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 Entered dateSiteRegionReactor typeSystemScramEvent description
ENS 5744426 November 2024 14:14:00PerryNRC Region 3GE-6High Pressure Core Spray
Emergency Core Cooling System
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.
ENS 5743823 November 2024 02:42:00Watts BarNRC Region 2Westinghouse PWR 4-LoopThe 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 5743722 November 2024 12:50:00FitzPatrickNRC Region 1GE-4The 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.
ENS 5743219 November 2024 16:02:00SeabrookNRC Region 1Westinghouse PWR 4-LoopSteam Generator
Feedwater
Decay Heat Removal
Manual Scram
Automatic Scram
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 5743119 November 2024 16:02:00SeabrookNRC Region 1Westinghouse PWR 4-LoopSteam Generator
Feedwater
Decay Heat Removal
Manual Scram
Automatic Scram
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 5742615 November 2024 16:29:00MonticelloNRC Region 3GE-3Emergency Diesel GeneratorThe 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.
ENS 5742515 November 2024 12:14:00Saint LucieNRC Region 2CEFeedwater
Steam Bypass Control System
Control Rod
Manual ScramThe 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.
ENS 5742414 November 2024 10:58:00SeabrookNRC Region 1Westinghouse PWR 4-LoopThe 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 5742213 November 2024 13:38:00MillstoneNRC Region 1Westinghouse PWR 4-LoopSecondary containmentThe 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 5742012 November 2024 01:21:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopMain Steam Isolation ValveThe 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.
ENS 5741911 November 2024 16:49:00PilgrimNRC Region 1GE-3The 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.
ENS 5741810 November 2024 07:59:00Grand GulfNRC Region 4GE-6Feedwater
Reactor Protection System
Main Steam Isolation Valve
Reactor Core Isolation Cooling
Safety Relief Valve
Main Condenser
Control Rod
Manual Scram

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.

  • * * UPDATE ON 11/10/24 AT 1236 CST FROM MELISSA LIMBECK TO KERBY SCALES * * *

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)

ENS 574124 November 2024 09:02:00MonticelloNRC Region 3GE-3The 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.
ENS 574113 November 2024 23:33:00CookNRC Region 3Westinghouse PWR 4-LoopThe 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 5740830 October 2024 10:46:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopThe 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.
ENS 5740729 October 2024 18:07:00PerryNRC Region 3GE-6The 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
ENS 5740629 October 2024 17:32:00PerryNRC Region 3GE-6High Pressure Core Spray
Emergency Core Cooling System
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.
ENS 5740429 October 2024 13:50:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopSteam Generator
Reactor Protection System
Auxiliary Feedwater
Control Rod
Main Steam
Automatic ScramThe 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.
ENS 5740228 October 2024 13:55:00CatawbaNRC Region 2Westinghouse PWR 4-LoopThe 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 5740128 October 2024 13:23:00OconeeNRC Region 2B&W-L-LPThe 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.
ENS 5739723 October 2024 16:24:00River BendNRC Region 4GE-6The 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.
ENS 5738717 October 2024 11:55:00BrunswickNRC Region 2GE-4Primary containment
Shutdown Cooling
Reactor Recirculation Pump
Residual Heat Removal
Reactor Water Cleanup
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.
ENS 5738416 October 2024 11:37:00BrunswickNRC Region 2GE-4The 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.
ENS 5738315 October 2024 16:12:00North AnnaNRC Region 2Westinghouse PWR 3-LoopThe 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.
ENS 5737914 October 2024 13:22:00South TexasNRC Region 4Westinghouse PWR 4-LoopThe 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 5737310 October 2024 11:39:00Davis BesseNRC Region 3B&W-R-LPEmergency Diesel GeneratorThe 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.
ENS 5737210 October 2024 08:45:00Calvert CliffsNRC Region 1CEReactor Protection System
Main Condenser
Control Rod
Automatic ScramThe 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.
ENS 573698 October 2024 18:12:00Arkansas NuclearNRC Region 4CEThe 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.
ENS 573667 October 2024 18:13:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopThe 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.
ENS 573614 October 2024 04:23:00VogtleNRC Region 2Westinghouse PWR 4-LoopThe 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 5735330 September 2024 16:37:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopThe 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.
ENS 5735129 September 2024 08:40:00SequoyahNRC Region 2Westinghouse PWR 4-LoopReactor Protection System
Auxiliary Feedwater
Automatic ScramThe 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 5734928 September 2024 00:35:00Grand GulfNRC Region 4GE-6Reactor Core Isolation Cooling
High Pressure Core Spray

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.

