Semantic search
Jump to navigation
Jump to search
| Start date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 57361 | 4 October 2024 02:35:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 2235 on 10/03/2024, the Vogtle 1 and 2 seismic monitoring panel experienced an electrical fault, rendering the panel nonfunctional. Compensatory measures for seismic event classification have been implemented in accordance with Vogtle procedures. This is an eight-hour, non-emergency notification for a loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the seismic monitoring panel is the method for evaluating that an operational basis earthquake (OBE) threshold has been exceeded following a seismic event. This is in accordance with Initiating condition `seismic event greater than OBE levels' and emergency action level HU2. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57343 | 27 September 2024 07:46:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone: At 0346 EDT on 9/27/24, with Unit 1 in mode 1 at 54 percent power, the reactor was manually tripped due to degrading condenser vacuum secondary to environmental conditions. The trip was not complex with all systems responding normally post trip. Closure of containment isolation valves (CIVs) in multiple systems occurred. Operations responded and stabilized the plant. The reactor protection system actuation while critical event is being reported as a 4-hour non-emergency notification per 10 CFR 52.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 52.72(b)(3)(iv)(a) as an event that results in a valid actuation of the 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: Decay heat removal is via steam bypass to the main condenser. Unit 2 was not affected. |
| ENS 57326 | 17 September 2024 05:27:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via phone and email: At 0127 EDT on 9/17/2024, with Unit 3 in mode 1 at 100% power, the reactor automatically tripped due to the passive residual heat removal heat exchanger outlet flow control valve failing open. A manual safeguards actuation was initiated due to the lowering pressurizer water level resulting from the reactor coolant system cooldown that was caused by the passive residual heat removal heat exchanger outlet flow control valve failing open. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected. Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). 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, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation. 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: The failure of the control valve does not inhibit the residual heat removal system from functioning as it is passive. The reactor coolant system maximum allowable cooldown rate was exceeded (Technical Specification 3.4.3). The limit is 100 degrees F per hour above 350 degrees F. The maximum observed cooldown rate was 226 degrees F per hour. At time 0458 EDT, reactor coolant system temperature is 369.1 degrees F, reactor pressure is 900 psig. Currently, the plant is cooling down and proceeding toward placing shutdown cooling online. |
| ENS 57291 | 26 August 2024 17:00:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: On 8/26/2024 at 1200 CDT, Farley Nuclear Plant Medical Services identified a false negative quality assurance test. (The contracted laboratory) was provided an adulterated sample of hydrocodone and hydromorphone that was part of a blind performance test. The results from the (contracted laboratory) returned a false negative. This false negative test result will be investigated, and the results reported as required. This event is being reported in accordance with 10 CFR 26.719(c)(3). The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officer Report Guidance: The contracted laboratory was a U.S. Department of Health and Human Service (HHS) certified laboratory.
The following update was provided by the licensee via email: Following further review of the event, it has been determined that this issue is not reportable under 10CFR26.719(c)(3) as the unsatisfactory test was not for a validity screening test. This event is reportable for testing errors in accordance with 10CFR26.719(c)(1) and a 30 day report will be submitted. Notified R2DO (Suber) and FFD Group (email). |
| ENS 57270 | 13 August 2024 15:49:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone: At 1200 EDT on August 13, 2024, with Unit 2 in Mode 1 at 100 percent power, Vogtle Unit 2 declared an ALERT per emergency action level (EAL) SA9 due to a fire that caused visible damage to a safety system component needed for the current operating mode. At 1151 EDT, the fire was extinguished. The equipment affected was the safety-related regulating 480V transformer which supplies power to the Unit 2 'B' engineered safety features chiller. There was no impact to the safety and health of the public or plant personnel. Units 1, 3, and 4 are unaffected. State and local officials were notified. The NRC resident inspector was notified. The NRC decided to remain in the Normal mode of operation at 1234 EDT. Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, EPA Emergency Ops Center, USDA Watch Officer, FDA Emergency Ops Center (email), FEMA NWC (email), DHS Nuclear SSA (email), DHS NRCC THD Desk (email), FEMA NRCC SASC (email), FERC RMC (email), CWMD Watch Desk (email). The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The fire alarm was received at 1145 EDT. A fire was confirmed at 1149 EDT. The switchgear was de-energized and a fire extinguisher was used to put out the fire.
