Semantic search
| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 57438 | 23 November 2024 02:42:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone or email: At 1937 EST on 11/22/2024, it was discovered that both trains of the control room emergency air temperature control system (CREATCS) were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). 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: Technical specification 3.7.11 conditions A and C were entered as a result of this event. The 'B' train of CREATCS was restored at 0130 EST on 11/23/24 and the plant exited condition C. The 'A' train remained out of service at the time of notification. Although CREATCS is a common system for both Units 1 and 2, Unit 1 was defueled and outside the mode of applicability during the timeframe of this event. |
| ENS 57431 | 19 November 2024 16:02:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1350 (EST) on 11/19/2024, with Unit 1 in mode 1 at 100% power, the reactor was manually tripped due to an automatic trip of the `B' main feedwater pump turbine. The reactor trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished by the steam dumps to the condenser. Emergency feedwater actuated due to low-low steam generator (water) level, as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(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 'B' main feedwater pump trip is under investigation. There was maintenance involving the 'B' main feedwater pump at the time of the scram. |
| ENS 57432 | 19 November 2024 16:02:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1350 (EST) on 11/19/2024, with Unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to an automatic trip of the `B' main feedwater pump turbine. The reactor trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished by the steam dumps to the condenser. Emergency feedwater actuated due to low-low steam generator level, as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(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 'B' main feedwater pump turbine trip is under investigation. |
| ENS 57424 | 14 November 2024 10:58:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: NextEra Energy Seabrook LLC. makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found in a GE - Hitachi Relay, CR120B (Model #DD945E118P0060) during pre-installation bench testing. During bench testing, the relay failed to energize and transfer all associated contacts. The relay was purchased from GE - Hitachi (GEH) as safety-related, GE CR-120B relays. All GE CR-120B relays that were purchased in the same batch as the failed relay were located and quarantined in order to be returned to GEH for forensic testing. NextEra Energy Seabrook, LLC has concluded that this defect constitutes a substantial safety hazard (SSH). A SSH exists because the nature of the defect was such that, if installed in certain safety-related applications and failed, it would have prevented the fulfillment of a safety function. On November 12, 2024, the Seabrook site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification will be provided in accordance with 10 CFR 21.21(d)(3)(ii). Because the defect was discovered prior to installation, there was no impact to safety-related equipment. The NRC Senior Resident Inspector has been informed. |
| ENS 57422 | 13 November 2024 13:38:00 | Millstone | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0902 EST, on 10/10/2024, with Millstone Unit 3 in mode 1 at 100 percent power, it was discovered that the secondary containment boundary was inoperable when the latch that secured a hatch that was part of the secondary containment boundary was not functional. The latch was repaired by 1115, on 10/10/2024, and the secondary containment boundary was declared operable at 1200, on 10/10/2024. The initial assessment of reportability concluded that an immediate report was not required. However, upon additional review, it has been determined that because the secondary containment boundary is a single-train system that performs a safety function, an 8-hour report was required in accordance with 10 CFR 50. 72 (b)(3)(v)(C) and (D). This report should have been made on 10/10/2024 and is late. There has been no impact to Unit 2, and Unit 3 continues to operate in mode 1 at 100 percent power. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57411 | 3 November 2024 23:33:00 | Cook | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On 11/03/2024, at 2242 EST, DC Cook Unit 2 received an annunciator indicating a fire in containment. Verification time of existence of a fire exceeded the threshold for an Unusual Event (UE), and a UE was declared at 2312 on 11/03/24. Subsequently, the alarm was determined to not be valid and the UE was exited at 2328. Berrien County and the State of Michigan were notified of the UE declaration and exit. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No actual fire existed. The emergency action level for this event is HU4.2. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). |
| ENS 57402 | 28 October 2024 13:55:00 | Catawba | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via email: The licensee received two alarms due to direct current (DC) output voltage fluctuating between 127.4 to 131.3 volts. After troubleshooting, the DC output voltage fluctuations were caused by the battery charger printed circuit board. The part has been sent to the vendor, Ametek, for evaluation. Catawba is the only plant known to have this issue at this time. The evaluation is expected to be completed on January 31, 2025. Catawba condition report 02526388 Ametek Part Number: 80-921-4031-90 Ametek failure analysis number: 24-006 |
