ML23103A470

NRR E-mail Capture - Susquehanna Unit 2 - Verbal Authorization of Relief Request 4RR-10

ML23103A470
Person / Time
Site:	Susquehanna
Issue date:	04/12/2023
From:	James Kim NRC/NRR/DORL/LPL1
To:	Brown K Susquehanna
References
L-2023-LRR-0014
Download: ML23103A470 (4)

Text

From: James Kim Sent: Wednesday, April 12, 2023 1:18 PM To: Brown, Katie Cc: Matthew Mitchell; Hipo Gonzalez

Subject:

Susquehanna Unit 2 - Verbal Authorization of Relief Request 4RR-10 Attachments: Susquenhanna 2 - Relief Request 4RR-10 Verbal Authorization.docx

Ms. Brown:

In accordance with NRR Office Instruction LIC-102, Relief Request Reviews, the NRR staff has provided verbal authorization for Susquehanna Unit 2 relief request 4RR-10, which requested to use an alternative to specific paragraphs of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, IWA-4000 requirements as described in your letter to NRC dated April 10, 2023, as supplemented by letter dated April 11, 2023.

Attached is the script for the verbal authorization of relief request 4RR-10 that was provided on April 12, 2023, by Matthew Mitchell and Hipolito Gonzalez. The NRC staff intends to follow-up this verbal authorization with a written safety evaluation within approximately 150 days. Please let me know if you have any questions. A copy of this email and verbal authorization will be made publicly available in ADAMS.

The following NRC and licensee personnel participated in the conference call:

NRC

Matthew Mitchell - Chief, Piping and Head Penetrations Branch, Office of Nuclear Reactor Regulation Hipolito Gonzalez - Chief, Plant Licensing Branch I, Office of Nuclear Reactor Regulation Christopher Highly - Senior Resident Inspector, Susquehanna, Region 1 James Kim - Licensing Project Manager, Plant Licensing Branch I, Office of Nuclear Reactor Regulation

Susquehanna Nuclear, LLC

Katie Brown - Regulatory Affairs Shane Jurek - Design Engineering Thomas Kupetz - ISI Engineering Ronald Vazquies - Design Engineering Ryan Sutliff - ISI Branch Manager Mark Jones - General Manager Engineering Jerry Lubinsky - Station Engineering Manager

Thank you.

James S. Kim US Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Hearing Identifier: NRR_DRMA Email Number: 2039

Mail Envelope Properties (DM6PR09MB471197C6353CA1FF83F94924E49B9)

Subject:

Susquehanna Unit 2 - Verbal Authorization of Relief Request 4RR-10 Sent Date: 4/12/2023 1:18:17 PM Received Date: 4/12/2023 1:18:00 PM From: James Kim

Created By: James.Kim@nrc.gov

Recipients:

"Matthew Mitchell" <Matthew.Mitchell@nrc.gov>

Tracking Status: None "Hipo Gonzalez" <Hipolito.Gonzalez@nrc.gov>

Tracking Status: None "Brown, Katie" <Katie.Brown@talenenergy.com>

Tracking Status: None

Post Office: DM6PR09MB4711.namprd09.prod.outlook.com

Files Size Date & Time MESSAGE 1881 4/12/2023 1:18:00 PM Susquenhanna 2 - Relief Request 4RR-10 Verbal Authorization.docx 82472

Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 4RR-10 REGARDING CORE SPRAY INJECTION TO REACTOR VESSEL VALVE TALEN ENERGY SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 DOCKET NO. 50-388 EPID: L-2023-LLR-0014

Technical Evaluation read by Matthew Mitchell, Chief of the Piping and Head Penetration Branch, Office of Nuclear Reactor Regulation

