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| Event date | Site | Region | Reactor type | Event description | Reporting criterion | |
|---|---|---|---|---|---|---|
| ENS 57444 | 26 November 2024 16:58:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: At 1158 EST on 11/26/24, the Division 3 diesel generator was declared inoperable due to failure of the right bank air start motor during a planned monthly surveillance run. Troubleshooting of the issue is in progress. This condition could prevent 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). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Division 3 diesel generator supports high pressure core spray, a single train system. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57438 | 23 November 2024 00:37:00 | 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. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57431 | 19 November 2024 18:50:00 | 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. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57432 | 19 November 2024 18:50:00 | 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. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57426 | 15 November 2024 19:05:00 | Monticello | NRC Region 3 | GE-3 | The following information was provided by the licensee via phone and email: At 1305 CST, on November 15, 2024, it was determined that division 2 cables for the '12' emergency diesel generator start circuitry are routed through a division 1 area without adequate fire barrier separation. 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. A fire impairment and hourly fire watch have been established for the affected fire zones. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72 (b)(3)(ii)(B). The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition |
| ENS 57425 | 15 November 2024 15:01:00 | Saint Lucie | NRC Region 2 | CE | The following information was provided by the licensee via phone and email: At 1001 EST, on November 15, 2024, with Unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to three control element assemblies fully inserting into the core. The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat is being removed by the steam bypass control system and main feedwater. Unit 2 was not affected. This event is being reported pursuant to 10 CFR 50.72 (b)(2)(iv)(B). 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 insertion of the three control rods is suspected to be caused by an electrical failure; however, the cause is still being investigated. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 57420 | 11 November 2024 22:31:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 2250 EST on November 11, 2024, a technical specification required shutdown was initiated at Beaver Valley Power Station Unit 2. The following technical specification limiting conditions of operation (LCOs) were entered at 1939 EST on November 11, 2024: LCO 3.6.3, containment isolation valves, condition C, one or more penetration flow paths with one containment isolation valve inoperable; required action C.1, isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. LCO 3.7.2, main steam isolation valves (MSIVs), condition C, one or more MSIVs inoperable in mode 2 or 3; required action C.1, close MSIV within 8 hours. These technical specification required actions will not be completed within the completion time; therefore, a technical specification 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). With one main steam isolation valve inoperable, this condition is also 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: The failure occurred during planned surveillance testing in preparation for reactor startup. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material |
| ENS 57419 | 11 November 2024 20:10:00 | Pilgrim | NRC Region 1 | GE-3 | The following is summary of information provided by the licensee via phone and email: On November 11, 2024, at 1510 EST, site personnel identified what appeared to be water bubbling up from the pavement adjacent to the sanitary lift station 'C' outside of the facility industrial area. Less than 100 gallons of non-radiological sanitary water ran to a catch basin connected to permitted outfall number 007. Visual inspection did not identify any odor or indication of flow at outfall number 007 discharge. By 1530, the lift station pumps had been secured, sources of influent to the lift station were removed from service, and efforts were underway to pump the tank. At 1611, an offsite notification was made to the Environmental Protection Agency's Enforcement and Compliance Assurance Division in accordance with Section B of the station's National Pollutant Discharge Elimination System (NPDES) Permit No. 0003557. The event was associated with leakage from underground sewage system piping from a non-radiological underground tank and lift station. The NRC Resident Inspector will be notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57418 | 10 November 2024 09:37:00 | Grand Gulf | NRC Region 4 | GE-6 | The following information was provided by the licensee via phone and email: On November 10, 2024, at 0337 CST, Grand Gulf Nuclear Station (GGNS) was operating in mode 1 at 100 percent power when a manual scram was initiated due to degrading main condenser vacuum. The cause of the degrading main condenser vacuum is not known at this time and is being investigated. All control rods fully inserted and there were no complications. Reactor pressure was initially maintained with main turbine bypass valves. Reactor water level was initially maintained with main feedwater and condensate. At 0457, operators transitioned pressure control to safety relief valves and began using reactor core isolation cooling (RCIC) to maintain reactor water level. This was performed using plant procedures due to degrading vacuum. GGNS is currently in mode 3. Reactor level is being maintained with RCIC and pressure is being maintained using the safety relief valves. The manual reactor protection system (RPS) actuation is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) and the RCIC actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time of the notification, main steam isolation valves had shut on low vacuum.
