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| Event date | Region | State | Site | Reactor type | Event description | |
|---|---|---|---|---|---|---|
| ENS 58224 | 31 March 2026 11:50:00 | NRC Region 2 | Georgia | Vogtle | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0750 EDT on March 31, 2026, the Vogtle 1 and 2 seismic monitoring panel was discovered (to be) nonfunctional while performing operator rounds in the control room. 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 58218 | 26 March 2026 09:00:00 | NRC Region 2 | Georgia | Vogtle | W-AP1000 | The following information was provided by the licensee via phone and email: At 0500 EDT on March 26, 2026, the presence of a prohibited substance was discovered within the protected area. Local law enforcement has been contacted. The NRC Resident Inspector has been notified. |
| ENS 58215 | 24 March 2026 21:47:00 | NRC Region 2 | North Carolina | McGuire | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On March 21, 2026, at approximately 0945 EDT, with Unit 2 in mode 4 during a planned refueling outage, personnel discovered boric acid on the mirror insulation on top of the `2B' reactor coolant system (RCS) crossover leg piping. On March 24, 2026, at 1747 EDT, it was identified to be a through-wall flaw on a three-quarter inch instrument line. The leakage is minor in nature and unquantifiable. The leakage is coming from a welded connection upstream of a low pressure flow transmitter root-valve connecting to the `2B' RCS crossover elbow. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 58181 | 2 March 2026 07:59:00 | NRC Region 2 | North Carolina | Brunswick | GE-4 | The following information was provided by the licensee via phone and email: At 0259 EST, on 03/02/2026, during a refueling outage at zero percent power while performing local leak rate testing on the reactor core isolation cooling (RCIC) isolation valves, which are 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 58177 | 26 February 2026 08:03:00 | NRC Region 2 | Georgia | Hatch | GE-4 | The following information was provided by the licensee via phone and email: At 0303 EST on 02/26/26, with Unit 1 in mode 2 at 1 percent power, the reactor was manually tripped as required by procedure due to balance of plant equipment not responding as expected. 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 the control rod drive (CRD) and reactor water cleanup (RWCU) systems. Decay heat is being removed by the RWCU system. Unit 2 is not impacted. 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 58166 | 14 February 2026 12:28:00 | NRC Region 2 | Tennessee Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: On February 14, 2026, at 0628 CST, it was determined that a licensed operator failed a test specified by the fitness for duty (FFD) testing program. The individual's authorization for site access has been terminated at all Tennessee Valley Authority facilities. The NRC Resident Inspector has been notified. |
| ENS 58152 | 4 February 2026 23:52:00 | NRC Region 2 | North Carolina | Harris | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1659 EST, offsite medical personnel (arrived on site and) declared a site employee deceased due to a non-work-related medical event inside the radiation control area. The individual was not contaminated. A notification to the Occupational Health and Safety Administration was made in accordance with regulatory requirements at 1852 EST. This is a four-hour non-emergency notification for notification of another government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. |
| ENS 58235 | 4 February 2026 04:15:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On February 3, 2026, Unit 3 experienced a loss of 'A' reactor protection system (RPS). The 3A RPS motor generator (MG) set was found tripped and coasting down. This condition resulted in a half-scram on channel 'A' as well as primary containment isolation system (PCIS) group 2, 3, 6, and 8 isolations. RPS 'A' was placed on alternate in accordance with 3-AOI-99-1. All systems responded as expected. Plant conditions which initiate PCIS group 2 and 8 actuations are reactor vessel low water level and high drywell pressure. Plant conditions which initiate PCIS group 3 actuations are reactor vessel low water level and reactor water cleanup area high temperature. Plant conditions which initiate PCIS group 6 actuations are reactor vessel low water level, high drywell pressure, or reactor building ventilation exhaust high radiation (reactor zone or refuel zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. Upon investigation, a conductor was found broken inside the crimp of a ring lug, most likely due to overtightening and high cyclic fatigue. The lug was on the conductor between the contacts of the thermal overload relays. This opened the circuit to the 1K relay, and the motor starter, which was the cause of the loss of the MG set. The lug was repaired, the condition was cleared, and all systems were realigned as necessary. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the corrective action program as condition report 2065520. The NRC Resident Inspector has been notified of this event. |