The licensee terminated the ALERT emergency action level at 1436 EDT. Notified R2DO (Lopez-Santiago), IR MOC (Grant), NRR EO (Felts), DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, EPA Emergency Ops Center, USDA Watch Officer, FDA Emergency Ops Center (email), FEMA NWC (email), DHS Nuclear SSA (email), DHS NRCC THD Desk (email), FEMA NRCC SASC (email), FERC RMC (email), CWMD Watch Desk (email). |
| ENS 57215 | 9 July 2024 01:25:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 2125 EDT on 07/08/2024, with Unit 3 in Mode 1 at 100 percent power, the reactor was manually tripped due to main feedwater pump `A' miniflow valve failing open, which resulted in lowering steam generator water level. Additionally, an automatic safeguards actuation occurred due to the cooldown of the reactor coolant system. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected. Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). 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, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation. 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: The cause of the main feedwater pump 'A' miniflow valve failing open was unknown and under investigation at the time of the notification of this event to the NRC. |
| ENS 57187 | 22 June 2024 11:28:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0728 EDT on 06/22/2024, with Unit 2 in Mode 3 at zero percent power and the reactor trip breakers closed, a manual actuation of the RPS was initiated during the withdrawal of the shutdown rods in preparation for Mode 2. This was procedurally directed due to a shutdown rod being misaligned from the other rods in the bank due to a malfunction. Units 1, 3 and 4 were not affected. Due to the manual actuation of the RPS, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 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 has been notified. |
| ENS 57164 | 5 June 2024 20:48:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 1648 EDT on 06/05/2024, with Unit 4 in Mode 1 at approximately 35 percent power, the reactor was manually tripped due to unexpected response of the turbine run back circuitry following a trip of main feedwater pump `C. The trip was not complex, with all safety 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 and main feedwater is supplying the steam generators. Units 1, 2 and 3 are not affected. 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. |
| ENS 57021 | 11 March 2024 17:37:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: On March 11, 2024, at 1337 EDT, with Unit 1 in Mode 1 at 35 percent power performing power ascension activities, the reactor was manually tripped due to the 'A' reactor feed pump (RFP) tripping on low suction pressure. Due to the power level at the time, the 'B' RFP had not been placed in service. Closure of containment isolation valves (CIVs) in multiple systems and actuation of high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) 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. The 'B' RFP was placed in service and is controlling reactor water level. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 2 is not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, 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, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. 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: The cause of the 'A' RFP is under investigation. The reactor electric plant remains in a normal lineup with both emergency diesel generators available. There were no temperature or pressure technical specification limits approached. |
| ENS 56975 | 18 February 2024 16:33:00 | Hatch | NRC Region 2 | GE-4 | The following is a synopsis of information was provided by the licensee via email and phone call: A non-licensed supervisor had a confirmed positive during a random fitness for duty test. The supervisor's access to the plant has been terminated. |
| ENS 56971 | 16 February 2024 11:34:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0048 CST on February 16, 2024, with Unit 2 in mode 1 at 100 percent power, the reactor was manually tripped due to a loss of 2A 125V DC distribution panel. The trip was complex due to the loss of components associated with A-train DC power. Operations responded and stabilized the plant. Decay heat is being removed by the atmospheric relief valves. Unit 1 is not affected. An automatic actuation of the auxiliary feedwater system (AFW) occurred due to low-low steam generator levels. The AFW auto-start is an expected response with low-low steam generator levels from the reactor trip. AFW is still currently controlling steam generator levels. 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 AFW System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56959 | 11 February 2024 15:11:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1011 EST on 02/11/2024, during a refueling outage at 0 percent power, while performing local leakage rate testing (LLRT) of the feedwater check valves (part of the containment boundary), it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56896 | 18 December 2023 07:23:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee email: At 0223 EST, on 12/18/2023, while Unit 2 was at 100 percent power in mode 1, the high pressure coolant injection (HPCI) outboard steam isolation valve closed resulting in the HPCI system being declared inoperable. The cause of the outboard steam isolation valve closing is under investigation. HPCI does not have a redundant system, therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The safety function was restored at 0512, on 12/18/23, and HPCI has been declared operable. Reactor core isolation cooling (RCIC) and low pressure emergency core cooling systems (ECCS) were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56852 | 14 November 2023 16:41:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1041 CST on 11/14/23 with Farley Unit 2 in Mode 1 at 10 percent power, the reactor was manually tripped due to rising steam generator levels. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Auxiliary feedwater (AFW) was manually initiated in accordance with plant procedures and is feeding the steam generators. Heat removal is being provided via the atmospheric relief valves. Unit 1 is not affected. 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 rods fully inserted. The licensee attempted to take manual control of the feedwater control valves to lower steam generator level but, due to reaching a steam generator level that requires a manual trip, the licensee manually tripped the reactor. |