The following update was provided by the licensee via phone and email: This update is being made to report the following occurrences which took place after the scram reported in event number 57418. On November 10, 2024, at 0545 CST, a group 1 containment isolation signal resulted in the closure of all MSIVs. The signal was due to continued degradation of condenser vacuum post-trip. At 0620, an automatic RPS actuation occurred when reactor water level lowered to level 3. This RPS actuation occurred with all control rods fully inserted. Reactor water level lowered following closure of an open safety relief valve and was recovered to within the established band. The events are being reported as specified system actuations in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Dixon) | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57412 | 4 November 2024 11:00:00 | Monticello | NRC Region 3 | GE-3 | The following information was provided by the licensee via phone and email: At 0500 CST, on November 4, 2024, the Monticello Nuclear Generating Plant was notified by Wright County dispatch of a spurious actuation of one emergency response siren that lasted approximately ten minutes. The cause of the actuation has not been determined and the vendor is investigating. The siren is no longer actuating. There was no impact to the health and safety of the public as a result of this event and the offsite response capabilities remain functional. No press release by the licensee is planned at this time. This event is being reported in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57411 | 4 November 2024 04:12:00 | 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). | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration |
| ENS 57408 | 30 October 2024 13:05:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone or email: At 0905 EDT, on October 30, 2024, a courtesy notification was made to OSHA for a contractor working at Turkey Point who was transported to an offsite medical facility for treatment of a personal medical condition. Upon arrival at that facility, medical personnel declared the individual was deceased. This event is being reported pursuant to 10 CFR 50.72(b)(2)(xi). The Resident Inspector has been notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57407 | 29 October 2024 21:30:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: At 1730 (EDT), on 10/29/2024, Vistra authorized a report to the State of Ohio in accordance with NEI 07-07, 'Industry Groundwater Protection Initiative,' of a liquid spill to the plant's outdoor ground from an outdoor cask holding radioactive waste. This spill contained radionuclides cobalt-60 and manganese-54 with activities greater than the site procedural limits for state reporting. The outdoor cask has been removed from the onsite storage area; remediation and causal investigation activities are being performed. This exceedance did not exceed any NRC regulations or reporting criteria. There was no impact on the health and safety of the public or plant personnel. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No EPA limits were exceeded | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57406 | 29 October 2024 15:52:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: At 1152 EDT, on 10/29/2024, the division 3 diesel generator was declared inoperable due to variations in steady state voltage during a planned monthly surveillance run. Troubleshooting of the issue is in progress. This condition could prevent 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). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Limiting condition for operations (LCO) 3.8.1.b was entered as a result of this event. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57404 | 29 October 2024 15:23:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via phone and email: At 1023 CDT, on 10/29/2024, with Unit 2 in Mode 1 at 100 percent (reactor) power, during the performance of surveillance procedure 2047, CONTROL ROD QUARTERLY EXERCISE, Unit 2 experienced an automatic reactor trip (followed by a) turbine trip. All control rods fully inserted into the core following the reactor trip. The cause of the Unit 2 reactor trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generators through the main steam dumps to the condenser. The Unit 2 reactor trip was not complex with all systems responding normally post trip. Due to the actuation of the reactor protection system (RPS) of Unit 2, this event is being reported as a RPS Actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B). Due to the actuation of the auxiliary feedwater system following the Unit 2 reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 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. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57401 | 28 October 2024 17:23:00 | Oconee | NRC Region 2 | B&W-L-LP | The following information was provided by the licensee via phone or email: On October 28, 2024, at 0533 EDT, it was discovered both trains of the unit 1 and unit 2 control room ventilation system booster fans were simultaneously inoperable due to trip of the supply breaker to the motor control center (MCC) supplying the normally closed motor operated intake dampers for both trains. