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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 51702 | 30 January 2016 06:23:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unanticipated Sfrcs Actuation While Restoring Main Feedwater to Steam Generators | At 0123 EST, with the unit shutdown in Mode 3 (Hot Standby), during the performance of procedure DB-OP-06910, 'Trip Recovery,' while attempting to restore main feedwater to the Steam Generators, Davis-Besse received a Steam Feedwater Rupture Control System (SFRCS) 'reverse delta pressure' signal to the Auxiliary Feedwater System (AFW). The Auxiliary Feedwater System was operating at the time, feeding the Steam Generators. The SFRCS signal did result in actuation/closure (of) several valves in the Main Steam System, as the SFRCS signal is designed to do. This SFRCS signal/valve actuation was not anticipated. The unit remained in Mode 3 and is stable. This actuation did not have any negative impact to the AFW system and the ability to feed the steam generators. The NRC Resident Inspector has been notified of the event. | Steam Generator Feedwater Auxiliary Feedwater Main Steam | 05000346/LER-2016-002 |
| 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 51483 | 21 October 2015 00:24:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Due to Suspicious Vehicle in Owner Controlled Area | Security condition in the Owner Controlled Area outside of the Protected Area. (There is) an unknown vehicle located on the South side of the Intake Canal. The vehicle is locked, the engine is not running, and the parking lights are on. Security is performing an inspection of the vehicle for explosives or other contraband in conjunction with local law enforcement. The Unusual Event was declared based on EAL HU-1. The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies. Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA.
At 2229 EDT, the Unusual Event for a Security Condition at Davis-Besse Nuclear Power Station was terminated. An inspection of the vehicle in question was performed and it was determined that no threat existed to the site at any time. The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies. Notified R3DO (Daley), IRD (Stapleton), NRR (Morris) and ILTAB (Tucker). Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA. | ||
| 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 50263 | 8 July 2014 13:25:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Due to Smoke Alarm Inside Containment | Unusual Event declared at 0935 EDT on July 8, 2014 due to a single fire alarm in containment. ELA HU4. No abnormal condition or plant impact at this time. The plant remains stable at 100% power. A single smoke detector inside containment alarmed at 0925 EDT. There are no other indications of smoke or fire at this time. Adjacent smoke detectors are not in alarm. A containment entry team is being assembled to verify that there is no fire inside containment. The smoke detector alarm cleared at 1006 EDT. The licensee notified the NRC Resident Inspector. The licensee will make appropriate notifications to state and local government agencies. Notified DHS, FEMA, and NICC via email and phone. Notified FEMA NWC and Nuclear SSA via email.
The Unusual Event was terminated at 1328 EDT on 7/8/14. A containment entry was performed and no indications of a fire existed. The fire alarm was spurious and the detector has been disabled. There was no notification of other government agencies and there is no media /press release planned. The plant maintained stable operations at 100% power. The licensee notified the NRC Resident Inspector, the state of Ohio and local officials. Notified R3DO (Hills) NRR EO (McGinty), IRC MOC (Grant). Notified DHS, FEMA, and NICC via email and phone. Notified FEMA NWC and Nuclear SSA via email. | ||
| ENS 50086 | 5 May 2014 18:56:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram with Rod Motion While Shutdown | During planned testing of the Control Rod Drive (CRD) system, the reactor trip breakers were opened via the manual reactor trip push buttons to de-energize a CRD motor in response to a high temperature. The partially withdrawn control rods fully inserted and all other rods remained in their initial positions. This manual Reactor Protection System (RPS) actuation while the reactor was not critical is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector, the State of Ohio, and Ottawa and Lucas Counties. | Reactor Protection System Control Rod | 05000346/LER-2014-002 |
| ENS 50097 | 4 May 2014 12:32:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Initiation of the Reactor Protection System While Shutdown | On 5/4/14 while the plant was in Mode 3 and the reactor not critical, unexpected position indications were observed on a Control Rod while withdrawing an Axial Power Shaping Rod (APSR). Due to the uncertainty of rod positions, the APSR was inserted into the core. The reactor trip breakers were then opened from the Control Room using the manual trip pushbuttons. All Control and Safety Rods were unlatched and fully inserted into the reactor core before the reactor trip breakers were opened. This manual initiation of the Reactor Protection System with the reactor not critical is being reported per 10 CFR 50.72(b)(3)(iv)(A). The reportability of this event was determined based on an extent of condition review for Event Number 50086 that occurred 5/5/14. The failure to meet the 8-hour reporting requirement has been entered into the Corrective Action Program. The licensee notified the NRC Resident Inspector, the State of Ohio, and Ottawa and Lucas Counties. | Reactor Protection System Control Rod | 05000346/LER-2014-001 |
| 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 47443 | 16 November 2011 07:22:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Alert Due to Fire in Electrical Bus Affecting Safety Related Equipment | At 0222 EST on November 16, 2011, an ALERT was declared due to an electrical fire in the auxiliary building which houses safety related equipment. The apparent cause of the fire was due to an unknown source of water leaking on a breaker, thus causing an arc. The electrical fire is out. The plant was at 0% power and will remain shutdown in Mode 5. There was no impact on core cooling, or emergency power supplies. The licensee has notified the NRC Resident Inspector and state and local agencies.
