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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55346 | 9 July 2021 01:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Main Turbine Trip | At 2154 EDT on 7/8/2021, with the Unit in Mode 1 at 100% power, the reactor automatically tripped due to trip of the main turbine, caused by failure of a non-safety related breaker during functional testing. Following the reactor trip the Steam Feed Rupture Control System automatically initiated on low Steam Generator 1 level, actuating both turbine-driven Auxiliary Feedwater Pumps. The operators subsequently started the high pressure injection pumps manually per procedure in response to overcooling indications. Operations responded and stabilized the plant. Decay heat was initially being removed via the Main Condenser. During post-trip response actions, while attempting to shut down the Auxiliary Feedwater Pumps, a low pressure condition was experienced in Steam Generator 2, resulting in isolation of the Main Condenser and steam being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage. 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 in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a four-hour, non-emergency notification of emergency core cooling system (ECCS) discharge into the reactor coolant system, and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an eight-hour, non-emergency notification of 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. | Steam Generator Reactor Coolant System Reactor Protection System Auxiliary Feedwater Main Turbine Emergency Core Cooling System Decay Heat Removal Main Condenser | |
| ENS 54611 | 25 March 2020 16:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Loss of Reactor Coolant Pumps | At 1240 (EDT) on March 25, 2020, with the Unit in Mode 2 at approximately 0% (zero percent) power starting up from a refueling outage, the reactor was manually tripped due to a trip of two of four Reactor Coolant Pumps. 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 to the main condenser. The cause of the Reactor Coolant Pump trips is under investigation. 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Protection System Main Condenser | |
| ENS 54263 | 7 September 2019 17:09:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | At 1309 EDT on September 7, 2019, with the unit in Mode 1 at approximately 95 percent power, the reactor automatically tripped during main turbine valve testing. The trip was not complex with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the turbine bypass valves discharging steam to the main condenser. 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). The cause of the reactor protection system actuation is under evaluation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Protection System Main Turbine Main Condenser | |
| ENS 52232 | 10 September 2016 07:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Unit Trip Due to Main Generator Lock-Out | At 0343 EDT, with the unit operating at approximately 100% full power, an automatic reactor trip occurred due to a Main Generator lock-out. The cause of the generator lock-out is being investigated at this time. All control rods fully inserted. Post trip, the Steam Feedwater Rupture Control System was actuated due to high Steam Generator 1 level. The cause of the high Steam Generator 1 level is being investigated at this time. The unit is currently in Mode 3 (Hot Standby) and stable, at approximately 550 degrees F and 2155 psig. Steam is being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage, and all safety systems functioned as expected. The NRC Resident Inspector has been notified of the event. The licensee notified the State of Ohio, Ottawa and Lucas County. | Steam Generator Feedwater Decay Heat Removal Control Rod | 05000346/LER-2016-009 |
| ENS 51696 | 29 January 2016 18:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Reactor Protection System Actuation | At 1322 EST, with the unit operating at approximately 100% full power, an automatic reactor trip occurred due to actuation of Reactor Protection System (RPS) Channel 4. The cause of the RPS actuation is being investigated at this time. Nuclear Instrumentation calibration for RPS Channel 2 was in progress at the time of the trip, with Channel 2 in bypass and Channel 1 in trip. All control rods fully inserted. Immediately post trip, the Steam Feedwater Rupture Control System actuated due to high Steam Generator 1 level due to unknown causes. The Main Steam Isolation Valves closed and Auxiliary Feedwater started as expected. Secondary side relief valves lifted in response to the trip, with two of the relief valves (one on each header) not properly reseating until operators manually lowered Main Steam Header pressure. The Bayshore 345 kV Offsite Electrical Distribution Circuit automatically isolated at the time of the unit trip. This was unexpected. The remaining offsite circuits remain in service. The unit is currently in Mode 3 (Hot Standby) and stable, at approximately 550 degrees F and 2155 psig. Steam is being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage, and all safety systems functioned as expected. Both primary Source Range nuclear instruments automatically energized, however, they were previously declared inoperable due to an administrative issue. Both Source Range instruments are functional and indicating properly. Both alternate Source Range instruments are operable, and all required Technical Specification actions have been completed. The NRC Resident Inspector has been notified of the event. | Steam Generator Feedwater Reactor Protection System Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Control Rod Main Steam | 05000346/LER-2016-001 |
| ENS 51061 | 9 May 2015 22:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to Steam Leak in the Turbine Building | At 1855 EDT, a steam leak from the #1 moisture separator reheater in the turbine building was reported to the control room. Operators performed a rapid down power to approximately 30% at which time the reactor was manually tripped. At 1910 EDT an Unusual Event was declared. The steam feed rupture control system was manually initiated (this includes actuation of both turbine-driven Auxiliary Feedwater Pumps) and the steam leak was isolated. Station air compressor #2 (non-safety related) tripped. Station air compressor #1 automatically started. The unit is currently in mode 3 (Hot Standby) and stable. Steam is being discharged through the atmospheric dumps as a means of decay heat removal. There is no known primary to secondary leakage. All systems functioned as expected. There were no reported injuries and personnel accountability is in progress. The licensee notified state and local agencies and informed the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, NICC Watch Officer and FEMA NWC and NuclearSSA via email.
