ML22112A089

NRR E-mail Capture - Draft RAIs for Farley LAR Re. TS 5.5.17

ML22112A089
Person / Time
Site:	Farley
Issue date:	04/22/2022
From:	Stephanie Devlin-Gill NRC/NRR/DORL/LPL2-1
To:	Sparkman W Southern Nuclear Operating Co
References
L-2021-LLA-0229
Download: ML22112A089 (4)

Text

From:

Devlin-Gill, Stephanie Sent:

Friday, April 22, 2022 9:39 AM To:

Sparkman, Wesley A.

Subject:

Draft RAIs for Farley LAR re. TS 5.5.17 (EPID: L-2021-LLA-0229)

Attachments:

Draft to SNC RAIs EPID L-2021-LLA-0229.docx Wes, Attached are the NRC staffs draft RAIs for SNCs license amendment request for the Joseph M.

Farley Nuclear Plant, Units 1 and 2 (Agency-wide Documents Access and Management System Accession No. ML21348A733), which proposes to revise Technical Specifications (TS) 5.5.17, Containment Leakage Rate Testing Program to update the peak calculated containment internal pressure for the design basis loss of coolant accident from 43.8 psig to 45 psig.

The questions are being transmitted to you to determine 1) If the question clearly conveys the NRC informational needs, 2) Whether the regulatory basis for the question is clear, and 3) If the information has already been provided in existing docketed correspondence. Additionally, review of the draft question will allow you to determine whether you are able to support a 30 day response time. Please get back to me at as soon as possible with your determinations so I can finalize issuance of the RAIs.

Thank you.

stephanie Stephanie Devlin-Gill (she/her)

Licensing Project Manager

w. 301-415-5301 U.S. NRC, NRR, DORL, LPL2-1

Hearing Identifier:

NRR_DRMA Email Number:

1604 Mail Envelope Properties (DM6PR09MB4806735D806BFB718CAC7412BDF79)

Subject:

Draft RAIs for Farley LAR re. TS 5.5.17 (EPID: L-2021-LLA-0229)

Sent Date:

4/22/2022 9:38:47 AM Received Date:

4/22/2022 9:38:00 AM From:

Devlin-Gill, Stephanie Created By:

Stephanie.Devlin-Gill@nrc.gov Recipients:

"Sparkman, Wesley A." <WASPARKM@southernco.com>

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DM6PR09MB4806.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1150 4/22/2022 9:38:00 AM Draft to SNC RAIs EPID L-2021-LLA-0229.docx 40717 Options Priority:

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REQUEST FOR ADDITIONAL INFORMATION REGARDING CHANGES TO TECHNICAL SPECIFICATION 5.5.17 TO INCREASE CALCULATED PEAK CONTAINMENT PRESSURE EPID: L-2021-LLA-0229 LICENSE AMENDMENT REQUEST JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-348 AND 50-364 By letter dated December 13, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21348A733), Southern Nuclear Operating Company (the licensee) submitted a license amendment request (LAR) for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP). The LAR proposes to revise the FNP Technical Specifications (TS) 5.5.17, Containment Leakage Rate Testing Program to update the peak calculated containment internal pressure for the design basis loss of coolant accident (LOCA) from 43.8 psig to 45 psig. The design limit for the containment internal pressure is 54 psig.

The U.S. Nuclear Regulatory Commission (NRC) staff has determined that the following requests for additional information (RAI) are needed.

RAI 01

REGULATORY BASIS:

Appendix A to Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), General Design Criteria for Nuclear Power Plants (GDC), Criterion 16, Containment design, states:

Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

RAI:

In Section 3.4 of the enclosure to the LAR, the licensee states that containment pressure analyses to determine the limiting LOCA were performed using Version 8.1 of the GOTHIC computer program. The NRC notes that Version 6.0 of GOTHIC code was used in the analysis of record (AOR) as described in the FNP Final Safety Analysis Report (FSAR), Section 6.2. In Section 3.4 of the enclosure to the LAR, the licensee states that [t]he change in code version does not result in a numerically significant departure from the previous analysis and does not yield a benefit to peak pressure and temperature results.

The AOR and the revised analysis performed in support of this LAR do not utilize the same version of GOTHIC. Since GOTHIC is not an NRC-approved code but is an accepted code for containment response analysis for FNP, the NRC staff requests that the licensee provide data to document that a verification analysis was performed to justify the assertion that the change in version does not result in a numerically significant departure from the previous analysis. The NRC staff requests that the licensee provide comparison results between the two versions of the GOTHIC code. For example, the licensee could provide a comparison of curves for the pressure and temperature response of the containment for the Double-Ended Hot Leg (DEHL) and Double-Ended Pump Suction (DEPS) cases documented in Figures 1 through 4 in Section 3.4.1 of the enclosure to the LAR.

RAI 02

REGULATORY BASIS:

Appendix A to 10 CFR 50, GDC, Criterion 16 Containment design, as stated above.

Appendix A to 10 CFR 50, GDC, Criterion 38 Containment heat removal states, in part:

A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

RAI:

Section 6.2.1.3.6 of the FNP FSAR states that [d]uring sump water recirculation, a third pressure peak occurs due to steam evolution from the reactor because of boiloff of the hotter core injection water. The FSAR states that the maximum sump water temperature of 260 degrees Fahrenheit (°F) occurs at 1252 seconds (s). The NRC staff requests the licensee provide the following information to ensure acceptability of the containment heat removal system, consistent with requirements of GDC 38:

(a) Updated sump water temperature profile and the value of the peak water temperature during the LOCA recirculation phase.

(b) Minimum net positive suction head (NPSH) available (NPSHA) for the pumps that draw water from the sump during the LOCA recirculation phase.

(c) In case containment accident pressure (CAP) is used for the minimum NPSHA calculation in (b), confirm that it is less than the vapor pressure at the sump water temperature.

(d) Maximum NPSH required (NPSHR) for the pumps in (b) at their respective applicable flow rates.

(e) Minimum NPSH margin = NPSHA-NPSHR For the items above, please confirm that the analysis reflects the latest sump strainer design installed to meet Generic Safety Issue 191, Assessment of Debris Accumulation on PWR Sump Performance.