Semantic search
| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 56858 | 16 November 2023 12:12:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 0906 Eastern Standard Time (EST) on November 16, 2023, it was determined that a non-licensed employee supervisor failed a test specified by the Fitness for Duty (FFD) testing program. The individual's authorization for site access has been removed. The NRC Resident Inspector has been notified. |
| ENS 56846 | 10 November 2023 03:14:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0118 EST, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually scrammed due to degrading main condenser vacuum. The scram was not complex, with all systems responding normally post-scram. The main turbine bypass valves opened automatically to maintain reactor pressure. Operations responded and stabilized the plant. Reactor water level is being maintained via feedwater pumps. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not impacted. Due to Reactor Protection System actuation while critical, this event is being reported as a four-hour and eight-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Unit 1 reactor is currently stable in mode 3. An investigation is in progress into the cause of the degrading condenser vacuum. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56826 | 1 November 2023 09:38:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0648 EDT on 11/1/23, with Unit 2 in MODE 1 at 56 percent power, the reactor was manually tripped due to a trip of the 'B' reactor feed pump (RFP). The 'A' RFP had been previously isolated due to a leak. Closure of containment isolation valves (CIVs) in multiple systems and the actuation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) occurred as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained with RCIC. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 was not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, 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, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. There was no impact on the health and safety of the public or plant personnel. The Resident Inspector was notified. |
| ENS 56822 | 30 October 2023 17:06:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56811 | 22 October 2023 16:40:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax and phone: On October 22, 2023, at 1149 CDT, with the reactor at 100 percent core thermal power and steady state conditions, the Cooper Nuclear Station secondary containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 limit of -0.25 inches water gauge. The condition existed for approximately 80 seconds until the reactor building ventilation system responded to restore differential pressure to normal. Investigations identified a hinged duct access hatch found open. The hatch was closed and latched, and ventilation system parameters were returned to normal. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material and mitigate the consequences of an accident. The NRC Senior Resident Inspector has been informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time the licensee notified the NRC Headquarters Operations Officer, the cause of the hinged access duct being open had not been determined. This event has been added to the licensee's corrective action program. |
| ENS 56797 | 15 October 2023 23:30:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee: At 2256 EDT on October 15, 2023, Brunswick declared a Notification of Unusual Event due to a fire not extinguished within 15 minutes. The licensee received fire alarms and indication of a halon discharge in the basement of the emergency diesel generator building. Due to the delay in the entry into the area, the licensee was not able to verify that the fire was out within 15 minutes. Upon entry into the room, the licensee noted an acrid odor near a transformer, but there was not a fire in the room. The fire was declared out at 2310 EDT. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via email: Termination of Unusual Event due to verification of no fire in the basement of the emergency diesel generator building." The licensee terminated the Unusual Event at 0045 on 10/16/23. The licensee notified the NRC Resident Inspector. Notified R2DO (Miller), IR-MOC (Grant), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). |
| ENS 56786 | 10 October 2023 19:44:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax: On October 10, 2023, at 1553 CDT, Cooper Nuclear Station (CNS) was notified of a spurious actuation of a single alert notification system siren in Nemaha, Nebraska. The CNS Emergency Alert System (EAS) was not activated. The actuation occurred during siren testing conducted at approximately 1545 CDT. No emergency conditions are present at Cooper Nuclear Station. A press release from Nebraska Public Power District is not planned at this time. This condition is reportable under 10CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Offsite notification was to local Nemaha County Emergency Management. |
| ENS 56753 | 21 September 2023 10:31:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A licensed operator had a confirmed positive test for alcohol during another entity's pre-access fitness-for-duty screening for unescorted access authorization. The individual's unescorted access at Peach Bottom Atomic Power Station has been denied. The NRC Resident Inspector has been notified. |
| ENS 56685 | 20 August 2023 18:30:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via email: On 8/20/2023 at 1600 EDT, during plant walkdowns in the drywell while in mode 3 to identify a cause of increasing unidentified leakage rate, reactor coolant system pressure boundary leakage (approximately 2 gpm) was identified on the reactor recirculation sample line between the reactor recirculation sample line inboard isolation valve (B3100F019) and where the sample line taps off the B reactor recirculation jet pump riser. This requires entry into technical specification 3.4.4 condition C, identification of pressure boundary leakage with a required action to be in mode 3 in 12 hours and mode 4 in 36 hours. At 1630 EDT, a technical specification required shutdown to mode 4, cold shutdown, was initiated. A press release by DTE is anticipated. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i), a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi), and an eight-hour, non-emergency notification 10 CFR 50.72(b)(3)(ii)(A) for the degraded condition of the pressure boundary. Investigation into the cause of the reactor coolant system pressure boundary leakage is still ongoing. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. |
| ENS 56667 | 8 August 2023 17:03:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: A licensed (non-active) individual failed to comply with fitness for duty testing policies. The individual's unescorted access was terminated. |
| ENS 56650 | 1 August 2023 15:53:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via email: On 08/01/2023 at 0955 EDT, the Fermi 2 active seismic monitoring system provided indication of a potential seismic activity event. Plant abnormal procedures were entered, and compensatory measure 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 nonfunctional to validate the calibration of the system. Femi 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 accelerometer was declared inoperable. The HPCI accelerometer is the sole 'trigger' for the seismic recording system, which outputs peak accelerations experienced during a seismic event. This is used in assessment of the magnitude of an earthquake for EAL 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. |
| ENS 56570 | 13 June 2023 06:02:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via email: At 2333 EDT on June 12, 2023, the division 2 Mechanical Draft Cooling Tower (MDCT) Fan `D' was declared inoperable due to a trip of the fan while running in high speed. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components, including the High-Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on a loss of the HPCI room cooler. The cause of MDCT Fan `D' trip is currently unknown with trouble shooting being developed for remediation of the condition. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Senior Resident Inspector has been notified.