At 0443 EST on November 16, 2011, Davis Besse, Unit 1 exited their ALERT. The electrical short affected the Control Room Emergency Ventilation Fan #1 Damper. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills) and Canada Nuclear Safety Commission (Jim Sandlef). | Control Room Emergency Ventilation | |
| ENS 46551 | 19 January 2011 07:32:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to a Fire and Explosions in the Protected Area | An electrical fire and explosions were reported near the Containment Access Facility construction area. An Unusual Event was declared based on HU4. Temporary electrical power at service disconnect DSLM3-3 was isolated. The fire was out at 0243 EST. The fire was extinguished using dry chemical. The fire was reported at 0232 EST on 1/19/11. The cause of the fire has not been determined at this time. The fire and explosions were initially reported by site security personnel. The licensee declared the NOUE at 0243 EST based on criteria HU4. The licensee initially called for offsite assistance in putting out the fire, however, the fire was extinguished by plant personnel and the offsite assistance was turned back. The licensee posted a reflash watch. The fire reportedly involved temporary cables and possibly a transformer supplying power to the construction area which is located inside the protected area outside the auxiliary building. The licensee notified the NRC Resident Inspector.
The licensee terminated the Notification of Unusual Event at 0358 EST. No additional information is available at this time. Notified R3DO (Bloomer), NRR EO (Skeen), IRD (Gott), DHS (Stringfield), and FEMA (Casto). | ||
| ENS 45162 | 25 June 2009 04:49:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Discovery of an After-The-Fact Alert Due to Catastrophic Failure of Ccpd | A transitory ALERT (condition was determined to have existed) based on Emergency Action Level 7.D.2 - 'Onsite Explosion Affecting Plant Operation'. At time 0049 on 06/25/09 a catastrophic failure-explosion of the Constant Current Potential Device (CCPD) on 'J' Bus near Air Circuit Breaker (ACB) 34563 resulted in a loss of switchyard 345 KV Bus 'J'. This event de-energized Startup Transformer 01 which is a tie from offsite sources to the Unit 13.8 KV Busses. The unit entered (Technical Specification) LCO 3.8.1, Condition A (due to the loss of one offsite power source). The Unit remains stable and in operation at 100% RTP (reactor thermal power). A problem solving decision making team is working on (the) troubleshooting/repair/restoration activities. This event did not impact any plant safety systems or result in any release of radioactive material. Failure of the CCPD caused automatic opening of the breakers on both sides of the 'J' bus which was configured as part of a switchyard ring bus at the time of the event. This resulted in the loss of one of the offsite power ties. However, startup Transformer 02 is still energized from offsite power and remains available for plant operations. Other than the de-energized startup transformer, onsite electrical configurations are normal including availability of emergency diesel generators. The licensee is in a 72 hour LCO per Tech Spec 3.8.1, Condition A, to restore the lost offsite power source. The licensee is inspecting the switchyard for collateral damage to other equipment from the failure of the CCPD. The licensee believes the CCPD failure is likely a result of equipment failure and not the result of any equipment tampering. The licensee stated that initially, the severity of the CCPD failure was not recognized because of the night time conditions and minimal lighting in the area. After daylight examination of the location of the event, it was determined that the failure of the CCPD should have been classified as an explosion affecting plant operation under EAL 7.D.2. Consequently, the licensee made the after-the-fact declaration. Licensee has notified the NRC Resident Inspector, and will be notifying State and local authorities. | Emergency Diesel Generator | |
| 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 41330 | 13 January 2005 14:49:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Entry Into Technical Specification Required Shut Down Due to Loss of Essential Bus | The following information was provided by the licensee via facsimile: At 0849 hours on January 13, 2005, during testing of Essential Bus D1 (Undervoltage Units monthly functional test) (the) Potential Transformer secondary fuse blew. This caused an Essential 4160 VAC Bus D1 undervoltage condition. Emergency Diesel Generator #2 auto started due to the undervoltage condition on Essential Bus D1. This auto start of EDG #2 is reportable IAW (in accordance with) 10CFR50.72(b)(3)(iv)(a). (The licensee entered) into Technical Specification 3.0.3 with a loss of Essential 4160 VAC Bus D1 (with) plant shutdown required within 1 hour if the Essential Bus