05000315/LER-2025-001, Two Main Steam Safety Valves Failed Setpoint Testing
| ML25135A102 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/15/2025 |
| From: | Ellis L Indiana Michigan Power Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| AEP-NRC-2025-28 LER 2025-001-00 | |
| Download: ML25135A102 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3152025001R00 - NRC Website | |
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INDIANA MICHIGAN POWER" An MP Company BOUNDLESS ENERGY-May 15, 2025 Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 1 LICENSEE EVENT REPORT 315/2025-001-00 Two Main Steam Safety Valves Failed Setpoint Testing Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 indianamichiganpower.com AEP-NRC-2025-28 10 CFR 50.73 In accordance with 10 CFR 50.73, Licensee Event Report (LER) System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant Unit 1, is submitting as an enclosure to this letter the following report:
LER 315/2025-001-00: Two Main Steam Safety Valves Failed Setpoint Testing There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.
Luke C. Ellis Plant Manager MPH/sjh
Enclosure:
Licensee Event Report 315/2025-001-00: Two Main Steam Safety Valves Failed Setpoint Testing
U.S. Nuclear Regulatory Commission Page 2 c:
EGLE - RMD/RPS J. B. Giessner -- NRC Region Ill NRC Resident Inspector N. Quilico -- MPSC R. M. Sistevaris -- AEP Ft. Wayne S. P. Wall, NRC Washington D.C.
A. J. Williamson -- AEP Ft. Wayne AEP-NRC-2025-28
Enclosure to AEP-NRC-2025-28 Licensee Event Report 315/2025-001-00: Two Main Steam Safety Valves Failed Setpoint Testing
Abstract
On March 20, 2025, with Donald C. Cook Unit 1 in Mode 1 at 63% reactor power, the station commenced Technical Specification (TS) surveillance testing on the Unit 1 Main Steam Safety Valves (MSSVs) setpoints just prior to the Unit 1 Cycle 33 (U1C33) Refueling Outage. The Unit 1 Steam Generator (SG) #4 MSSV, 1-SV-18-4, did not lift with 110% of set pressure applied with the test equipment. After reactor power was further reduced to 44.5%, SG#4 MSSV, 1-SV-28-4, also failed to lift.
Based on the guidance in NUREG-1022, Revision 3, the Unit 1 SG #4 MSSVs were considered to be inoperable longer than allowed by TS as the condition should be considered to have existed during plant operation and is reportable under 1 O CFR 50.73(a)(2)(i)(B) "Any operation or condition prohibited by the plant's Technical Specifications."
A cause evaluation is currently in progress and has determined that the preliminary cause of the inability of 1-SV-1B-4 and 1-SV-28-4 to lift was corrosion bonding. Immediate corrective action taken was to replace the valves. Future corrective action planned is to modify the tail pipe drain configuration.
EVENT DESCRIPTION
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- 2. DOCKET NUMBER
- 3. LER NUMBER I YEAR SEQUENTIAL REV 00315 NUMBER NO.
l202sl -I 001 1-G On March 20, 2025, with Donald C. Cook Unit 1 in Mode 1 at 63% reactor power, the station commenced Technical Specification (TS) surveillance testing on the Unit 1 Main Steam Safety Valves (MSSVs) setpoints just prior to the Unit 1 Cycle 33 (U1 C33) Refueling Outage. There were no structures, systems, or components that were inoperable at the start of the event which contributed to the event.
On 3/20/25 1-SV-1 B-4, Unit 1 Steam Generator (SG) #4 Safety Valve 1 B, did not lift with 110% of set pressure applied with the trevi test equipment.
On 3/21/25, after Reactor Power was further reduced to 44.5% to support continued MSSV testing, 1-SV-2B-4, Unit 1 Steam Generator #4 Safety Valve 2B, did not lift with 110% of set pressure applied with the trevi test equipment.
Testing was discontinued in each occurrence and the valves were shipped to the vender for further inspection, testing, and refurbishment.
TS Limiting Condition For Operation (LCO) 3.7.1 requires a minimum of 20 MSSVs (5 per SG) to be operable in Modes 1, 2, and 3. On 3/21/25, since both valves exhibited similar discrepancies, these discrepancies are likely to have arisen over a period of time, per the guidance in NUREG-1022, and therefore are considered to have occurred earlier than the point of discovery during testing. Therefore the condition must be considered to have existed during plant operation and is reportable under 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition prohibited by the plant's Technical Specifications."
COMPONENT 1-SV-1B-4, S/G 1-OME-3-4 SAFETY VALVE 1B 1-SV-2B-4, S/G 1-OME-3-4 SAFETY VALVE 2B
CAUSE OF THE EVENT
A cause evaluation is currently in progress and has determined that the preliminary cause of the inability of 1-SV-1 B-4 and 1-SV-2B-4 to lift was corrosion bonding. Some corrosion from normal aging, wear, and minor periodic seat leakage is expected. The wear identified was much more than expected as other valves have been in service for a significant number of years between refurbishments with no significant corrosion identified. The outlet elbow and tailpipe drain pan drains were clogged for an extended period of time for the failed valves, which prevented condensed seat leakage draining from the valves when seat leakage occurred. This resulted in the valve bodies periodically filling with water and corroding at a faster rate than would be expected for a valve of a similar age with similar seat leakage history and clear drains. If the final evaluation results differ from this preliminary conclusion, then this report will be supplemented with the updated information.
CORRECTIVE ACTIONS
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- 2. DOCKET NUMBER
- 3. LER NUMBER 00315 I
YEAR SEQUENTIAL REV NUMBER NO.
~-I 005 1-G Both MSSVs were removed for refurbishment in U1C33 and replaced with pre-tested valves.
Future corrective action planned is to modify the tail pipe drain configuration to eliminate the potential for additional corrosion beyond normal aging caused by the clogged drains.
ASSESSMENT OF SAFETY CONSEQUENCES
PROBABILISTIC RISK ASSESSMENT (PRA)
The modeling of the MSSVs in the PRA was reviewed to estimate the potential impact of these failures on Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). The PRA modeled function was not impacted by these two failures of the MSSVs to open at their prescribed set points, therefore this condition did not result in a significant impact on the models' estimated CDF or LERF values. The MSSVs also perform a function to relieve secondary side pressure, which is not modeled in PRA. However, due to the amount of redundancy for this function, it's unlikely that the failure of these two MSSVs would have a significant impact on this unmodeled function. For this reason, it can be concluded this event is of very low significance.
NUCLEAR SAFETY There is an in-progress evaluation to determine past operation impacts on the safety analysis with the 1-SV-1 B-4 and 1-SV-2B-4 MSSVs on loop 4 both assumed to remain closed. With credit for better-estimate assumptions as necessary, including credit for steam relief through steam generator power operated relief valves, preliminary information from the in-progress evaluation supports the conclusion that all Cook Unit 1 Nuclear Plant UFSAR safety analyses would have met their applicable safety analysis limits. Therefore, there was no actual nuclear safety hazard. If the results of the final evaluation differ from these preliminary conclusions, then this licensee event report will be supplemented with the updated information.
INDUSTRIAL SAFETY There was no actual or potential industrial safety hazard resulting from this event.
RADIOLOGICAL SAFETY There was no actual or potential radiological safety hazard resulting from this event.
PREVIOUS SIMILAR EVENTS
A review of Licensee Event Reports for the past three years found no similar events. Page 3
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