| ENS 58141 | 31 January 2026 02:55:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 2155 EST, on January 30, 2026, with unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to a steam leak in the feedwater heater system. 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 emergency feedwater system. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) resulting from a valid actuation of the reactor protection and emergency feedwater systems. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 58122 | 18 January 2026 20:45:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: At 1445 CST on 1/18/2026, Browns Ferry Unit 1 high pressure coolant injection (HPCI) was declared inoperable due to 1-FCV-073-0006A, (HPCI steam line condensate inboard drain valve), and 1-FCV-073-0006B, (HCPI steam line condensate inboard drain valve), being found closed. This condition is being reported as an 8-hour non-emergency notification per 10-CFR 50.72(b)(3)(v). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The HCPI steam line condensate inboard and outboard drain valves are air operated solenoid valves that failed closed. The cause is under investigation and restoration to the normal open position is currently in process. Technical Specification 3.3.5.1 Condition C was entered with a 14-day limiting condition for operation (LCO). |
| ENS 58115 | 14 January 2026 12:44:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 0744 EST on January 14, 2026, it was determined that an individual in a non-licensed supervisory role failed a test specified by the FFD testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 58101 | 29 December 2025 17:29:00 | NRC Region 2 | North Carolina | McGuire | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1229 EST on 12/29/2025, with Unit 1 in mode 1 at approximately 100 percent power, the reactor automatically tripped due to a lockout of the main generator. The trip was not complex, with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater pumps due to anticipation of automatic actuation and has stabilized Unit 1. Decay heat is being removed by the condenser. Unit 2 was not affected. Due to the reactor protection system actuation while critical and actuation of the motor driven auxiliary feedwater pumps, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8-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 58042 | 17 November 2025 14:11:00 | NRC Region 2 | Tennessee | Sequoyah | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0911 EST, on 11/17/2025, it was discovered that both trains of the control room emergency ventilation system were simultaneously inoperable due to an unauthorized door breach of a door in the control room envelope; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). At 0913 EST, the door was closed, and both trains of the control room emergency ventilation system were restored to operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via phone and email: Based on further engineering analysis, sufficient time margin existed to allow the control room emergency ventilation system (CREVS) to successfully mitigate the impacts of an accident. The duration of the breach was less than the analyzed time for accident conditions to impact the control room envelope. Operations has revised and retracted the technical specification entries after evaluating that CREVS remained operable during the breach event. Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(v)(D) report was not required and NRC event notification 58042 is retracted. No licensee event report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The NRC Resident Inspector has been notified. Notified R2DO (Nielsen) |
| ENS 58041 | 15 November 2025 03:54:00 | NRC Region 2 | South Carolina | Oconee | B&W-L-LP | The following information was provided by the licensee via phone and email: At 2254 EST, on November 14, 2025, with Unit 2 in no mode at zero percent power, an actuation of the emergency AC power system occurred during verification of electrical system alignment. The reason for the Keowee Hydro units actuating was due to loss of power to the Unit 2 main feeder busses when a potential transformer drawer for the Unit 2 main feeder busses was opened, resulting in breakers supplying power to the Unit 2 main feeder busses opening. The Keowee Hydro Units 1 and 2 automatically started as designed when a main feeder bus undervoltage signal was received. There was no impact to Unit 1 or Unit 3. This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the emergency AC power system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 58020 | 3 November 2025 17:15:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On November 3, 2025, at 1215 EST, the 'A' reserve station service transformer (RSST) pilot wire lockout actuated while restoring the 'A' RSST to service. This resulted in the electrical isolation of the 'A' RSST and the Unit 1 'J' emergency bus. The #3 EDG automatically started and loaded onto the Unit 1 'J' emergency bus, as designed. Operations entered the appropriate abnormal procedures and ensured stable conditions. All safety systems functioned as designed and all electrical parameters remained stable. No radiological consequences resulted from this event. This event is being reported pursuant to 10CFR50.72(b)(3)(iv)(A) due to actuation of the #3 EDG. The NRC Resident Inspector was notified. |