By letter dated April 10, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession ML23100A128), as supplemented by letter dated April 11, 2023 (ML23101A153), Susquehanna Nuclear, LLC, (the licensee) submitting Relief Request 4RR-10 which requested to use an alternative to specific paragraphs of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, IWA-4000 requirements.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee proposed to leave the existing leakoff port plug of core spray injection-to-reactor vessel valve 252F007B in service for one additional cycle of operation in lieu of replacing the plug with an ASME Code-compliant, seal-welded plug during the units Spring 2023 refueling outage. The licensee requested this alternative for one cycle of operation (i.e., Cycle 22 through Spring 2025) at the Susquehanna Steam Electric Station (Susquehanna), Unit 2. Because Cycle 22 of operation will span portions of both the fourth and fifth 10-year inservice inspection (ISI) intervals, the licensee requested that the proposed alternative cover the remainder of the fourth and the initial portion of the fifth 10-year ISI interval. The Susquehanna, Unit 2, fourth 10-year ISI interval is scheduled to end on May 31, 2024, and the fifth 10-year ISI interval is scheduled to begin on June 1, 2024.

The licensees basis for hardship to not repair the subject valve in the ongoing refueling outage is the valves proximity to the reactor vessel and the size of the line, which makes it not possible to isolate the valve from the reactor vessel. During replacement of the plug, if complications involving leakage through the port or valve packing were to occur, the only way to prevent water leaking through valve 252F007B would be to place the disc on the backseat of the valve, which may not eliminate all leakby. Therefore, replacement of the plug could require a full core offload and draining the reactor vessel water level below the core spray injection lines. This activity would result in unnecessary fuel moves which present an increase in risk of damaging fuel.

Deferral of the plug replacement for valve 252F007B for one operating cycle allows for additional planning and helps to reduce the risk of human performance errors during potential full core offload and vessel draindown activities. The NRC staff finds the licensees hardship justification acceptable because concerns from the risk to plant safety as well as the risk of personnel exposure to occupational hazards and excessive radiation constitute hardship without a compensating increase in the level of quality and safety.

The NRC staff verified that: (1) the licensee has performed the appropriate inspection of the existing leakoff port plug of valve 252F007B and confirmed no evidence leakage at the plug, (2) the licensee will perform the ASME Code,Section XI, Table IWB-2500-1 required system leakage test and associated visual examination of valve 252F007B during the units Spring 2023 refueling outage to ensure that no leakage is evident prior to restart of plant operation, and (3) the licensee will monitor the drywell for sources of unidentified leakage in accordance with Technical Specifications (TS) Surveillance Requirement (SR) 3.4.4.1, which requires verification that the reactor coolant system (RCS) leakage limits specified in Limiting Condition of Operation (LCO) 3.4.4 are met. In its supplement dated April 11, 2023, the licensee explained how existing plant procedures and required actions in response to potential unidentified leakage inside of containment will ensure the timely identification of leakage from the leakoff port plug of valve 252F007B should it occur during Cycle 22 of operation. Further, the staff understands that in the event of potential failure of the port plug of valve 252F007B, the loss of coolant would be within the capacity of normal plant makeup and support an orderly shutdown of the unit.

Therefore, based on above, the NRC staff finds that (1) there is reasonable assurance that the licensees proposed alternative has a minimal impact on safety and (2) the licensees hardship justification is acceptable.

Authorization read by Hipólito J. González, Chief of the Plant Licensing Branch I, Office of Nuclear Reactor Regulation

As Chief of the Plant Licensing Branch I, Office of Nuclear Reactor Regulation, I agree with the conclusions of the Piping and Head Pene trations Branch.

The NRC staff concludes that the proposed alternative will provide reasonable assurance of structural integrity and leak tightness of the subject valve until the next scheduled refueling outage when the valves leakoff port plug is replaced with the ASME Code-required, seal-welded plug. The NRC staff finds that performing the ASME Code-required repair during the units current Spring 2023 refueling outage could result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).

Therefore, effective April 12, 2023, the NRC staff authorizes the use of the Relief Request 4RR-10 at Susquehanna, Unit 2 for one cycle of operation until Spring 2025.

All other requirements in ASME Code,Section XI, for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding the proposed relief while subsequently preparing the written safety evaluation.