the capability to remove heat from the reactor coolant system during emergency conditions when the main feedwater system is not available. The system is comprised of two full-sized pumps, one motor-driven and one turbine-driven.

Primary Component Cooling (PCCW) Water System The PCCW system supplies flow to the safeguard components that are required for safe shutdown or to mitigate the consequences of an accident. The system consists of two independent and redundant flow loops.

Service Water (SW) I Ultimate Heat Sink (UHS)

The UHS employs two independent and redundant cooling loops. Each loop can be supplied by either of two full-capacity SW pumps (four pumps total) drawing water from the Atlantic Ocean or alternatively, each loop can be supplied by a full-capacity cooling tower pump (two pumps total) drawing water from a mechanical draft cooling tower.

Diesel Generators (DG)

Two redundant diesel generators are provided to automatically connect to the two trains of redundant emergency buses when a loss of all offsite power

sources occurs. Each emergency bus and associated load group has sufficient redundancy to assure that the safety functions are performed.

2.2 Proposed TS Changes The licensee proposed changes to SRs 4.5.2.e, 4.6.2.1.c, 4.6.2.2.c, 4.6.3.2, 4.7.1.2.1.c, 4.7.3.b, 4.7.4.1.b, 4.7.4.2.b, and 4.8.1.1.2.g to remove any wording that required the SR to be performed while shut down. The licensee also proposed changes to TSs 6.1.1, 6.2.1.b, 6.2.1.c, 6.12.b, 6.13.b, 6.14.1.a, 6.14.1.a.8, and 6.14.1.b of the administrative section of the TSs to revise plant-specific titles. Finally, the licensee proposed changes to TSs 6.1.2 and 6.2.2 and to delete Table 6.2-1 and TS 6.2.4 to align with NUREG-1431, Volume 1, "Standard Technical Specifications - Westinghouse Plants: Specifications," Revision 4:

The following are the proposed changes to the TS SRs as described in Section 2.4, "Description of the Proposed Change," of the LAR:

1. SR 4.5.2.e:

Each ECCS subsystem shall be demonstrated OPERABLE:

In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and

2) Verify~ng that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.

2. SR 4.6.2.1.c:

Each Containment Spray System shall be demonstrated OPERABLE:

In accordance with the Surveillance Frequency Control Program during shutdown, 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.

3. SR 4.6.2.2.c:

The Spray Additive System shall be demonstrated OPERABLE:

In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal:

4. SR 4.6.3.2:

Each containment isolation valve shall be demonstrated OPERABLE during shutdown 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,

c. Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.

5. SR 4.7.1.2.1.c:

Each auxiliary feedwater pump shall be demonstrated OPERABLE:

In accordance with the Surveillance Frequency Control Program during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal;

2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal;

3) Verifying that with all manual actions, including power source and valve alignment, the startup feedwater pump starts within the required elapsed time; and

4) Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.

6. SR 4.7.3.b:

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

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

7. SR 4.7.4.1.b:

Each service water loop shall be demonstrated OPERABLE:

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

8. SR 4.7.4.2.b:

Each service water cooling tower loop shall be demonstrated OPERABLE:

In accordance with the Surveillance Frequency Control Program during shutdown, 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,

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.

9. SR 4.8.1.1.2.g:

In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously from standby condition, during shutdown, and verifying that both diesel generators achieve:

, 1) A generator voltage and frequency greater than or equal to 3740 volts and 58.8 Hz [Hertz] within 10 seconds after the start signal, and

2) A steady-state generator voltage and frequency of 4160 +/- 420 volts and 60 +/- 1.2 Hz.

Proposed changes to administrative controls Revise plant-specific titles

10. TS 6.1.1:

The Station DiFeotor plant manager shall be responsible for overall station operation and shall delegate in writing the succession to this responsibility during his absence.

11. TS 6.2.1.b:

The Station DiFeotor plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

12. TS 6.2.1.c:

The Site Vioe PFesident A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

13. TS 6.12:

Changes to the Process Control Program (PCP):

b. Shall become effective after review and acceptance by the SORG Onsite Review Group and approval of the Station DiFeotor plant manager.