| ENS 56897 | 2 November 2023 01:11:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email and phone: At 2011 EDT on 11/01/23, with Unit 2 in Mode 3 at 0 percent power, Unit 2 received multiple spurious actuations. These actuations consisted of a partial group 1 and a partial group 5 primary containment isolation and a partial secondary containment isolation. The partial Group 1 isolation resulted in the closure of two main steam isolation valves (MSIVs); all other MSIVs were already closed. The partial group 5 isolation auto closed one of the reactor water cleanup (RWCU) isolation valves. The partial secondary containment isolation resulted in the closure of the inboard refueling floor and reactor building secondary containment isolation valves (SCIVs). Additionally, at 2238 EDT, Unit 2 again received multiple spurious actuations. These actuations consisted of a partial group 5 primary containment isolation and a partial secondary containment isolation. The partial group 5 isolation auto closed one of the RWCU isolation valves The partial secondary containment isolation resulted in the closure of the inboard refueling floor and reactor building SCIVs. And again, at 2354 EDT, Unit 2 received spurious actuations which consisted of a partial secondary containment isolation which resulted in the closure of the inboard refueling floor and reactor building SCIVs. The spurious actuations seen on 11/1/23 are triggered at -35 inches reactor water level (RWL) for group 5 and secondary containment isolations and at -101 inches RWL for group 1 isolations. It was determined that a combination of the RWL fluctuating above and below the wide range instrument reference leg tap, the reactor vessel pressure being lowered, and reactor core isolation cooling introducing colder water conditions near the reference leg tap of the wide range instrument caused the spurious actuations. Using multiple RWL indications for each of the instances mentioned above, the actuations were confirmed to be spurious as RWL was being controlled in a band of +55 inches to +85 inches at the time of the actuations. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of a partial group 1, a partial group 5, and partial secondary containment logic. The NRC Resident has been notified. |
| ENS 56826 | 1 November 2023 10:48:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0648 EDT on 11/1/23, with Unit 2 in MODE 1 at 56 percent power, the reactor was manually tripped due to a trip of the 'B' reactor feed pump (RFP). The 'A' RFP had been previously isolated due to a leak. Closure of containment isolation valves (CIVs) in multiple systems and the actuation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) 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. Reactor water level is being maintained with RCIC. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 was not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, 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, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. There was no impact on the health and safety of the public or plant personnel. The Resident Inspector was notified. |
| ENS 56784 | 10 October 2023 00:10:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: A non-licensed employee supervisor failed a test specified by the fitness for duty testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56694 | 23 August 2023 13:39:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: A non-licensed contract supervisor failed a test specified by the FFD testing program. The employee's access to the plant has been terminated. The NRC Resident Inspectors have been notified |
| ENS 56692 | 22 August 2023 21:24:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 1724 EDT, on August 22, 2023, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to a failure of the non-safety heater drain pump 'B' and the failure of the non-safety condensate pump 'A' to automatically or manually start. At 1735 EDT, a fire was identified on heater drain pump 'B' and was extinguished by the onsite fire brigade at 1807 EDT. Operations responded and stabilized the plant. The trip was not complex, with all safety systems responding normally post-trip. Decay heat is being removed by the main steam system to the main condenser using the steam dumps. There was no impact to Units 2, 3, or 4. An automatic actuation of the auxiliary feedwater system (AFW) also occurred, as expected, due to lo-lo steam generator levels resulting from the reactor trip. AFW is currently controlling all steam generator levels at their normal levels. 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). Due to the notification of another government agency, the Burke County Fire Department, this event is being reported as a four-hour, non-emergency notification under 10 CFR 50.72(b)(2)(xi). The Burke County Fire Department was not needed to extinguish the fire. 