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| {{#Wiki_filter:OFFICIAL USE ONLY PROPRIETARY INFORMATION March 10, 2021 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038 | | {{#Wiki_filter:}} |
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| ==SUBJECT:==
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| HOPE CREEK GENERATING STATION ISSUANCE OF AMENDMENT NO. 226 RE: REVISE EMERGENCY CORE COOLING SYSTEM TECHNICAL SPECIFICATION FOR HIGH PRESSURE COOLANT INJECTION SYSTEM INOPERABILITY (EPID L-2020-LLA-0131)
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| ==Dear Mr. Carr:==
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| The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 226 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station in response to your application dated June 15, 2020, as supplemented by {{letter dated|date=November 4, 2020|text=letter dated November 4, 2020}}.
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| The amendment revises Technical Specification 3/4.5.1, ECCS [Emergency Core Cooling System] - Operating, Limiting Condition for Operation 3.5.1, Action c, to clarify the entry conditions for the action and to add a new action to address the condition where the high pressure coolant injection system is inoperable, coincident with inoperability of a low pressure coolant injection subsystem and a core spray system subsystem.
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| The NRC staff has determined that the related safety evaluation contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, requests for withholding. The proprietary information is indicated by text enclosed with double brackets. The proprietary version of the safety evaluation is provided as . The NRC staff has also prepared a non-proprietary version of the safety evaluation, which is provided in Enclosure 3.
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| Enclosure 2 to this letter contains proprietary information. When separated from Enclosure 2, this document is DECONTROLLED.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY - PROPRIETARY INFORMATION E. Carr A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
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| Sincerely,
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| /RA/
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| James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354
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| ==Enclosures:==
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| : 1. Amendment No. 226 to Renewed License No. NPF-57
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| : 2. Safety Evaluation (Proprietary)
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| : 3. Safety Evaluation (Non-Proprietary) cc without Enclosure 2: Listserv OFFICIAL USE ONLY - PROPRIETARY INFORMATION
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| PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 Renewed License No. NPF-57
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| : 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
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| A. The application for amendment filed by PSEG Nuclear LLC dated June 15, 2020, as supplemented by {{letter dated|date=November 4, 2020|text=letter dated November 4, 2020}}, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
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| Enclosure 1
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| : 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-57 is hereby amended to read as follows:
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| (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| : 3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days.
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| FOR THE NUCLEAR REGULATORY COMMISSION
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| /RA/
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| James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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| ==Attachment:==
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| Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 10, 2021
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| ATTACHMENT TO LICENSE AMENDMENT NO. 226 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
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| Remove Insert 3 3 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
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| Remove Insert 3/4 5-3 3/4 5-3
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| reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) PSEG Nuclear 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) PSEG Nuclear 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) PSEG Nuclear 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. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.
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| (7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.
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| 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:
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| (1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3902 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
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| (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| Renewed License No. NPF-57 Amendment No. 226
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| EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION ACTION: (Continued)
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| : c. For the HPCI system:
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| : 1. With the HPCI system inoperable, provided the Core Spray System, the LPCI system, the ADS and the RCIC system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to < 200 psig within the following 24 hours.
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| : 2. With the HPCI system inoperable, provided the ADS and the RCIC system are OPERABLE:
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| : a. With either one LPCI subsystem or one CSS subsystem inoperable, restore the HPCI system to OPERABLE status within 72 hours or restore the LPCI subsystem/CSS subsystem to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to < 200 psig within the following 24 hours.
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| : b. With one LPCI subsystem and one CSS subsystem inoperable, restore the HPCI system to OPERABLE status within 8 hours or restore the LPCI subsystem or CSS subsystem to OPERABLE status within 8 hours or be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to < 200 psig within the following 24 hours.
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| : d. For the ADS:
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| : 1. With one of the above required ADS valves inoperable, provided the HPCI system, the core spray system and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 100 psig within the next 24 hours.
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| : 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure to 100 psig within the next 24 hours.