and DC battery chargers 2P and 2N are not restored. At 0949 hours, (the licensee) commenced plant shutdown to comply with Technical Specification 3.0.3. The initiation of the plant shutdown is reportable IAW 10CFR50.72(b)(2)(i). (At) 1049 hours, (the license) re-energized Essential 4160 VAC Bus D1 and verified battery chargers 2P and 2N energized. (At)1051 hours, (the licensee) exited Technical Specification 3.0.3 and stopped plant shutdown. (The licensee also noted that) at 0855 hours, EDG #2 was shutdown when the #2 Service Water pump did not start. (At) 1144 hours, Service Water Pump #2 (was started and the licensee) declared Service Water Loop 2 OPERABLE. The license has notified the NRC Resident Inspector, will notify local government agencies, and expects to make a press release. | Service water Emergency Diesel Generator | 05000346/LER-2005-001 |
| 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 | |
| ENS 40208 | 1 October 2003 01:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Specified System Actuation | At 2134 EDT on 9/30/03, a reactor trip occurred on RPS (Reactor Protection System) shutdown high pressure. The plant was in Mode 3 performing a cooldown to Mode 4. RCS (Reactor Coolant System) pressure was 1788 psig on the cooldown pressure indicator. Group 1 safety rods were withdrawn prior to the trip. All other safety and regulating rods were already fully inserted. All systems performed as required, and the plant is stable in Mode 3. RCS pressure is 1755 psig, and RCS temperature is 526 degrees F. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Reactor Protection System | |
| ENS 40161 | 16 September 2003 04:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Coolant System Pressure Increase Caused a Valid Signal | At 1150 hours on 9/15/03 during Reactor Coolant System (RCS) pressure increase to approximately 700 psig, the Core Flood (CF) Tank 1 outlet motor-operated valve (CF1B) opened when the breaker for the valve was closed. Because RCS pressure was not high enough to actuate the pressure switch setpoint of 770 psig to open this valve, it was believed this was an invalid signal, and therefore the valve opening was not reportable under the criteria of 10 CFR50.72 After further review, at 0030 hours on 9/16/03 it was determined the sensed pressure was within the setpoint range of the switch, making this a valid signal to open the valve. The CF Tank pressure at the time (approximately 600 psig) was less than the RCS pressure of approximately 700 psig, therefore no discharge into the RCS occurred, and this event is not reportable per 10 CFR50.72(b)(2)(iv). The CF System is a passive system, but is used in conjunction with other Emergency Core Cooling Systems to mitigate significant events. Therefore, this event involving the opening of the CF Tank discharge valve CF1B is being conservatively reported in accordance with 10 CFR50.72(b(3)(2)(iv)(A) as a valid actuation of a train of the Emergency Core Cooling System. Upon opening of this valve, CF Tank 1 level and pressure were observed lowering with a corresponding rise in the Reactor Coolant Drain Tank, most likely due to leakage past a check valve. CF1B was closed according to procedures and the source breaker for the valve opened. The NRC Resident Inspector was notified by the licensee.
The actuation of the interlock for the Core Flood Tank 1 Outlet Isolation Valve did not constitute an actuation of the Core Flood System. The valve opened automatically as designed to ensure the Core Flood System was capable of performing its safety prior to being manually opened by the Operators. As part of the Emergency Core Cooling Systems, the primary function of the Core Flood System is to deliver cooling water to the reactor core in the event of a Loss of Cooling Accident per the Updated Safety Analysis Report. However, plant conditions (namely, Reactor Coolant System pressure lowering below Core Flood Tank pressure) did not exist that would have required the Core Flood System to perform its safety function. Therefore, the automatic opening of this valve does not constitute the actuation of the Emergency Core Cooling System, so this event is not reportable per 10CFR50.72(b)(3)(iv)(A). Therefore, NRC Event Number 40161 is retracted." R3DO (Anne Marie Stone) & NRR EO (Herb Berkow) notified. The NRC Resident Inspector will be informed of this retraction by the licensee. | Reactor Coolant System | |
| ENS 40070 | 14 August 2003 20:10:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to a Loss of Offsite Power. | All systems operated as expected. Decay Heat pumps are available if needed. NRC Resident Inspector was notified of this event by the licensee.
The plant exited the NOUE at 1940 on 08/15/03. The grid is stable, the plant is on off site power and the emergency diesels are in standby. Notified FEMA (Steindurf), Reg 3 RDO (S. Burgess) and EO (S. Richards) | 05000346/LER-2003-009 |