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 resulted 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 56779 | 7 August 2023 18:39:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via phone and email: At 1439 EDT on August 7, 2023, a spurious level spike on the unit 4 reactor coolant system (RCS) level instrument (4-RCS-LT160A, 'Hot Leg 1 Level') caused actuation of containment isolation, reactor trip, automatic depressurization system (ADS) stage 4, and in containment refueling water storage tank (IRWST) isolation signals. The spurious level changes caused an invalid signal based on the incidental response of the 4-RCS-LT160A instrumentation due to water spray that was being used for reactor vessel cleaning (being performed prior to initial fuel loading). The level fluctuations resulted in engineered safety features actuation signals (containment isolation, ADS stage 4, and IRWST isolation signals) and a reactor trip signal, with the reactor trip signal already present. Three containment isolation valves closed due to the containment isolation signal that was generated. These valves were: 4-CAS-V014, 'instrument air supply containment isolation, air-operated valve,' 4-SFS-V034, 'spent fuel pool cooling system suction header containment isolation, motor-operated valve,' and 4-SFS-V035, 'spent fuel pool cooling system suction header containment isolation, motor-operated valve.' The other automatic containment isolation valves were either already closed at the time of the event or properly removed from service. All affected equipment functioned properly. The other actuation signals that were observed during this event (ADS stage 4, IRWST isolation, and reactor trip) did not result in any equipment changing position or automatically operating (i.e., the actuation signals occurred while the systems were properly removed from service). Units 1, 2, and 3 were not affected. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified. |
| ENS 56648 | 31 July 2023 19:06:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 1506 EDT on July 31, 2023, it was determined that a contractor supervisor failed a test specified by the fitness for duty testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56632 | 21 July 2023 15:48:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 1148 EDT on 07/21/2023, with Unit 3 in Mode 1 at 32 percent power, the reactor automatically tripped on low reactor coolant pump (RCP) speed due to decaying RCP motor voltage during power ascension testing. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are not affected. 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. |
| ENS 56614 | 9 July 2023 17:28:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 1328 EDT on 07/09/2023, with Unit 3 in Mode 1 at 45 percent power, the reactor automatically tripped during power ascension testing due to low reactor coolant flow from decaying voltage to the reactor coolant pumps. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are not affected. 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. |
| ENS 56497 | 2 May 2023 08:23:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0423 EDT on 05/02/2023, with Unit 3 in Mode 1 at 14 percent power, the reactor was manually tripped due to securing all main feed pumps, due to sudden high differential pressure on their suction strainers. The trip was not complex, with all safety-related systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the reactor trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dumps, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 were not affected. 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. |
| ENS 56460 | 10 April 2023 04:48:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0048 EDT on 4/10/2023, with Unit 3 in Mode 1 at 18 percent power, the reactor automatically tripped due to low reactor coolant flow due to voltage decaying to the reactor coolant pumps during main generator testing activities. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to atmosphere. Units 1, 2, and 4 are not affected. 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. |
| ENS 56446 | 31 March 2023 18:32:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1432 EDT on 03/31/23, with Unit 2 in mode 1 at 97 percent power, the reactor was manually tripped due to a loss of both recirculation pumps. The cause of the recirculation pump trips is under investigation. Additionally, closure of CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. 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). It is also reportable under 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. |
| ENS 56415 | 16 March 2023 12:45:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0845 EDT on March 16, 2023, it was determined that a contract employee supervisor failed a for-cause FFD test. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56414 | 16 March 2023 01:57:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 2157 EDT on 03/15/2023, with Unit 3 in Mode 1 at 18 percent power, the reactor automatically tripped due to the loss of two reactor coolant pumps when their electrical buses failed to transfer after a main generator excitation protective relay tripped. Operations responded and stabilized the plant. Decay heat is being removed by steam generator power operated relief valves. Units 1, 2, and 4 are not affected. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, nonemergency 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. |