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| : e. With a CSS and/or LPCI header P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 7 days or determine the ECCS header P locally at least once per 12 hours; otherwise, declare the associated ECCS subsystem inoperable.
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| : f. The discharge line "keep filled" alarm instrumentation associated with a LPCI and/or CSS subsystem(s) may be in an inoperable status for up to 6 hours for required surveillance testing* provided that the "keep filled" alarm instrumentation associated with at least one LPCI or CSS subsystem serviced by the affected "keep filled" system remains OPERABLE; otherwise, perform Surveillance Requirement 4.5.1.a.1.a.
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| : g. In the event an ECCS system is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 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.
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| * This includes testing of the "Reactor Coolant System Interface Valves Leakage Pressure Monitors" associated with LPCI and CSS in accordance with Surveillance 4.4.3.2.3 HOPE CREEK 3/4 5-3 Amendment No. 226
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| ENCLOSURE 3 (NON-PROPRIETARY)
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| SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 226 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Proprietary information has been redacted from this document pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations.
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| Redacted information is identified by blank text enclosed within ((double brackets)).
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 226 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354
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| ==1.0 INTRODUCTION==
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| By {{letter dated|date=June 15, 2020|text=letter dated June 15, 2020}} (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20167A190), as supplemented by {{letter dated|date=November 4, 2020|text=letter dated November 4, 2020}} (ADAMS Accession No. ML20310A219), PSEG Nuclear LLC (PSEG, the licensee) submitted a license amendment request for the Hope Creek Generating Station (Hope Creek). The amendment would revise Technical Specification (TS) 3/4.5.1, ECCS [Emergency Core Cooling System] - Operating. Specifically, the proposed TS changes would add a new Action 2.b to TS Action 3.5.1.c allowing continued operation up to 8 hours for the condition where the high pressure coolant injection (HPCI) system is inoperable, coincident with an inoperable core spray system (CSS) subsystem and an inoperable low pressure coolant injection (LPCI) subsystem. Also, the proposed TS would rephrase the existing TS Action 3.5.1.c to improve clarity relative to entry conditions.
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| The supplemental {{letter dated|date=November 4, 2020|text=letter dated November 4, 2020}}, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 28, 2020 (85 FR 45448).
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| ==2.0 REGULATORY EVALUATION==
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| The requirements in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.46, Acceptance criterial for emergency core cooling systems for light-water nuclear power reactors, state, in part, that the ECCS must be designed such that an evaluation performed using an acceptable evaluation model (EM) demonstrates that acceptance criteria, set forth in 10 CFR 50.46(b), including peak cladding temperature, cladding oxidation, hydrogen generation, maintenance of coolable core geometry, and long-term core cooling are met for a variety of hypothetical loss-of-coolant accidents (LOCAs), including the most severe hypothetical LOCA.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION The requirements for TS are set forth in 10 CFR 50.36,Technical specifications. Specific categories of TS are provided in 10 CFR 50.36(c) and include limiting conditions for operation (LCOs). If an LCO is not met, the facility must be shut down, or other acceptable remedial action must be taken. The regulation at 10 CFR 50.36(c)(2), Limiting conditions for operations, states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
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| ==3.0 TECHNICAL EVALUATION==
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| The licensee proposed revisions to TS 3.5.1 for operation of the ECCS. Specifically, the changes would rephrase the Action statements, c.1 and c.2, and renumber c.2 as c.2.a to improve clarity relative to entry conditions.
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| Current TS 3.5.1.c.1 permits the HPCI system to be inoperable for up to 14 days provided that the CSS, LPCI, automatic depressurization system (ADS), and the reactor core isolation cooling (RCIC) systems are operable. The licensee would add a new Action statement c.2.b to TS 3.5.1. The proposed addition avoids an unnecessary entry into TS Action 3.0.3 (i.e.,
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| eliminates the need for immediate plant shutdown) in the event HPCI, and one of the other ECCS systems or RCIC are inoperable. Instead, it would allow HPCI to be inoperable for up to 8 hours with one CSS subsystem and one LPCI subsystem also inoperable. Otherwise, the plant is required to be in hot shutdown in the following 12 hours and cold shutdown in the next 24 hours.