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). At 1059 EDT, one train of the unit 1 and unit 2 control room ventilation system booster fans was restored to operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Although the control room ventilation affects both units 1 and 2, unit 1 is currently defueled and outside the mode of applicability during the timeframe of this event. For clarification regarding train separation, both trains for control room ventilation have separate power supplies for their booster fans. However, power for both of their inlet dampers and one of the booster fans are all fed from the same MCC supply breaker that tripped. Therefore, both trains were simultaneously inoperable at 0533 EDT. The loss of power to the MCC is still under investigation. Procedurally, Oconee was able to restore one train to service, the train that still had power to its booster fan, at 1059 EDT, by manually opening its inlet damper in the fail-safe position. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57397 | 23 October 2024 16:34:00 | River Bend | NRC Region 4 | GE-6 | The following information was provided by the licensee via phone and email: A non-licensed contract supervisor violated the station's fitness-for-duty program. The employee's access to the plant has been terminated. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57384 | 16 October 2024 12:30:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: A non-licensed contract supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57383 | 15 October 2024 16:06:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access has been terminated. The NRC Resident Inspector has been notified. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57379 | 14 October 2024 14:34:00 | 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. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57373 | 10 October 2024 14:45:00 | Davis Besse | NRC Region 3 | B&W-R-LP | The following information was provided by the licensee via phone and email: On October 10, 2024, at 1045 EDT, it was determined that an additional failure mechanism existed for a previous issue with the emergency diesel generator (EDG) speed switches. As a result, for some initiating events such as a fire or high energy line break potentially causing a ground on the negative buses on both trains of the station direct current (DC) system, both EDG speed switches could have been affected, resulting in a failure of the EDGs to function. 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, as the EDG speed switches were replaced in January 2022 with a different design that is not susceptible to grounds on the station DC system. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition |
| ENS 57422 | 10 October 2024 14:02:00 | 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. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material |
| ENS 57372 | 10 October 2024 09:57:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via phone and email: At 0557 EDT on 10/10/2024, with Unit 2 in mode 1 at 100 percent power, the reactor automatically tripped due to turbine generator loss of field. The trip was not complex 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). Operations responded using EOP-0, post trip immediate actions, and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 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 additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The exciter is suspected to being the cause and is under investigation. All control rods fully inserted. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| ENS 57369 | 8 October 2024 19:31:00 | Arkansas Nuclear | NRC Region 4 | CE | The following information was provided by the licensee via phone and email: At 1431 CDT, on October 8, 2024, Arkansas Nuclear One, Unit 2 (ANO-2) completed the analysis related to an indication revealed on head penetration '71' during reactor vessel closure head inspections. It was determined that the indication is not acceptable under the American Society of Mechanical Engineers (ASME) code requirements. The indication displays characteristics of abnormal degradation of a barrier that requires taking corrective actions to ensure the barriers capability. No leak path signal was identified during ultrasonic testing or bare metal visual inspections. The plant was in cold shutdown at zero percent power and defueled for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. This is the only indication that is currently present; however, if additional indications are found, they will also be repaired prior to the plant startup. The NRC Senior Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 57366 | 7 October 2024 21:46:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: On October 7, 2024 at 1444 EDT, a contract worker at Turkey Point was transported off-site for treatment at an off-site medical facility. On October 7, 2024 at 1746 EST, a courtesy notification was made to OSHA for an individual who was transported to an offsite medical facility for treatment of a personal medical condition. Upon arrival at that facility, medical personnel declared the individual was deceased. This event is being reported pursuant to accordance 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 53456 | 7 October 2024 09:43:00 | Arkansas Nuclear | NRC Region 4 | On June 12, 2018, at 1500 CDT, a Reactor Coolant System (RCS) Pressure Boundary leak was identified during a Mode 3, hot shutdown walkdown on a High Pressure Injection Line (HPI) to Reactor Coolant Pump (P32C) drain line weld near MU-1066A HPI Line Drain Valve and MU-1066B HPI Line Drain Valve. The 3/4 inch drain line containing drain valves MU-1066A and MU-1066B on the 'C' HPI header (CCA-5 pipe class) has a through-wall defect on the pipe stub or welds between the sockolet and valve MU-1066A. The leak location is in the ASME Class I RCS Pressure Boundary. The hot shutdown walkdown was being performed as part of a planned outage to investigate excessive Reactor Building Sump inleakage. Total unidentified RCS leakage prior to the investigation was determined to be at 0.165 gpm. After the initial investigation of the leakage, the following Tech Specs (TS) were determined be applicable: TS 3.4.5 - RCS Loops Mode 3, TS 3.4.13 - RCS Leakage, TS 3.5.2 - ECCS. Unit 1 is currently in Mode 3 and in progress of an RCS cooldown to comply with Tech Spec requirements. The licensee notified the NRC Resident Inspector. | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | |
