| | Site | Start date | Title | Description |
|---|
| 05000259/LER-2025-001, Reactor Scram Due to Low Reactor Water Level | Browns Ferry | 30 September 2025 | Reactor Scram Due to Low Reactor Water Level | | | ENS 57928 | Hatch | 13 September 2025 21:04:00 | Manual Reactor Scram | The following information was provided by the licensee via phone and email:
On September 13, 2025, at 1704 EDT, with Unit 2 in mode 1 at 70 percent power performing main turbine testing, the Unit 2 reactor was manually tripped due to loss of both reactor recirculation pumps. Due to the power level at the time, closure of containment isolation valves (CIVs) in multiple systems occurred, as a result of reaching the actuation setpoint on reactor water level, as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Normal reactor level and pressure control systems are controlling as expected. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 1 is not affected.
The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs.
There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
Main turbine control valve testing was in progress when the reactor recirculation pumps tripped. | | ML25253A274 | | 12 September 2025 | TLR-RES/DE/REB-2025-018 Assessment of Condition Monitoring Methods and Technologies for Inservice Inspection and Testing of Nuclear Power Plant Components | | | ML25238A063 | | 20 August 2025 | Transcript of Advisory Committee on Reactor Safeguards - BWRX-300 Design-Centered & Clinch River Subcommittee Meeting, August 20, 2025, Pages 1-188 (Open) | | | IR 05000373/2025002 | LaSalle | 11 August 2025 | County Station Integrated Inspection Report 05000373/2025002 and 05000374/2025002 | | | IR 05000397/2025002 | Columbia | 11 August 2025 | Integrated Inspection Report 05000397/2025002 | | | IR 05000440/2025002 | Perry | 7 August 2025 | Integrated Inspection Report 05000440/2025002 | | | IR 05000259/2025002 | Browns Ferry | 6 August 2025 | Integrated Inspection Report 05000259/2025002; 05000260/2025002, 050000296/2025002, and 07200052/2025001 | | | ENS 57847 | Browns Ferry | 2 August 2025 04:50:00 | Automatic Reactor Scram | The following information was provided by the licensee via phone and email:
On 8/1/2025, at 2350 CDT, Browns Ferry Unit 1 experienced an automatic scram due to low reactor water level. A low reactor water level of `+2' inches resulted in a valid actuation of the reactor protection system which caused all of the rods to insert. During the scram response, there was a valid actuation of the primary containment isolations systems groups 2, 3, 6 and 8. Upon receipt of these signals, all components actuated as required. Following the scram, reactor water level lowered below the '-45' inches setpoint, actuating high pressure coolant injection and reactor core isolation coolant and tripped both reactor recirculation pumps as required. Operations responded and stabilized the plant. Reactor water level is being maintained via the condensate system. Decay heat is being removed by bypass valves to the main condenser. There was no impact on Units 2 and 3.
This event requires a 4-hour non-emergency report per 10 CFR 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
This event requires a 4-hour non-emergency report per 10 CFR 50.72(b)(2)(iv)(A), any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
This event requires an 8-hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B): 1) Reactor protection system including: reactor scram or reactor trip. 2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).
All safety systems operated as expected. At no time were public health and safety at risk. The NRC Resident Inspector has been notified. |
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