05000364/LER-2023-001, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits

Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits

ML23271A132
Person / Time
Site:	Farley
Issue date:	09/28/2023
From:	Erb D Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-23-0747 LER 2023-001-00
Download: ML23271A132 (1)

LER-2023-001, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3642023001R00 - NRC Website

text

A Southern Nuclear Delson Erb Joseph M. Farley Nuclear Plant Vice President - Farley 7388 North State Hwy 95 Columbia, Alabama 3631 9 334.661.2100 tel 334.661.2512 fax

dcrb@southcmco.com

September 28, 2023

Docket No.: 50-364 NL-23-0747

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001

Joseph M. Farley Nuclear Plant-Unit 2 Licensee Event Report 2023-001-00 Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits

Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), Southern Nuclear Company is submitting the enclosed Licensee Event Report for Unit 2.

This letter contains no NRC commitments. If you have any questions regarding this submittal, please contact Gene Surber, Licensing Manager, at (334) 661-2265.

Delson Erb Vice President-Farley

DE/rgs/cbg

Enclosure: Unit 2 Licensee Event Report 2023-001-00

Cc: Regional Administrator, Region II NRR Project Manager-Farley Nuclear Plant Senior Resident Inspector - Farley Nuclear Plant RTYPE: CFA04.054 Joseph M. Farley Nuclear Plant - Unit 2 Licensee Event Report 2023-001-00 Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits

Enclosure

Unit 2 Licensee Event Report 2023-001-00

Abstract

On June 14, 2023, Farley Unit 2 entered Mode 5 to conduct planned maintenance on the 2A Pressurizer Code Safety Valve (PSV) which had indications of leaking. The PSV was removed from the system and delivered to a vendor to test in accordance with Technical Specification (T/S) 3.4.10, Pressurizer Safety Valves.

On August 1, 2023, while in Mode 1 at 100% power level, Farley Unit 2 was informed by the vendor that the set pressure results were outside of the T/S as-found acceptance criteria of 2423 - 2510 psig. During as-found testing the 2A PSV failed to lift.

Additional testing by the vendor using in-situ device identified that 2A PSV lifted at approximately 2599 psig. Although this value is greater than T/S limits, the 110% ASME Code limitation and Safety Limit were not exceeded. The failure to lift within the acceptable band was determined to be the result of internal steam cutting of the disc insert and nozzle. The 2A PSV was replaced during the planned outage with a pre-tested spare.

EVENT DESCRIPTION

On July 8, 2022, it was identified that the 2A Pressurizer Code Safety Valve (PSV) [EEIS:AB/RV] began leaking as evidence by elevated tailpipe temperatures and Pressurizer Relief Tank (PRT) parameters. The 2A PSV (Serial#

N56963-01-0001, Manufacturer: Crosby, Model: HB-86-BP) was installed on Unit 2 during refuel 27 (2R27) in the fall of 2020. The 2A PSV was monitored throughout the cycle, and a maintenance outage was planned based on the expectation that pre-identified operational triggers would be met.

On June 14, 2023, Farley Unit 2 entered Mode 5 to conduct planned maintenance on the 2A PSV. Following the shutdown, the 2A PSV was removed from the system on June 16, 2023, and delivered to a vendor to test in accordance with Technical Specification (T/S) 3.4.10. The 2A PSV was replaced with a pre-tested spare. Unit 2 completed the planned outage and reached Mode 1 on June 19, 2023.

On August 1, 2023, while in Mode 1 at 100% power level, Farley was informed by the vendor that the set pressure results were outside of the T/S as-found acceptance criteria of 2423 - 2510 psig. During as-found testing at the vendor facility the 2A PSV failed to lift. Additional testing by the vendor using an in-situ device identified that 2A PSV lifted at approximately 2599 psig which is a 4.6% error of set pressure (2485 psig).

EVENT ANALYSIS

Leakage past the seat was also identified during testing at the off-site facility. The cause of the 2A PSV failure to lift was determined to be the result of excessive steam cutting of the valve disc insert and nozzle.

REPORTABILITY AND SAFETY ASSESSMENT :

Although the T/S pressure limit was surpassed, the ASME Code limitation and Safety Limit of 110% design pressure was not exceeded. This failure constitutes a condition that is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications." Although seat leakage was evident while online beginning in July of 2022 there is no firm evidence of when the failure to meet the lift setting requirements occurred prior to the time of discovery at the test facility. There were no actual safety consequences for this event. No safety system responses occurred. There was no release of radioactivity.

During operation, all three PSVs are required to be operable per T/S 3.4.10. The combined relief capacity of all three valves is greater than the maximum surge rate resulting from a complete loss of steam flow to the turbine assuming no reactor trip until the first Reactor Trip System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and assuming no operation of the pressurizer power-operated relief valves (PORVs) or steam dump valves. Each PSV is designed to relieve 345,000 lbs per hour of saturated steam at the valve setpoint. The Reactor Coolant System (RCS) is protected from overpressure regardless of a reactor trip, assuming all PSVs function properly. Should the reactor trip at the first protection grade trip (High Pressurizer Pressure), then only 40% of the total valve capacity is required. This 40%

readily falls within that provided by the other two PSVs and protection from over pressurization can be expected. With the two other operable PSVs this issue had very low safety significance. Additionally, this issue did not prevent the PSVs from fulfilling their design safety function and the existing configuration risk monitor, Phoenix Risk Model (PRM), showed little to no increase in risk due to the PSV being out of service.

CORRECTIVE ACTIONS

1. The 2A PSV was replaced during the June 2023 planned outage. The as-left setpoints were within +/- 1 % tolerance.

2. The 2A PSV (SN N56963-01-0001) will be disassembled and inspected to include the replacement of the disc and nozzle prior to re-installation in any future plant location.

3. Site annunciator procedures will be revised to provide operational guidance when PSV leakage occurs.

PREVIOUS SIMILAR EVENTS

This LER is the first PSV to have failed to lift or lift high based on a 10-year search; previous reports were based on PSVs lifting early.

OTHER SYSTEMS AFFECTED :

No other systems were affected by this event.