The purpose of this notification is to retract a previous event notification (EN) 56570 reported on June 13, 2023, at 0602 EDT. The cause of the fan trip was a failed vibration switch. At 0429 EDT on June 14, 2023, the vibration switch was replaced, the MDCT fan "D" was tested satisfactory for operability, and the UHS, emergency diesel generator 13/14, and MDCT were declared operable. Following the initial EN, further analysis of the condition was performed utilizing a previously performed gothic analysis model (to perform HPCI room heat-up calculations) which bounded this condition. Based on the initial conditions at the time of the indication loss, specifically HPCI room and suppression pool temperature, it was determined that the resulting worst case post-accident room temperature was sufficiently low enough to provide margin to HPCI operability without the room cooler in service for the required mission time. No other concerns were noted during the event. HPCI remained operable and there was no loss of safety function. The fan trip did not involve a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). Therefore, the NRC non-emergency 10CFR50.72(b)(3)(v)(D) report was not required and the NRC report 56570 can be retracted, and no licensee event report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The licensee notified the NRC Resident Inspector. Notified R3DO (Nguyen) |
| ENS 56542 | 26 May 2023 14:16:00 | Fermi | NRC Region 3 | GE-4 | The following is a summary of the information provided by the licensee via email: As previously reported under Fermi LER 2023-001-00, submitted on May 22, 2023, at 1145 EDT on March 23, 2023, it was determined that all mechanical draft cooling tower (MDCT) fan brakes would not perform their design function during a tornado due to the speed switch not functioning over its published voltage and frequency ranges. The MDCT fan brakes are required to prevent fan overspeed from a design basis tornado. On May 25, 2023, Fermi completed its 10 CFR Part 21 discovery process and determined the need to perform a 10 CFR Part 21 evaluation. The vendor (Engine Systems Inc. (ESI)) was contacted and the purchaser (Fermi) assumed responsibility for performing the Part 21 evaluation for the supplied mechanism. This Part 21 evaluation is being tracked by Fermi CARD 23-20075. It has been determined the direct cause of the event was due to the Dynalco speed switch model SST-2400A-1, supplied by ESI, not functioning over its published voltage and frequency ranges. Corrective actions were taken to develop a design change to correct MDCT fan speed control system returning the MDCT fans, ultimate heat sink, and the service water subsystems to service on March 24, 2023. The root cause evaluation is ongoing, and written follow-up will be provided in 30 days by providing a supplement to the original LER by June 24, 2023. No new commitments are being made in this submittal. |
| ENS 56532 | 22 May 2023 16:24:00 | Susquehanna | NRC Region 1 | 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 to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56531 | 20 May 2023 07:45:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: On 5/20/2023 at 0315 CDT, Browns Ferry Unit 1 was at 80 percent reactor power performing, 'Turbine control valve fast closure turbine trip and RPT (recirculation pump trip) initiate logic testing'. During performance of this test, Unit 1 received a full reactor scram. An investigation is in progress to determine the cause of the scram. All systems responded as expected, and Unit 1 is stable at zero percent power in mode 3. All control rods fully inserted into the core. Main steam isolation valves remained open with main turbine bypass valves controlling pressure. Reactor feedwater pumps remained in service to control reactor water level. Primary containment isolation signals groups 2, 3, 6, and 8 were received with expected system actuations. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. The event is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. The NRC Resident has been notified. |
| ENS 56513 | 9 May 2023 17:41:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 1455 (EST) on Tuesday May 9, 2023, Peach Bottom Atomic Power Station (PBAPS) technical support center (TSC) ventilation system lost power. Power loss was caused by a tree down on the 361 transmission line. Power was not able to be restored within an hour. At 1639 (EST), power was restored to TSC ventilation, and capability was restored. This report is being submitted pursuant to 10 CFR 50.72(b)(3)(xiii) as a major loss of emergency preparedness capabilities due to a reduction in the effectiveness of the onsite TSC. NRC Resident has been notified. |
| ENS 56509 | 8 May 2023 06:25:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax: At time 0207 CDT, Cooper Nuclear Station (CNS) entered Technical Specification (Limiting Condition for Operation) LCO 3.0.3 due to declaring core spray subsystems A and B inoperable. This declaration was based on an issue with relays installed from the same manufacturing batch. The ability of the relays to function correctly to annunciate loss of logic power was called into question and they were declared inoperable. The plant has initiated actions to repair/replace affected relays. This event is reportable under 10 CFR 50.72(b)(2)(i) as an initiation of any nuclear plant shutdown required by Technical Specifications. In addition, this event Is reportable under 10 CFR 50.72(b)(3)(v) as a condition that could have prevented the fulfillment of a safety function for the core spray systems. NRC Resident Inspector was notified.