| ENS 57361 | 4 October 2024 02:35:00 | Vogtle | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 2235 on 10/03/2024, the Vogtle 1 and 2 seismic monitoring panel experienced an electrical fault, rendering the panel nonfunctional. Compensatory measures for seismic event classification have been implemented in accordance with Vogtle procedures. This is an eight-hour, non-emergency notification for a loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the seismic monitoring panel is the method for evaluating that an operational basis earthquake (OBE) threshold has been exceeded following a seismic event. This is in accordance with Initiating condition `seismic event greater than OBE levels' and emergency action level HU2. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 57424 | 1 October 2024 12:38:00 | 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. | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect |
| ENS 57353 | 30 September 2024 18:40:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via phone and email: At 1340 CDT on 9/30/2024, Prairie Island Nuclear Generating Plant was informed that the Minnesota State Duty Officer and the Environmental National Response Center had been notified of a coolant leak from a diesel driven cooling pump expansion tank that had reached the waters of the state. The estimated quantity is 0.7 gallons of NALCO LCS-60. The leak has been isolated. The NRC Resident Inspector has been notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57351 | 29 September 2024 09:56:00 | 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. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57349 | 27 September 2024 23:22:00 | Grand Gulf | NRC Region 4 | GE-6 | The following information was provided by the licensee via phone and email: At 1822 CST on September 27, 2024, Grand Gulf Nuclear Station (GGNS) was conducting surveillance testing on the high pressure core spray (HPCS) division Ill diesel generator. Following initiation of the test signal, the HPCS pump room cooler start time exceeded the surveillance procedure allowance of less than or equal to 20 seconds. The HPCS pump room cooler started in 26.2 seconds. HPCS was already inoperable for performance of the surveillance testing. The event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition which could have prevented the fulfillment of a safety function. Troubleshooting is in progress. HPCS, a single-train system, will remain inoperable until the condition is corrected. All sources of offsite power are available. No other safety systems are inoperable. Reactor core isolation cooling was verified to be operable per GGNS technical specification 3.5.1.B.1. The NRC Senior Resident Inspector has been notified.
Investigation of the delayed start time of the HPCS pump room cooler indicated that the condition would not have challenged the ability of the room cooler to maintain temperatures less than the temperature limit of 150 degrees Fahrenheit. As a result, HPCS remained capable of fulfilling its safety function. Therefore, EN 57349 is being retracted. The NRC senior resident inspector has been notified of this retraction. Notified R4DO (O'Keefe) | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57344 | 27 September 2024 11:46:00 | 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. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57343 | 27 September 2024 07:46:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone: At 0346 EDT on 9/27/24, with Unit 1 in mode 1 at 54 percent power, the reactor was manually tripped due to degrading condenser vacuum secondary to environmental conditions. The trip was not complex with all systems responding normally post trip. Closure of containment isolation valves (CIVs) in multiple systems occurred. Operations responded and stabilized the plant. The reactor protection system actuation while critical event is being reported as a 4-hour non-emergency notification per 10 CFR 52.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 52.72(b)(3)(iv)(a) as an event that results in a valid actuation of the CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat removal is via steam bypass to the main condenser. Unit 2 was not affected. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57437 | 26 September 2024 22:01:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: This notification is a 10 CFR 21.21(a)(2) interim report for General Electric thermal overload relay, model CF124G011, part number DD317A7861P003. A sample of overload relays were sent to PowerLabs for parts quality initiative testing. The results were reviewed by James A. FitzPatrick Nuclear Power Plant (JAF) and a deviation in one relay component was discovered. Testing identified a failure to latch on trip, which is a deviation from the performance characteristics of the relay. Under normal operation, the relay would latch in the tripped state requiring a manual reset of the relay. If the relay with the deviation were installed, the relay would trip when required; however, it would automatically reset. The unexpected reset could result in unintended cycling of associated equipment including repeated exposure to inrush current and potential damage. Bench testing would be expected to identify this condition prior to installation. Based on a review, this potential condition does not affect installed equipment. The affected relay was stored at JAF since July 1998. The cause of the deviation cannot be investigated because the part is not available; however, the evaluation of the potential effect of the condition on equipment where the relay could have been used at JAF is ongoing, and it is expected to be completed by February 28, 2025. This notification is being submitted as an interim report per 10CFR21.21(a)(2). The NRC resident inspector has been notified. | 10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component |
| ENS 57338 | 25 September 2024 03:04:00 | Grand Gulf | NRC Region 4 | GE-6 | The following information was provided by the licensee via phone or email: On September 24, 2024, at 2204 CDT, Grand Gulf Nuclear Station (GGNS) was conducting surveillance testing on the high pressure core spray (HPCS) division III diesel generator. During testing, the HPCS pump breaker unexpectedly tripped after the HPCS diesel generator started and powered the safety bus. The breaker performed its motor protection function and tripped due to an over-frequency indication. The event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition which could have prevented the fulfillment of a safety function. Troubleshooting is in progress. HPCS, a single-train system, will remain inoperable until the condition is corrected. All sources of offsite power are available. No other safety systems are inoperable. Reactor core isolation cooling was verified to be operable per GGNS Technical Specification 3.5.1.B.1. The NRC Senior Resident has been notified. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57333 | 23 September 2024 11:20:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 0720 EDT on September 23, 2024, James A. FitzPatrick was at 100 percent power when an automatic scram occurred as a result of a main turbine trip due to an automatic trip of the generator output breakers; the cause is still under investigation. The scram was not complex. The automatic scram inserted all control rods. A subsequent reactor pressure vessel (RPV) low water level resulted in a group 2 isolation and initiation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems. RCIC did inject, but HPCI did not inject, as expected, based on RPV water level recovery with the feedwater system. Reactor pressure is being maintained by main steam line bypass valves. The plant is stable in Mode 3 with the 'A' reactor feed pump maintaining RPV water level. The initiation of the reactor protection system (RPS) due to the automatic scram signal while critical is reportable per 10 CFR 50.72(b)(2)(iv)(B). The general containment Group 2 isolations and HPCI and RCIC system actuations are reportable 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. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The group 2 containment isolation affects multiple systems.
On 9/23/2024 at 1156 EDT Constellation communications provided a media statement to Oswego area news media contacts summarizing the events that had occurred at Nine Mile Point Unit 2 and FitzPatrick Unit 1. This is a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified. Notified R1DO (Dimitriadis) | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57334 | 23 September 2024 11:20:00 | Nine Mile Point | NRC Region 1 | GE-5 | The following information was provided by the licensee via phone and email: On 9/23/2024 at 0720 EDT, with Unit 2 in mode 1 at 100 percent power, the reactor automatically scrammed due to turbine stop valve closure on a turbine trip. The scram was not complex. Due to the reactor protection system (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Following the scram, reactor water level dropped below level 2 (108.8 inches), starting high pressure core spray (HPCS) and reactor core isolation cooling (RCIC); both injected into the reactor. RCIC is being used with turbine bypass valves to remove decay heat. Due to the emergency core cooling systems HPCS and RCIC discharging into the reactor coolant system, this event is being reported a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A), and an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). In addition, with reactor water level below level 2 (108.8 inches), primary containment isolation signals actuated resulting in group 2 recirculation sample system isolation, group 3 traveling in-core probe (TIP) isolation valve isolation, group 6 and 7 reactor water cleanup isolation, group 8 containment isolations, and group 9 containment purge isolations. This event is being reported as an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). Operations responded using procedure N2-EOP-RPV (1-3) and stabilized the plant in mode 3. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was informed. There was no impact on Unit 1.
