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 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 5679013 October 2023 01:27:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Manual Reactor TripThe following information was provided by the licensee via email: On 10/12/23 at 2127 EDT, with the Unit 1 in Mode 1 at 100% Power, operators identified degrading condenser vacuum and manually tripped the reactor. All control rods inserted as expected. The trip was not complex, and all systems responded normally post-trip. The cause of the degraded condenser vacuum was an unexpected closure of the condenser air ejector regulator. The cause of the air ejector regulator going closed is not fully understood and is being investigated. Following the SCRAM, Operators responded and stabilized the plant. Decay heat is being removed by the Main Steam System through the Atmospheric Relief Valves (ARVs) and Auxiliary Feed Water (AFW) systems. 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) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.Reactor Protection System
Control Rod
Auxiliary Feed Water
Main Steam
ENS 567319 September 2023 15:43:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Automatic Reactor TripThe following information was provided by the licensee via email: On 9/9/23 at 1143 EDT, with the Unit 1 in Mode 1 at 100 percent power, all 4 turbine control valves closed resulting in a reactor protection system (RPS) automatic reactor trip on over temperature differential temperature. All control rods inserted as expected. The trip was not complex and all systems responded normally post-trip. The cause of the control valve closure has not been determined. Following the SCRAM, operators responded and stabilized the plant. Decay heat is being removed by the main steam system through the atmospheric relief valves and auxiliary feed water systems. Due to the 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) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.Reactor Protection System
Control Rod
Auxiliary Feed Water
Main Steam
ENS 555044 October 2021 04:31:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid System ActuationThe 'A' Steam Generator Narrow Range Water Level went less than 17 percent causing an Auxiliary Feed Water System valid actuation signal. The Auxiliary Feed Water System was in service at the time of the event providing decay heat removal. There was no adverse effect on plant systems. The Steam Generator Narrow Range Water Level was restored to normal operating band. This is being reported per 10 CFR 50.72(b)(3)(iv)(A), which states, '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.' (Reactor Coolant System) RCS Pressure 340 pounds and RCS Temperature 340 Degrees F. The NRC Resident Inspector was notified.Steam Generator
Auxiliary Feedwater
Decay Heat Removal
ENS 5369626 October 2018 04:00:0010 CFR 50.72(b)(3)(iv)(A), System Actuation
10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness		Automatic Start of the Emergency Diesel Generator and Loss of Radiation MonitorRCS (Reactor Coolant System) Pressure: vented to containment, refueling cavity greater than 23ft. (above reactor vessel). RCS temperature: 96 degrees Fahrenheit. The 12A bus de-energized, 'A' EDG (Emergency Diesel Generator) automatically started and loaded on (emergency) buses 14 and 18. The RCS configuration is refueling cavity level greater than 23ft. above the reactor flange with no impact to shutdown cooling. Radiation monitor R-1, Control Room radiation monitor, lost power for 2 hrs 10 min. This placed Ginna in a major loss of emergency preparedness capabilities. A temporary radiation monitor has been installed in the Control Room. Prior to the notification, the licensee had restored the 12A bus from offsite power and the R-1 monitor was re-energized. The licensee notified the NRC Resident Inspector.Emergency Diesel Generator
Shutdown Cooling
ENS 5173012 February 2016 04:05:0010 CFR 50.72(b)(3)(iv)(A), System ActuationEmergency Diesel Generator Auto-Started Due to the Loss of a Station Service TransformerOn 02/11/2016, at 2305 (EST), Ginna Station experienced a loss of Station Service Transformer 12A causing Emergency Diesel Generator 1A to automatically start due to under-voltage signals to safeguards buses 14 and 18. All plant systems responded as designed. Control room operators stabilized the plant per abnormal operating procedures. The plant is currently in a 100/0 electrical lineup (supplied by the 12B Service Station Transformer) on the off site circuit 767 with the 1A Emergency Diesel Generator secured. The loss of the station service transformer is currently under investigation. This is reportable as a valid system actuation that was not part of a pre-planned sequence during testing. The NRC Resident Inspector has been notified.Emergency Diesel Generator
