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| Event date | Region | State | Site | Reactor type | Title | Event description | Reporting criterion | |
|---|---|---|---|---|---|---|---|---|
| ENS 58349 | 3 July 2026 22:27:00 | NRC Region 3 | Michigan | Fermi | GE-4 | Automatic Reactor Scram | The following information was provided by the licensee via phone and email: At approximately 1827 EDT on July 3, 2026, Fermi 2 experienced a lockout of bus `101' resulting in the loss of transformer `64' followed by a momentary loss of both reactor protection system (RPS) bus power sources resulting in a reactor scram. Both level `3' and level `2' isolations were received due to the reactor scram. The scram was non-complicated. Operations responded and stabilized the plant. Reactor water level is being maintained at normal level. The lowest water level observed was approximately 94.1 inches. Decay heat is being removed by the main steam system to the main condenser using automatic operation of the turbine bypass valve system. All control rods inserted into the core. 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 NRC Resident Inspector was notified.
The following information was provided by the licensee via phone and email: It was determined that reactor protection system (RPS) 'A' experienced a loss of power, while RPS 'B' remained energized throughout the transient. The automatic reactor scram was initiated on reactor pressure vessel (RPV) level 3 signal. Reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) automatically initiated on an RPV level 2 signal. Primary containment and reactor isolation functions occurred as expected in response to the low reactor water level condition. Emergency diesel generators 11 and 12 automatically started and supplied their respective safety-related buses following the loss of power. Notified R3DO (Bentancourt-Roldan). | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58345 | 1 July 2026 15:23:00 | NRC Region 4 | Texas | South Texas | Westinghouse PWR 4-Loop | Loss of Communication Capability | The following information was provided by the licensee via phone and email: At 1023 CDT on July 1, 2026, it was determined that all emergency response organization (ERO) notification methods were lost. This event constitutes a major loss of offsite communications capability and is reportable per 10 CFR 50.72(b)(3)(xiii). 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 Everbridge system utilized to notify offsite responders was identified as out of service. The Everbridge system has since been restored. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58334 | 27 June 2026 02:55:00 | NRC Region 2 | South Carolina | Summer | Westinghouse PWR 3-Loop | Loss of Communication Capability | The following information was provided by the licensee via phone and email: At 2255 EDT on June 26, 2026, with the unit operating at 100 percent power, V. C. Summer staff was notified by the nuclear licensing and emergency preparedness group, that a nationwide Dominion Energy Emergency Notification System (DEENS) Everbridge outage resulted in the loss of the primary and alternate methods for emergency response organization (ERO) notifications. This event constitutes a major loss of off-site communications capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector was notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58337 | 27 June 2026 02:42:00 | NRC Region 1 | Pennsylvania | Susquehanna | GE-4 | Loss of Communication Capability | The following information was provided by the licensee via phone and email: On June 26, 2026, at 2242 EDT, Susquehanna Steam Electric Station control room personnel were notified of a nationwide outage impacting the mass notification system (Everbridge). The outage resulted in complete loss of all notification methods to emergency response organization (ERO) personnel. Control room personnel were subsequently notified that Everbridge was restored at 2300. This event constitutes a major loss of offsite communications capability between the site and ERO personnel assigned offsite and is being reported under 10 CFR 50.72(b)(3)(xiii), loss of emergency preparedness capabilities. This event had no adverse impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58336 | 27 June 2026 01:21:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | Loss of Communication Capability | The following information was provided by the licensee via phone and email: At 2121 EDT on June 26, 2026, with Unit 1 and Unit 2 operating at 100 percent power, Surry Power Station Operations was notified by the corporate emergency preparedness organization that a nationwide Dominion Energy Emergency Notification System (DEENS) Everbridge outage resulted in a loss of the primary, backup, and alternate methods for emergency response organization (ERO) notifications per VPAP-2802. This event constitutes a major loss of off-site communications capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector was notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58335 | 27 June 2026 01:15:00 | NRC Region 2 | Virginia | North Anna | Westinghouse PWR 3-Loop | Loss of Communication Capability | The following information was provided by the licensee