| ENS 58104 | 2 November 2025 14:15:00 | NRC Region 2 | Georgia | Hatch | GE-4 | The following information was provided by the licensee via phone and email: On November 2, 2025, at 0915 EST, and on November 3, 2025 at 0203 EST, while (Hatch) Unit 1 was at 100-percent power, the normal supply breaker from the `1D' 600V switchgear to the 120/208V essential transformer `1C' tripped on overcurrent due to a ground resulting in the loss of the `1B' essential bus and the `1B' instrument bus. The loss of power impacted the logic of the primary and secondary containment isolation systems resulting in the invalid actuation of containment isolation valves (CIVs) in multiple systems. The ground within the `1D' 600V switchgear was cleared by racking (out) the associated breaker and by replacing a degraded wire. A future action includes performing a detailed cubical and bus bar inspection of the `1D' 600V switchgear. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) because it was not part of a pre-planned sequence and resulted in the invalid, partial actuation of CIVs in both the primary containment and secondary containment systems, with all systems responding normally. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 58010 | 27 October 2025 18:51:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 0920 EDT on 10/28/25, the supervisor of nuclear site safety contacted the area director of OSHA (Occupational Safety and Health Administration) to notify them of a work-related injury which resulted in the employee being admitted to a hospital. The individual was not contaminated and was transported offsite to Chippenham Hospital in Richmond, VA. This was a 24-hour notification in accordance with 29 CFR 1904.8. The NRC Resident Inspector has been notified. |
| ENS 57997 | 22 October 2025 01:08:00 | NRC Region 2 | Virginia | North Anna | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email and phone: On October 21, 2025, at 2108 EDT, Unit 1 automatically tripped from 74 percent power due to a negative rate trip. The unit has been stabilized in mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated, and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater pumps actuated as designed because of the reactor trip and is reportable per 10 CFR 50.72(b)(3)(iv)(A) for a valid engineered safety feature (ESF) actuation. Decay heat is being removed via the steam generator power-operated relief valves and Unit 1 is in a normal shutdown electrical lineup. Unit 2 was not affected by this event. The NRC Resident Inspector has been notified. |
| ENS 58087 | 21 October 2025 15:01:00 | NRC Region 2 | North Carolina | Harris | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a licensee event report submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 1001 EDT on October 21, 2025, an invalid actuation of the 'B' train motor-driven auxiliary feedwater (MDAFW) pump occurred with Harris Nuclear Plant while in mode 6. After investigation (it was determined that) the pump automatic start signal was from the anticipated transient without scram mitigation system actuation circuitry (AMSAC). Restoration from functional testing of AMSAC was in progress at the time of the pump start. Separate surveillance testing of the 'B' train emergency safeguards sequencer was being prepared for and enabled the 'B' train MDAFW pump start while restoring 'B' MDAFW pump control power prior to the event. The 'B' train MDAFW pump started and ran with no abnormalities noted. The overlapping testing activities allowed for the AMSAC start signal to be sent to the 'B' MDAFW pump. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. This event did not impact the health and safety of the public. The NRC Resident Inspector has been notified. |
| ENS 57987 | 14 October 2025 19:52:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1552 EDT on October 14, 2025, 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 systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the emergency feedwater system. Due to the reactor protection system actuation while critical, this event is being reported as a 4-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for valid actuation of the reactor protection and emergency feedwater systems. 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: A generator field breaker lockout occurred. In addition, a small fire was observed in the generator field breaker and extinguished in less than 15 minutes by the onsite fire brigade. The unit is stable in hot standby with emergency feedwater and main condenser steam dumps in-service. |
| ENS 57983 | 13 October 2025 17:11:00 | NRC Region 2 | South Carolina | Oconee | B&W-L-LP | The following information was provided by the licensee via phone and email: At 1311 EDT on October 13, 2025, it was determined that a non-licensed supervisor had a confirmed positive test as specified by the FFD testing program. The individual's authorization for site access has been terminated at all Duke Energy facilities. The NRC Resident Inspectors have been notified. |
