05000364/LER-2023-003, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits
| ML23341A209 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/07/2023 |
| From: | Dean E Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-23-0883 LER 2023-003-00 | |
| Download: ML23341A209 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3642023003R00 - NRC Website | |
text
.t,, Southern Nuclear Edwin Dean Ill Vice President - Farley December 7, 2023 Docket No.:
50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Ladies and Gentlemen:
Joseph M. Farley Nuclear Plant - Unit 2 Licensee Event Report 2023-003-00 Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits Joseph M. Farley Nuclear Plant 7388 North State Hwy 95 Columbia, Alabama 36319 334.661.2100 tel 334.661.2512 fax EDDEANil@southcmco.com NL-23-0883 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.
Respectfully submitted, fl_;_D&_~
Edwin Dean Ill Vice President-Farley SD/rgs/cbg Enclosure: Unit 2 Licensee Event Report 2023-003-00 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 2023-003-00 Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits Enclosure Unit 2 Licensee Event Report 2023-003-00
Abstract
On October 24, 2023, while at 0% power level and Mode 6 (refueling), it was discovered that a Unit 2 pressurizer code safety valve (PSV), which had been removed during the refueling outage (2R29) and shipped off-site for testing, failed its as-found lift pressure test. The PSV lifted above the Technical Specification (TS) 3.4.10 allowable lift setting value. Setpoint drift of the PSV is the most likely cause of the failure.
It is likely that the PSV was outside of the TS limits longer than allowable by the Required Action Statement ( 15 minutes) during the previous operating cycle in all applicable modes of operation. Therefore, this condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS.
The PSV was replaced during the October 2023 refueling outage.
EVENT DESCRIPTION
I
- 2. DOCKET NUMBER
- 3. LER NUMBER 364 I
YEAR SEQUENTIAL REV NUMBER NO.
- - 1 003 1-0 On October 24, 2023, while Unit 2 was at 0% power level and in Mode 6 (Refueling), with the Reactor Coolant System (RCS) [AB] at atmospheric pressure and 91 degrees Fahrenheit, the 2C Pressurizer Code Safety Valve (PSV) [RV] (Serial #
N56963-01-0003, Manufacturer: Crosby; Model: HB-86-BP) was removed as part of the routine In-Service Testing (1ST) program and sent to an off-site testing facility. The as-found lift pressure was discovered to be 2522 psig which was outside of the Technical Specification (TS) 3.4.10 allowable lift pressure settings of>/= 2423 psig and</= 2510 psig. The tested valve was within the ASME code acceptance band of+/- 3% (2411-2559 psig). Based on the lift pressure meeting the 1ST program (ASME code) monitored requirements, there was no 1ST scope expansion for the PSV.
EVENT ANALYSIS
Setpoint drift of the PSV was determined to be the most likely cause of the failure.
REPORTABILITY AND SAFETY ASSESSMENT
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." There is no firm evidence, prior to the time of discovery at the test facility, of when the failure occurred. Since the as-found lift setpoint was within ASME Code allowance and less than 110%
design pressure of the RCS, the condition did not have an adverse impact on its over-pressurization function. The as-found lift pressure was 2522 psig, and the valve re-closed following the lift. There were no actual safety consequences for this event. No safety system responses occurred. There was no release of radioactivity.
CORRECTIVE ACTIONS
The PSV was replaced during the October 2023 refueling outage. The as-left setpoints were within +/- 1 % tolerance.
Previous corrective actions for PSV failures included a spring changeout campaign for all PSV's and TS 3.4.10 was also previously revised. The 2C PSV that was removed was the last PSV in the spring changeout campaign.
PREVIOUS SIMILAR EVENTS
Similar events have been reported for Unit 1 and Unit 2.
LER 2023-01-00 LER 2020-02-00 LER 2019-01-00 LER 2018-01-00 LER 2017-03-00 OTHER SYSTEMS AFFECTED:
No other systems were affected by this event.
NRC FORM 388A (10-01*2023)
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