The following information was provided by the licensee via email: Technical Specification LCO 3.0.3 was exited at 0805 CDT on May 8, 2023. A reasonable expectation of operability was developed for the core spray subsystems A and B. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Shutdown was initiated and power was reduced approximately 45 percent. Reactor power is currently at 55 percent at the time of notification. Notified R4DO (Werner) via email.
The following information was provided by the licensee via email: CNS is retracting the 8-hour 10 CFR 50.72(b)(3)(v) non-emergency notification, for a condition that could have prevented the fulfillment of a safety function, made on May 8, 2023, at 0207 CDT (EN# 56509). Subsequent evaluation concluded that the core spray subsystems remained operable in accordance with the Technical Specifications Requirements 3.5.1, ECCS - Operating. As a result of the core spray system remaining operable, no loss of safety function occurred. The NRC Senior Resident Inspector has been notified. Notified R4DO (Werner). |
| ENS 56505 | 5 May 2023 16:00:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 05/04/2023 at 2034 CDT, a Browns Ferry Nuclear Plant non-licensed employee supervisor had a confirmed positive drug test identified during random fitness-for-duty medical testing. Employee's unescorted access has been suspended. A review of the employee's work has been completed. The (NRC) Resident Inspector has been notified. |
| ENS 56502 | 4 May 2023 10:27:00 | Limerick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A non-licensed, non-supervisor contractor was found to be in possession of alcohol in the protected area. The individual's site access has been terminated. The NRC Senior Resident Inspector has been notified. |
| ENS 56495 | 30 April 2023 02:30:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0200 EDT on 04/30/23, a Technical Specification required shutdown was initiated at Hope Creek Unit 1. Technical Specification Action 3.6.1.1 Primary Containment Integrity was entered on 04/30/23 at 0100 with a required action to restore primary containment integrity within 1 hour. This required action was not completed within the allowed outage 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56494 | 30 April 2023 02:30:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0100 EDT on 04/30/23, it was determined that the primary containment integrity did not meet (Technical Specification) TS 4.6.1.1.d requirement, suppression chamber in compliance with TS 3.6.2.1 due to the inability to establish test conditions for the bypass leakage test in accordance with TS 4.6.2.1.f. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D) & 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 56478 | 20 April 2023 05:24:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 0148 Eastern Daylight Time (EDT) on April 20, 2023, with Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a turbine trip. Turbine Bypass valves did not open on the trip due to Turbine Protection system power supply failure; the Safety Relief Valves (SRVs) opened automatically to control reactor pressure. Reactor Pressure reached approximately 1095 psig on the trip; exceeding the 1060 psig RPS trip setpoint. Operations responded and stabilized the plant. Operations was able to transition from SRVs to main steam line drains to the condenser. Reactor water level is being maintained via the Condensate / Feedwater system. Decay heat is being removed by discharging steam to the main condenser using the main steam line drains. 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). Reactor water level reached low level 1 (LL1) following the reactor trip. The LL1 signal causes Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves), and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the valid Primary Containment Isolation System (PCIS) actuation and RPS actuation from the reactor pressure signal, this event is also being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 is not affected by this event. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56464 | 11 April 2023 11:05:00 | Limerick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid specific system actuation of the Emergency Service Water (ESW) System. On 2/16/2023, while performing a calibration planned maintenance (PM) for a jacket water pressure indicator during a D13 diesel generator system outage window, the 'C' ESW pump unexpectedly auto-started. Subsequent investigation identified that the affected jacket water pressure indicator shares a common sensing line with a jacket water pressure switch that provides a back-up to the engine speed switch for the engine running signal. At the time the jacket water pressure indicator calibration PM was being performed, the power circuits for D13 diesel generator instrumentation were energized. Pressurization of the energized jacket water pressure switch during the pressure indicator calibration activity resulted in initiation of a false engine running signal to the `C' ESW pump start logic. This event is considered an invalid system actuation because the 'C' ESW pump started in response to a false signal that the D13 EDG was running when the D13 EDG did not start. The actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. The ESW system functioned as expected in response to the actuation. The affected ESW pump was shut down in accordance with plant procedures. There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector. |