| ENS 56367 | 16 February 2023 12:43:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0743 EST on February 16, 2023, it was determined that a contract supervisor failed a random fitness-for-duty (FFD) test. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56342 | 7 February 2023 22:38:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1738 EST on 02/07/2023, while in mode 5 at 0 percent power, it was determined during local leak rate testing (LLRT) that the primary containment leakage rate exceeded the allowable limit defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.' Both primary containment isolation valves in a penetration failed LLRT requirements which represents a failure to maintain primary containment integrity. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56329 | 1 February 2023 15:56:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0956 CST with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The trip was not complex with all safety related systems responding normally post-trip. During the trip, the non safety related '1A' 4160V bus lost power resulting in the loss of one Reactor Coolant Pump (RCP-1A). Operations responded and stabilized the plant. The '1A' 4160V bus was re-energized at 1031 CST. Decay heat is being removed by steam dumps to the main condenser. Farley Unit 2 is not affected. An automatic actuation of (Auxiliary Feedwater) AFW 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 report 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 56311 | 14 January 2023 12:21:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0721 EST on 01/14/2023, with Unit 3 in Mode 3 at 0 percent power and reactor trip breakers open, a manual actuation of the RPS occurred while conducting pre-criticality testing. The RPS manual actuation was procedurally driven in response to low gland steam pressure, resulting in the necessity to break condenser vacuum following a trip of the auxiliary boiler. The reactor trip breakers were in an open state at the time of the event. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56294 | 4 January 2023 03:59:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2259 EST on 1/3/2023, with Unit 1 and Unit 2 in Mode 1 at 100 percent power, an actuation of the Unit 1 B and Unit 2 A emergency diesel generator (EDG) systems, as well as an actuation of the associated auxiliary feedwater (AFW) systems on each unit occurred. The reason for the EDG auto-starts was due to a loss of an offsite power source (loss of one of the two reserve auxiliary transformers (RAT) on each unit) to the Unit 1 B and Unit 2 A safety related buses. The EDG and AFW systems automatically started as designed when the valid undervoltage signal on the affected safety related bus was received. This 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 the Emergency Diesel Generator and the Auxiliary Feedwater Systems for both Unit 1 and Unit 2. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56231 | 17 November 2022 15:09:00 | Vogtle | NRC Region 2 | W-AP1000 | A contract supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56175 | 23 October 2022 08:05:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0405 EDT on 10/23/2022, with (Vogtle) Unit 3 in Mode 6 and the reactor subcritical for greater than 28 hours, it was discovered that all three required flow paths for the stage four ADS were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The diverse actuation system was operable for manual stage four ADS during this time period. At 0432 EDT on 10/23/2022, two of the three required flow paths were restored to operable status, which exited the reportable condition. All required flow paths were operable at 0447 EDT. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56147 | 6 October 2022 06:44:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via email: At 0244 EDT on 10/06/2022, with Unit 3 Defueled at 0 percent power, an actuation of the RPS occurred during restoration of Division B Class 1E DC and uninterruptible power supply system. The reason for the RPS actuation was due to the opening of the Division B passive residual heat removal (PRHR) heat exchanger outlet flow control valve. The reactor trip breakers were in an open state at the time of the event when the RPS signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56112 | 17 September 2022 03:57:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the Southern Nuclear Company via email: At 2257 EDT on 09/16/2022, it was determined that there was a programmatic vulnerability of the Fleet FFD program. Specifically, it was determined that some individuals were not placed into the follow-up pool for additional screening when required by the program. All identified personnel were in the random FFD pool, and were subject to the behavioral observation program. This is reportable in accordance with 10CFR26.719(b)(4) for all Units and 10CFR26.417(b)(1) for Vogtle Units 3&4. The NRC Resident Inspectors have been notified. See EN#s 56113, 56114, and 56115. |
| ENS 56114 | 17 September 2022 02:57:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by Southern Nuclear Company via email: At 2257 EDT on 09/16/2022, it was determined that there was a programmatic vulnerability of the Fleet FFD program. Specifically, it was determined that some individuals were not placed into the follow-up pool for additional screening when required by the program. All identified personnel were in the random FFD pool, and were subject to the behavioral observation program. This is reportable in accordance with 10CFR26.719(b)(4) for all Units and 10CFR26.417(b)(1) for Vogtle Units 3 and 4. The NRC Resident Inspectors have been notified. See EN#s 56112, 56113, and 56115. |
| ENS 56113 | 17 September 2022 02:57:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the Southern Nuclear Company via email: At 2257 EDT on 09/16/2022, it was determined that there was a programmatic vulnerability of the Fleet FFD program. Specifically, it was determined that some individuals were not placed into the follow-up pool for additional screening when required by the program. All identified personnel were in the random FFD pool, and were subject to the behavioral observation program. This is reportable in accordance with 10CFR26.719(b)(4) for all Units and 10CFR26.417(b)(1) for Vogtle Units 3 and 4. The NRC Resident Inspectors have been notified. See EN#s 56112, 56114, and 56115. |
| ENS 56115 | 17 September 2022 02:57:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the Southern Nuclear Company via email: At 2257 EDT on 09/16/2022, it was determined that there was a programmatic vulnerability of the Fleet FFD program. Specifically, it was determined that some individuals were not placed into the follow-up pool for additional screening when required by the program. All identified personnel were in the random FFD pool, and were subject to the behavioral observation program. This is reportable in accordance with 10CFR26.719(b)(4) for all Units and 10CFR26.417(b)(1) for Vogtle Units 3 and 4. The NRC Resident Inspectors have been notified. See EN#s 56112, 56113, and 56114. |