via phone and email: At 2115 EDT on June 26, 2026, with Unit 1 and Unit 2 operating at 100 percent power, North Anna Power Station operations was notified by the corporate emergency preparedness organization that a nationwide Dominion Energy Emergency Notification System (DEENS) Everbridge outage resulted in a loss of the primary, backup, and alternate methods for emergency response organization (ERO) notifications per VPAP-2802. This event constitutes a major loss of off-site communications capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector was notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58333 | 27 June 2026 00:30:00 | NRC Region 1 | Connecticut | Millstone | CE Westinghouse PWR 4-Loop | Loss of Communication Capability | The following information was provided by the licensee via phone and email: At 2030 EDT on June 26, 2026, it was determined that the primary, backup, and alternate methods for site emergency response organization (SERO) notifications were lost. This event constitutes a major loss of off-site communications capability and is reportable per 10 CFR 50.72(b)(3)(xiii). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58326 | 22 June 2026 14:09:00 | NRC Region 1 | Connecticut | Millstone | CE Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: At 1009 EDT on 6/22/2026, it was determined that a supervisor 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58322 | 18 June 2026 04:25:00 | NRC Region 4 | Texas | South Texas | Westinghouse PWR 4-Loop | Automatic Reactor Trip | The following information was provided by the licensee via phone and email: At 2325 (CDT) on 06/17/2026, (South Texas Project) Unit 2 reactor automatically tripped following a lockout of the Unit main and (auxiliary) transformers. This trip was not complex with all systems responding as normal post trip. Auxiliary feedwater actuated and standby diesel generators (DG) 21 and 23 started to carry their respective engineered safety features (ESF) busses. The Unit 2 balance of plant DG was out of service at the time of the trip. This issue resulted in no power to the Unit 2 instrument air compressors. The unit was cooled down utilizing steam generator pressure operated relief valves (PORV). 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 to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Unit 1 was unaffected by this event. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58320 | 15 June 2026 21:49:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: On June 15, 2026, at approximately 1749 (EDT), positive test results were confirmed on a random fitness for duty (FFD) test for a (contract) supplemental supervisor. The supplemental supervisor held unescorted access at Surry and North Anna Power Stations. No work on safety-related components was performed during this access period at Surry or North Anna. Individual was onsite to perform a walk down. Individual was denied unescorted access at Surry and North Anna Power Stations. FFD Evaluation will be submitted by the Dominion Energy Supervisor or designee at Surry and North Anna Power Stations. The NRC Resident Inspector was notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58315 | 11 June 2026 13:09:00 | NRC Region 2 | North Carolina | Brunswick | GE-4 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0909 EDT, on June 11, 2026, it was determined that a non-licensed contract supervisor failed a test specified by the fitness-for-duty (FFD) testing program. The individual's authorization for site access has been removed. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58314 | 11 June 2026 07:41:00 | NRC Region 1 | Connecticut | Millstone | CE | Manual Reactor Trip | The following information was provided by the licensee via phone and fax: At 0339 EDT on June 11, 2026, control element assembly '14' dropped to the bottom of the core. To stabilize plant conditions, the reactor was manually tripped at 0341. This resulted in the actuation of the reactor protection system (RPS) when the reactor was critical and 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. There has been no impact to Unit 3 which remains at 100 percent power. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58312 | 10 June 2026 19:26:00 | NRC Region 4 | Arizona | Palo Verde | CE | Fitness for Duty | The following information was provided by the licensee via phone and email: At 1226 MST on June 10, 2026, it was determined that a non-licensed 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58310 | 9 June 2026 17:21:00 | NRC Region 4 | Texas | South Texas | Westinghouse PWR 4-Loop | Part 21 - Relay Failure | The following information was provided by the licensee via phone and email: On June 9, 2026, STP (South Texas Project) Nuclear Operating Company (STPNOC) completed a Part 21 evaluation which identified a defect with TE Connectivity MDR styled relays (MDR 131-1 and MDR 134-1) manufactured after mid-year 2020. Since 2023, eleven (11) of these relays in both Unit 1 and Unit 2 have experienced premature failures resulting in essential chillers being inoperable and incapable of fulfilling their intended safety function. The relays were procured as commercial grade items and were dedicated by STPNOC solely