| ENS 56458 | 8 April 2023 00:59:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 2052 EDT on April 7, 2023, during routine system preventative maintenance functional testing, the Unit 1 HPCI turbine stop valve, FV-15612, remained in the intermediate position. This failure resulted in the Unit 1 HPCI system being inoperable. This is being reported as a loss of an entire safety function condition in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Unit 1 HPCI inoperability places Unit 1 in a 14-day Technical Specification (TS) Limiting Condition for Operation (LCO). |
| ENS 56452 | 4 April 2023 19:55:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 1915 EDT, Susquehanna Nuclear Control Room was notified of a reportable oil release from a spare transformer outside of its secondary containment of an unknown quantity. Oil staining was observed on the ground outside of the designed containment vault stored on the grounds of Susquehanna Nuclear. The quantity and duration of the oil leak is unknown and thus poses a potential pollution risk to groundwater. Spill response measures are in-progress and as of 1500 on 4/4/2023 a field walkdown reported no visible oil outside of the containment vault. The spill event is reportable under Pennsylvania Department of Environmental Protection (PADEP) Clean Streams Law (PACSL) per PA Code 91.33 and 25 PA code 92a.41. This notification is being written to notify the US Nuclear Regulatory Commission within 4 hours of determination of a required report to another government agency per 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. |
| ENS 56446 | 31 March 2023 15:54:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1432 EDT on 03/31/23, with Unit 2 in mode 1 at 97 percent power, the reactor was manually tripped due to a loss of both recirculation pumps. The cause of the recirculation pump trips is under investigation. Additionally, closure of CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56434 | 26 March 2023 19:41:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: On 03/26/2023 at 1603 EDT, while performing Appendix J local leak rate testing, it was determined that the Secondary Containment Bypass Leakage (SCBL) limit had been exceeded for Unit 2. During performance of the leak rate test, SE-259-027 for X-9B penetration, it was determined that the combined SCBL limit of 15 standard cubic feet per hour for the as-found minimum pathway was exceeded, as specified in Technical Specification, Surveillance Requirement 3.6.1.3.11. This event is being reported pursuant to 10CFR50.72(b)(3)(ii). The Resident Inspector has been notified. |
| ENS 56430 | 23 March 2023 20:40:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 1736 EDT on March 23, 2023, during overcurrent testing of the '2B' (Emergency Safeguards System) ESS Bus, the work group was re-installing tested relays and inadvertently caused a '2B' ESS Bus lockout. This resulted in the '2B' ESS Bus deenergizing and a valid start signal provided to the 'B' Emergency Diesel Generator (EDG). The 'B' EDG started and functioned as designed. This is being reported as an unplanned actuation of systems that mitigate the consequences of significant events in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. |
| ENS 56429 | 23 March 2023 18:30:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via fax: While in Mode 1 at 100 percent power at 1145 EDT on March 23, 2023, it was determined that all mechanical draft cooling Tower (MDCT) fan brakes would not perform their design function during a tornado due to a design flaw with the control system. The MDCT fan brakes are required to prevent fan overspeed from a design basis tornado. The MDCT fans are required to support the operability of the ultimate heat sink (UHS). At the time of discovery, the provisions of LCO 3.0.9 were being utilized for loss of the 'D' MDCT fan brake (barrier loss). When it was identified the condition was a design flaw common to all MDCT fan brakes, the 24-hour allowance for restoration was entered. A design change is currently being implemented to restore MDCT fan brake operability. This condition is reportable per 10 CFR 50.72(b)(3)(v)(A), (B), & (D). The NRC Resident Inspector has been notified. |
| ENS 56411 | 15 March 2023 04:27:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 2257 (CDT) on 3/14/2023 during the 2R22 refueling outage on Browns Ferry Nuclear Plant Unit 2, it was determined there was RCS boundary leakage from five of eight sensing lines that pass through containment penetrations X-30 and X-34 that did not meet the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The condition will be resolved prior to plant startup. 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.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56411 reported on 3/14/23. Following the initial notification, further analysis of the condition was performed. It was determined that the leaking pipe weld was ASME Section XI Code Class 2 piping which falls under the requirements of ASME Section XI Subsection IWC and not Subsection IWB. Therefore, this condition does not represent a serious degradation of the nuclear power plant, including its principle safety barriers. Based upon the above, the leaks identified on the ASME Section XI Code Class 2 equivalent Main Steam sense lines are not reportable under 10 CFR 50.72(b)(3)(ii). Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(ii) report was not required and the NRC report 56411 can be retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(ii) is required to be submitted. Notified R2DO (Miller) |
| ENS 56409 | 14 March 2023 15:52:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 1000 EDT on March 14, 2023, during valve diagnostic testing, the high pressure core injection (HPCI) lube oil cooling water supply isolation valve did not stroke open. This failure resulted in the Unit 2 HPCI system being inoperable. This is being reported as a loss of an entire safety function condition in accordance with 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.