14.TS6.13:

Changes to the Offsite Does Calculation Manual (ODCM):

b. Shall become effective after review and acceptance by the SORG Onsite Review Group and the approval of the Station DiFeotor plant manager.

15. TS 6.14:

Changes to the ODCM:

b. Shall become effective after review and acceptance by the SORG Onsite Review Group and the approval of the Station DiFeotor plant manager.

16. TS 6.1.2:

The Shift Manager (or during his allsenoe from the oontrol room, a designated individual) shall Ile responsible for the oontrol room oommand funotion. A management direotive to this effeot, signed by the Site Vioe President shall be reissued to all station peFSonnel on an annual basis. The Shift Manager (SM) shall be responsible for the control room command function. During any absence of the SM from the control room while the unit is in MODE 1, 2 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the SM from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

Delete TS 6.2.2.a through TS 6.2.2.e. and replace with the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4.

b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 6.2.2.a and 6.2.2.d for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

c. A radiation protection technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

d. An individual (Shift Technical Advisor (STA)) shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

The STA position shall be manned in MODES 1, 2, 3, and 4 unless the SM or the individual with a Senior Operator license meets ~he qualifications for the STA.

e. While the unit is in MODE 1, 2, 3 or 4, a licensed senior operator, either the SM or SRO, shall be on shift having had at least 6 months of hot operating experience.

Delete Table 6.2-1 Delete TS 6.2.4

2.3 Regulatory Requirements and Guidance The following are the regulatory requirements and guidance that the NRC staff considered in its review of the LAR.

Section 50.36, "Technical specifications," of 10 CFR establishes the regulatory requirements related to the content of TSs.

Section 50.36(c)(3) of 10 CFR states, in part, that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Section 50.36(c)(5) of 10 CFR states, in part, that the TSs will include provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

Section 50.54, "Conditions of licenses," of 10 CFR, in part, discusses the conditions in every nuclear power reactor operating license issued under 10 CFR Part 50. Paragraph 10 CFR 50.54(m) discusses reactor operators and senior reactor operators licensed under 10 CFR Part 55.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Revision 7, Chapter 13, "Conduct of Operations," Section 13.1.2 -

13.1.3, "Operating Organization," provides guidance for the review of the structure, functions, and responsibilities of the onsite organization established to safely operate and maintain the facility.

NUREG-1431, Volume 1, Revision 4, is the NRC guidance document for format and content of TSs for Westinghouse plants.

3.0 TECHNICAL EVALUATION

3.1 Removal of the "During Shutdown" Limitation The proposed changes to SRs 4.5.2.e, 4.6.2.1.c, 4.6.2.2.c, 4.6.3.2, 4.7.1.2.1.c, 4.7.3.b, 4.7.4.1.b, 4.7.4.2.b, and 4.8.1.1.2.g are shown in Section 2.2, Numbers 1 through 9, of this SE.

In the application, the licensee proposes removing the condition "during shutdown" to the SRs above. The licensee states that removing the condition that these SRs be performed "during shutdown" will eliminate the need to perform duplicate testing. The license states that many of the systems or components associated with these SRs have other SRs that require the system or component to be tested at power, but that test at power cannot be credited because of the "during shutdown" restriction. The licensee also states that the proposed changes are consistent with the standard found in NUREG-1431, since the SRs in NUREG-1431 that correspond to the Seabrook SRs above do not restrict performance to shutdown conditions.

The licensee also points out that removing the condition "during shutdown" does not prevent the SR from being performed during shutdown; rather, it provides the flexibility of it being done at power or shutdown. The licensee states that it will continue to evaluate the risk impact of performing SRs as required by 10 CFR 50.65(a)(4), and SRs previously performed during shutdown will be performed during operation only when it is safe to do so.