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| The NRC staffs evaluation addresses the acceptability of the above proposed changes to TS 3.5.1.c as follows.
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| 3.1 Current TS LCO 3.5.1.c Requirements The current Hope Creek TS affected by the proposed change is associated with Action c to TS LCO 3.5.1 for the condition of an inoperable HPCI system coincident with inoperability of low pressure ECCS subsystems. Current TS Action Statement 3.5.1.c states:
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| : c. For the HPCI system, provided the Core Spray System, the LPCI system, the ADS and the RCIC system are OPERABLE:
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| : 1. With the HPCI system inoperable, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig within the following 24 hours.
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| : 2. With the HPCI system inoperable and either one LPCI subsystem or one CSS subsystem inoperable, restore the HPCI system to operable status within 72 hours or restore the LPCI subsystem/CSS subsystem to operable status within 72 hours.
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| Otherwise be in in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig in the next 24 hours.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2 Proposed TS LCO 3.5.1.c Requirements The revised TS Action 3.5.1.c is to improve clarity of the Action and to provide an allowable outage time (AOT) and explicit direction for the condition when both an LPCI and a CSS subsystem are inoperable coincident with the HPCI system being inoperable.
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| The proposed changes to Hope Creek TS Action 3.5.1.c stated in Section 3.1 of this safety evaluation (SE) are identified below. Proposed deletions are shown in strikeout and proposed additions are underlined.
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| : c. For the HPCI system: , provided the Core Spray System, the LPCI system, the ADS and the RCIC systems are OPERABLE:
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| : 1. With the HPCI system inoperable, provided the Core Spray System, the LPCI system, the ADS and the RCIC system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig within the following 24 hours.
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| : 2. With the HPCI system inoperable, provided the ADS and the RCIC system are OPERABLE: and either one LPCI subsystem or one CSS subsystem inoperable, restore the HPCI system to operable status within 72 hours or restore the LPCI subsystem/CSS subsystem to operable status within 72 hours.
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| Otherwise, be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig in the next 24 hours.
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| : a. With either one LPCI subsystem or one CSS subsystem inoperable, restore the HPCI system to OPERABLE status within 72 hours or restore the LPCI subsystem/CSS subsystem to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig within the following 24 hours.
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| : b. With one LPCI subsystem and one CSS subsystem inoperable, restore the HPCI system to OPERABLE status within 8 hours or restore the LPCI subsystem or CSS subsystem to OPERABLE status within 8 hours or be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig within the following 24 hours.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3 TS Changes 3.3.1 Action Statementsc.1 and c.2 and Renumbering c.2 to c.2.a The changes to relocate information in existing action statement c to action statements c.1 and c.2 to clarify entry conditions, and to renumber c.2 as c.2.a are editorial in nature. As identified in Section 3.2 of this SE, proposed deletions are shown in strikeout and proposed additions are underlined. The NRC staff found that the changes would improve clarity relative to entry conditions, and do not change the corresponding current TS requirements. Therefore, the NRC staff concludes that the changes are acceptable.
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| 3.3.2 Added Action 2.b to TS Action 3.5.1.c The added TS Action 3.5.1.c.2.b would allow continued operation when the HPCI is inoperable, and one CSS subsystem and one LPCI subsystem are also inoperable. The NRC staff evaluated whether the proposed action conditions are within the NRC-approved analyzed conditions and determines the acceptability of the changes as follows.
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| 3.3.2.1 Emergency Core Cooling System Operating Requirements The ECCS is designed to protect against the effects of a postulated LOCAs in compliance with the requirements of 10 CFR 50.46. The ECCS is described in Section 6.3 of the Hope Creek Updated Final Safety Analysis Report (UFSAR). The TS 3.5.1 Actions allow operation for various cases when ECCS subsystems are inoperable. Current Action a covers cases when the CSS is inoperable; Action b specifies situations when the LPCI is inoperable; and Action c defines conditions when the HPCI is inoperable.