| ENS 57379 | 14 October 2024 13:22:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On October 14, 2024, a licensed employee violated the station's fitness for duty (FFD) policy. The employee's unescorted access to the site has been terminated. The event was determined to be reportable under 10 CFR 26.719(b)(2)(ii). The NRC Resident Inspector has been notified. |
| ENS 57361 | 4 October 2024 04:23:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 2235 on 10/03/2024, the Vogtle 1 and 2 seismic monitoring panel experienced an electrical fault, rendering the panel nonfunctional. Compensatory measures for seismic event classification have been implemented in accordance with Vogtle procedures. This is an eight-hour, non-emergency notification for a loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the seismic monitoring panel is the method for evaluating that an operational basis earthquake (OBE) threshold has been exceeded following a seismic event. This is in accordance with Initiating condition `seismic event greater than OBE levels' and emergency action level HU2. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57351 | 29 September 2024 08:40:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 05:56 EDT on 09/29/2024, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The motor driven auxiliary feedwater (MDAFW) level control valves (LCV) for loop 1 failed to respond from the main control room. All others systems responded normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 2 is currently stable in Mode 6 for a maintenance outage and was 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) 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 AFW system. The AFW system started automatically and is operating as designed with the exception of the MDAFW LCVs for loop 1. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57344 | 27 September 2024 11:25:00 | Catawba | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0748 EDT on 9/27/24, Catawba Unit 2 was manually tripped due to loss of condenser vacuum. The Unit 2 auxiliary feedwater (AFW) system started automatically as expected. Decay heat is being removed by the steam generators and discharging to the condenser. Due to the Unit 2 reactor protection system actuation while critical, this event is being reported as a four-hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). The automatic start of the Unit 2 AFW system is being reported as an eight-hour, nonemergency notification per 10 CFR 50.72(b)(3)(iv)(A). On Unit 1, it was determined that at 0746, all trains of the Unit 1 AFW were inoperable when the Unit 1 hotwell temperature exceeded the operability limit for the AFW system. Therefore, this condition is being reported as an eight-hour, nonemergency notification per 10 CFR 50.72(b)(3)(v). The affected safety function was restored on 9/27/24 at 0851 EDT when the Unit 1 hotwell temperature returned below the operability limit for the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The loss of vacuum for both Unit 1 and Unit 2 was due to loss of power to cooling tower fans. The suspected cause of loss of cooling tower fans was due to water intrusion due to Hurricane Helene. |
| ENS 57336 | 23 September 2024 21:46:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 2127 EDT, on 8/01/2024, with Unit 1 in mode 1 at 98 percent power, a complete actuation of the 'A' train containment ventilation isolation (CVI) occurred. The 'A' train CVI resulted from the failure of a radiation monitor providing input to the isolation circuitry. The CVI removes containment purge from operation should it be in service and secures other radiation monitors which measure reactor coolant system leakage. In accordance with the station's procedures and technical specifications, a restoration from the CVI was made. Troubleshooting revealed that replacement of this obsolete radiation monitor was justified; a design change to perform this replacement is in progress. This report is being made under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of the 'A' train CVI. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified of the event. |
| ENS 57324 | 16 September 2024 16:30:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1248 EDT on July 22, 2024, with Unit 1 in mode 1 at 100% power, a complete actuation of the 'A' train containment ventilation isolation (CVI) occurred. The 'A' train CVI resulted from the failure of a radiation monitor providing input to the isolation circuitry. This radiation monitor was subsequently repaired and a restoration from the CVI was made. The CVI removes containment purge from operation should it be in service and secures other radiation monitors which measure reactor coolant system leakage. This report is being made under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of the 'A' train CVI. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified of the event. |