for use at STP. An investigation determined that the relay failures resulted from overheating of the relay coils during operation. The vendor informed STP that a subcontractor responsible for rotor grinding produced relay rotors that were undersized and outside of specification. The reduced rotor dimensions resulted in a looser fit between the rotor and stator, reducing eddy currents and increasing coil current consumption. The increased coil wattage then generated excessive heat that caused the relay windings to short at the finish lead, resulting in relay failure. A written notification in accordance with 10 CFR 21.21 (d)(3)(ii) will be provided within 30 days. The NRC Resident Inspector has been notified, June 10, 2026, at 1330 CDT. | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect |
| ENS 58306 | 9 June 2026 14:34:00 | NRC Region 4 | Kansas | Wolf Creek | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0934 CDT on June 9, 2026, it was determined that a non-licensed 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58301 | 4 June 2026 19:35:00 | NRC Region 1 | Connecticut | Millstone | CE | Technical Specification Required Shutdown | The following information was provided by the licensee via phone and email: On June 4, 2026, at 1535 EDT, a technical specification (TS) required shutdown was initiated at Millstone Unit 2. TS 3.1.3.1 action 'A' was entered at 1234 EDT with a required action to reduce thermal power to less than 70 percent of the maximum allowable thermal power within 1 hour and restore control element assembly misalignment within 2 hours. This required action was not completed within the completion time, therefore, a TS required shutdown was initiated and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. There has been no impact to Unit 3, which remains at 100 percent power. | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown |
| ENS 58293 | 28 May 2026 16:09:00 | NRC Region 4 | Texas | South Texas | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: On May 28, 2026, at 1109 CDT, it was determined that a non-licensed supervisor failed a test specified by the FFD testing program. The individual's authorization for site access has been terminated at South Texas Project. This has been determined to be reportable under 10 CFR 26.719(b)(2). The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58288 | 21 May 2026 12:00:00 | NRC Region 2 | Georgia | Vogtle | W-AP1000 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0800 EDT on May 21, 2026, it was determined that a 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58286 | 18 May 2026 22:07:00 | NRC Region 3 | Michigan | Fermi | GE-4 | Manual Reactor Scram | The following information was provided by the licensee via phone and email: At approximately 1803 EDT on May 18, 2026, the operating crew responded to a fire below the main generator. The fire was identified on the Y-phase of the isophase ducts. At 1807, the reactor was manually scrammed, and a level 3 isolation signal was received due to the scram. The scram was non-complicated. The fire was confirmed extinguished at 1810. Operations responded and stabilized the plant. Reactor water level is being maintained at normal level. The lowest water level observed was approximately 139.5 inches. Decay heat is being removed by the main steam system to the main condenser using automatic operation of the turbine bypass valves. All control rods inserted into the core. 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). | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58309 | 18 May 2026 02:49:00 | NRC Region 4 | Missouri | Callaway | Westinghouse PWR 4-Loop | Invalid Specified System Actuation | The following information was provided by the licensee via phone and email: This report is being made in accordance with 10 CFR 50.73(a)(2)(iv)(A) using the telephone option in 50.73(a)(1) in lieu of a written LER (licensee event report) concerning an event in which Callaway experienced an automatic start of the turbine driven auxiliary feedwater pump (TDAFP) in response to an invalid auxiliary feedwater actuation signal. The reactor protection system was not involved. This TDAFP actuation was complete, and the TDAFP started and functioned successfully in accordance with the initiating actuation signal. This actuation occurred at 2149 CDT on May 17, 2026. The most probable cause of the actuation was a voltage perturbation caused by switching inverters in preparation for tagging out the in-service inverter, resulting in the TDAFP receiving a start signal. This signal did not result from plant conditions or parameters satisfying the requirements for initiation of the system. Therefore, the actuation was invalid. The NRC Resident Inspector was notified. | 10 CFR 50.73(a)(1), Submit an LER |