| ENS 57977 | 10 October 2025 05:00:00 | NRC Region 2 | North Carolina | McGuire | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0100 EDT, on 10/10/2025, with McGuire Unit 2 in mode 1 at 5 percent power, the reactor was manually tripped due to an approximate 4 gpm leak associated with the chemical and volume control system. The motor driven auxiliary feedwater pumps were manually started due to anticipation of automatic actuation. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical and manual start of the motor driven auxiliary feedwater pumps, 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 system actuation per 50.72(b)(3)(iv)(A). At 0211 on 10/10/25, it was determined the required train of the refueling water storage tank was inoperable due to insufficient borated water volume; therefore, this condition is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The safety function was restored at 0249 on 10/10/2025 and the required train has been declared operable. There was no impact to the other unit. 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 is being removed via steam dump valves to the main condenser, and the auxiliary feedwater system. The cause of the leak is being investigated. |
| ENS 57968 | 6 October 2025 10:27:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On October 6, 2025, at 0627 EDT, a notification to the Occupational Safety and Health Administration (OSHA) was initiated due to a Dominion employee experiencing a non-work-related medical event that resulted in the employee passing. When the issue was identified, the station first aid team responded to administer first aid. Upon arrival, the employee was nonresponsive with no pulse. The employee was pronounced deceased on site at 0627 EDT. A report to OSHA will be made in accordance with federal requirements. This event is reportable to the NRC per 10 CFR 50.72(b)(2)(xi) since another government agency will be notified of this fatality. The employee was in the plant protected area and was not contaminated. The NRC Resident Inspector has been notified. |
| ENS 58069 | 3 October 2025 14:48:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: As allowed by 10 CFR 50.73(a)(1), VC Summer Nuclear Station (VCSNS) is making a 60-day notification of an invalid actuation under 10 CFR 50.73(a)(2)(iv)(A). At 0948 EDT, on October 3, 2025, VCSNS commenced a shutdown to repair a steam line leak. On October 4 at 2116 EDT, VCSNS was in mode 3, with the 'A' motor-driven emergency feedwater (EFW) pump providing decay heat removal. Main feedwater (MFW) had been secured per standard operating procedures. The deaerator storage tank (DAST) experienced a level decrease due to secondary-side valve leak-by. When the DAST level reached low-low level, an automatic signal was generated to secure the already secured MFW Pumps, which automatically sent an actuation signal to EFW. The low-low DAST signal is intended to secure MFW pumps to prevent damage due to a potential loss of net positive suction head at lower DAST levels. This sequence resulted in the invalid actuation of the previously secured 'B' EFW pump. This actuation was not in response to an actual engineered safety feature (ESF) condition and does not meet the criteria for a valid ESF actuation. This event is being reported as a 60-day report in accordance with 10 CFR 50.73(a)(1) for an invalid actuation of the EFW System. The NRC Senior Resident Inspector has been notified. |
| ENS 57942 | 22 September 2025 19:00:00 | NRC Region 2 | Tennessee | Watts Bar | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: "On 9/22/25, Watts Bar Nuclear (WBN) Operations was informed that a WBN licensed operator had tested positive for a controlled substance during a pre-screening test to regain unescorted access, in violation of the Tennessee Valley Authority (TVA) fitness for duty policy. "A pre-access screening was completed on 9/16/25 to regain unescorted access. The results were sent to the TVA medical review officer on 9/22/25. The test was declared positive for a controlled substance and WBN Operations was notified at 1500 EDT on 9/22/25. "The individual's unescorted access remains revoked. "The NRC Resident Inspector has been notified. |
| ENS 57928 | 13 September 2025 21:04:00 | NRC Region 2 | Georgia | Hatch | GE-4 | The following information was provided by the licensee via phone and email: On September 13, 2025, at 1704 EDT, with Unit 2 in mode 1 at 70 percent power performing main turbine testing, the Unit 2 reactor was manually tripped due to loss of both reactor recirculation pumps. Due to the power level at the time, closure of containment isolation valves (CIVs) in multiple systems occurred, as a result of reaching the actuation setpoint on reactor water level, as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Normal reactor level and pressure control systems are controlling as expected. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 1 is not affected. The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Main turbine control valve testing was in progress when the reactor recirculation pumps tripped. |
| ENS 57917 | 10 September 2025 19:40:00 | NRC Region 2 | South Carolina | Oconee | B&W-L-LP | The following information was provided by the licensee via phone and email: At 1207 EDT on September 10, 2025, Oconee Nuclear Station Unit 1 declared an unusual event (Event number 57914). A small reactor coolant system (RCS) leak occurred during routine RCS letdown filter maintenance. The letdown filters were bypassed, and the leak was isolated. The unusual event was terminated at 1353 EDT. This notification is being made solely as a four-hour, non-emergency notification for a news release. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57914 | 10 September 2025 16:07:00 | NRC Region 2 | South Carolina | Oconee | B&W-L-LP | The following information was provided by the licensee via phone and email: At 1207 EDT on September 10, 2025, Oconee Nuclear Station Unit 1 declared an Unusual Event (SU5.1) due to identified reactor coolant system (RCS) leakage from Unit 1 for greater than 15 minutes. The calculated leak rate was 26 gallons per minute (gpm). The leakage was identified to be from letdown filter 1A, which was subsequently isolated. Units 2 and 3 remain at 100 percent power and are unaffected by this event. This event was terminated at 1353 after isolating the source of the leak. The NRC Resident Inspector has been notified. Notified DHS SWO, FEMA Operations Center, CISA Central Watch Officer, FEMA NWC, DHS Nuclear SSA (email), CWMD Watch Desk (email). |