ENS 4921424 July 2013 18:19:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Automatic Reactor Trip from Full PowerAt 1419 EDT on 7/24/2013, the reactor tripped due to a reactor protection system (RPS) actuation signal from a turbine trip, which was caused by a generator trip. All control rods inserted on the trip and reactor coolant system (RCS) pressure is currently 2235 psig and stable with RCS temperature stable at 547 degrees F. Decay heat is being removed by steam dumps (to the main condenser) and auxiliary feedwater which auto started as expected. The cause of the generator trip is under investigation. The plant will remain in Mode 3 until the cause of the trip is determined. The plant notified the NRC Resident Inspector.Reactor Coolant System
Reactor Protection System
Auxiliary Feedwater
Control Rod
ENS 479883 June 2012 06:39:0010 CFR 50.72(b)(3)(iv)(A), System ActuationAutostart of "B" Edg Due to Loss of Offsite Power Circuit 767At approximately 0239 hours on June 3, 2012 the 'B' Emergency Diesel Generator (EDG) automatically started when offsite power circuit 767 was de-energized. The EDG started and re-energized Safe Guards busses 16 and 17. The selected Service Water (SW) pump 'B' automatically started to supply cooling to the EDG. The operators responded to the loss of circuit 767 using abnormal operating procedure AP-ELEC.1 'Loss of 12A and/or 12B Busses'. Offsite power was restored to 12B bus using ER-ELEC.1 'Restoration of Offsite Power' on circuit 7T at 0318 hours. The 'B' EDG was shutdown at 0445 hours. The initial investigation of the loss of circuit 767 indicates that the likely cause was due to wildlife, e.g., raccoon. The licensee informed the NRC Resident Inspector.Service water
Emergency Diesel Generator
ENS 4733812 October 2011 03:28:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Automatic Reactor Trip Due to Turbine Auto Stop Valve ClosureAutomatic Reactor Trip due to Turbine Auto Stop Valve Closure and Actuation of Auxiliary Feedwater System. At 2328 on 10/11/2011, the reactor tripped due to a RPS actuation Signal from a turbine trip, which was caused by a Turbine Auto Stop signal. All control rods inserted on the trip, RCS pressure is currently 2235 psig and stable, and RCS average temperature is 547 degrees and stable. Decay heat removal is being controlled by auxiliary feedwater which auto started as expected and steam generator atmospheric relief valves. The licensee is investigating the cause of the Auto Stop Signal. The plant will be maintained in MODE 3 until the cause of trip is determined. The licensee has notified the NRC Resident Inspector. There is no primary to secondary leakage. Offsite power is normal and all EDG's are available.Steam Generator
Auxiliary Feedwater
Decay Heat Removal
Control Rod
ENS 469173 June 2011 04:39:0010 CFR 50.72(b)(3)(iv)(A), System ActuationUnexpected Emergency Diesel Generator Actuation

On 6/3/2011 at 0039 hours, during the performance of a work order to test components associated with Service Water Isolation, Emergency Diesel Generator (EDG) 'A' unexpectedly started automatically and its supply breaker to Safeguards Bus 14 closed. The Control Room staff observed normal voltage on Diesel Generator 'A'. Bus 14 voltage was never lost during this event, however, they also noted an associated Bus 14 undervoltage annunciator on the Main Control Board. Seconds later, Emergency Diesel Generator 'A' tripped on Reverse Power and its supply breaker to Bus 14 tripped open. The initiating action was the removal of the Bus 14 Normal Feed Breaker Control Power Fuses as part of the work order package. The Ginna EDG's have the following automatic start signals and logic: manual, safety injection signal (1/2 trains), undervoltage on respective safeguards bus, 'A' EDG Bus 14 or 18 (1 out of 2 degraded voltage + 1 out of 2 loss of voltage), 'B' EDG Bus 16 or 17 (1 out of 2 degraded voltage + 1 out of 2 loss of voltage). Investigation has commenced to determine the cause of the EDG start and undervoltage signal. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 7/26/11 AT 1214 EDT FROM SLABY TO HUFFMAN * * *

The purpose of this report is to retract the event discussed in Emergency Notification System report #46917 submitted on June 3rd, 2011. The ENS notification reported an unexpected start of Emergency Diesel Generator `A' during testing of a service water valve isolation circuit. As reported, Emergency Diesel Generator 'A' unexpectedly started and its supply breaker to Bus 14 closed. Seconds later, the Emergency Diesel Generator tripped on reverse power and its output breaker to Bus 14 opened. At the time of the event it was not understood why the diesel generator started. Subsequent troubleshooting and causal investigation identified that the signal was caused by a degraded control relay that unexpectedly changed state when control power was removed. This relay was expected to remain mechanically latched and would have remained in the desired position had control power not been removed as part of the test. Bus 14 voltage remained in the normal operating range throughout the event. Since this was not a valid undervoltage signal, the June 3rd, 2011 event is being retracted. A follow-up report will be made in accordance with 10CFR50.73(a)(1) and 10CFR50.73(a)(2)(iv). The NRC Resident Inspector has been notified. R1DO(Henderson) notified. See related EN #47094.