The following information was provided by the licensee via email: The purpose of this notification is to retract event notification (EN) 56409 reported on 03/14/2023. On March 09, 2023, Susquehanna Unit 2 entered a routine high pressure core injection (HPCI) maintenance outage. In support of this system outage, Technical Specification (TS) 3.5.1, Condition D was entered for an inoperable HPCI system. On March 14 as reported in EN 56409, the HPCI lube oil cooling water supply isolation valve did not electrically stroke open following engagement of manual clutch lever. Specifically, to support the maintenance evolution, electricians declutched the valve actuator to move it from the motor/electric operational mode to the manual operational mode as part of planned valve diagnostic data collection. In this testing configuration (i.e., manual operational mode), an attempt to electrically stroke the valve was made, resulting in the valve failure to stroke. Prior to this maintenance evolution, the HPCI lube oil cooling water supply isolation valve was found in the expected full-closed position with the motor/electric operational mode enabled, meaning prior to the HPCI maintenance outage, the affected valve was operating as designed and capable of performing all design functions. The described condition was therefore determined to be the result of the maintenance activity. NUREG-1022, Section 3.2.7, states: 'reports are not required when systems are declared inoperable as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that would have resulted in the system being declared inoperable).' Following completion of investigation and repair, Susquehanna determined that, per NUREG-1022, Section 3.2.7, the event was not reportable. HPCI was declared inoperable as part of a maintenance evolution which was done in accordance with an approved procedure and the TS. The described condition was not a pre-existing condition that would have resulted in the system being declared inoperable prior to the planned maintenance activity. Notified R1DO (Schroeder) |
| ENS 56385 | 2 March 2023 17:52:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1312 CST on March 2, 2023, Browns Ferry Nuclear Plant Units 1, 2, and 3 initiated voluntary communication to the state of Alabama and local officials as part of the Nuclear Energy Institute (NEI) Groundwater Protection Initiative (GPI), after receiving analysis results for leakage from a demineralized water storage tank that contained activity above the GPI voluntary communication threshold. All these results are significantly less than the limits established by the Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) for effluents from the station. Further samples obtained of the water prior to entering the Tennessee River were less than detectable. The leakage source has been isolated and additional corrective actions are in progress. This condition did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56373 | 19 February 2023 08:56:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via fax or email: At 0105 EST on February 19, 2023, with the James A. FitzPatrick Nuclear Power Plant (JAF) at 100 percent power, a valid high main steam line radiation signal was received. An actuation of a fire protection foam system caused migration of high conductivity water into a low conductivity sump. Organic compounds were introduced into the primary coolant and resulted in a temporary increase in nitrogen-16 which was detected by main steam line radiation monitors and actuated primary containment isolation signals in more than one system. The reactor water recirculation sample system isolated. The signal also went to the normally isolated main steam line drain system and condenser air removal system. The event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The elevated radiation condition no longer exists. Health and safety of the public was not impacted by this event. The NRC Resident Inspector was notified. |
| ENS 56371 | 18 February 2023 11:25:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On February 17, 2023 during the planned U2R22 outage on Browns Ferry Nuclear Plant Unit 2, personnel entered the Unit 2 drywell for leak identification. Personnel discovered a cracked weld on the 2A recirculation pump discharge isolation valve drain line. At 0439 CST on February 18, 2023, following engineering evaluation, this drain line was determined to be ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. |
| ENS 56346 | 9 February 2023 15:15:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax: A licensed operator had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56342 | 7 February 2023 21:10:00 | Hatch | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1738 EST on 02/07/2023, while in mode 5 at 0 percent power, it was determined during local leak rate testing (LLRT) that the primary containment leakage rate exceeded the allowable limit defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.' Both primary containment isolation valves in a penetration failed LLRT requirements which represents a failure to maintain primary containment integrity. 