| ENS 57908 | 8 September 2025 17:35:00 | NRC Region 2 | Tennessee | Sequoyah | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On 09/08/2025 at 1335 EDT, a Sequoyah Nuclear Plant contract 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 57906 | 5 September 2025 13:09:00 | NRC Region 2 | South Carolina | Catawba | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0909 EDT on September 5, 2025, with the unit shut down for a scheduled refueling outage it was determined the reactor coolant system barrier had a through wall flaw with leakage. The leakage is minor in nature and unquantifiable. The leakage is coming from a welded connection upstream of a 1-inch vent valve on the chemical and volume control system letdown line. The leak was isolated in accordance with plant technical specifications. 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 57893 | 29 August 2025 16:04:00 | NRC Region 2 | Tennessee | Sequoyah | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On August 29, 2025, Sequoyah Nuclear (SQN) operations was informed that a SQN licensed (employee) had tested positive for alcohol, in violation of the Tennessee Valley Authority fitness for duty policy. The individual's access has been placed on a 14-day denial. The NRC Resident Inspector has been notified. |
| ENS 57859 | 11 August 2025 06:12:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On August 11, 2025, at 0212 EDT, VC Summer Unit 1 was in Mode 1 at 100 percent power when the 'A' emergency diesel generator (XEG0001A) automatically actuated in response to undervoltage indications on Bus `1 DA'. Bus `1 DA' remained energized from its normal offsite power source. The XEG0001A was secured at 0302 EDT and realigned for auto start. Dominion Energy South Carolina (DESC) is actively investigating to determine the cause of the transient at this time. This event is being reported as eight-hour notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the 'A' emergency diesel generator. The NRC Resident Inspector was notified. |
| ENS 57857 | 9 August 2025 15:13:00 | NRC Region 2 | Florida | Turkey Point | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1113 EDT on August 9, 2025, with Unit 3 in mode 1 at 100 percent power, the reactor automatically tripped due to a turbine control system failure. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in mode 3. Decay heat is being removed by steam generators. Unit 4 is not affected. This event is being reported pursuant to 10CFR50.72(b)(2)(iv)(B). An actuation of the auxiliary feed water system (AFW) occurred during the Unit 3 reactor trip. The AFW system automatically started as designed. This event is being reported pursuant to 10CFR50.72(b)(3)(iv)(A). 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 turbine control system failure is under investigation. |
| ENS 57854 | 6 August 2025 22:47:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: Surry Unit 1 reactor automatically tripped at 1847 EDT on August 6, 2025, due to a spurious actuation of high consequence limiting safeguards train `B' (due to a false high containment pressure trip signal). Reactor coolant temperature is being maintained at 547 degrees Fahrenheit on the main steam dumps with main feedwater supplying the steam generators. All systems operated as required. The trip was uncomplicated and all control rods fully inserted into the core. Reactor protection system, emergency core cooling system, auxiliary feedwater system, emergency diesel generators, phase I and phase II containment isolation signals all actuated as designed. Offsite power remains available. There is no impact to Surry Unit 2. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of reactor protection system activation and 10 CFR 50.72(b)(3)(iv)(A) for 8-hour notification of specified system actuation. The NRC Resident Inspector has been notified via cell phone. |
| ENS 57847 | 2 August 2025 04:50:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: On 8/1/2025, at 2350 CDT, Browns Ferry Unit 1 experienced an automatic scram due to low reactor water level. A low reactor water level of `+2' inches resulted in a valid actuation of the reactor protection system which caused all of the rods to insert. During the scram response, there was a valid actuation of the primary containment isolations systems groups 2, 3, 6 and 8. Upon receipt of these signals, all components actuated as required. Following the scram, reactor water level lowered below the '-45' inches setpoint, actuating high pressure coolant injection and reactor core isolation coolant and tripped both reactor recirculation pumps as required. Operations responded and stabilized the plant. Reactor water level is being maintained via the condensate system. Decay heat is being removed by bypass valves to the main condenser. There was no impact on Units 2 and 3. This event requires a 4-hour non-emergency report per 10 CFR 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event requires a 4-hour non-emergency report per 10 CFR 50.72(b)(2)(iv)(A), any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event requires an 8-hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B): 1) Reactor protection system including: reactor scram or reactor trip. 2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). All safety systems operated as expected. At no time were public health and safety at risk. The NRC Resident Inspector has been notified. |