Service water
Emergency Diesel Generator
ENS 4324317 March 2007 02:09:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Automatic Reactor Trip Caused by Safety Injection SignalAt 2209 on 3/16/07, the plant tripped on a Safety Injection (SI) signal initiated because of low main steam line pressure in the 'A' main steam loop. The licensee is currently conducting a post trip review but believes that the low main steam line pressure in 'A' loop was caused by a spurious isolation of the 'B' Main Steam Line Isolation Valve (MSIV). The isolation in the 'B' main steam loop lead to high main steam flow and low main steam line pressure in the 'A' main steam loop. The 'A' MSIV then also auto-closed on the SI signal. The plant is currently stable in Mode 3 at about 2235 psig pressure and 547 degrees average Reactor Coolant System (RCS) temperature. All control rods fully inserted on the trip. Decay heat is currently being removed by auxiliary feedwater feeding the steam generators and steaming out the plant atmospheric steam valves. Since plant pressure did not decrease below 1500 psig, SI did not actually inject into the RCS. The licensee secured SI. No primary PORV's or safety valves lifted. No main steam safeties lifted according to plant closure indicators. There are no primary to secondary steam generator tube leaks. All electrical safeguards buses are powered by offsite power. The Emergency Diesel Generators (EDG) started but did not load and were shut down. The EDGs are operable and available if needed. The licensee notified the NRC Resident Inspector.Steam Generator
Reactor Coolant System
Emergency Diesel Generator
Auxiliary Feedwater
Main Steam Line
Control Rod
Main Steam
ENS 4312828 January 2007 01:40:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Automatic Reactor Trip Due to Loss of Load Transient Caused by Ehc ProblemOn January 27, 2007 at approximately 2040 hours an automatic reactor trip occurred. The cause of the trip was Over Temperature Delta T (2/4). All systems functioned as designed. All control rods inserted on the trip. Decay heat removal is via condenser steam dump and Auxiliary Feedwater. The initial cause of the trip appears to be from a loss of load due to a turbine electro-hydraulic system issue. This is still under investigation. RCS Temperature is 547 Degrees F and stable RCS Pressure is 2235 psig and stable Both pressurizer PORVs momentarily opened and then closed during the transient. For 8Hr Non Emergency 10 CFR 50.72(b)(3) RPS actuation occurred. Auxiliary Feed Water actuation occurred. There was no testing or maintenance in progress at the time of the transient. The licensee informed the NRC Resident Inspector.Auxiliary Feedwater
Decay Heat Removal
Control Rod
ENS 4176510 June 2005 17:51:0010 CFR 50.72(b)(3)(iv)(A), System ActuationLoss of Offsite Bus Due to a Lighting Strike with Actuation of Emergency Diesel GeneratorAt 1351 hours, a lightning strike resulted in a loss of Offsite Circuit #751. Loss of Circuit #751 resulted in a momentary loss of Bus 16 until its re-energization by Emergency Diesel Generator B. Safeguards Bus 17 remained de-energized as its Supply Breaker from Emergency Diesel Generator B failed to close as expected. Additionally, while responding per applicable Abnormal Procedures, Service Water Pump A was manually started. Service Water Pump A tripped approximately 2 minutes later with a report of smoke and sparks emanating from the Service Water Pump A motor. Due to the loss of Safeguards Bus 17 and Service Water Pump A, the plant ran on Service Water Pump C only. Emergency Diesel Generator B was transferred to alternate cooling during the event. The plant remained stable in Mode 1, 100% power, Tavg at 561�F, and RCS pressure at 2235 psig during the entire event. Emergency Buses 16 and 17 are now powered by Offsite Circuit #767. Service Water Pumps B and C are now operating. Offsite Circuit #751 has been restored to operable status. Emergency Diesel Generator B is secure and remains inoperable per Technical Specifications while work continues on the Emergency Diesel Generator B Supply Breaker To Bus 17. Service Water Pump A remains inoperable while work continues on it's motor. Licensee stated that they are in a 7 day LCO for restoring the equipment. The licensee notified the NRC Resident Inspector.Emergency Diesel Generator