| ENS 57321 | 13 September 2024 13:54:00 | Byron | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On 9/13/2024 at 0823 CDT, during the Byron Unit 1 refueling outage, it was determined that a previous overlay repair on penetration number 31 of the reactor vessel closure head was degraded because the results of a planned liquid penetrant test did not meet applicable acceptance criteria. Any required repairs will be completed in accordance with the ASME code of record prior to returning the vessel head to service. 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 57313 | 10 September 2024 08:28:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A). The event was an invalid actuation of the Unit 1 containment ventilation isolation (CVI) system. On August 3, 2024, at 0840 EDT, with Unit 1 at 100 percent power, train 'A' of the CVI system actuated due to an invalid signal from 1-RM-90-130, containment purge air exhaust monitor. 1-RM-90-130 was out of service for maintenance testing at the time of the invalid signal. The cause of the signal was determined to be the result of an installed multimeter timing out, creating a short in the actuation circuitry. The train 'A' CVI signal was a full actuation of that train and the system functioned as designed. Prior to and following the CVI alarm, all other radiation monitors were stable at their normal values; therefore, the CVI (actuation) was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. Subsequent completion of the maintenance instruction was successful. This event was entered into the corrective action program as CR 1948103. The NRC Resident Inspector was notified. |
| ENS 57285 | 23 August 2024 13:44:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 1219 EDT, 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 auxiliary feedwater (AFW) and steam dump systems. Unit 2 is currently in a refueling outage (U2R26) and was 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) 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 AFW system. The AFW system started automatically and is operating as designed. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57284 | 23 August 2024 09:02:00 | Wolf Creek | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0500 CDT on 8/23/24, Wolf Creek entered technical specification limiting condition for operation (LCO) 3.7.5 required action D.1 which requires shutdown to mode 3 within 6 hours. The turbine driven auxiliary feedwater pump discharge valve to the 'B' steam generator was not successfully restored to operable prior to expiration of the 72 hour completion time. At 0800 CDT, the shutdown to mode 3 was initiated, which is being reported in accordance with 10CFR50.72(b)(2)(i). The NRC Resident Inspector has been notified. |
| ENS 57282 | 21 August 2024 19:30:00 | Millstone | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax and phone: At 1200 EDT on 8/21/2024, with Millstone unit 3 in mode 1 at 100 percent power, it was discovered that the secondary containment boundary was inoperable while maintenance activities on the system were in progress. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) and (D). There is no impact on the health and safety of the public and plant personnel. The NRC Resident Inspector has been notified. Unit 3 continues to operate in mode 1 at 100 percent power with actions in progress to restore the system to operable within the technical specification allowed outage time. There has been no impact to unit 2, which remains at 100 percent power. The state of Connecticut and local towns were notified. |
| ENS 57279 | 21 August 2024 01:34:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax or email: At 2010 CDT on 08/20/2024, with Unit 2 in Mode 1 at 100% power, Train B Essential Chiller was unable to maintain the Chilled Water Outlet Temperature in the required band. Train B Essential Chilled Water and the associated cooled components were declared inoperable. Train B Control Room Makeup and Cleanup Filtration System was declared inoperable due to the unavailability of cooling. Train C Control Room Makeup and Cleanup Filtration System was previously inoperable for planned maintenance for reasons other than loss of cooling. This resulted in the inoperability of two of the three 50-percent capacity safety trains required for accident mitigation. This event is reportable under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. 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: Train C has been placed back in service and is awaiting testing to verify its ability to perform its safety function. Testing is expected to be completed by 1500 CDT on 08/21/2024. Plant is on a 72 hr S/D clock until restoration is verified. |
| ENS 57274 | 19 August 2024 15:55:00 | Callaway | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Senior Resident Inspector has been notified. |
| ENS 57270 | 13 August 2024 12:22:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone: At 1200 EDT on August 13, 2024, with Unit 2 in Mode 1 at 100 percent power, Vogtle Unit 2 declared an ALERT per emergency action level (EAL) SA9 due to a fire that caused visible damage to a safety system component needed for the current operating mode. At 1151 EDT, the fire was extinguished. The equipment affected was the safety-related regulating 480V transformer which supplies power to the Unit 2 'B' engineered safety features chiller. There was no impact to the safety and health of the public or plant personnel. Units 1, 3, and 4 are unaffected. State and local officials were notified. The NRC resident inspector was notified. The NRC decided to remain in the Normal mode of operation at 1234 EDT. Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, EPA Emergency Ops Center, USDA Watch Officer, FDA Emergency Ops Center (email), FEMA NWC (email), DHS Nuclear SSA (email), DHS NRCC THD Desk (email), FEMA NRCC SASC (email), FERC RMC (email), CWMD Watch Desk (email). The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The fire alarm was received at 1145 EDT. A fire was confirmed at 1149 EDT. The switchgear was de-energized and a fire extinguisher was used to put out the fire.