On 9/23/2024 at 1156 EDT Constellation communications provided a media statement to Oswego area news media contacts summarizing the events that had occurred at Nine Mile Point Unit 2 and FitzPatrick Unit 1. This is a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified. Notified R1DO (Dimitriadis), NRR EO (Felts), and IR MOC (Grant). | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge |
| ENS 57329 | 18 September 2024 15:15:00 | Columbia | NRC Region 4 | GE-5 | The following information was provided by the licensee via email: On September 18, 2024, Columbia Generating Station determined that lubricating oil was likely released into the plant service water system due to a failed heat exchanger on the reactor feed turbine alpha. Isolation of the heat exchanger from the plant service water system is in progress. The plant service water system returns water to a water basin that contains, at a minimum, 300,000 gallons of water. The water basin is connected to the Columbia River via a blowdown line. The blowdown line was secured at 0739 PDT on 9/18/2024. A visual inspection of the basin did not identify any oil sheen or film, but a sample downstream of the affected heat exchanger revealed an oily sheen in the sample bottle. It does not appear the oil released poses a threat to human health or the environment. However, there could have been a discharge of an unknown quantity of oil into the Columbia River. This matter is immediately reportable under Revised Code of Washington 90.56.280 to the U.S. Coast Guard National Response Center, Washington State Department of Ecology, and to the Energy Facility Site Evaluation Council per National Pollutant Discharge Elimination System (NPDES) permit section S3.F.2.b.i. This condition is being reported pursuant to 10 CFR 50.72(b)(2)(xi) for notification of other government agencies concerning an event related to the health and safety of the public or protection of the environment. Notifications to off-site agencies were performed at 0850 PDT on 9/18/2024. The NRC Resident Inspector has been notified. United States Coast Guard National Response Center Incident Report #141183 Washington State Department of Ecology #733853
The following information was provided by the licensee via email: On September 18, 2024, Columbia Generating Station notified the NRC (event notification 57329) under 10 CFR 50.72(b)(2)(xi) of an offsite notification for potential oil release to the plant service water system and potentially the Columbia River. Further troubleshooting efforts refuted the possibility of leakage from the heat exchanger to the plant service water system. Engineering determined the most likely path of oil leakage was through a seal in the reactor feed turbine 'A' pump. This leakage would exit with seal water leak-off into the marine oil separator into collection barrels which are collected and processed. Since troubleshooting efforts determined oil leakage was not from the heat exchanger to the plant service water system, the offsite notification retractions were completed as of October 10, 2024, and therefore event notification 57329 is being retracted. The NRC Resident Inspector was notified. Notified R4DO (Gaddy). | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57328 | 18 September 2024 00:05:00 | Summer | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email and phone: At 2005 EDT on 9/17/2024, it was discovered that steam propagation door DRCB/501 would not latch properly; thus making the door inoperable. Door DRCB/501 is required as a steam propagation barrier to protect both trains of engineered safety feature equipment from effects of a postulated steam line break. Due to this inoperability, the plant 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. Steam propagation door DRCB/501 was repaired and maintained in the closed and latched position at 2032 EDT on 9/17/2024. | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| ENS 57345 | 17 September 2024 22:00:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57346 | 17 September 2024 22:00:00 | Monticello | NRC Region 3 | GE-3 | The following information was provided by the licensee via email: On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed. | 10 CFR 26.719, FFD Reporting requirements |