| ENS 57833 | 28 July 2025 17:33:00 | NRC Region 2 | Tennessee | Watts Bar | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On July 28, 2025, Watts Bar Nuclear (WBN) operations was informed that a WBN licensed reactor operator had tested positive for a controlled substance, in violation of the Tennessee Valley Authority (TVA) fitness for duty policy. A random screening was completed on July 17, 2025. The results were sent to the TVA medical review officer on July 28, 2025. The test was declared positive for a controlled substance and WBN operations was notified at 1333 EDT on July 28, 2025. The individual's unescorted access has been revoked. The NRC Resident Inspector has been notified. |
| ENS 57816 | 17 July 2025 15:00:00 | NRC Region 2 | Virginia | North Anna | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1100 EDT on 07/17/2025, North Anna Power Station reported elevated levels of tritium in a ground water monitoring well to the State of Virginia as a non-voluntary reporting of tritium. An investigation is currently ongoing to identify the cause of the elevated tritium levels. The tritium levels in this location do not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification of the other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57810 | 13 July 2025 18:18:00 | NRC Region 2 | Tennessee | Watts Bar | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1418 EDT on 7/13/2025, with Unit 2 in mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complicated with all systems responding normally post trip. Operators responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dump system and the auxiliary feedwater (AFW) system. 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). The expected actuation of the AFW system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the main turbine trip was due to the loss of both main feed pumps due to a loss of secondary power. The cause of the loss of secondary power is still being investigated. |
| ENS 57800 | 3 July 2025 00:22:00 | NRC Region 2 | Alabama | Farley | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1922 CDT on 07/02/2025, the Farley Unit 1 and Unit 2 seismic monitoring panel experienced a fault, rendering the panel nonfunctional. Compensatory measures for seismic event classification have been implemented in accordance with Farley procedures. This report is being submitted as 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 if an operational basis earthquake (OBE) threshold has been exceeded following a seismic event. This is in accordance with the initiating condition for a 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. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee confirmed alternative means of recognizing a seismic event for emergency plan entry are available.
The following information was provided by the licensee via phone and email: It was determined through event investigation that two additional unplanned events occurred within a 3-year period. The seismic computer experienced an unplanned event on 9/29/24 at 1306 CDT and was restored on 9/30/24 at 1742. The other unplanned event occurred on 3/19/25 at 1120 and was restored to service on 3/24/25 at 1626. The NRC Resident Inspector has been notified. Notified R2DO (Davis) |
| ENS 57789 | 28 June 2025 03:09:00 | NRC Region 2 | Alabama | Farley | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On June 27, 2025, at 2209 CDT, with Unit 2 in mode 1 at 100 percent power, the reactor was manually tripped due to degrading condenser vacuum. The cause of the low vacuum is under investigation. All safety related systems responded normally post trip. Decay heat is being removed by atmospheric relief valves. Farley Unit 1 is not affected. An automatic actuation of auxiliary feedwater system occurred, which is an expected response from the reactor trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been informed. |