  • * * RETRACTION ON 11/25/2024 AT 1333 EST FROM JEFF HARDY TO IAN HOWARD * * *

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)

ENS 5734627 September 2024 12:10:00MonticelloNRC Region 3GE-3The 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.
ENS 5734527 September 2024 11:43:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopThe 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.
ENS 5734427 September 2024 11:25:00CatawbaNRC Region 2Westinghouse PWR 4-LoopSteam Generator
Reactor Protection System
Auxiliary Feedwater
Manual ScramThe 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 5734327 September 2024 07:02:00HatchNRC Region 2GE-4Reactor Protection System
Decay Heat Removal
Main Condenser
Manual ScramThe following information was provided by the licensee via phone: At 0346 EDT on 9/27/24, with Unit 1 in mode 1 at 54 percent power, the reactor was manually tripped due to degrading condenser vacuum secondary to environmental conditions. The trip was not complex with all systems responding normally post trip. Closure of containment isolation valves (CIVs) in multiple systems occurred. Operations responded and stabilized the plant. The reactor protection system actuation while critical event is being reported as a 4-hour non-emergency notification per 10 CFR 52.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 52.72(b)(3)(iv)(a) as an event that results in a valid actuation of the CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat removal is via steam bypass to the main condenser. Unit 2 was not affected.
ENS 5733825 September 2024 02:06:00Grand GulfNRC Region 4GE-6Reactor Core Isolation Cooling
High Pressure Core Spray
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.
ENS 5733623 September 2024 21:46:00Watts BarNRC Region 2Westinghouse PWR 4-LoopReactor Coolant SystemThe 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 5733423 September 2024 11:08:00Nine Mile PointNRC Region 1GE-5Reactor Coolant System
Reactor Protection System
Reactor Core Isolation Cooling
Primary containment
High Pressure Core Spray
Reactor Water Cleanup
Emergency Core Cooling System
Automatic Scram

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.

  • * * UPDATE ON 9/23/2024 AT 1550 EDT FROM RYAN LOOMIS TO IAN HOWARD * * *

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).

ENS 5733323 September 2024 11:02:00FitzPatrickNRC Region 1GE-4Feedwater
High Pressure Coolant Injection
Reactor Protection System
Reactor Core Isolation Cooling
Reactor Pressure Vessel
Main Steam Line
Control Rod
Automatic Scram

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.

  • * * UPDATE ON 9/23/2024 AT 1540 EDT FROM RYAN PERRY TO SAMUEL COLVARD * * *

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)

ENS 5732918 September 2024 13:31:00ColumbiaNRC Region 4GE-5Service water

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

  • * * RETRACTION ON 10/10/24 AT 1513 EDT FROM CHASE WILLIAMS TO KAREN COTTON * * *

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).

ENS 5733018 September 2024 13:30:00Point BeachNRC Region 3Westinghouse PWR 2-LoopReactor Protection SystemThe 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
ENS 5732817 September 2024 23:38:00SummerNRC Region 2Westinghouse PWR 3-LoopThe 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.
ENS 5732617 September 2024 04:48:00VogtleNRC Region 2W-AP1000Reactor Coolant System
Reactor Protection System
Shutdown Cooling
Residual Heat Removal
Automatic ScramThe following information was provided by the licensee via phone and email: At 0127 EDT on 9/17/2024, with Unit 3 in mode 1 at 100% power, the reactor automatically tripped due to the passive residual heat removal heat exchanger outlet flow control valve failing open. A manual safeguards actuation was initiated due to the lowering pressurizer water level resulting from the reactor coolant system cooldown that was caused by the passive residual heat removal heat exchanger outlet flow control valve failing open. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected. Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The failure of the control valve does not inhibit the residual heat removal system from functioning as it is passive. The reactor coolant system maximum allowable cooldown rate was exceeded (Technical Specification 3.4.3). The limit is 100 degrees F per hour above 350 degrees F. The maximum observed cooldown rate was 226 degrees F per hour. At time 0458 EDT, reactor coolant system temperature is 369.1 degrees F, reactor pressure is 900 psig. Currently, the plant is cooling down and proceeding toward placing shutdown cooling online.
ENS 5732516 September 2024 20:24:00FermiNRC Region 3GE-4Reactor Pressure VesselThe 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.
ENS 5732416 September 2024 16:30:00Watts BarNRC Region 2Westinghouse PWR 4-LoopReactor Coolant SystemThe 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 5732216 September 2024 13:02:00BrunswickNRC Region 2GE-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).

  • * * UPDATE ON 9/17/2024 AT 1411 EDT FROM DAVID MACDONALD TO ROBERT THOMPSON * * *

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).

ENS 5732113 September 2024 13:54:00ByronNRC Region 3Westinghouse PWR 4-LoopThe 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 5731410 September 2024 23:19:00ColumbiaNRC Region 4GE-5Service waterThe 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.