| ENS 57326 | 17 September 2024 05:27:00 | Vogtle | NRC Region 2 | W-AP1000 | The following information was provided by the licensee via phone and email: At 0127 EDT on 9/17/2024, with Unit 3 in mode 1 at 100% power, the reactor automatically tripped due to the passive residual heat removal heat exchanger outlet flow control valve failing open. A manual safeguards actuation was initiated due to the lowering pressurizer water level resulting from the reactor coolant system cooldown that was caused by the passive residual heat removal heat exchanger outlet flow control valve failing open. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected. Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The failure of the control valve does not inhibit the residual heat removal system from functioning as it is passive. The reactor coolant system maximum allowable cooldown rate was exceeded (Technical Specification 3.4.3). The limit is 100 degrees F per hour above 350 degrees F. The maximum observed cooldown rate was 226 degrees F per hour. At time 0458 EDT, reactor coolant system temperature is 369.1 degrees F, reactor pressure is 900 psig. Currently, the plant is cooling down and proceeding toward placing shutdown cooling online. | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge |
| ENS 57330 | 17 September 2024 05:00:00 | Point Beach | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by NextEra Energy Point Beach, LLC (NextEra) via phone and email: NextEra makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found in a Westinghouse relay, model NBFD31S, during pre-installation bench testing. Specifically, the relay was found to not function as required due to its internal plunger not operating properly. This malfunctioning caused the plunger to not fully extend and cause the normally open contacts to remain closed. Investigations completed by Westinghouse determined that the plunger would not function properly because its kickout spring was misaligned due to human error. This relay was procured from Westinghouse for safety related nuclear applications. NextEra has concluded that this defect constitutes a substantial safety hazard (SSH). A SSH exists because of the nature of the defect was such that the relay would not be able to perform its safety function if installed, and would result in a loss of redundancy in a safety related system, in this case, the reactor protection system. On September 17, 2024, the Point Beach 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 in accordance with 10 CFR 21.21(d)(3(ii) will be provided within 30 days. Since this defect was discovered prior to installation, in accordance with station requirements for bench testing, and the vendor has concluded that this event is an isolated case, there were no actual impacts on safety related equipment. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Responsible corporate officer: Michael Durbin Site Vice President (920) 755-7854 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect |
| ENS 57325 | 16 September 2024 17:29:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via phone and email: On September 16, 2024, at 1329 EDT, the Fermi 2 active seismic monitoring system provided indication of a potential seismic activity event. Plant abnormal procedures were entered and compensatory measures were met and remain in place. Neither the United States Geological Survey (USGS), nor the next closest nuclear power plant could confirm or validate the readings obtained at Fermi. The seismic monitoring system was declared inoperable to validate the calibration of the system. Fermi 2 has two active seismic monitors. One on the reactor pressure vessel pedestal and one in the high pressure core injection (HPCI) room. Only the HPCI room seismic monitor was declared inoperable. The HPCI room accelerometer is the sole 'trigger' for the seismic recording system (which outputs peak accelerations experienced during a seismic event) and the associated control room alarm. This is used in assessment of the magnitude of an earthquake for emergency action level HU 2.1. The loss of the active seismic monitoring system is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). No seismic activity has been felt onsite and the USGS recorded no seismic activity in the area. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee confirmed alternative means of recognizing a seismic event for emergency plan entry are available. | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness |
| ENS 57322 | 16 September 2024 16:40:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax and phone: On September 16, 2024, at 1240 EDT, with Unit 1 in mode 1 at 100 percent power and Unit 2 in mode 1 at 100 percent power, an Unusual Event was declared due to roads in the area leading to the plant being flooded and having the potential to prohibit plant staff from accessing the site via personal vehicles (Emergency Action Level HU3.4). Current onsite plant staff is sufficient for plant operation. This event is being reported in accordance with 10 CFR 50.72(a)(1)(i) due to the declaration of an emergency classification as specified in the approved Emergency Plan. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The NRC decided to remain in the Normal mode of operation at 1320 EDT. Notified DHS SWO, FEMA Ops Center, CISA Central Watch Officer, FEMA NWC (email), DHS Nuclear SSA (email), CWMD Watch Desk (email).