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 56358 | 6 February 2023 13:26:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by Constellation via email: On 02/06/2023 at 0416 EST, the Constellation Emergency Response Organization (ERO) Notification Database System uploaded data files into the Mass Notification System (Everbridge) which is used to notify ERO personnel when activated. At 0630, the individual reviewing the uploaded files discovered that the data files did not upload properly and that Everbridge may not notify all ERO individuals within the required 10 minutes of system initiation. Constellation resolved the issue by 0752. During the time period of 0416 to 0752, control room operators would have been unaware that the ERO notification was not successful. Therefore, this issue constitutes a loss of offsite communications capability and is reportable under 10 CFR 50.72(b)(3)(xiii), 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' This loss of offsite communications capability affected all Constellation nuclear stations. There was no impact on the health and safety of the public or plant personnel. Each affected station NRC Resident Inspectors have been or will be notified. |
| ENS 56362 | 6 February 2023 13:26:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by Constellation via email: On 02/06/2023 at 0416 EST, the Constellation Emergency Response Organization (ERO) Notification Database System uploaded data files into the Mass Notification System (Everbridge) which is used to notify ERO personnel when activated. At 0630, the individual reviewing the uploaded files discovered that the data files did not upload properly and that Everbridge may not notify all ERO individuals within the required 10 minutes of system initiation. Constellation resolved the issue by 0752. During the time period of 0416 to 0752, control room operators would have been unaware that the ERO notification was not successful. Therefore, this issue constitutes a loss of offsite communications capability and is reportable under 10 CFR 50.72(b)(3)(xiii), 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' This loss of offsite communications capability affected all Constellation nuclear stations. There was no impact on the health and safety of the public or plant personnel. Each affected station NRC Resident Inspectors have been or will be notified. |
| ENS 56360 | 6 February 2023 13:26:00 | Limerick | NRC Region 1 | GE-4 | The following information was provided by Constellation via email: On 02/06/2023 at 0416 EST, the Constellation Emergency Response Organization (ERO) Notification Database System uploaded data files into the Mass Notification System (Everbridge) which is used to notify ERO personnel when activated. At 0630, the individual reviewing the uploaded files discovered that the data files did not upload properly and that Everbridge may not notify all ERO individuals within the required 10 minutes of system initiation. Constellation resolved the issue by 0752. During the time period of 0416 to 0752, control room operators would have been unaware that the ERO notification was not successful. Therefore, this issue constitutes a loss of offsite communications capability and is reportable under 10 CFR 50.72(b)(3)(xiii), 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' This loss of offsite communications capability affected all Constellation nuclear stations. There was no impact on the health and safety of the public or plant personnel. Each affected station NRC Resident Inspectors have been or will be notified. |
| ENS 56321 | 24 January 2023 08:43:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0121 CST on 01/24/2023, it was discovered that the Unit 1 High Pressure Coolant Injection System (HPCI) was inoperable; therefore, the condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. 1-FCV-073-0006B, HPCI Steam Line Condensate Outboard Drain Valve, failed closed during normal plant configuration. This valve is normally open. The HPCI steam line is not being drained with the valve in the current position. The Unit 1 Nuclear Unit Senior Operator entered Unit 1 Technical Specifications LCO 3.5.1 Condition C with required actions C.1 to immediately verify by administrative means that the Reactor Core Isolation Cooling (RCIC) system is operable and C.2 to restore HPCI to operable status in 14 days. RCIC has been verified operable by administrative means. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56295 | 4 January 2023 08:28:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via email: At 0148 EST on January 4, 2023 it was identified that P4400F603B, Division 2 Emergency Equipment Cooling Water (EECW) Supply Isolation Valve, lost position indication. Division 2 EECW System was declared inoperable due to the potential that this valve may not be capable of performing its safety function to automatically isolate the safety related Division 2 EECW system from the non-safety related Reactor Building Closed Cooling Water (RBCCW) system. Because the Division 2 EECW system provides cooling to the High Pressure Coolant Injection (HPCI) room cooler, HPCI was also declared inoperable; therefore, this condition is being reported as an eight-hour, non--emergency notification per 10 CFR 50.72(b)(3)(v)(D). At 0240 EST, position indication was restored and Division 2 EECW and HPCI was returned to operable following inspection of the associated motor control center (MCC) and testing of the associated fuses. The cause of the loss of indication is under investigation. The Senior NRC resident inspector has been notified.