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| 3.3.2.1.1 Available ECCS Subsystems under Action 2.b to TS 3.5.1.c According to the TS 3.5.1 requirements, an operable ECCS requires that the HPCI system, two CSS subsystems, four LPCI subsystems, and ADS be operable. The added Action 2.b to TS 3.5.1.c would allow continued operation when the HPCI is inoperable with concurrent inoperability of an LPCI subsystem and a CSS subsystem. Under the proposed Action 2.b condition, the following ECCS subsystems are still available for use in the event of a LOCA:
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| * 1 CSS subsystem, 3 LPCI subsystems, and ADS.
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| 3.3.2.1.2 ECCS Subsystems Assumed in the LOCA Analysis of Record The licensee states that for both large-break and small-break LOCA analyses, the above available ECCS subsystems are sufficient to provide adequate core cooling. The licensee confirmed by the {{letter dated|date=November 4, 2020|text=letter dated November 4, 2020}} (ADAMS Accession No. ML20310A219), that the current LOCA analysis of record (AOR) was discussed in the Hope Creek UFSAR. As stated in Section 6.3.3.7 of the Hope Creek UFSAR, the LOCA AOR at Hope Creek was performed with the previously NRC-approved SAFER/GESTR - LOCA methodology and the analysis results were discussed in Reference 6.3-4 of the UFSAR, which is a plant-specific LOCA analysis report applicable to Hope Creek, NEDC-33172-P, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis for Hope Creek Generating Station at Power Uprate, March 2005 (ADAMS Accession No. ML053250469; not publicly available). The LOCA analysis in NEDC-33172-P was performed in accordance with NRC requirements OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION demonstrating conformance with the requirements of 10 CFR 50.46 for ECCS performance acceptance criteria.
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| NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR [Light-Water Edition] Edition, (SRP) Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, March 2007 (ADAMS Accession No. ML070550016), provides the review guidance for the LOCA analysis. SRP Section 15.6.5 states that the LOCA analysis should consider piping breaks postulated to occur at various locations and include a spectrum of break sizes, up to a maximum pipe break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant pressure boundary. In addition, the analysis should address the effects of single failures on the ECCS performance. Following the SRP Section 15.6.5 guidance, the licensee performed the AOR for break sizes ranging from ((
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| ))
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| The analysis included the effects of the following credible single failures: (1) channel A direct current (Battery) source, (2) a standby diesel generator, (3) an LPCI injection valve, and (4) the HPCI. For each assumed single failure, the licensee identified the remaining operable ECCS subsystems in Table 4-4 of NEDC-33172-P. The analysis covered for breaks located in ((
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| )) As a result, the analyses identified that ((
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| )) and demonstrated that the results met the 10 CFR 50.46 requirements for the ECCS performance.
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| The LOCA AOR included the results of the analyzed LOCA cases as cases A.1 through C.5 in Appendix A of NEDC-33172-P. For each case, the following available ECCS subsystems assumed in the analysis.
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| (1) For the breaks in non-ECCS piping system, the available ECCS subsystems were:
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| (( ))
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| (2) For the breaks in the ECCS piping system, there involved the following two cases:
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| (a) for the CSS line break, the available ECCS subsystems were:
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| (( )) and (b) for the LPCI line break, the available ECCS subsystems were:
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| (( ))
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| A comparison of the ECCS subsystems assumed in the previously NRC-approved LOCA analysis (discussed in Section 3.3.2.1.2 of this SE) with the available ECCS systems for the proposed TS Action 3.5.1.c.2.b. (1 CSS subsystem, 3 LPCI subsystems, and ADS discussed in Section 3.3.2.1.1 of this SE) shows that the ECCS capability retained by the proposed change
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| (( )) The proposed TS Action 3.5.1.c.2.b. bounds the NRC-approved analyzed conditions that meet the 10 CFR 50.46 requirements for the ECCS performance acceptance criteria. Therefore, the NRC staff concludes the proposed Action is acceptable.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.2.2 Allowable Outage Time of 8 Hours and Shutdown Conditions The proposed TS Action 3.5.1.c.2.b. allows for conditions of loss of the HPCI, one CSS subsystem, and one LPCI system, an AOT of 8 hours for restoring the HPCI, or the CSS subsystem, or the LPCI subsystem to operable status, or to bring the plant to HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 pounds per square inch gauge (psig) within the following 24 hours.