The licensee terminated the ALERT emergency action level at 1436 EDT. Notified R2DO (Lopez-Santiago), IR MOC (Grant), NRR EO (Felts), DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, EPA Emergency Ops Center, USDA Watch Officer, FDA Emergency Ops Center (email), FEMA NWC (email), DHS Nuclear SSA (email), DHS NRCC THD Desk (email), FEMA NRCC SASC (email), FERC RMC (email), CWMD Watch Desk (email). |
| ENS 57253 | 30 July 2024 18:52:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1641 EDT, with Unit 2 in Mode 1 at 94 percent power and increasing in power after a forced outage, the reactor automatically tripped due to an electrical trouble 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 auxiliary feedwater (AFW) and steam dump systems. 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) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. |
| ENS 57241 | 24 July 2024 20:58:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On July 24, 2024, a licensed operator violated the station's fitness for duty (FFD) policy. The employee's unescorted access to South Texas Project has been terminated. The event was determined to be reportable under 10 CFR 26.719(b)(2)(ii). The NRC resident inspector has been notified. |
| ENS 57237 | 24 July 2024 09:31:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: A Notification of Unusual Event was declared by South Texas Project Unit 1 at 0718 CDT for emergency action level (EAL) SU.1, loss of all offsite power for greater than 15 minutes, following a fire in the switchyard. Unit 1 tripped following the loss of power and is stable in Mode 3. Unit 2 reduced power to 90 percent but was otherwise unaffected by this event. Offsite services responded to the switchyard fire. The fire was extinguished at 0925 CDT. There is no radioactive release and no threat to public safety. The licensee notified state and local authorities and the NRC senior resident inspector. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via phone and email: At 0702 CDT on 7/24/2024, with Unit 1 in Mode 1 at 100 percent power, the Unit 1 reactor automatically tripped due to loss of offsite power. The trip was not complex, with all systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the scram. Operations responded and stabilized the plant. Decay heat is being removed by steam generator power operated relief valves (PORV). Unit 2 was reduced in power to approximately 90 percent power due to conditions in the switchyard. 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. Notified R4DO (Azua)
The following is a summary of information provided by the licensee via phone: At 1146 CDT, South Texas Project Unit 1 terminated the previously declared Notification of Unusual Event due to restoration of an offsite source of electrical power. Notified R4DO (Azua), NRR EO (McKenna), IR MOC (Crouch), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via phone and email: For the 10 CFR 50.72(b)(3)(iv)(A) reporting requirements: At 0702 CDT on 7/24/2024, with both Unit 1 and 2 in Mode 1 at 100 percent power, the South Texas Project (STP) north and south switchyard electrical buses were de-energized. In Unit 1, all emergency diesel generators (EDGs) 11, 12, and 13 automatically started in response to loss of offsite power on train `A', `B', and `C' engineered safety feature (ESF) buses. Also in Unit 1, trains `A', `B', and `C' of the auxiliary feedwater (AFW) system automatically started. In Unit 2, EDG 22 automatically started in response to loss of offsite power on the train `B' ESF bus. Also in Unit 2, train `B' of the AFW system automatically started. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of a pressurized water reactor auxiliary feedwater system (50.72(b)(3)(iv)(B)(6)) and emergency alternating current (AC) electrical power system (50.72(b)(3)(iv)(B)(8)). There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. For the 10 CFR 50.72(b)(2)(xi) reporting requirement: A news release was completed at 1140 CDT on 7/24/2024, by South Texas Project on the declaration of the Unusual Event. This media release is being reported in accordance with 10 CFR 50.72(b)(2)(xi): Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Notified R4DO (Azua)
The following information was provided by the licensee via phone and email: After a review of station logs, it was determined that there was not a loss of all offsite AC power to Unit 1 related to the event that occurred on July 24, 2024. An offsite AC power source was available through the 138 kV transmission line. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At this time the licensee is not retracting the declaration of emergency action level 'SU.1'. Notified R4DO (Werner) |
| ENS 57225 | 12 July 2024 19:37:00 | McGuire | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email and phone: On July 12, 2024, at 1337 EDT, operations discovered that the technical support center (TSC) ventilation system was non-functional, which resulted in an unplanned loss of the TSC that could not be restored within seventy-five minutes. If an emergency had been declared requiring TSC activation during this period, the TSC would have been staffed and activated using existing emergency planning procedures. If relocation of the TSC had been necessary, the emergency coordinator would have relocated the TSC staff to an alternate location in accordance with applicable site 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 loss of the TSC ventilation system affected 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. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The TSC ventilation system remains out-of-service at the time of the notification. In the event of an emergency, the licensee will use the alternate TSC facility per applicable site procedures until the ventilation system is restored. Repair of the ventilation system is being worked around-the-clock. |
| ENS 57214 | 8 July 2024 18:24:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1521 EDT on July 8, 2024, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dump system and the auxiliary feedwater (AFW) system. 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). 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 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 specific cause of the turbine trip is under investigation by the licensee. |