The following information was provided by the licensee via phone and email: At approximately 1400 EDT on September 17, 2024, the Unusual Event at Brunswick was terminated due to the flood waters receding and roads to the plant becoming passable. The NRC resident inspector has been notified. Notified R2DO (Suber), NRR EO (Felts), IR MOC (Crouch), DHS SWO, FEMA Ops Center, CISA Central Watch Officer, FEMA NWC (email), DHS Nuclear SSA (email), CWMD Watch Desk (email). | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration |
| ENS 57321 | 13 September 2024 13:23:00 | 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. | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| ENS 57314 | 10 September 2024 23:30:00 | Columbia | NRC Region 4 | GE-5 | The following information was provided by the licensee via email: On September 10, 2024, Columbia Generating Station determined that no more than approximately thirty (30) gallons of silicone oil was inadvertently released into a plant service water system due to a failed heat exchanger on a plant installed air compressor. The plant service water system returns water to a water basin that contains at a minimum 300,000 gallons of water. The water basin is connected to the Columbia River via a blowdown line. The compromised heat exchanger and compressor have been isolated and secured from the plant service water system. Although not confirmed, it is suspected that an unknown quantity of silicone oil may have been released to the Columbia River. A visual inspection of the basin did not identify any oil sheen or film, and there are no additional actions needed to mitigate this issue. It does not appear the oil released poses a threat to human health or the environment, however, because there could have been a discharge of an unknown quantity of silicone oil into the Columbia River, this matter is immediately reportable under (Revised Code of Washington) 90.56.280 to the U.S. Coast Guard National Response Center, Washington State Department of Ecology, and to the State of Washington Energy Facility Site Evaluation Council (EFSEC) under the site's National Pollutant Discharge Elimination System (NPDES) permit section S3.F.2.b. This condition is being reported pursuant to 10 CFR 50.72(b)(2)(xi) for notification of other government agencies concerning an event related to the health and safety of the public or protection of the environment. Notifications to off-site agencies are in progress. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The NRC resident inspector will be notified. | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release |
| ENS 57387 | 10 September 2024 17:42:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a licensee event report (LER) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 1342 EDT, on September 10, 2024, the reactor water cleanup (RWCU) inboard primary containment isolation valve (PCIV), and the reactor recirculation pump sample inboard PCIV, unexpectedly closed. At the time of this event, work was in progress replacing a control relay in the residual heat removal (RHR) shutdown cooling inboard isolation PCIV circuitry. This relay replacement required lifting the leads of several wires. The neutral side of the relay was electrically connected with the actuation logic for the inboard RWCU and reactor recirculation pump sample PCIVs; the lifting of this lead resulted in the unexpected closure of these PCIVs. The actuation was not initiated in response to actual plant conditions, nor an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. During this event the PCIVs functioned successfully, and the actuations were complete. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 2 was not affected. | 10 CFR 50.73(a)(1), Submit an LER |
| ENS 57298 | 30 August 2024 15:51:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 1051 CDT on 8/30/2024, during transfer of 4KV shutdown bus 1 to support Unit 1 shutdown activities, the alternate feeder breaker failed to close resulting in 4KV shutdown boards 'A' and 'B' experiencing an under voltage condition. This resulted in 'A' and 'B' diesel generators automatically starting and tying to their respective boards. This condition also caused a loss of reactor protection system (RPS) channel 'A' on Units 1 and 2, resulting in invalid actuation of primary containment isolation system Groups 2, 3, 6, and 8. The failure of the board to transfer was identified during preparation for the evolution, contingency actions were prepared and implemented as planned. The breaker failure to close has been corrected and 4KV shutdown bus 1 is energized on alternate. 4KV shutdown boards 'A' and 'B' have been restored to offsite power supplies and the diesel generators are secured. All systems responded as expected for the loss of voltage. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The change in reactor power from 70 percent to 40 percent was not as a result of the failed breaker, rather Browns Ferry Unit 1's change in reactor power was due to a scheduled reactor shutdown which was in progress. In regards to the Unit 2 loss of channel 'A' RPS, this was not a specified system actuation. The actuation of the 'A' and 'B' diesel generators were the specified system actuation. Although the 'A' and 'B' diesels are common to both Units 1 and 2, only Unit 1 credits these specific diesel generators for accident mitigation. As such, this event is only reportable from Unit 1. Unit 2 did not experience a specified system actuation. | 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| ENS 57291 | 26 August 2024 17:00:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: On 8/26/2024 at 1200 CDT, Farley Nuclear Plant Medical Services identified a false negative quality assurance test. (The contracted laboratory) was provided an adulterated sample of hydrocodone and hydromorphone that was part of a blind performance test. The results from the (contracted laboratory) returned a false negative. This false negative test result will be investigated, and the results reported as required. This event is being reported in accordance with 10 CFR 26.719(c)(3). The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officer Report Guidance: The contracted laboratory was a U.S. Department of Health and Human Service (HHS) certified laboratory.
The following update was provided by the licensee via email: Following further review of the event, it has been determined that this issue is not reportable under 10CFR26.719(c)(3) as the unsatisfactory test was not for a validity screening test. This event is reportable for testing errors in accordance with 10CFR26.719(c)(1) and a 30 day report will be submitted. Notified R2DO (Suber) and FFD Group (email). | 10 CFR 26.719, FFD Reporting requirements |