The licensee exited the Unusual Event at 2121 EDT based on the following: At 2121 hours EDT, the Unusual Event at the Davis-Besse Nuclear Power Station was terminated. The steam leak has been isolated and plant conditions are stable. Cooling continues to be maintained via the auxiliary feedwater system. The initiation of auxiliary feedwater at the start of the event is reportable as a Specified System Actuation per 10CFR50.72(b)(3)(iv)(A). The licensee notified state and local agencies and informed the NRC Resident Inspector. Notified R3DO (Skokowski), NRR EO (Morris) and IRD (Grant). Notified DHS SWO, FEMA Ops Center, NICC Watch Officer and FEMA NWC and NuclearSSA via email. | Auxiliary Feedwater Decay Heat Removal | 05000346/LER-2015-002 |
| ENS 49159 | 30 June 2013 01:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Reactor Coolant Pump Trip | Automatic trip of Reactor Coolant Pump 1-2 due to an electrical differential current fault resulted in an RPS actuation on Flux/Delta Flux/Flow. Startup Feedwater Valve 1 did not respond as expected post-trip and has been placed in manual control. All secondary side steam reliefs initially re-seated following reactor trip. Subsequent Main Steam Line #1 Safety Valve leakage mitigated during post-trip recovery actions. All other systems have functioned as expected. The plant is stable in Mode 3 - Hot Standby. All rods inserted into the core during the trip. Decay heat is being removed via turbine bypass valves to the main condenser with normal feedwater to the steam generators. The plant is in its normal shutdown electrical lineup. The licensee characterized the trip as uncomplicated. The licensee will be notifying Lucas and Ottawa counties, the State of Ohio and will be issuing a press release. They have notified the NRC Resident Inspector. | Steam Generator Feedwater Main Steam Line Main Condenser | |
| ENS 42828 | 6 September 2006 06:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip on Loss of Condenser Vacuum | Observed degrading condenser pressure. Entered abnormal procedure DB-OP-02518, High Condenser Pressure and reduced reactor power. At <280 Mwe and > 5 inches Hg (mercury) A (absolute) , manually tripped the reactor at approximately 45% power in accordance with procedure. Normal post-trip response. Condenser pressure is slowly recovering. Still trying to determine the source of the condenser air in-leakage. Notified Ottawa County Sheriff of main steam safety / atmospheric vent valve operation at 0231 hours per procedure. All control rods fully inserted on the trip. Decay heat is being removed using the turbine bypass valves and the motor driven feed pump. There is no steam generator tube leakage. The atmospheric vent valves / main steam safety valves lifted for a few seconds following the trip and fully reseated after the initial lifting. Plant electrical power if from the grid backfeeding to the station. The electric grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Main Steam Safety Valve Control Rod Main Steam | |
| ENS 40921 | 4 August 2004 14:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unexpected Reactor Trip During Maintenance Activities | At 1024 EDT a reactor trip occurred during maintenance activities involving the control rod drive trip breakers. All control rods fully inserted. The unit is currently stable with decay heat removal via the main steam system through the turbine bypass valves to the main condenser. Post trip response was normal with the following exceptions noted: 1. #4 main turbine stop valve may not have fully closed. 2. A #2 Steam Generator Safety Valve may be lifting early (lifted at 1010 psi rather than the 1050 psi setpoint). 3. Turbine Bypass valve SP13A3 stuck slightly open - isolated. All offsite power lines have been verified operable and both EDGs are available in standby, if needed. The licensee informed the local sheriff's department as required whenever a Steam Generator Safety valve lifts and the NRC Resident Inspector. | Steam Generator Main Turbine Decay Heat Removal Main Condenser Control Rod Main Steam |