| ENS 57211 | 7 July 2024 23:12:00 | Byron | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1440 CDT on 7/7/2024, it was discovered that both trains of the control room ventilation temperature control system were simultaneously inoperable. Due to this inoperability, the system was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. One train of control room ventilation temperature control was restored to operable status at 1634 CDT on 7/7/2024. |
| ENS 57201 | 30 June 2024 16:13:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1222 CDT on 06/30/24, South Texas Project notified the Texas Commission on Environmental Quality (TCEQ) and the National Response Center regarding an oil spill of 280 gallons to the ground inside the protected area. This event was recorded as TECQ Event number: 20242394 and Incident Response Center incident number: 1403420. The spill occurred due to the overflow of an oily waste sump. The influent to the sump was stopped. The spill was confined to a 15 feet by 15 feet area on the ground and did not enter any waterway. This notification is being made solely as a four-hour, non-emergency notification for a notification to another government agency. This was determined to be reportable as required by 10CFR50.72(b)(2)(xi). There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified. |
| ENS 57195 | 27 June 2024 14:03:00 | Cook | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee email: On June 27, 2024, at 0804 (EDT), D.C. Cook Unit 2 had an automatic start of the turbine driven auxiliary feedwater pump (TDAFP) following a controlled down power and manual reactor trip at approximately 17 percent power. The automatic start of the TDAFP was due to a steam generator water level 'low low' signal following the reactor trip. The down power and trip were performed in accordance with normal shutdown procedures to comply with the required action C.1 of technical specification 3.4.13, 'reactor coolant system operational leakage.' Reference event notification number EN57194. An automatic start of the TDAFP is an eight hour report per 10CFR 50.72(b)(3)(iv)(A). Unit 2 is being supplied by offsite power. All control rods fully inserted. Steam generators are being fed by both motor driven auxiliary feedwater pumps. Decay heat is being removed via the steam dump system to the main condenser. Preliminary evaluation indicates plant systems functioned normally following the reactor trip. D.C. Cook Unit 2 remains in Mode 3 to repair the previously reported reactor coolant system leakage through valve 2-QRV-251, 'CVCS (chemical and volume control system) charging pumps discharge flow control' valve packing. The NRC Resident Inspector has been notified. |
| ENS 57194 | 27 June 2024 04:30:00 | Cook | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0159 EDT on 6-27-24, DC Cook Unit 2 commenced a controlled shutdown to address identified leakage through 2-QRV-251, CVCS (chemical and volume control system) charging pumps discharge flow control valve, valve packing. The leakage started at 2303 EDT on 6-26-24 and was determined to be greater than the 10 gallons per minute limit per technical specification 3.4.13 allowable leakage. The leakage was determined to not be repairable at power. Per 10CFR 50.72(b)(2) this is a 4 hour non-emergency report for Technical Specification required shutdown." The NRC Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: A normal shutdown is expected to be completed by 0900 EDT. There are no radiological releases. Unit 1 was not affected. There is no estimate for the time of the repairs. |
| ENS 57187 | 22 June 2024 11:44:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0728 EDT on 06/22/2024, with Unit 2 in Mode 3 at zero percent power and the reactor trip breakers closed, a manual actuation of the RPS was initiated during the withdrawal of the shutdown rods in preparation for Mode 2. This was procedurally directed due to a shutdown rod being misaligned from the other rods in the bank due to a malfunction. Units 1, 3 and 4 were not affected. Due to the manual actuation of the RPS, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57153 | 30 May 2024 17:43:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email and phone: On May 30, 2024, at 1200 CDT, South Texas Project (STP) FFD management identified from industry operating experience (OE) a programmatic failure, degradation, or discovered a vulnerability of the fitness for duty (FFD) program that may permit undetected drug or alcohol use or abuse by individuals within a protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. A review of the personnel in-processed and placed into the follow-up program by STP and external utilities since the implementation of the Illuminate software (07/31/2023) was completed. The issue affecting individuals placed into the follow-up program by external utilities was bound to in-processing of individuals (between) 02/22/2024 and 04/09/2024. One other individual processed in November of 2023, was also affected by this event. This event did not impact STP personnel that were either placed or had an existing record in the follow-up program. Compensatory measures were implemented and an extent of condition review was completed. This is a 24-hour reportable event per 10 CFR 26.719(b)(4). The NRC Resident Inspector has been notified. |
| ENS 57146 | 26 May 2024 13:52:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee by phone and email: At 0210 CDT 5/24/24, essential chiller A train and cascading equipment was declared inoperable for maintenance to correct a temperature control malfunction. At 0720 CDT 5/26/24, essential cooling water B train and cascading equipment (including B train essential chiller) was declared inoperable due to a through wall leak discovered on the essential cooling water return header temperature element thermal well. This condition resulted in an inoperable condition on two out of three safety trains for the accident mitigating function, including the train A and train B high head safety injection, low head safety injection, containment spray, electrical auxiliary building heating ventilation and air conditioning (HVAC), and essential chilled water. All C train safety related equipment remains operable. This was determined to be reportable within 8 hours as required by 10CFR50.72(b)(3)(v)(D). NRC Resident Inspector has been notified.