ENS 415581 April 2005 19:02:0010 CFR 50.72(b)(3)(iv)(A), System ActuationEmergency Diesel Generator Start Due to a Partial Loss of Offsite PowerThe following information was provided by the licensee via fax (licensee text in quotes): With the plant in mode 6 and refueling operations in progress, a loss of offsite power circuit 751 occurred. This loss of power caused an undervoltage condition on safeguards busses 16 and 17 and an automatic start of Emergency Diesel Generator 1B. Both of these busses were subsequently energized by the diesel. Refueling operations in progress were immediately halted. Core cooling was momentarily interrupted and restored upon safeguards bus reenergization. The spent fuel pool cooling loop in operation was powered from the opposite train and hence was not interrupted. RCS temperature was maintained at 74 degrees Fahrenheit and spent fuel pool temperature was also at 74 degrees Fahrenheit throughout the event. This partial loss of power was due to a substation transformer fault (one of two offsite power sources were lost). The licensee notified the NRC Resident Inspector.Emergency Diesel Generator
ENS 4141417 February 2005 02:12:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		Reactor Trip Due to Turbine Trip as a Result of Failed Power SupplyGinna Station received a reactor trip from Turbine Trip at 2112 hrs on 02/16/05. The Turbine Trip signal was generated from the ATWS Mitigation System Actuation Circuitry (AMSAC) related to a failed power supply in the Advanced Digital Feedwater Control System. All control rods inserted on the reactor trip. The plant is currently stable in Mode 3, RCS Pressure 2235 psig, Temperature 540 deg F. AFW (Auxiliary Feed Water) did actuate as designed after the trip. For the transient, min & max Temperatures, Pressures & Levels are: RCS Temperature: Max - 561 deg F Min - 538 deg F RCS Pressure: Max - 2250 psig Min - 2218 psig Pressurizer Level: Max - 50% Min - 28% The electrical grid is stable. Decay heat is being rejected to the Main Condenser. The licensee notified the NRC Resident Inspector and the State Public Service Commission.Feedwater
Main Condenser
Control Rod
ENS 4031913 November 2003 14:54:0010 CFR 50.72(b)(3)(iv)(A), System ActuationAuto Start and Loading of the "B" Emergency Diesel GeneratorWhile the plant was in "ER-SC.1", Adverse Weather Plan, due to high winds and offsite power in a 50/50 alternate lineup, a momentary loss of offsite power circuit "751" occurred. This resulted in an undervoltage condition on safeguard busses "16" and "17". The "B" emergency diesel Generator automatically started and energized buses "16" and "17" . Entry was made into AP.Elec.1, Loss of 12A and/or 12B busses. The offsite electrical lineup was changed to 100/0 on circuit "767". Safeguard busses were re-energized from offsite power and the "B" diesel generator was secured and returned to automatic. This sequence of events took 22 minutes. The NRC Resident Inspector was notified.Emergency Diesel Generator
ENS 4024815 October 2003 14:24:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation		a Manual Reactor Trip from 1% Power as Per Abnormal Operating Procedures OccurredThe plant was in mode 2 (1% power) making preparations to place the unit on-line per procedures. At 0921, the plant had entered ER-SC.1, Adverse Weather Plan, due to sustained winds greater than 55 mph. At 1024, offsite power circuit "751" was lost due to offsite storm damage(tree down on powerline). The loss of circuit "751" led to the loss of the "B" reactor coolant pump on undervotltage. Procedure AP-RCS.2, loss of Reactor Coolant flow was entered. At 1026 a manual Reactor Trip was initiated as directed by procedures. All rods fully inserted, no ECCS injection or safety valves lifted and all ESF systems functioned as designed. Due to the partial loss of offsite power, the "B" emergency diesel auto started and loaded to supply safeguards equipment. The "A" and "B" aux feedwater pumps also auto started. Offsite power was restored to safeguards equipment being supplied power by the "B" emergency diesel within 15 minutes from the second offsite power line. The NRC Resident inspector was notified.Feedwater
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