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| The current TS Action 3.8.3.1.a related to onsite power distribution systems specifies that with any of the four alternating current distribution system channels de-energized, component operability will be restored within 8 hours or the plant shall be shut down. The effects of loss of any alternating current distribution channel include loss of one CSS subsystem, an LPCI subsystem, and safety related supporting systems powered by that channel. The proposed TS Action 3.5.1.c.2.b is similar to TS Action 3.8.3.1.a because both the proposed TS Action 3.5.1.c.2.b and TS Action 3.8.3.1 involve an AOT of 8 hours to restore multiple inoperable safety related structures, systems and components including a loss of one CSS subsystem, an LPCI subsystem. Additionally, as discussed in Section 3.3.2.1.2 of this SE, a sufficient number of the ECCS subsystems remain available to adequately mitigate a design-basis LOCA. The licensee also noted that its risk analysis shows that, with reasonable assurance, the proposed AOT of 8 hours is within the current risk acceptance guidelines in Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ADAMS Accession No. ML17317A256), and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, May 2011 (ADAMS Accession No. ML100910008).
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| Based on its findings that the proposed AOT in TS Action 3.5.1.c.2.b is similar to the relevant TS Action 3.8.3.1.a, and that a sufficient number of ECCS subsystems remain available to adequately mitigate a design-basis LOCA, the NRC staff concludes that the AOT of 8 hours is acceptable.
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| As for the shutdown conditions and the associated AOTs, the NRC staff found that they are consistent with that specified in the acceptable TS Action 3.5.1.c.2.a discussed in Section 3.2 and Section 3.3.1 of this SE. Therefore, the NRC staff concludes the shutdown conditions and the associated times are acceptable.
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| 3.3.3 Technical Evaluation Conclusion The NRC staff has evaluated the safety implications of the proposed TS Revision for ECCS operability when HPCI, one CSS, and one LPCI subsystem are inoperable. The proposed change will leave one CSS subsystem, three LPCI subsystems, and ADS available. This complement of equipment ((
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| )) that meets the 10 CFR 50.46 insofar as they relate to the ECCS performance acceptance criteria. Based on its review of the Hope Creek LOCA analysis and the licensee's evaluation, the NRC staff found that the ((
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| )) ECCS equipment is sufficient to furnish core cooling in the event of a postulated LOCA. Therefore, as the proposed TS conditions are bounded by the existing LOCA analysis, as discussed in Section 3.3.2.1.2 of this SE, the NRC staff determined that the proposed TS changes meet the requirements of 10 CFR 50.46.
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION
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| OFFICIAL USE ONLY PROPRIETARY INFORMATION Since the LCOs remain unchanged, and the proposed change for the AOT of 8 hours is consistent with the associated TS requirements for onsite power distribution system operability, and a sufficient number of ECCS subsystems remain available to adequately mitigate a design-basis LOCA, the NRC staff concludes that the proposed changes are consistent with the requirements of 10 CFR 50.36(c)(2). Therefore, the NRC staff concludes that the proposed TS changes are acceptable.
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| ==4.0 STATE CONSULTATION==
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| In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendment on October 6, 2020. The State official had no comments.
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| ==5.0 ENVIRONMENTAL CONSIDERATION==
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| 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 and changes SRs.
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| 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, as published in the Federal Register (85 FR 45448; July 28, 2020), and there has been no public comment on such finding.
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| 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.
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| ==6.0 CONCLUSION==
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| The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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| Principal Contributors: S. Sun M. Hamm Date: March 10, 2021 OFFICIAL USE ONLY PROPRIETARY INFORMATION}}
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