The following information was provided by the licensee by phone and email: This is a communication to retract the 8-hour notification Event Notification (EN) 57146 reported to the NRC pursuant to 10 CFR 50.72(b)(3)(v)(D) on 05/26/2024. Based on a subsequent engineering review of the conditions that existed at the time of discovery, it was determined that: 1) The maximum postulated leak rate, conservatively estimated, from the B Train essential cooling water return piping thermowell in the mechanical auxiliary building sump room would have been less than the administrative allowable limit for leakage in this room during a design basis accident, 2) No adjacent safety related components or functions would have been adversely affected, and 3) the return line leakage represented a negligible impact regarding essential cooling water system inventory and the system ability to cool required components. Therefore, it was recommended that the B Train essential cooling water system with the as-found leakage condition be considered operable. Therefore, this event notification is being retracted. The NRC Resident Inspector has been notified. Notified R4DO (Taylor) |
| ENS 57145 | 25 May 2024 15:26:00 | Millstone | NRC Region 1 | CE Westinghouse PWR 4-Loop | The following information was provided by the licensee by phone and email: A 50 ml bottle of vodka was found in the Unit 3 debris basket on the exterior of the intake structure. The bottle likely came from the ultimate heat sink (Niantic Bay) during normal backwash operations by the system that collects debris. Security has discarded the contraband. 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 bottle was found unsealed. |
| ENS 57128 | 15 May 2024 05:50:00 | Cook | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On May 15, 2024 at 0427 EDT, DC Cooks Unit 2 reactor was manually tripped due to difficulty maintaining steam generator water levels. DC Cook Unit 2 had removed the main turbine from service at approximately 0354 EDT during a planned down-power to repair a steam leak on the high pressure turbine right outer steam/stop control valve upstream drip pot. Stable steam generator water levels were unable to be maintained. As a result, DC Cook Unit 2 was manually tripped with reactor power stabilizing at approximately 20 percent. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System actuation as a four hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight hour report. The reactor trip was not complicated and all plant systems functioned normally. The DC Cook NRC Resident Inspector was notified. |
| ENS 57126 | 13 May 2024 16:40:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0917 EDT on May 13, 2024, a control room operator erroneously rendered the `B train of the Unit 2 residual heat removal (RHR) system inoperable. This occurred while the `A train of the Unit 2 RHR system was out of service for preplanned maintenance. RHR serves as the low head safety injection (LHSI) subsystem for the emergency core cooling system (ECCS) and because of this, Unit 2 was without a required train of ECCS from 0917 EDT to 0921 EDT. No other equipment issues were identified. The LHSI subsystem is credited by the analysis for a large break loss of coolant accident at full power. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC resident inspector has been notified. There is no release of radioactive material associated with this event. |
| ENS 57124 | 12 May 2024 20:46:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email and phone: At 1641 CDT on May 12, 2024, with Unit 2 in Mode 1 at 15 percent power, the reactor automatically tripped due to a unit auxiliary transformer lockout. During the trip, all control rods fully inserted. The cause of the transformer lockout is currently unknown. Emergency diesel generator (EDG) 21 and 23 actuated and all three engineered safety feature (ESF) busses were energized. All equipment responded as expected except for steam generator power operated relief valve (PORV) 2C which failed to open when required in automatic, and the load center (LC) E2A output breaker which failed to close automatically but was closed manually. Steam generator PORV 2C did open when placed in manual, although it subsequently failed to full open and was then closed. Primary system temperature and pressure are currently being maintained at 567 degrees/2235 psig following start of reactor coolant pumps 2A and 2D. Due to the reactor protection system actuation (RPS) while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the emergency diesel generators. 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: South Texas Project Unit 2 was in Mode 1 at 15 percent power due to performance of testing and analysis on the main turbine prior to the RPS actuation.
The following information was provided by the licensee via email and phone: South Texas Project is submitting the following correction to the event notification: The steam generator (SG) power operated relief valve (PORV) 2C did not fail to open automatically. System pressure during this event did not reach the automatic setpoint for the PORV (1225 psi), and there was no demand for it to open automatically. During the event, SG PORV 2C was taken to manual and it went full open when the up button was pushed slightly. It went closed when the down button was pressed to close it manually. In addition, the load center E2A output breaker initially failed to close automatically, however, after operations placed it in pull-to-lock and returned the hand switch to automatic, it closed automatically. Notified R4DO (Dixon). |
