| | Site | Start date | Title | Description |
|---|
| ENS 57769 | Farley | 20 June 2025 01:27:00 | Automatic Reactor Trip and Automatic Actuation of Auxiliary Feedwater System | The following information was provided by the licensee via phone and email:
On June 19, 2025, at 2027 CDT, with Unit 2 in mode 1 at 100 percent power, the reactor automatically tripped due to an `A' steam generator (SG) water level low signal. The low level in the SG was caused by a feedwater control system malfunction. All safety related systems responded normally post-trip. Operations responded and stabilized the plant.
Decay heat is being removed by steam dumps to the main condenser. Farley Unit 1 is not affected.
An automatic actuation of auxiliary feedwater system also occurred, which is an expected response from the reactor trip.
Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and 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 the auxiliary feedwater system.
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | | ENS 57763 | Grand Gulf | 17 June 2025 09:26:00 | Manual Reactor Scram | The following information was provided by the licensee via phone and email:
On June 17, 2025, at 0426 CDT, Grand Gulf Nuclear Station (GGNS) was operating at 70 percent power when a manual scram was initiated due to degraded main condenser vacuum caused by the loss of the 'A' circulating water pump.
All control rods fully inserted, there were no complications, and all plant systems responded as designed. Immediately after the scram, an expected reactor water Level 3 isolation signal was received. Reactor pressure is being maintained via the turbine bypass valves. Reactor level is being maintained via main feedwater. GGNS is currently in Mode 3. No radiological releases have occurred due to the event.
The cause of the circulating water pump trip is under investigation.
The manual reactor protection system actuation is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B). The expected reactor water Level 3 isolation signal is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A).
The NRC Resident Inspector was notified. | | ENS 57737 | Browns Ferry | 31 May 2025 03:47:00 | Manual Reactor Trip | The following information was provided by the licensee via phone and email:
At 2247 CDT on 5/30/25, with Unit 2 operating in mode 1 at 39 percent reactor power, the reactor was manually tripped due to a trip of the only operating reactor recirculation pump (2B). Approximately 44 minutes prior at 2203 CDT, the 2A reactor recirculation pump tripped.
Operations responded and stabilized the plant. Primary containment isolation systems (PCIS) received an actuation signal for groups 2, 3, 6 and 8 on reactor water level at +2 inches. All primary containment systems that received an actuation signal performed as designed. All other systems functioned as designed. Reactor water level control is via condensate and feedwater, and reactor cooldown is in progress using turbine bypass valves to the main condenser.
Due to the reactor protection system (RPS) actuation while critical, this event is being reported as a four-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The actuation of RPS and PCIS also requires an eight-hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A).
There was no impact to 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:
The site reduced power from 100 percent following the loss of the 2A recirculation pump. | | JAFP-25-0034, License Amendment Request for Proposed Changes to the Technical Specification Primary Containment Isolation Instrumentation Tables and New Main Steam Tunnel Area Temperature Technical Specification 3.7.8 | FitzPatrick | 29 May 2025 | License Amendment Request for Proposed Changes to the Technical Specification Primary Containment Isolation Instrumentation Tables and New Main Steam Tunnel Area Temperature Technical Specification 3.7.8 | | | ENS 57734 | Vogtle | 28 May 2025 16:53:00 | Manual Reactor Trip and Auxiliary Feedwater Actuation | The following information was provided by the licensee via phone and email:
At 1253 EDT, on May 28, 2025, with Unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to a loss of main feedwater pump `A'. The trip was not complex, with all systems responding normally post-trip.
Operations responded and stabilized the plant. Decay heat is being removed by discharging steam through the steam dumps to the main condenser. Units 2, 3, and 4 are not affected. An automatic actuation of auxiliary feedwater (AFW) also occurred. The AFW auto-start is an expected response from the reactor trip.
Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also 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 the auxiliary feedwater system.
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:
All control rods inserted on the trip. A main feedwater pump lube oil evolution was in progress at the time of the trip. | | ML25085A416 | 05200050 | 22 May 2025 | SDAA - FSER Chapter 10 - Steam and Power Conversion System | | | ML25094A183 | 05200050 | 22 May 2025 | SDAA - FSER Appendix C - Abbreviations | | | ML25085A417 | 05200050 | 22 May 2025 | SDAA - FSER Chapter 11 - Radioactive Waste Management | | | ML25129A077 | Grand Gulf | 13 May 2025 | Integrated Inspection Report 05000416.2025001 | | | ENS 57705 | Browns Ferry | 12 May 2025 18:53:00 | Automatic Reactor Trip | The following information was provided by the licensee via phone and email:
At 1353 CDT on May 12, 2025, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of the electrohydraulic control (EHC) system. The trip was not complex, with all systems responding normally post-trip with the exception of the turbine bypass valves due to the loss of EHC.
Operations responded and stabilized the plant. At 1407 with Unit 2 in Mode 3, there was a second automatic reactor trip due to a low reactor water level transient caused by manually opening and closing a main steam relief valve. Reactor water level is being maintained via feed water pump. Decay heat is being removed by discharging steam (via main steam line drains) to the main condenser. Units 1 and 3 are not affected.
Due to the reactor protection system (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B).
Additionally, due to the actuation of RPS and groups 2, 3, 6, and 8 of the primary containment isolation system, this event is being reported as an eight-hour, non-emergency notification per 50.72(b)(3)(iv)(A).
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified. |
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