| ENS 58280 | 15 May 2026 14:06:00 | NRC Region 3 | Illinois | LaSalle | GE-5 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0906 CDT on 05/15/2026, it was determined that a non-licensed supervisor tested positive in accordance with the fitness for duty testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58273 | 11 May 2026 10:00:00 | NRC Region 3 | Michigan | Cook | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: On May 11, 2026, at 0600 EDT, a fitness for duty (FFD) employee had a confirmed positive test result for a controlled substance during a random fitness for duty test. The employee's access to the plant has been denied. This notification is being made under 10 CFR 26.719 (b)(2)(ii), significant FFD policy violations or programmatic failures, as a twenty-four (24) hour report. The licensee has notified the NRC Resident Inspector. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58272 | 9 May 2026 20:29:00 | NRC Region 4 | Nebraska | Cooper | GE-4 | Loss of the Public Prompt Notification System | The following information was provided by the licensee via phone and email: On May 9, 2026, at 1529 CDT, the National Weather Service reported to Cooper Nuclear Station that the National Warning System radio tower near Shubert, Nebraska was not functioning. The Shubert tower transmitter activates the emergency alert system and tone alert radios used for public notification. Additional information from the National Weather Service received at 1244 CDT on May 10, 2026, determined that the Shubert Tower transmitter would not likely be repaired within 24 hours. The backup notification system has been verified as available. This condition is reportable under 10 CFR 50.72(b)(3)(xiii) as a major loss of the public prompt notification system capability since the backup alert methods do not meet the primary system's design objective. The NRC Senior Resident Inspector has been informed. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58269 | 6 May 2026 15:18:00 | NRC Region 4 | Mississippi | Grand Gulf | GE-6 | Fitness for Duty | The following information was provided by the licensee via phone and email: On May 6, 2026, at 1018 CDT, Grand Gulf Nuclear Station was notified that a non-licensed supervisor violated the station's fitness for duty policy. The employee's unescorted access has been terminated at all Entergy facilities. This event was determined to be reportable under 10 CFR 26.719(b)(2)(ii). The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58263 | 1 May 2026 20:08:00 | NRC Region 4 | Texas | South Texas | Westinghouse PWR 4-Loop | Valid Emergency Diesel Generator Actuation | The following information was provided by the licensee via phone and email: On May 1, 2026, at 1508 CDT, with both units at full power, the South Texas Project (STP) switchyard north bus was de-energized (locked out). This loss of power de-energized the standby '1' transformer, which was aligned to supply power to the `1B' 4KV engineered safety feature (ESF) bus. This resulted in a specified system actuation of emergency diesel generator (EDG) `1B' on the undervoltage condition, to automatically start the '1B' EDG, power the '1B' 4KV ESF bus, and sequence on select loads. The cause of the (switchyard north bus) lockout is under investigation at this time. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electric power system per 10 CFR 50.72(b)(3)(iv)(B)(8). There was no failure of automatic equipment to start. 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: Offsite power had been restored to the `1B' 4KV ESF bus and the '1B' EDG returned to standby by the time of notification. | 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 58257 | 23 April 2026 18:08:00 | NRC Region 2 | Georgia | Hatch | GE-4 | Automatic Reactor Scram | The following information was provided by the licensee via phone and email: On April 23, 2026, at 1408 EDT, with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped on high reactor pressure. The cause of the high reactor pressure is currently under investigation. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained by the feedwater 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). 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 rise in reactor pressure was a result of slow closure of the main turbine control valves over approximately one minute due to an unknown cause. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58256 | 23 April 2026 12:50:00 | NRC Region 3 | Michigan | Cook | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: On April 23, 2026, at 0850 EDT, a non-licensed supervisor had a confirmed positive test result for alcohol during a random fitness for duty (FFD) test. The employee's access to the plant has been denied. This notification is being made under 10 CFR 26.719 (b)(2)(ii), significant FFD policy violations or programmatic failures, as a twenty-four-hour report. The licensee has notified the NRC Resident Inspector. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58248 | 21 April 2026 16:40:00 | NRC Region 2 | South Carolina | Catawba | Westinghouse PWR 4-Loop | Steam Generator Tube Degradation | The following information was provided by the licensee via phone and email: At 1240 EDT on 4/21/2026, with the unit at 100 percent power, it was determined that a tube within the '2B' steam generator did not meet the performance criteria per plant technical specifications. The subject tube was removed from service via plugging during a previous plant shutdown. 