| ENS 57107 | 5 May 2024 08:11:00 | Braidwood | NRC Region 3 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email and phone: At 0338 CDT, with the unit 1 in mode 1 at 6 percent power, the reactor automatically tripped due to lowering steam generator water level. The trip was uncomplicated with all systems responding normally post-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) for an actuation of the auxiliary feedwater system. Operations responded using procedure 1BwEP-0 and stabilized the plant in mode 3. Decay heat is removed by steam dumps via the main condenser. 1A and 1B auxiliary feedwater pumps were actuated manually prior to the reactor trip in an attempt to restore steam generator water level. Unit 2 is not affected. 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 email and phone: At 0338 CDT, with the unit 1 in mode 2 at 3 percent power, the reactor automatically tripped due to lowering steam generator water level. The trip was uncomplicated with all systems responding normally post-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 10CFR50.72(b)(3)(iv)(A) for an actuation of the auxiliary feedwater system, eight-hour notification. Operations responded using procedure 1BwEP-0 and stabilized the plant in mode 3. Decay heat is being removed by steam dumps via the main condenser. 1A and 1B auxiliary feedwater pumps were actuated manually prior to reactor trip in an attempt to restore steam generator water level. Unit 2 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Notified R3DO (Hartman) |
| ENS 57091 | 25 April 2024 13:38:00 | Callaway | NRC Region 4 | Westinghouse PWR 4-Loop | 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), under the provision of 10 CFR 50.73(a)(1), detailing the event in which an unplanned actuation of the turbine driven auxiliary feedwater pump (TDAFP) at the Callaway plant occurred in response to an invalid actuation signal. The actuation occurred at 2056 (CDT) on 3/21/2024 during restoration from maintenance on the NN12 inverter. The actuation signal was received while closing breaker NK0211 (for connecting the inverter to its associated 125-VDC bus). In response to the TDAFP actuation, operators closed the flow control valves and reduced turbine load by approximately 10 MW electrical. Initial investigation showed that a spurious manual actuation signal had been received and cleared 5 seconds later. The direct cause of the event was due to a voltage transient generated on the NK02 125-VDC bus during closure of the NK0211 breaker. The actuation occurred due to degradation of a 48-VDC power supply (PS1) within engineered safety features actuation system (ESFAS) logic cabinet SA036C. The power supply exhibited elevated ripple during testing as part of troubleshooting efforts, which was indicative of degradation of the regulation circuitry within the supply. This degradation prevented the power supply from sufficiently filtering the transient that occurred on the 125-VDC bus associated with the NN12 inverter. The power supply was replaced. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee originally submitted this event under 10 CFR 50.72(b)(3)(iv)(A) in EN 57043. The licensee has retracted EN 57043. |
| ENS 57077 | 15 April 2024 14:38:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in Mode 1 at 100 percent power, an invalid start of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The 1A-A, 1B-B, and 2B-B EDGs started and functioned successfully. The start signal for the 1A-A, 1B-B, and 2B-B EDGs was generated from the common emergency start of the 2A-A EDG. The signal was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design (basis) safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. This event was originally reported under EN 56970 on February 16, 2024, at 0205 EST in accordance with 10 CFR 50.72(b)(3) (iv)(A) as an event that results in a valid actuation of the emergency diesel generator system. This EN was retracted on February 21, 2024, at 1549 EST. This event is being reported in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) as an event that results in an invalid actuation of the emergency diesel generator system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57043 | 22 March 2024 01:46:00 | Callaway | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2056 on 3/21/24, Callaway Plant was in Mode 1 at approximately 100 percent power when an automatic start of the turbine driven auxiliary feedwater pump occurred. The event occurred while restoring inverter NN12 from maintenance. NN12 is the normal in-service inverter for the group 2 120-VAC instrument bus (NN02). The actuation occurred while swapping from the swing inverter (NN18) to the normal in-service inverter (NN12). All safety systems responded as expected. At 2334, the turbine driven auxiliary feedwater pump was secured. The plant is being maintained in a stable condition, in mode 1. The NRC Resident Inspector was notified The licensee is investigating the cause of the automatic start.
Event Notification (EN) 57043, made on 03/21/2024 pursuant to 10 CFR 50.72(b)(3)(iv)(A), is being retracted based upon further investigation into the cause of the turbine driven auxiliary feedwater pump (TDAFP) actuation. The TDAFP received an invalid manual initiation signal caused by a voltage transient that was generated on the NK02 125-VDC bus upon closure of downstream breaker NK0211 (while restoring inverter NN12 from maintenance). This actuation signal was due to degradation of a 48-VDC power supply (PS1) within engineered safety features actuation system (ESFAS) logic cabinet SA036C. This degradation likely prevented the power supply from sufficiently filtering the transient that occurred on the 125-VDC bus associated with the NN12 inverter. Notified R4DO (Warnick) |
| ENS 57033 | 17 March 2024 17:59:00 | Comanche Peak | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On March 17, 2024, at 1515 CDT, the Comanche Peak Unit 2 reactor was manually tripped due to an anticipated automatic trip due to lo-lo steam generator (SG) water levels. Prior to the trip, main feedwater pump '2B' tripped and an auto runback to 700 MW (60 percent power) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs. Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedures IPO-007B. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the '2B' main feed pump trip was due to loss of primary and redundant power to the servo control valve. The loss of power to the servo control valve is under investigation. |
| ENS 57026 | 13 March 2024 02:29:00 | Catawba | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On March 12, 2024, at 2111 EDT, a valid containment ventilation isolation train 'A' and 'B' signal was received due to a spurious loss of power to 1EMF-38 (containment particulate radiation monitor) and 1EMF-39 (containment gas radiation monitor). The power to 1EMF-38 and 1EMF-39 was restored. 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: There were no plant evolutions ongoing at the time of the event and the cause of the loss of power is under investigation. There was no impact to Unit 2.