During its review, the NRC staff noted that the SRs in NUREG-1431 that correspond to the Seabrook SRs do not have a condition to be performed during shutdown. The NRC staff also notes that removal of the "during shutdown" condition will not change the test in such a way that it no longer provides assurance that the necessary quality of the systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The NRC staff notes that removing the condition "during shutdown" does not eliminate the licensee's requirement to meet 10 CFR 50.65(a)(4) for evaluating the risk impact prior to performance of the test, and that removing the condition does not prevent the licensee from performing the test during shutdown, if needed. Therefore, the NRC staff finds the proposed changes to the SRs above continue to meet the requirements of 10 CFR 50.36(c)(3) and are acceptable.

3.2 TS Section 6.1, "Responsibility" 3.2.1 TS 6.1.1 The proposed changes to TS 6.1.1 are shown in Section 2.2, Number 10, of this SE. The TS discusses the individual responsible for the overall station operation and the responsibility of that person during his or her absence. Specifically, the licensee proposes to delete the title "Station Director and replace it with "plant manager." The use of the term "plant manager" is consistent with its use in NUREG-1431, Volume 1, Revision 4, the Standard Technical Specifications for Westinghouse plants. The licensee considers this change administrative in nature, as it replaces a plant-specific title with a generic title. The NRC staff finds this to be acceptable, as the proposed title will not change or reassign the responsibility for overall station operation.

3.2.2 TS 6.1.2 The proposed changes to TS 6.1.2 are shown in Section 2.2, Number 16, of this SE. TS 6.1.2 discusses the control room command function. The proposed change specifies the designation of control to an individual with an active SRO license or a licensed Reactor Operator during various operating modes and deletes the annual reissuance of a management directive signed by the Site Vice President. The licensee states the management directive will restate the requirements of the control room command function in the new proposed language; therefore, the deletion will remove the redundancy. The proposed language will be consistent with Section 5.1.2, "Responsibility," of NUREG-1431, and will be consistent with the changes proposed to Section 6.2.2, "Station Staff," for the Seabrook TSs. The NRC staff has reviewed the change and finds it to be acceptable because the responsibility for the control room command function does not change, and furthermore, the change will allow it to be consistent with the standard TSs for Westinghouse plants.

3.3 TS 6.2, "Organization" 3.3.1 TS 6.2.1.b The proposed changes to TS 6.2.1.b are shown in Section 2.2, Number 11, of this SE. TS Section 6.2.1.b discusses the individual who is responsible for overall unit safe operation and his or her role. The licensee proposes to delete the title "Station Director and replace it with "plant manager." The use of the term "plant manager is consistent with the use in NUREG-1431, and the licensee finds this change to be administrative in nature, as it replaces a plant-specific title with a generic title. The NRC staff finds the proposed change to be

acceptable because the proposed title does not change or reassign the responsibility for overall station operation. Additionally, the replacement of the proposed term is reflected throughout the rest of the TSs.

3.3.2 TS 6.2.1.c The proposed changes to TS 6.2.1. b are shown in Section 2.2, Number 12, of this SE. TS Section 6.2.1.c discusses the individual who will have corporate responsibility for overall plant nuclear safety and his or her role. The licensee proposes to delete "Site Vice President" and replace it with "specified corporate officer." The term "specified corporate officer" is a term consistent with its use in NUREG-1431, and the licensee finds this change to be administrative in nature, as it replaces a plant-specific title with a generic title. The NRC staff finds this proposed change to be acceptable because the proposed title change does not change or reassign the designated individual's corporate responsibility.

3.3.3 TS 6.2.2 and Deletion of Table 6.2-1 The proposed changes to TS 6.2.2 and the deletion of Table 6.2-1 are shown in Section 2.2, Numbers 17 and 18, of this SE, respectively. Current TS 6.2.2 discusses plant staff and refers to Table 6.2-1, "Minimum Shift Crew Composition," which specifies the positon and the number of individuals required to fill the position based on the mode. The licensee proposes several changes to this section. Table 6.2-1 and its reference will be deleted because the proposed revision, Section 6.2.2.b, will include the reference to 10 CFR 50.54(m)(2)(i), which specifies the minimum requirements per shift for onsite staffing of nuclear power units by Operators and Senior Operators licensed under 10 CFR Part 55. The licensee also plans to revise the rest of the section with language similar to that in NUREG-1431, Section 5.2.2, "Unit Staff," with the exception of Section 6.2.2.d, which is discussed in Section 3.3.4 of this SE. The proposed language will not change the level of staffing or reduce responsibilities or changes to the technical qualifications for each position; therefore, the NRC staff finds the changes to be acceptable.

