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 Discovered dateReporting criterionTitleEvent description
ENS 558215 April 2022 06:23:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge		Scram on LOW LevelThe following information was provided by the licensee via telephone and email: On 4/5/2022, at time 0223, during maintenance on Feedwater Level Control Valve 2FWS-LV10B, a Feedwater transient occurred resulting in an RPS Automatic Reactor Scram on Low Level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 Recirculation Sample System Isolation, Group 3 TIP ((Traversing Incore Probe)) Isolation Valve Isolation, Group 6 and 7 Reactor Water Cleanup Isolation and Group 9 Containment Purge Isolations. All control rods inserted as expected. High Pressure Core Spray and Reactor Core Isolation Cooling initiated and injected as expected. ECCS Systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable in Mode 3. These 4 hour and 8-hour non-emergency ENS ((Emergency Notification System)) reports are being made in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: There was no impact on Unit 1.
ENS 5611619 September 2022 06:32:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentSafety System InoperabilityThe following information was provided by the licensee via email: At 0132 CDT on September 19, 2022, River Bend Station (RBS) was operating at 100% power when the high pressure core spray (HPCS) system was declared inoperable in accordance with technical specification 3.8.9, condition E (declare HPCS and standby service water system pump 2C inoperable immediately) due to a E22-S003, HPCS transformer feeder malfunction. The HPCS is a single train system at RBS, therefore this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfilment of a safety function. The reactor core isolation cooling system has been verified to be operable. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: RBS has entered a 14-day limiting condition for operation due to the loss of HPCS and they have upgraded their on-line plant risk model to "yellow".
ENS 5628220 December 2022 03:01:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Manual Scram Due to Loss of Feedwater PumpThe following information was provided by the licensee via email: At 2101 (CST) on December 19, 2022, a manual reactor scram was initiated at Grand Gulf Nuclear Station (GGNS). Following the reactor scram, the high pressure core spray (HPCS) system was used to maintain reactor water level. The manual (reactor protection system) RPS actuation is being reported in accordance with 10 CFR 50.72(b)(2) and the HPCS actuation is being reported in accordance with 10 CFR 50.72(b)(3). At 2058, GGNS experienced a loss of a condensate booster pump. At 2101, the `A' reactor feedwater pump tripped and the reactor was manually scrammed. All control rods were fully inserted into the core. At 2104, the `B' reactor feedwater pump tripped and HPCS was manually started. HPCS was manually injected to maintain reactor water level at 2121. The `A' reactor feedwater pump was successfully restarted at 2126. GGNS is currently in Mode 3. Reactor level is being maintained with the `A' reactor feedwater pump and pressure is being maintained with the turbine bypass valves. The NRC Resident Inspector was notified.
ENS 562985 January 2023 17:42:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge		Automatic Reactor ScramThe following information was provided by the licensee via phone and email: At 1242 (EST) on 05 January 2023, with the Unit in Mode 1 at 99 percent power, the reactor automatically tripped on low Reactor Pressure Vessel level while restoring power to Digital Feedwater Control Stations when there was a perturbation to the level controls. The reason for perturbation is unknown at this time. The trip was not complex, with all systems responding normally post trip. Operations responded and stabilized the plant. High pressure core spray was manually initiated in accordance with site procedures. Reactor water level is being maintained via the Feedwater System. Decay heat is being removed by the Main Condenser. 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)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5662517 May 2023 10:39:0010 CFR 50.73(a)(1), Submit an LER60-DAY Optional Telephonic Notification for an Invalid Actuation of an Emergency AC Electrical Power SystemThe following information was provided by the licensee email: At 0339 CDT on May 17, 2023, diesel generator 3 (DG3) had an auto-start during a surveillance test of excess flow check valves in containment atmosphere instrument sensing lines. During the surveillance, workers failed to recognize residual pressure in the system from the test. Per procedure, MS-PS-47C (main steam pressure switch) was placed back in service, resulting in initiation logic for both the high pressure core spray (HPCS) system and DG3 auto-start. Because the HPCS system was tagged out of service for maintenance it did not actuate. The auto-start of DG3 was an expected response to the high drywell pressure indication. The signals cleared, and DG3 was shutdown per procedure. As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv)(A), the licensee may, at its option, provide a telephonic notification to the NRC Operations Center within 60 days of discovery of the event instead of submitting a written licensee event report. This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) for invalid actuations reported under 10 CFR 50.73 (a)(2)(iv)(A). This actuation was invalid since it was caused by programmatic issues in quality of procedural guidance and not the result of actual plant conditions warranting auto-start of DG3. The actuations were not initiated in response to actual plant conditions, this was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation. Therefore, this event has been determined to be an invalid actuation. Diesel generator 3 system responded as designed to the actuation signal. The HPCS system did not actuate since it was tagged out of service. There was no impact on the health and safety of the public or plant personnel. The following information is provided as specified in NUREG-1022: (a) The diesel generator 3 was actuated. (b) The actuation of DG3 was complete. (c) The DG3 train was started and functioned successfully. The NRC Resident Inspector has been notified.
ENS 567102 September 2023 10:32:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge		Automatic Reactor Scram Due to Feedwater TransientThe following information was provided by the licensee via email: On 9/2/2023 at 0632 EDT, a feedwater transient occurred resulting in an reactor protection system (RPS) automatic reactor scram on low level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 recirculation sample system isolation, Group 3 traveling in-core probe (TIP) isolation valve isolation, Group 6 and 7 reactor water cleanup isolation, and Group 9 containment purge isolations. All control rods inserted as expected. High pressure core spray and reactor core isolation cooling initiated and injected as expected. ECCS systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable and in Mode 3. These 4 hour and 8 hour non-emergency reports are being made in accordance with 10 CFR 50.72(b)(2) (iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. There was no impact on Unit 1.
ENS 5693829 January 2024 16:05:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentHigh Pressure Core Spray FailureThe following information was provided by the licensee via email: At 1005 CST on January 29, 2024, Grand Gulf Nuclear Station was conducting surveillance testing on the high pressure core spray system. During testing, the 1E22F012 minimum flow valve failed to return to the full closed position. The valve went from full open indication to dual 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. The NRC Senior Resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All off-site power is available. No other systems are out of service and there are no compensatory measures taken. There is no increase to plant risk.
ENS 5718118 June 2024 20:40:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentInoperability of Division 3 Diesel Generator Supporting High Pressure Core SprayThe following information was provided by the licensee via phone and email: At 1640 EDT on 06/18/2024, the division 3 diesel generator was declared inoperable. 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)(D). 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 division 3 diesel generator was declared inoperable due to potential water intrusion into the electrical generator. Inspection of the generator is in progress. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This event resulted in Perry Unit 1 entering a 72 hour limiting condition for operation (LCO) in accordance with Technical Specification 3.8.1. condition 'B'.
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