LR-N20-0069, Response to Final Request for Additional Information Regarding Revise Technical Specifications Requirements for High Pressure Coolant Injection System Inoperability

Response to Final Request for Additional Information Regarding Revise Technical Specifications Requirements for High Pressure Coolant Injection System Inoperability

ML20310A219
Person / Time
Site:	Hope Creek
Issue date:	11/04/2020
From:	Casulli E Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H20-01, LR-N20-0069
Download: ML20310A219 (6)

Text

10 CFR 50.90 LR-N20-0069 LAR H20-01 November 4, 2020 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Response to Final Request for Additional Information Regarding Revise Technical Specifications Requirements for High Pressure Coolant Injection System Inoperability

References:

1. PSEG letter to NRC, License Amendment Request: Revise Emergency Core Cooling System Technical Specifications With Respect to High Pressure Coolant Injection System Inoperability, dated June 15, 2020 (ADAMS Accession No. ML20167A190)

2. NRC letter to PSEG, Final Request for Additional Information Regarding Revise Technical Specifications Requirements for High Pressure Coolant Injection System Inoperability (L-2020-LLA-0131), dated October 21, 2020 (ADAMS Accession No. ML20295A492)

In the Reference 1 letter, PSEG Nuclear LLC (PSEG) submitted a license amendment request (LAR) for Hope Creek Generating Station (HCGS). The proposed amendment would revise the HCGS Technical Specifications to clarify the entry conditions for Limiting Condition for Operation (LCO) 3.5.1 Action c relative to High Pressure Coolant Injection (HPCI) system inoperability. The LAR also proposed to add a new TS Action, 3.5.1.c.2.b to address the condition where the HPCI system is inoperable coincident with inoperability of a Low Pressure Coolant Injection (LPCI) subsystem and a Core Spray System (CSS) subsystem.

In the Reference 2 letter, the U.S. Nuclear Regulatory Commission staff provided PSEG a Request for Additional Information (RAI) to support the NRC staffs detailed technical review of Reference 1. The requested information is provided in Attachment 1.

PSEG has determined that the information provided in this submittal does not alter the conclusions reached in the 10 CFR 50.92 no significant hazards determination previously submitted. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

LR-N20-0069 10 CFR 50.90 Page 2 If you have any questions or require additional information, please contact Mr. Michael Wiwel at 856-339-7907.

I declare under penalty of perjury that the foregoing is true and correct.

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Respectfully, Edward T. Casulli Vice President - Hope Creek Generating Station PSEG Nuclear LLC Response to Request for Additional Information EPID L-2020-LLA-0131 cc: Administrator, Region I, NRC NRC Project Manager NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Station Commitment Tracking Coordinator

LR-N20-0069 LAR H20-01 Response to Final Request for Additional Information Regarding Revise Technical Specifications Requirements for High Pressure Coolant Injection System Inoperability L-2020-LLA-0131

LR-N20-0069 LAR H20-01 Regulatory Analysis Basis Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(b) requires each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications. The technical specification will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34.

10 CFR 50.46 contains requirements for the design and cooling performance of the Emergency Core Cooling System (ECCS).

10 CFR 50, Appendix A, General Design Criteria (GDC) CRITERION 35 -

EMERGENCY CORE COOLING states:

"A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

Request for Additional Information Question #1 Section 3.1, page 5 of the LAR, contains the following statement: Therefore, instead of two LPCI subsystems, Hope Creek has four independent subsystems of low pressure injection capability in addition to the two low pressure CSS subsystems, providing an increased level of redundancy to that assumed in NUREG-1433 relative to the LPCI function.

Please provide information related to the capacity of each LPCI subsystem. Provide an evaluation of whether 3 LPCI subsystems can supply a combined flow rate needed to mitigate a LOCA.

Response to RAI Question #1:

The Residual Heat Removal (RHR) pumps used at Hope Creek Generating Station are constant speed, motor driven pumps with a Low Pressure Coolant Injection (LPCI) flow rate that is dynamically dependent on the pressure of the reactor pressure vessel (RPV) following depressurization either through actuation of the Automatic Depressurization System (ADS) and/or through the break itself.

LR-N20-0069 LAR H20-01 The in-service testing performed to satisfy Technical Specification Surveillance Requirement 4.5.1.b.2 assures each LPCI pump can develop a flow of at least 10,000 gpm against a test line pressure corresponding to a RPV to primary containment differential pressure of >20 psid. This required LPCI flow capacity per RHR pump supports the Emergency Core Cooling System-Loss of Coolant Accident (ECCS-LOCA)

Analyses of Record (AOR) for Hope Creek and is sufficient to mitigate a LOCA as discussed in the response to the second question in this Request for Additional Information.

Request for Additional Information Question #2 The proposed TS Action 3.5.1.c.2.b would allow continued operation up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the condition where the HPCI system is inoperable, coincident with an inoperable core spray system (CSS) subsystem and an inoperable low pressure coolant injection (LPCI) subsystem. According to the TS 3.5.1 requirements, an operable ECCS requires that HPCI, two CSS subsystems, four LPCI subsystems and ADS be operable. Under the proposed Action 3.5.1.c.2.b condition, the following ECCS subsystems are still available for use in the event of a LOCA: 1 CSS subsystem, 3 LPCI subsystems, and ADS. The licensee states in its LAR that for both large-break and small-break LOCA events, the above available ECCS subsystems are sufficient to provide adequate core cooling. The licensee does not provide a discussion of an applicable LOCA analysis to support its statement above.

The ECCS is designed to protect against the effects of a postulated LOCA in compliance with the requirements of 10 CFR 50.46, "Acceptable Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

Please provide a discussion of the analysis to support the statement that in the event of a LOCA, 1 CSS subsystem, 3 LPCI subsystems, and ADS are sufficient to provide adequate cooling and address how the analysis meets the 10 CFR 50.46 requirements insofar as they relate to the ECCS performance acceptance criteria.

Response to RAI Question #2:

The current ECCS-LOCA Analyses of Record (AOR) and the tables in Hope Creek UFSAR Section 6.3 identify the limiting design basis LOCA as well as concurrent worst case single failures. The ECCS-LOCA AOR and Hope Creek UFSAR Section 6.3.3.3 identify that for both large and small break LOCAs, failure of the channel A dc source is the most severe single active failure relative to its challenge to LOCA mitigation. Per the ECCS-LOCA AOR and UFSAR Table 6.3-6, the assumed failure of the Channel A dc source and an additional assumed failure of an ADS valve, leaves the following remaining ECCS subsystems which provide sufficient mitigation of the limiting design basis LOCA event to meet 10 CFR 50.46 criteria:

1 Core Spray Loop + 3 LPCI + 4 ADS Valves LR-N20-0069 LAR H20-01 The remaining contingent of ECCS injection sources following failure of the Channel A dc source reflects the state of available ECCS injection sources during the new eight hour Action Statement 3.5.1.c.2.b proposed in Reference 1.