After further review of the event, it was determined the actuation of the associated containment ventilation isolation train 'A' and 'B' was not valid. This is due to the loss of power being associated with the control room modules for 1EMF-38 and 1EMF-39, and not a result of an actual sensed parameter or plant condition. Therefore, this event notification is being retracted. The NRC Resident Inspector has been notified. Notified R2DO (Miller) |
| ENS 57024 | 12 March 2024 12:16:00 | Comanche Peak | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: On March 12, 2024, at 0816 CDT, Comanche Peak Unit 2 reactor automatically tripped on lo-lo level in the 2-03 steam generator (SG). Prior to the trip, main feedwater pump (MFP) 2A speed reduced and a manual runback to 700 MW (60 percent) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs. Concurrent with the loss of speed on MFP 2A, a servo filter swap was in progress on MFP 2A. Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedure IPO-007A. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the loss of the MFP is under investigation. Unit 1 was unaffected. |
| ENS 57019 | 10 March 2024 12:05:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On 3/9/2024 at 2126 CST, train C essential cooling water was declared inoperable due to a through-wall leak on the discharge vent line. This would also cascade and cause train C essential chilled water to be inoperable. On 3/10/2024 at 0353 CDT, train B essential chilled water was declared inoperable due to chilled water outlet temperature greater than 52 degrees F following startup of essential chiller 12B. Chilled water outlet temperature was adjusted to less than 52 degrees F at 0440 CDT, and train B essential chilled water was declared operable. This condition resulted in the inoperability of two of the three safety trains required for the accident mitigating functions including: high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chilled water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector. |
| ENS 57010 | 5 March 2024 19:14:00 | Comanche Peak | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email:
On March 5, 2024, at 0720 CST, the X-02 118V uninterruptible power supply air conditioning (UPS A/C) unit tripped with the associated emergency fan coil units (EFCUs) shut down for planned maintenance in the area. The X-01 UPS A/C unit was declared inoperable upon discovery due to a scheduled outage of support systems (Unit 1 station service water) via the safety function determination process. This placed the site in technical specification 3.7.20 condition A, B, and C to restore the UPS A/C system within one hour. The EFCUs were restarted at 0729 which satisfied condition B and C, and X-01 UPS A/C unit was aligned to Unit 2 cooling water at 0801, exiting condition A. The condition that could have prevented the fulfillment of the safety function lasted for approximately nine minutes. Area temperatures had no notable change based on field observations during the condition. The UPS HVAC system provides temperature control for the safety related UPS and distribution rooms during all normal and accident conditions. The UPS HVAC system consists of (a) a dedicated UPS room EFCU in each safety-related UPS and distribution room, and (b) two electrically independent and redundant A/C trains either of which can support all four safety related UPS and distribution rooms; each train consists of an air conditioning unit, ductwork, dampers, and instrumentation. The NRC Resident Inspector has been notified. |
| ENS 57006 | 5 March 2024 04:14:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. 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 auxiliary feedwater 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 on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. |
| ENS 56983 | 21 February 2024 10:33:00 | Wolf Creek | NRC Region 4 | Westinghouse PWR 4-Loop | The following is a synopsis of information that was provided by the licensee via email and phone call: A non-licensed supervisor had a confirmed positive during a fitness for duty test. The supervisor's access to the plant has been terminated. |
| ENS 56970 | 16 February 2024 02:05:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in mode 1 at 100 percent power, an actuation of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The reason for the emergency diesel generator system auto-start was clearance removal sequencing errors. The emergency diesel generator system automatically started as designed when the common emergency start signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the emergency diesel generator system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: In accordance with NUREG-1022, Section 2.8 and Section 4.2.3, Watts Barr is retracting the previous report EN 56970 pursuant to 10 CFR 50.72(b)(3)(iv)(A). The start signal for the 1A-A, 1B-B, and 2B-B emergency diesel generators (EDG)s was from activation of the common emergency start of the 2A-A EDG. The actuation was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. Since the starts were not initiated via an automatic signal from a LOOP, SI, or traditional operator action, the signal is not a valid actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). Therefore, EN 56970 is being retracted. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Miller) |
| ENS 56968 | 15 February 2024 05:45:00 | Callaway | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0247 CST on 2/15/2024, Callaway Plant was in mode 1 at approximately 100 percent power when a turbine trip and reactor trip occurred. All safety systems responded as expected with the exception of an indication issue on the feedwater isolation valves, which were confirmed closed. A valid feedwater isolation signal and auxiliary feedwater actuation signal were also received as a result of the reactor trip. The plant is being maintained stable in mode 3. All control rods fully inserted from the reactor trip signal and decay heat is being removed via the auxiliary feedwater system and steam dumps. The NRC Resident Inspector was notified. |