| ENS 57779 | 24 June 2025 07:51:00 | NRC Region 2 | Tennessee | Sequoyah | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0351 EDT on 6/24/25, Sequoyah Unit 2 operators performed a manual reactor trip. All safety systems responded normally, and the plant is currently stable in mode 3 (hot standby) at normal operating temperature and pressure. Unit 2 was manually tripped due to lowering level in the loop 1 steam generator caused by main feedwater regulating valve malfunction. 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 in accordance with 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 57778 | 24 June 2025 04:14:00 | NRC Region 2 | Tennessee | Sequoyah | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0014 EDT on 6/24/2025, Sequoyah Unit 1 operators performed a manual reactor trip. All safety systems responded normally, and the plant is currently stable in mode 3 (hot standby) at normal operating temperature and pressure. Unit 1 was manually tripped due to lowering steam generator water levels caused by a secondary control system malfunction. Unit 2 is not impacted and remains at 100 percent power in mode 1. 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 in accordance with 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 57774 | 23 June 2025 02:04:00 | NRC Region 2 | Virginia | North Anna | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 2204 EDT on 6/22/2025, with unit 1 and unit 2 operating at 100 percent power, the 'A' reserve station service transformer (RSST) was lost which resulted in the loss of power to the '1J' emergency bus. As a result of the power loss, the '1J' emergency diesel generator (EDG) automatically started as designed and restored power to the '1J' emergency bus. During the event, the unit 1 'A' charging pump (1-CH-P-1A) automatically started as designed due to the loss of power event. Security footage reported a bright white flash associated with the 'A' RSST at the time of the event. The valid actuation of these ESF components due to loss of electrical power is reportable per 10 CFR 50.72(b)(3)(iv)(A). The '1J' emergency bus off-site power source was restored to service via an alternate source and the '1J' EDG was secured and returned to automatic. Restoration of offsite power to operable is complete. 1-CH-P-1A was secured and returned to automatic. Both units are currently stable, and an investigation is underway to determine the cause of the loss of the 'A' RSST." Unit 2 was unaffected. The NRC resident inspector was notified. |
| ENS 57773 | 21 June 2025 18:38:00 | NRC Region 2 | Florida | Turkey Point | Westinghouse PWR 3-Loop | The following information is a summary of the information provided by the licensee via phone and email: On 6/21/2025, at 1438 EDT, Turkey Point experienced an unplanned reactor trip and a spurious safety injection signal when the '4A' 4 kV bus locked out. An Unusual Event, SU8.1, was declared at 1453 EDT due to two open, motor-operated steam generator sample containment isolation valves not closing on the phase 'A' containment isolation signal due to the loss of power. There were no abnormal parameters that would require a safety injection signal. Turkey Point unit 3 was unaffected and remains at 100 percent power. Turkey Point unit 4 is stable in mode 3. The state and one county were notified, and the other county will be notified. The NRC Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee manually closed the valves due to loss of power to the bus, and they are investigating the cause of the reactor trip.
Turkey Point unit 4 exited the Unusual Event at 1609 EDT. Notified R2RA (Lara), NRR (Bowman), NSIR (Erlanger), R2DO (Blamey), NRR EO (Felts), IR MOC (Whited), R2 PAO (Gasperson). Notified DHS SWO, FEMA Operations Center, CISA Central Watch Officer, FEMA NWC, DHS Nuclear SSA (email), CWMD Watch Desk (email).
The following information was provided by the licensee via phone and email: On June 21, 2025, at 1438 EDT, while Turkey Point Unit 4 was in mode 1 at 100 percent power, the reactor automatically tripped due to lockout of the '4A' 4 kV bus. The cause of the bus lockout is unknown. The trip was complicated with all systems responding normally post-trip, except for containment isolation valves powered from the '4A' 4 kV bus. Operations stabilized the plant in mode 3. Decay heat is removed by discharging steam from the steam generators to atmosphere. Unit 3 is not affected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B). An actuation of the auxiliary feedwater (AFW) system occurred during the reactor trip and safety injection signal. The AFW pumps automatically started as designed when the low steam generator, safety injection, and '4A' 4 kV bus undervoltage signals were received. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified." Notified R2DO (Blamey). |
| ENS 57769 | 20 June 2025 01:27:00 | NRC Region 2 | Alabama | Farley | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On June 19, 2025, at 2027 CDT, with Unit 2 in mode 1 at 100 percent power, the reactor automatically tripped due to an `A' steam generator (SG) water level low signal. The low level in the SG was caused by a feedwater control system malfunction. All safety related systems responded normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by steam dumps to the main condenser. Farley Unit 1 is not affected. An automatic actuation of auxiliary feedwater system also occurred, which is an expected response from the reactor trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57850 | 10 June 2025 23:10:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On June 10, 2025, Unit 2 operations personnel received an `A' channel half scram and entered 2-AOI-99-1. Motor generator set 2A was shut down and reactor protection system `A' was swapped to alternate. This resulted in primary containment isolation system (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of standby gas treatment (SGT) trains `A', `B', and `C' and control room emergency ventilation system (CREV) train `A'. All affected safety systems responded as expected. Plant conditions which initiate PCIS groups 2 and 8 actuations are reactor vessel low water level and high drywell pressure. Plant conditions which initiate PCIS group 3 actuations, are reactor vessel low water level and reactor water cleanup area high temperature. Plant conditions which initiate PCIS group 6, CREV and SGT actuations, are reactor vessel low water level, high drywell pressure, or reactor building ventilation exhaust high radiation (reactor zone or refuel zone). At the time of the event, these conditions did not exist; the actuation was due to a loss of power and not due to a low reactor water level or drywell pressure. Therefore, the actuation of the PCIS, CREV, and SGT was invalid. Upon investigation, the 2A2 circuit protector was found to have charred wire on the top right lug. The terminal block was replaced. