05000254/LER-2025-003, Main Steam Isolation Valves Local Leak Rate Test Exceed Technical Specification Limits
| ML25129A031 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/09/2025 |
| From: | Hild D Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| SVP-25-026 LER 2025-003-00 | |
| Download: ML25129A031 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2542025003R00 - NRC Website | |
text
Constellation.
SVP-25-026 May 9, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Nuclear Power Station, Unit 1 Renewed Facility Operating License No. DPR-29 NRC Docket No. 50-254 10 CFR 50.73
Subject:
Licensee Event Report 254/2025-003-00 "Main Steam Isolation Valves Local Leak Rate Test Exceed Technical Specification Limits" Enclosed is Licensee Event Report 254/2025-003-00 "Main Steam Isolation Valves Local Leak Rate Test Exceed Technical Specification Limits".
This report is being submitted in accordance with 10 CFR 50. 73(a)(2)(i)(B) for any operation or condition prohibited by Technical Specifications.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this report, please contact Conner Bealer at 779-231-6207.
Res~ully, tJ OougHi~
Site Vice President Quad Cities Nuclear Power Station cc:
Regional Administrator - NRC Region Ill NRC Senior Resident Inspector-Quad Cities Nuclear Power Station
Abstract
On 03112/2025 at 22:56, the as-found Local Leak Rate Tests (LLRT) on various Main Steam Isolation Valves (MSIV) exceeded the Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.10 limits of ~62.4 standard cubic feet per hour (scfh) through each leakage path, and 156 scfh for the combined leakage rate for all MSIV leakage paths.
The C Main Steam Line (MSL) exceeded the leakage limit on both the inboard and outboard valves, resulting in a failed leakage path. The collective set of 4 MSLs exceeded the 156 scfh combined leakage rate limit.
Corrective actions included flushing, disassembling, inspecting, repairing, and retesting the valves. The cause is attributed to pilot rotation causing stem nut wear which leads to a leakage path.
This is being reported in accordance with 10 CFR 50. 73(a)(2)(i)(B), any operation or condition prohibited by Technical Specifications.
PLANT AND SYSTEM IDENTIFICATION
- 2. DOCKET NUMBER YEAR 00254 2025
- 3. LER NUMBER SEQUENTIAL NUMBER 003 REV NO.
00 General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION Main Steam Isolation Valves Local Leak Rate Test Exceed Technical Specification Limits
A. CONDITION PRIOR TO EVENT
Unit: 1 Reactor Mode: 5 Event Date: March 12, 2025 Mode Name: Refueling Event Time: 2256 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.58408e-4 months <br /> CST Power Level: 000%
No systems, structures, or components that were inoperable at the start of the event contributed to the event.
8. DESCRIPTION OF EVENT
On 03/12/2025 at 22:56, the as-found Local Leak Rate Tests (LLRT) on various Main Steam [SB] Isolation Valves (MSIV)[ISV] were found to exceed the Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.10 limits of~ 62.4 standard cubic feet per hour (scfh) through each leakage path, and 156 scfh for the combined leakage rate for all MSIV leakage paths. Individual MSIV tests were performed beginning on 3/10/2025. An actual date cannot be determined for the transition beyond the allowable leakage. As a result, TS 3.6.1.3 Condition D, MSIV leakage rate not within limit, Required Action D.1, completion time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was conservatively not met.
There are four main steam lines (MSL), designated A through D. Each steam line contains an inboard and outboard MSIV, designed number 1 and 2 respectively. For example, 1 B would be the inboard MSIV on the B MSL, and 2D would be the outboard MSIV on the D MSL. The full equipment number would be 1-0203-1 B, but this report will include only the two digit representations.
The C MSL did not meet the TS SR 3.6.1.3.10 limit of~ 62.4 scfh through each leakage path:
C MSL - greater than 80 scfh The combined leakage rate for all MSIV leakage paths(</= 156 scfh) is determined by taking the valve in the MSL with the lesser leakage rate:
A MSL-4.3 scfh 8 MSL - 54.6 scfh C MSL - 120.2 scfh D MSL - 36 scfh Total - 215.1 scfh, which is above the TS limit~ 156 scfh
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO.