The following retraction was received from the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56295, reported on 1/4/2023. Following the initial EN, further analysis of the condition was performed utilizing a gothic analysis model to perform HPCI room heat-up calculations. Based on the initial conditions at the time of the indication loss, specifically HPCI room and Suppression Pool temperature, it was determined that the resulting worst case post-accident room temperature was sufficiently low enough to provide margin to HPCI operability without the room cooler in service for the required mission time. No other concerns were noted during the event. Therefore, HPCI remained operable and there was no loss of safety function. The event did not involve a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(v)(D) report was not required and the NRC report 56295 can be retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The NRC Senior Resident Inspector has been notified. Notified R3DO (Ruiz). |
| ENS 56287 | 28 December 2022 09:18:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER (Licensee Event Report) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0906 Eastern Time (EST) on November 9, 2022, an invalid actuation of Group 6 Primary Containment Isolation Valves (PCIVs) (i.e., Containment Atmospheric Control/Monitoring and Post Accident Sampling isolation valves) occurred. In addition, per design, Reactor Building Ventilation isolated and Standby Gas Treatment started. It was determined that this condition was caused by faulty test equipment that was being used during preparation for the Main Stack Radiation Monitor High Radiation Response Time test. This test requires connecting a recording device to monitor for the test start signal on a Unit 2 relay associated with the Main Stack High Radiation signal. The recorder faulted which caused the associated fuse to blow and resulted in Unit 2 receiving a Main Stack High Radiation signal and Group 6 PCIV actuation. It was verified that the radiation monitor was not in trip electrically (i.e., there was no high radiation condition). The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. 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. |
| ENS 56283 | 21 December 2022 13:32:00 | Limerick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid specific system actuation of the Emergency Service Water System (ESW). On 11/2/2022, during normal reactor operations, multiple main control room alarms were received for D12 Emergency Diesel Generator (EDG) running and Unit 1 Division 2 Safeguard Battery Ground. The D12 EDG did not start; however, the 'B' ESW Pump auto started. Subsequent troubleshooting determined that the cause of the D12 EDG running alarms and the inadvertent auto start of the 'B' ESW Pump was a malfunction on the D12 EDG speed switch. This event is considered an invalid system actuation because the 'B' ESW Pump started in response to a false signal that the D12 EDG was running when D12 EDG did not start. This was a complete actuation of the ESW System and the system functioned as expected in response to the actuation. The affected ESW Pump was shut down in accordance with plant procedures and the degraded D12 EDG speed switch was replaced. There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector. |
| ENS 56281 | 19 December 2022 13:12:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0735 EST on December 19, 2022, it was determined that a non-licensed employee supervisor failed a test specified by the Fitness-for-Duty (FFD) testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56280 | 19 December 2022 12:50:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) for an invalid actuation of a primary containment isolation signal affecting more than one system. On November 11, 2022, at 2333 hours EST, Peach Bottom experienced an unplanned loss of the #343 Off-Site Startup Source. Due to the temporary loss of power during automatic bus transfers, several systems experienced Primary Containment Isolation System (PCIS) Group II and Group III (GP II/III) isolation signals. Plant Systems impacted by isolation valve closure included: Reactor Water Clean Up (RWCU), Containment Atmospheric Control (CAC), Traversing In-Core Probe (TIP) Purge, Primary Containment Floor and Equipment Drains, and the Instrument Nitrogen system. All equipment responded as designed. Plant conditions which initiate PCIS GP II isolation signals are Reactor Vessel Low Water Level, High Drywell Pressure, RWCU system High Flow or RWCU Non-Regenerative Heat Exchanger High Outlet Temperature. The PCIS GP III actuations are initiated by the Reactor Vessel Low Water Level, Primary Containment High Pressure, Reactor Building Ventilation High Radiation or Refuel Floor Ventilation High Radiation. At the time of the event, none of these actual plant conditions existed; therefore, the actuation of the PCIS was invalid. The loss of the #343 Off-Site Startup Source was caused by a failed printed circuit card in the programable logic controller (PLC) for the 3435 breaker. There is no time-based maintenance strategy for PLC replacement. The PLC circuit card was replaced, and the breaker restored to full qualification and service. Preventive maintenance strategy will be enhanced to address the identified vulnerability. The licensee has notified the NRC Resident Inspector. |