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the corrective action program as condition report 2019406. |
| ENS 57752 | 10 June 2025 16:30:00 | NRC Region 2 | Florida | Saint Lucie | CE | The following information is a summary provided by the licensee via phone and email: At 1230 EDT on June 10, 2025, a non-licensed supervisor failed a fitness for duty test. The NRC Resident Inspector has been notified. |
| ENS 57751 | 4 June 2025 18:43:00 | NRC Region 2 | Tennessee Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: On June 4, 2025, the Tennessee Valley Authority (TVA) determined there are manufacturing non-conformances associated with the stem failure on a 10-inch, Class 900 Anchor Darling double-disc gate valve, used as a high pressure coolant injection system (HPCI) isolation valve in Browns Ferry Nuclear Plant, Unit 3 (vendor drawing: W0025604; serial number: E125T-2-2). On May 9, 2024, the vendor, Flowserve, was contacted and assumed responsibility for performing the Part 21 Evaluation for this valve. On October 28, 2024, Flowserve provided a 10 CFR 21.21(b) notification to TVA, stating that they were not capable of evaluating the existence of a defect. TVA procured additional engineering expertise to complete the required evaluation. These evaluations were tracked by TVA under CR 1942523. An independent failure analysis by BWXT was provided to Flowserve. BWXT concluded that 'the most likely cause of failure was brittle overload fracture due to a combination of tensile and bending forces that were exacerbated by the presence of shallow outer diameter initiated cracks and a significant loss of material ductility due to thermal embrittlement.' TVA also procured a second independent technical evaluation from MPR Associates, Inc., and provided their report to Flowserve to help with their evaluation. This report concluded that the event was apparently caused by an improper upper wedge-to-stem joint, and the resulting mismatch in mating surface diameters resulted in the bending stress which led to the valve failure, in conjunction with thermal embrittlement and excessive torques. TVA is providing notification of the existence of the defect and its evaluation. This event was entered into the corrective action program as condition report 1914295. The NRC Resident Inspector has been notified of this event, and a written report will be submitted within 30 days. Previous interim reports regarding this issue were submitted on June 23, 2024; August 22, 2024; and November 27, 2024. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The non-conforming part is no longer in service. There are similar parts in service at the Browns Ferry site, but it has been determined that the risk is low. Discussion will follow in the 30-day report. |
| ENS 57741 | 4 June 2025 01:37:00 | NRC Region 2 | Georgia | Hatch | GE-4 | The following information was provided by the licensee via phone and email: At 2137 EDT on 06/03/2025, while Unit 1 was at 98 percent power in mode 1, the high-pressure coolant injection (HPCI) turbine stop valve failed to open as required when testing the auxiliary oil pump, resulting in the HPCI system being declared inoperable. The cause of the turbine stop valve failing to open 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). Reactor core isolation cooling 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. There was no impact to Unit 2. |
| ENS 57737 | 31 May 2025 03:47:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | The following information was provided by the licensee via phone and email: At 2247 CDT on 5/30/25, with Unit 2 operating in mode 1 at 39 percent reactor power, the reactor was manually tripped due to a trip of the only operating reactor recirculation pump (2B). Approximately 44 minutes prior at 2203 CDT, the 2A reactor recirculation pump tripped. Operations responded and stabilized the plant. Primary containment isolation systems (PCIS) received an actuation signal for groups 2, 3, 6 and 8 on reactor water level at +2 inches. All primary containment systems that received an actuation signal performed as designed. All other systems functioned as designed. Reactor water level control is via condensate and feedwater, and reactor cooldown is in progress using turbine bypass valves to the main condenser. Due to the reactor protection system (RPS) actuation while critical, this event is being reported as a four-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The actuation of RPS and PCIS also requires an eight-hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The site reduced power from 100 percent following the loss of the 2A recirculation pump. |