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 58246 | 20 April 2026 13:02:00 | NRC Region 1 | New York | Nine Mile Point | GE-5 GE-2 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0902 EDT on 4/20/26, it was determined that a supervisor tested positive in accordance with the fitness-for-duty (FFD) testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58244 | 17 April 2026 15:08:00 | NRC Region 3 | Illinois | LaSalle | GE-5 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 1008 (CDT) on 4/17/26, it was determined that a licensed operator tested positive in accordance with the fitness-for-duty (FFD) testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58243 | 16 April 2026 20:30:00 | NRC Region 2 | Alabama | Browns Ferry | GE-4 | Technical Specification Required Shutdown | The following information was provided by the licensee via phone and email: On April 16, 2026, at 2102 CDT, Browns Ferry Unit 1 was operating at 100 percent reactor power when a plant shutdown was initiated due to technical specification (TS) requirements. At 1530 CDT on April 16, 2026, while investigating a leak in the reactor steam vault, a reactor coolant system (RCS) pressure boundary leak was identified. TS limiting condition of operation 3.4.4 condition 'A' was entered with a required action to isolate the affected component, pipe, or vessel from the reactor coolant water by use of a closed manual valve, closed and deactivated automatic valve, blind flange, or check valve within 4 hours. Due to the leak being unisolable, TS 3.4.4 condition 'D' was entered with required action to be in mode 3 within 12 hours and mode 4 within 36 hours. This event requires an 8-hour report in accordance with (IAW) 50.72(b)(3)(ii)(A) Degraded Condition for the flaw in the RCS boundary, and a 4-hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's technical specifications.' There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown |
| ENS 58242 | 15 April 2026 13:20:00 | NRC Region 1 | Pennsylvania | Beaver Valley | Westinghouse PWR 3-Loop | Degraded Condition | The following information was provided by the licensee via phone and email: On April 15, 2026, at 0920 EDT, while performing examinations of the 66 reactor vessel head penetrations, it was determined that two penetrations did not meet ASME Code Section XI (requirements). Penetrations 54 and 59 will require repair prior to returning the vessel head to service. The indications were not through-wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME code case N-729-6 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 58298 | 12 April 2026 17:35:00 | NRC Region 2 | Virginia | Surry | Westinghouse PWR 3-Loop | Invalid Specified System Actuation | 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 specified system actuation of the Surry Power Station Unit 2 emergency core cooling system (ECCS). On 04/12/2026 at 1335 EDT, during Unit 2 cooldown, failure of 02-SI-RLY-SIB2B-RELAY (Unit 2 safety injection block 2 relay, train 'B') prevented the 'B' train of safety injection (SI) from being blocked from initiation signals caused by high steam flow. A SI initiation occurred during reactor coolant system cooldown due to signals from steam flow detectors spiking above the bistable setpoint, which completed the safety injection matrix. Based on then current plant conditions, an actual high steam flow condition could not exist, and therefore, a hi steam flow coincident with a low average coolant temperature or low steam line pressure safety injection should not occur. This event is considered an invalid system actuation because the actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. All portions of the ECCS functioned as would be expected were the actuation to be valid. Unit 2 entered cold shutdown at 1418. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | 10 CFR 50.73(a)(1), Submit an LER |
| ENS 58228 | 1 April 2026 13:24:00 | NRC Region 1 | Pennsylvania | Peach Bottom | GE-4 | Unanalyzed Condition | The following information was provided by the licensee via phone and email: At 0924 EDT on April 1, 2026, it was determined that Unit 3 was in an unanalyzed condition because two emergency diesel generators (EDGs) were inoperable. One EDG was inoperable due to scheduled maintenance, and the second EDG was inoperable due to blocking for an emergent plant issue. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The second EDG was returned to service, and the condition was exited. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 2 was unaffected by this condition. No other limiting conditions exist. The second EDG was blocked due to a failed transformer, but the transformer has been repaired. | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition |
| ENS 58224 | 31 March 2026 11:50:00 | NRC Region 2 | Georgia | Vogtle | Westinghouse PWR 4-Loop | Loss of Seismic Monitoring | 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. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58254 | 27 March 2026 00:03:00 | NRC Region 2 | Georgia | Hatch | GE-4 | Lost Source | The following information was provided by the licensee via phone and email: On March 26, 2026, at 2003 EDT, during a required semiannual source inventory, Hatch discovered that a radioactive source was missing. The source had been positively accounted for during the September 2025 inventory. The material involved is a National Institute of Standards and Technology traceable mixed-gamma calibration source (source ID 0091-00-00). It is a solid epoxy source contained in a 125 mL plastic bottle and was decay-corrected to approximately 0.22 microcuries at the time of discovery. The lost radioactive material includes americium-241 in excess of 10 times the quantity specified in appendix C to part 20. Therefore, this is reported as a nonemergency loss of licensed material under 10 CFR 20.2201(a)(1)(ii). The NRC Resident Inspector has been notified. THIS MATERIAL EVENT CONTAINS A 'Less than Cat 3' LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf | 10 CFR 20.2201(a)(1)(ii) |
| ENS 58218 | 26 March 2026 09:00:00 | NRC Region 2 | Georgia | Vogtle | W-AP1000 | Fitness for Duty | 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58217 | 26 March 2026 00:06:00 | NRC Region 3 | Michigan | Cook | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: On March 25, 2026, at 2006 EDT, a security supervisor had a confirmed positive test result for alcohol during a random fitness-for-duty (FFD) test. The employee's access to the plant has been denied. This notification is being made under 10 CFR 26.719 (b)(2)(ii), significant FFD policy violations or programmatic failures, as a twenty-four (24) hour report. The licensee has notified the NRC Resident Inspector. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58215 | 24 March 2026 21:47:00 | NRC Region 2 | North Carolina | McGuire | Westinghouse PWR 4-Loop | Degraded Condition | 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 58211 | 19 March 2026 15:19:00 | NRC Region 4 | Arizona | Palo Verde | CE | Unusual Event Declared for RCS Leakage for Greater than 15 Minutes | The following is a summary of information provided by the licensee via phone: At 0819 MST on 3/19/2026, with Palo Verde Unit 2 at 95 percent power, the site declared an Unusual Event under emergency action level 'SU5.1' due to identified reactor coolant system leakage for greater than 15 minutes. The leak was isolated by securing letdown. A manual reactor trip was inserted, and Unit 2 is currently shutdown in mode 3. Palo Verde Units 1 and 3 were unaffected. State and local authorities were notified. The NRC Senior Resident Inspector has been notified. Notified DHS SWO, FEMA Ops Center, FEMA NWC, CISA Central, CWMD Watch Desk, DHS Nuclear SSA (email), and DHS NRCC THD Desk (email).
The following update was provided by the licensee via phone: This update is a 10 CFR 50.72(b)(2)(iv)(B) required report of a valid RPS actuation while critical due to operators tripping Palo Verde Unit 2 at 0841 MST. Notified R4DO (Dixon), NRR EO (Feliz Adorno), IR MOC (Whited), R4RA (Monninger), NRR (Groom), R4 PAO (Smith), and NSIR (Williams).
The following information was provided by the licensee via phone: Palo Verde Unit 2 terminated the notice of unusual event 'SU5.1' at 1410 MST. The leak was isolated, the plant is stable, and there was no release. State and local authorities were notified. The NRC Senior Resident Inspector has been notified. Internal notifications: R4DO (Dixon), NRR EO (Feliz Adorno), IR MOC (Whited), R4RA (Monninger), NRR (Groom), R4 PAO (Smith), and NSIR (Williams). External notifications: DHS SWO, FEMA Ops Center, FEMA NWC, CISA Central, CWMD Watch Desk, DHS Nuclear SSA (email), and DHS NRCC THD Desk (email).
This notification constitutes a retraction of the 1-hour immediate notification (EN 58211) of the unusual event made to the NRC by Palo Verde Nuclear Generating Station (PVNGS) on March 19, 2026. Palo Verde is (partially) retracting this event notification in accordance with the guidance provided in NUREG 1022, Revision 3, `Event Reporting Guidelines 10 CFR 50.72 and 50.73,' specifically Section 2.8, `Retraction and Cancellation of Event Reporting,' and Section 4.2.3, `ENS Notification Retraction.' The leakage reported on March 19, 2026, was subsequently determined to have originated from a failed O-ring on a seal injection filter housing. The leakage was isolated by closing installed isolation valves on the seal injection filter housing. Following evaluation, this condition was determined not to meet the criteria for reactor coolant system (RCS) identified leakage as defined in SU5.1 of the PVNGS emergency plan. Clarification: Leakage was not isolated by securing letdown, as originally reported in the ENS notification submitted on March 19, 2026. This update applies solely to the retraction of the unusual event declaration. The NRC Resident Inspector has been notified. Notified R4DO (Deese). | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration |
| ENS 58202 | 17 March 2026 12:54:00 | NRC Region 1 | Connecticut | Millstone | Westinghouse PWR 4-Loop | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0854 (EDT) on March 17, 2026, it was determined that a licensed operator failed a test specified by the fitness for duty (FFD) testing program. The employee's unescorted access has been placed on hold in accordance with the licensee's FFD policy. The event was determined to be reportable under 10 CFR 26.719(b)(2)(ii). The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58200 | 14 March 2026 11:00:00 | NRC Region 1 | Pennsylvania | Beaver Valley | Westinghouse PWR 3-Loop | Loss of Power to the Emergency Operations Facility | The following information was provided by the licensee via phone and email: At 0700 EDT on March 14, 2026, the emergency operations facility (EOF) lost normal and alternate electrical power. At the time of the event, the normal power source to the EOF (offsite power) was lost due to the effects of weather in the wider area and attempts to establish the alternate power generator were unsuccessful. At the time of this notification, power is not yet restored to the facility but is anticipated to be restored within the next few hours. The cause of the backup generator failure to run is unknown at this time. If an emergency is declared requiring EOF activation while there is no available power, compensatory measures have been established to ensure that all assessment and communications functions will be available, which include staffing the EOF at a location within the on-site emergency response facility and activating using existing emergency planning 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 loss of normal and alternate electrical power affects the functionality of an emergency response facility. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 58190 | 10 March 2026 18:34:00 | NRC Region 4 | Arizona | Palo Verde | CE | Fitness for Duty | The following information was provided by the licensee via phone and email: On March 10, 2026, at 1134 MST, a reactor operator's test results were confirmed positive for alcohol following a for-cause, fitness for duty screening test. The individual's unescorted access has been immediately placed on administrative hold until the denial process has been completed per station procedures. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58187 | 9 March 2026 13:33:00 | NRC Region 3 | Illinois | Clinton | GE-6 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 0833 CDT on 03/09/2026, it was determined that a licensed operator tested positive in accordance with the FFD testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58186 | 8 March 2026 22:17:00 | NRC Region 3 | Illinois | Byron | Westinghouse PWR 4-Loop | Degraded Reactor Vessel Closure Head Penetration | The following information was provided by the licensee via phone and email: At 1717 CDT on March 8, 2026, during the Byron Unit 1 refueling outage, it was determined that the control rod drive mechanism (CRDM) penetration 31 was degraded because liquid penetrant testing performed on the embedded flaw repair weld identified an unacceptable rounded indication, in accordance with American Society of Mechanical Engineers (ASME) Section III acceptance standards and an NRC-approved licensee relief request for a previously performed embedded flaw repair. Required repairs will be completed in accordance with the ASME code of record prior to returning the vessel head to service. Therefore, 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 58225 | 5 March 2026 17:24:00 | NRC Region 4 | Louisiana Mississippi | River Bend | GE-6 | Lost Radioactive Material in Transit | The following information was provided by the licensee via phone: On March 5, 2026, at 1224 CST, River Bend declared a radioactive materials shipment lost after it was supposed to be delivered by February 2, 2026. The last time the shipment was scanned in the tracking system was on February 11, 2026, at the (common carrier) distribution warehouse in Pearl, MS. The device is a digital acquisition unit inside a Pelican case which contains 122 microcuries of Co-60. River Bend plans on notifying the State of Mississippi regarding the lost material. Because the quantity of radioactive material lost exceeds 10 times the quantity specified in appendix C to part 20, this is being reported as a non-emergency loss of licensed material. The NRC Resident Inspector has been notified. THIS MATERIAL EVENT CONTAINS A 'Less than Cat 3' LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf | 10 CFR 20.2201(a)(1)(ii) |
| ENS 58181 | 2 March 2026 07:59:00 | NRC Region 2 | North Carolina | Brunswick | GE-4 | Degraded Condition - Primary Containment | 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 58179 | 27 February 2026 23:23:00 | NRC Region 3 | Illinois | Clinton | GE-6 | Fitness for Duty | The following information was provided by the licensee via phone and email: At 1723 CST on February 27, 2026, it was determined that a non-licensed supervisor tested positive in accordance with the FFD testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 58177 | 26 February 2026 08:03:00 | NRC Region 2 | Georgia | Hatch | GE-4 | Manual Reactor Scram | 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. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 58170 | 23 February 2026 15:34:00 | NRC Region 3 | Michigan | Fermi | GE-4 | Fitness for Duty Event | The following information was provided by the licensee via fax or email: On February 23, 2026, at 1034 EST, it was determined that a non-licensed supervisor failed a test specified by the fitness for duty (FFD) testing program. The individual's unescorted site access has been terminated. The NRC Resident and Region 3 Inspectors had been notified. | 10 CFR 26.719, FFD Reporting requirements |