| ENS 56028 | 3 August 2022 17:58:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 1258 CDT on August 3, 2022, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to the supply breakers of the 1B startup transformer opening. The fast dead bus transfer for the reactor coolant pumps did not occur during the event. Currently the plant is in Mode 3 on natural circulation. Operations responded and stabilized the plant. Decay heat is being removed by steaming with atmospheric relief valves. Unit 2 is not affected. An automatic actuation of the 1B diesel occurred because of the power loss to the 1G 4160V bus. Additionally, the actuation of motor driven and turbine driven auxiliary feedwater pumps (AFW) also occurred. AFW auto-start is an expected response from this 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 1B diesel and 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 55875 | 3 May 2022 19:41:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 1541 EDT on May 3, 2022, with Unit 1 in Mode 1 at 100 power, the reactor was manually tripped due to the loss of one of the main feed pumps. The trip was not complex, with all systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the scram. Operations responded and stabilized the plant. Decay heat is being removed by Auxiliary Feedwater through the steam dumps to the condenser. Unit 2 is not affected. An automatic actuation of the Auxiliary Feedwater System (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, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour, nonemergency 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 55735 | 7 February 2022 14:14:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via telefone: A non-licensed contractor superintendent had a confirmed positive for alcohol during a for-cause fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 55679 | 29 December 2021 20:52:00 | Hatch | NRC Region 2 | GE-4 | This following information was conveyed by the licensee via phone and email: At 1552 EST on 12/29/21, with Unit 1 in Mode 1 at 90 percent power, the reactor was manually tripped due to reactor pressure perturbations. The cause of the reactor pressure perturbations is under investigation. Additionally, closure of (containment isolation valves) CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not affected. 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). It is also reportable under 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. |
| ENS 55552 | 29 October 2021 14:10:00 | Vogtle | NRC Region 2 | W-AP1000 | At 1010 EDT on October 29, 2021, Southern Nuclear Operating Company (SNC) determined a contractor supervisor attempted to subvert a random Fitness-for-Duty (FFD) test. The employee's badge has been confiscated. The NRC Resident Inspector has been notified. |
| ENS 55490 | 24 September 2021 05:00:00 | Farley | NRC Region 1 | Based upon a 10 CFR 21.21(b) transfer notification from General Electric Hitachi (GEH), Southern Nuclear Operating Company's (SNC) Joseph M. Farley Nuclear Plant (FNP) has determined there is evidence a Substantial Safety Hazard could have been created by the failure of the rivets installed on certain EC Trip Units if they were left uncorrected. These EC Trip Units are a subcomponent of all five (5) emergency diesel generator control panel supply breakers at FNP. This defect was identified and the components were repaired by GEH before being installed in the plant. These defective EC Trip Units never posed a challenge to the safe operation of FNP. The NRC Senior Resident Inspector at FNP has been notified. | |
| ENS 55462 | 15 September 2021 11:58:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | At 0658 CDT on 09/15/2021 a non work-related death occurred of a site employee. The individual was outside of the Radiological Controlled Area. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. |
| ENS 55448 | 8 September 2021 05:59:00 | Hatch | NRC Region 2 | GE-4 | At 0159 EDT on 09/08/2021, the HPCI pump discharge valve failed to reopen during a valve surveillance, resulting in the HPCI system being declared INOPERABLE. HPCI does not have a redundant system; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). Reactor Core Isolation Cooling system and low pressure Emergency Core Cooling Systems were OPERABLE during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55420 | 20 August 2021 12:43:00 | Hatch | NRC Region 2 | GE-4 | A licensed operator failed a pre-access authorization test specified by the FFD testing program test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 55394 | 3 August 2021 14:26:00 | Hatch | NRC Region 2 | GE-4 | At 1026 EDT on 8/3/21, with Unit 1 in MODE 1 at 100 percent power, the reactor automatically tripped due to low reactor water level. The low reactor water level condition was due to a loss of both reactor feed pumps. The cause of the loss of feed pumps is under investigation. Additionally, the low reactor water level resulted in the automatic actuation of High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, and Containment Isolation Valves (CIVs) in multiple systems. All safety systems responded normally. Operations responded and stabilized the plant. Reactor water level is being maintained via RCIC system. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not affected. 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). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the HPCI and RCIC systems and CIVs. There was no impact on the health and safety of the public or plant. The Licensee notified the NRC Resident Inspector. The Unit will proceed to Mode 4 while the cause of the loss of feed pumps is under investigation. |