3.3.4 TS 6.2.4 The proposed changes to TS 6.2.4 are shown in Section 2.2, Number 19, of this SE. TS Section 6.2.4 discusses the STA function. The licensee proposes to delete this section and discuss the STA function in Section 6.2.2.d of the proposed TS. The proposed language in Section 6.2.2.d will not alter the role of the STA in providing advisory technical support and folds in the staffing requirement from the proposed deletion of Table 6.2-1. The NRC staff finds the proposed change to be acceptable because the change does not alter the requirements or function of an STA.

3.4 TS 6.12, "Process of Control Program (PCP)"

The proposed changes to TS 6.12 are shown in Section 2.2, Number 13, of this SE. TS 6.12.b discusses when changes to the PCP become effective. The current TS states that the PCP shall become effective after the review and acceptance by the SORC and approval of the Station Director. The licensee proposes to delete "SORC" and replace that title with "Onsite Review Group." This proposed title change to the review organization is considered an administrative change as it will not modify the composition or responsibilities of the review organization. In addition, the licensee proposes to delete "Station Director" and replace that positon tiUe with "plant manager." As discussed earlier in this SE, the use of the term "plant

manager is also acceptable because the proposed title does not change or reassign the responsibility for overall station operation. Therefore, the NRC staff finds the proposed changes to TS section 6.12.b to be acceptable as the titles will not change how and when the PCP becomes effective.

3.5 TS 6.13, "Offsite Dose Calculation Manual (ODCM}"

The proposed changes to TS 6.13 are shown in Section 2.2, Number 14, of this SE. TS 6.13.b discusses when changes to the ODCM becomes effective. The current TS states that the ODCM shall become effective after the review and acceptance by the Station Operation Review Committee (SORC) and the approval of the Station Director. The licensee proposes to delete "SORC" and replace that title with "Onsite Review Group." This proposed title change to the review organization is considered an administrative change, as it will not modify the composition or responsibilities of the review organization. In addition, the licensee proposes to delete "Station Director and replace that positon title with "plant manager." As previously discussed in this SE, the use of the term "plant manager is also acceptable because the proposed title does not change or reassign the responsibility for overall station operation. Therefore, the NRC staff finds the proposed changes to TS Section 6.13.b acceptable, as the titles will not change how and when the ODCM becomes effective.

3.6 TS 6.14, "Major Change to Liquid, Gaseous, and Solid Radwaste Treatment Systems*"

The proposed changes to TS 6.14 are shown in Section 2.2, Number 15, of this SE. TS 6.14.1 discusses the licensee-initiated major changes to the Radwaste Treatment Systems and responsibilities of the review organization. The licensee proposes to delete the references to "SORC" and replace that title with "Onsite Review Group." This proposed title change to the review organization is considered an administrative change, as it will not modify the composition or responsibilities of the review organization; therefore, the NRC staff finds this to be acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Hampshire State and Commonwealth of Massachusetts officials were notified of the proposed issuance of the amendment on September 24, 2018. The State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on February 13, 2018 (83 FR 6227). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22( c)(9).

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

6.0 CONCLUSION

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

Principal Contributors: D. Ki M. Hamm J. Poole Date: November 27, 2018

ML18247A538 *b memorandum **b e-mail OFFICE NRR/DORULPL 1/PM NRR/DORULPL 1/LA NRR/DRA/APHB/BC* NRR/DSS/STSB/BC**

NAME JPoole LRonewicz MKichlin_e for CFong CTilton for VCusumano DATE 10/17/2018 09/18/2018 08/08/2018 10/19/2018 OFFICE OGC - NLO** NRR/DORULPL 1/BC NRR/DORULPL1/PM NAME BHarris JDanna JPoole DATE 10/29/2018 11/15/2018 11/27/2018