05000354/LER-2024-001-01, Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT) Supplement

Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT) Supplement

ML24253A265
Person / Time
Site:	Hope Creek
Issue date:	09/05/2024
From:	Mclaughlin R Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N24-0054 LER 2024-001-01
Download: ML24253A265 (1)

LER-2024-001, Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT) Supplement

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3542024001R01 - NRC Website

text

Robert D. McLaughlin Plant Manager, Hope Creek Generating Station, PSEG Nuclear LLC PO Box 236 Hancocks Bridge, New Jersey 08038-0221 10 CFR 50.73 LR-N24-0054 September 5, 2024 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 Docket No. 50-354 Subject: Licensee Event Report 2024-001-01, Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT) Supplement PSEG Nuclear LLC submits licensee event report (LER) 2024-001-01 per 10 CFR 50.73(a)(2)(ii)(A) and 10 CFR 50.73(a)(2)(v)(C).

There are no regulatory commitments contained in this letter.

Please contact Ms. Tammy Morin at (856) 339 -2286 if you have questions.

Sincerely, Robert D. McLaughlin Plant Manager, Hope Creek Generating Station Attachment: Licensee Event Report cc: USNRC Regional Administrator Region 1 USNRC NRR Project Manager - Hope Creek USNRC Senior Resident Inspector - Hope Creek NJ Department of Environmental Protection, Bureau of Nuclear Engineering o PSEG I NUCLEAR

Abstract

On May 3, 2024, during a refueling outage, Hope Creek Generating Station (HCGS) performed Primary Containment Integrated Leak Rate Testing (ILRT) Type A test as required by Technical Specification 4.6.1.2. The as-found test was not completed successfully because the stabilization criteria was not met, and the test was invalidated. The as-left test was completed successfully prior to plant start-up.

This licensee event report (LER) supplement provides the results of a Root Cause Evaluation (RCE). The Direct Cause was inadequate bolt seating torque that led to loss of O-ring seal compression. The Root Cause was identified as inadequate procedural guidance for validating and measuring flange to drywell airlock surface gap.

This event was reported to the NRC as an eight-hour, non-emergency per 10 CFR 50.72(b)(3)(ii)(A) in Event Notification 57103 on May 3, 2024. Subsequently, LER 2024-001-00 was submitted on July 2, 2024, reporting the event per 10 CFR 50.73(a)(2)(ii)(A). This LER supplement also identifies additional reporting criterion, 10 CFR 50.73(a)(2)(v)(C).

Tammy Morin (856)-339-2286

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3. LER NUMBER YEAR SEQUENTIAL NUMBER REV NO.

052 050 PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor (BWR/4)

System - Primary Containment IDENTIFICATION OF OCCURRENCE Discovery Date: May 3, 2024 CONDITIONS PRIOR TO OCCURRENCE OPCON 4, Cold Shutdown. No other structures, systems, or components that could have contributed to the event were inoperable.

DESCRIPTION OF OCCURRENCE On May 2, 2024, at 10:00 am, the Primary Containment Integrated Leak Test (ILRT) commenced. At 1:08 pm, the stabilization period started. At 6:30 pm there was evidence of leakage and walkdowns were initiated to determine the location of any leaks. At 8:30 pm a leak was found at the drywell airlock hatch to primary containment flange. At 11:00 pm, in accordance with procedures, an attempt to quantify the leak was made.

On May 3, 2024, at 3:05 am, a second attempt to quantify the leak was made and at 4:11 am, it was determined that a step change in the leakage rate had invalidated the as-found test.

On May 4, 2024, at 10:00 am, the Type A as-left ILRT was completed with satisfactory results.

CAUSE OF EVENT The Direct Cause of the event was identified as inadequate bolt seating torque. This led to loss of O-ring seal compression, which created a leakage path through the air-lock flange. The RCE identified one Root Cause and one Contributing Cause.

The Root Cause was inadequate procedural guidance for validating and measuring flange to drywell airlock surface metal to metal contact. The procedure also did not provide steps to validate the torque values after rigging is relaxed or removed.

A Contributing Cause was identified which led to the invalid test. Individuals involved in the test did not fully assess the risk, consequences, or potential leakage adjustment, of installing a procedurally allowed pressure block on a Type B penetration without knowing the condition of the component.

SAFETY CONSEQUENCES AND IMPLICATIONS The event occurred when HCGS was in OPCON 4, Cold Shutdown. The test was performed in an operating mode where the safety function of the primary containment was not required, and the as-left test was performed successfully prior to entering a mode when the primary containment is required. This is reportable as a degraded condition of a principal safety barrier because the stabilization criteria were not met.

An engineering analysis was performed. The analysis demonstrates the dose for the control room, exclusion area boundary, and low population zone were within regulatory dose limits under all conditions. Therefore, there were no safety consequences to site personnel or the public.

CORRECTIVE ACTIONS A satisfactory Type A as-left ILRT was performed to verify Primary Containment integrity exists. The procedure for re-installation of the drywell has been revised to ensure metal to metal surface contact after installation of the airlock and an as-left LLRT on the joint will be performed. The Primary Containment ILRT procedure has been revised to require the program owner and test director to approve the installation of pressure blocks to Type B penetrations.

COMMITMENTS There are no regulatory commitments in this LER.

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