| ENS 56278 | 17 December 2022 04:03:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via email: On December 16, 2022 at 2351 CST, with the Unit in Mode 1 at 13 percent power, a manual scram was inserted due to lowering Reactor Pressure Vessel (RPV) pressure, which occurred following an unexpected opening of Main Turbine Bypass Valve 1. All control rods fully inserted. Following actuation of the manual scram, RPV pressure lowered, resulting in an automatic Primary Containment lsolation (PCIS) Group 1 isolation (expected response). The main steam isolation valves and steam line drain valves all closed. The Group 1 (isolation) has been reset allowing RPV pressure control with steam line drains to the main condenser. All systems responded as designed. The plant is stable in Mode 3. Investigation of the bypass valve opening is ongoing. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation and 50.72(b)(3)(iv)(A) Specified System Actuation. There was no impact on health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. |
| ENS 56257 | 3 December 2022 13:05:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 12/2/2022 at 2330 (CST) during the planned F311 outage on Browns Ferry Nuclear Plant Unit 3, personnel entered the Unit 3 drywell for leak identification. Personnel discovered a through-wall piping leak on a 0.75 inch test line between the two test line isolation valves. This 0.75 inch test line is located on the residual heat removal (RHR) loop 1 shutdown cooling and RHR return line to the reactor vessel. On 12/3/2022 at 1000 CST, Engineering determined this location is classified as ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. |
| ENS 56241 | 28 November 2022 08:38:00 | Fermi | NRC Region 3 | GE-4 | The following information was provided by the licensee via email: At 0400 EST on November 28, 2022, during the performance of Division 2 Residual Heat Removal (RHR) cooling tower fan operability and RHR Service Water valve lineup verification, it was reported that the Mechanical Draft Cooling Tower (MDCT) Fan 'B' was making a loud metallic noise. The cause of the metallic noise is unknown at this time. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components including the High Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on inoperable cooling water to the HPCI room cooler, per LCO 3.0.6. Investigation into the Division 2 MDCT Fan 'B' abnormal noise is in progress. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification 56241 reported on 11/28/2022. On 11/28/22, an event notification to the NRC was made when mechanical draft cooling tower (MDCT) Fan B was declared inoperable and issued Limited Condition of Operation (LCO) 2022-0428 for Division 2 MDCT Fan B abnormal noise. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS) (Technical Specification (TS) 3.7.2). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system (TS 3.7.2), which cools various safety related components, including the High-Pressure Coolant Injection (HPCI) system room cooler (TS LCO 3.0.6). Subsequent inspection and evaluation determined that the brake noise is expected while fans are running at low speeds. This is supported by plant technical procedure, 24.205.10 `Div. 2 RHR Cooling Tower Fan Operability and RHRSW Valve Line-up Verification' (line item 2.2 in Precautions and Limitations) which states `Chatter from the brakes of the MDCT Fans is expected and no cause for discontinuing the test.' The equipment vendor stated that brake chatter is possible and common given that the internal components are free to move along the splined connections. Internal Operating Experience from experienced station operators and maintenance technicians confirmed that the condition is normal and expected. Both Division 2 MDCTs exhibited the same behavior at low speed and passed surveillance testing satisfactorily. No other concerns were noted during fan operation. Therefore, HPCI remained operable and there was no loss of safety function. The event did not involve a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). EN 56241 is retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The NRC Resident Inspector has been notified. Notified R3DO (Stoedter). |
| ENS 56220 | 13 November 2022 04:01:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee email: On November 12, 2022, at 2319 CST, an actuation of the reactor protection system (RPS) initiated a full scram. The plant was in Mode 2, reactor pressure was 149 pounds. The high pressure coolant injection (HPCI) injection valve, HPCI-MOV-MO19, opened and injected cold water into the reactor vessel while HPCI system testing was in progress. The cause is still under investigation. All control rods inserted. Plant is currently in Mode 3 and stable. All systems operated as designed with no Primary Containment Isolation System group isolations. This event is being reported under two event classifications: 50. 72(b)(2)(iv)(B) -- "Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 50. 72(b)(3)(iv)(A) -- "Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The NRC Resident has been informed. |
| ENS 56174 | 22 October 2022 17:36:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax: During Mode 5 Refueling operations, while attempting to establish flow through the Fuel Pool Cooling system filter demineralizers, an air operated valve to a radioactive waste tank failed to close automatically. This caused the Fuel Pool Cooling system to pump water from the Skimmer Surge Tanks (SST) to the radioactive liquid waste system. In response to the loss of inventory from the SSTs, the Control Room operating crew started Core Spray Pump A to restore normal operating level In the SST. This prevented the loss of the Fuel Pool Cooling/Alternate Decay Heat Removal system which was the only in service system meeting the safety function of decay heat removal. Core Spray Pump A was used for Injection for less than 3 minutes. This is reportable as a discharge of ECCS into the RCS in response to an event, but not part of a pre-planned sequence under 10 CFR 50.72(b)(2)(iv)(A) and actuation of a specified system under 10 CFR 50.72(b)(3)(Iv)(A). The resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Licensee reported approximately 6000-7000 gallons of water was injected into the RCS. The stuck open air operated valve was closed. Proceeding with refueling outage operations. |