05000287/LER-2023-001, Condition Prohibited by Technical Specifications Due to Isolation Valve Exceeding Inservice Testing Leakage Requirements
| ML23068A470 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/09/2023 |
| From: | Denise Wilson Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RA-23-0082 LER 2023-001-00 | |
| Download: ML23068A470 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2872023001R00 - NRC Website | |
text
David A. Wilson Manager, Nuclear Support Services Oconee Nuclear Station Duke Energy ON0193 l 7800 Rochester Hwy Seneca, SC 29672 o. 864.873.3451 f: 864.873.5791 David.Wilson2@duke-energy.com RA-23-0082 March 9, 2023 10 CFR 50.73 Attn: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2746 Duke Energy Carolinas, LLC Oconee Nuclear Station Unit 3 Docket Number: 50-287 Renewed Operating License: DPR-55
Subject:
Licensee Event Report 287/2023-001, Revision 00 - Condition Prohibited by Technical Specifications due to Isolation Valve Exceeding Inservice Testing Leakage Requirements Licensee Event Report 287/2023-001, Revision 00, is being submitted pursuant to the requirements of 10 CFR 50.73 to provide notification of the subject event.
There are no regulatory commitments associated with this LER.
There are no unresolved corrective actions necessary to restore compliance with NRC requirements.
If there are questions, or further information is needed, contact Sam Adams, Regulatory Affairs, at (864) 873-3348.
Sincerely, David A. Wilson Manager, Nuclear Support Services Oconee Nuclear Station Enclosure: Licensee Event Report 287-2023-001 Rev.00 dc=com, dc=duke-energy, dc=ent, dc=nam, ou=Accounts, ou=Personal, ou=PNTransitional, cn=DAWilso (252825) 2023.03.09 18:20:20 -05'00'
RA-23-0082 March 9, 2022 Page 2 cc (w/Enclosure):
Ms. Laura Dudes, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Shawn Williams, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop O-08B1A Rockville, MD 20852-2738 Mr. Jared Nadel NRC Senior Resident Inspector Oconee Nuclear Station
Abstract
Oconee Nuclear Station Unit 3 00287 3
Condition Prohibited by Technical Specifications due to Isolation Valve Exceeding Inservice Testing Leakage Requirements 01 08 2023 2023 001 00 03 09 2023 1
100 Samuel Adams, Oconee Regulatory Affairs 864-873-3348 X
DA ISV C684 N
On January 8, 2023, with Unit 3 in Mode 1 operating at approximately 100% full power, a technical specification (TS) required surveillance test was performed on a manual isolation valve in the Unit 3 Spent Fuel Pool (SFP) Cooling Purification System. The surveillance, conducted in accordance with the Inservice Testing Program, tests the valve for leakage. Test results showed that the leakage through valve 3SF-53 did not meet the leakage requirements of TS Surveillance Requirement (SR) 3.7.19.2 and the valve was declared inoperable and isolated per TS 3.7.19. Subsequent investigation of the valve revealed the presence of foreign material (FM) that was impeding the valve disc from fully closing. The FM was removed and 3SF-53 successfully passed SR 3.7.19.2 testing thereby restoring operability of the valve. Further review of operation of the valve showed the valve had not been repositioned since May 17, 2022. Based on the presence of the foreign material and the length of time since the valve had been operated, it was determined that the valve had exceeded the outage time permitted by TS 3.7.19. Therefore, this condition is being reported under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition Prohibited by Technical Specifications.Page of
- 3. LER NUMBER YEAR SEQUENTIAL NUMBER REV NO.
052 050 EVENT DESCRIPTION (cont.)
The design basis dose analysis assumes no more than 6 gph of inter-system leakage through 3SF-53 and the other SFC system isolation valves within the scope of TS 3.7.19. While 3SF-53 exceeded this value, the total back leakage from ECCS to the BWST (i.e. the potential source of any post-accident RBES fluid that could leak through 3SF-53) was measured at 6 gph. Therefore, the design basis dose analysis assumptions regarding total potential external leakage from SFC system components were not exceeded.
CAUSAL FACTORS The cause of this event was the presence of foreign material in the valve seat thereby preventing that valve from disc from fully closing.
COMPLETED CORRECTIVE ACTIONS
- 1. 3SF-53 was disassembled and inspected.
- 2. All foreign material removed from the 3SF-53 internals.
- 3. An inspection of the upstream and downstream piping was performed with no further foreign material noted.
- 4. 3SF-53 was cleaned and rebuilt.
- 5. A blue check was performed of 3SF-53 valve internals and 100% contact was observed on both sides of the disc and seats.
- 6. TS SR 3.7.19.2 inservice leakage test was re-performed with satisfactory results.
PLANNED CORRECTIVE ACTION For valves utilized in refueling outage that are in the scope of TS 3.7.19, implement programmatic changes to ensure inservice leakage testing is completed prior to entering the Mode of Applicability for TS 3.7.19.
SAFETY ANALYSIS
The 3SF-53 leakage event was evaluated and determined to have had no impact to public health and safety. Overall, no design basis dose analysis assumptions concerning RBES fluid leakage to the Auxiliary Building and back leakage to the BWST were exceeded. Further, this event had no impacts on systems or equipment credited in the Oconee PRA and therefore had no impact on plant core damage risk.
ADDITIONAL INFORMATION
A review of Duke Energy's Corrective Action Program did not identify any additional events that involved similar underlying concerns or reasons as this event. This event is considered NOT IRIS Reportable.
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