00254 2025 -
003 00 Valve overhauls and repairs were performed on 1 C, 2A, 2C and 20 MSIVs. Post maintenance testing was performed and all valves met the TS SR limits.
Since both the TS limit for minimum path leakage in its MSL, and the combined total leakage of all MSLs exceeded the TS limits, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of past operation or condition which was prohibited by the plant Technical Specifications. This event is self-identified.
C. CAUSE OF EVENT
The cause is attributed to pilot rotation causing stem nut wear which leads to a leakage path.
Three of the affected valves had not yet been modified to include anti-rotation pilot keys. One valve included this modification. This modification is typically installed as valve leakage results warrant overhauls, and not all MS IVs have required overhaul since the development of this modification.
D. SAFETY ANALYSIS
System Design
The design of the MS IVs is to prevent reactor coolant [AD] inventory loss and protect plant personnel in the event of steam line breakage outside the isolation valves, and to complete the containment boundary after a Loss of Coolant Accident (LOCA). The MSIVs are 20-inch air/spring-operated, balanced "Y"-type globe valves mounted inboard and outboard of the containment. The inboard valve air is supplied from the containment drywell pneumatic system. The outboard valve is supplied by the normal instrument air system. This valve combines a full port design with straight-line flow to provide a very good flow pattern. These valves use upstream pressure to aid in closure by tilting the actuator toward the upstream side of the valve.
For lines that extend the primary containment boundary, the boundary includes the piping to the last (i.e.,
outboard) isolation valve. A primary containment pathway must be capable of being isolated and as such is tested in accordance with the Primary Containment Leakage Rate Testing Program. Penetration leak rate testing verifies the capability of the penetrations to maintain overall containment leakage (La) within the limits established by 10 CFR 50 Appendix J. Technical Specification 3.6.1.3 provides the operability requirements for primary containment isolation valves.
Safety Impact There were no safety consequences as a result of this event. There was no radiation release associated with this event. No operator actions were required based on being in the refueling mode. The overall Q1 R28 primary containment as-found minimum pathway leakage (all other leak pathways, excluding MSIVs) was 129.4485 scfh. This leakage is well within the allowed leakage limit of 1373.34 scfh. Therefore, the safety
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO.
00254 2025 -
003 00 significance of the "C" MSLs leakage contribution and the total combined MSL contribution to the overall primary containment leakage was minimal.
An informal modeling of the as found MSL leakage utilizing RadTrad and the Unit 1 Alternate Source Term (AST) Loss of Coolant Accident (LOCA) analysis was performed. The dose limits from this informal model analysis showed Control Room, Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) dose rates below the regulatory required dose limits for a design basis LOCA analysis.
This event is not considered a safety system functional failure.
E. CORRECTIVE ACTIONS
Immediate:
- 1. Valve overhauls and repairs were performed on 1 C, 2A, 2C and 20 MS IVs. Post maintenance testing was performed and all valves met the TS SR limits.
Follow-up:
- 1. Installation of anti-rotation modification on remaining MSIVs.
F. PREVIOUS OCCURENCES
The station events database, LERs and IRIS were reviewed for similar events at Quad Cities Nuclear Power Station.
LER (265/2016-001-00) Main Steam Isolation Valve Local Leak Rate Tests Exceed Technical Specification Limits, 03/21/2016. Two MSIVs exceeded the individual TS limit and combined minimum path MSIV leakage exceeded the TS limit. The cause was due to the MSIV design being susceptible to degraded main plug/seat interface during valve closure, and non-optimal valve design. Since this event all the Unit 1 and Unit 2 MSIV plugs have been re-designed and are in progress to be installed on all Unit 1 and Unit 2 MSIVs. There is also a modification to the pilot of the valve to include an Anti-Rotation key, which will be installed on all Unit 1 and Unit 2 MSIVs. The four valves refurbished and overhauled did not have the anti-rotation modification prior to 01 R28 but has now been installed.
G. COMPONENT FAILURE DATA
Failed Equipment:
Component Manufacturer: Crane Nuclear Component Model Number: Model 20 inch Y Pattern Globe Valve Component Part Number: n/a This event has been reported to IRIS. Page_4_ of _4_