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| Start date | Report date | Site | Reporting criterion | System | Event description | |
|---|---|---|---|---|---|---|
| ENS 57033 | 17 March 2024 20:15:00 | Comanche Peak | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via phone and email: On March 17, 2024, at 1515 CDT, the Comanche Peak Unit 2 reactor was manually tripped due to an anticipated automatic trip due to lo-lo steam generator (SG) water levels. Prior to the trip, main feedwater pump '2B' tripped and an auto runback to 700 MW (60 percent power) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs. Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedures IPO-007B. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the '2B' main feed pump trip was due to loss of primary and redundant power to the servo control valve. The loss of power to the servo control valve is under investigation. | |
| ENS 57032 | 16 March 2024 19:49:00 | Waterford | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Main Steam Isolation Valve Control Rod Main Steam | The following information was provided by the licensee via phone and email: At 1449 CDT, Waterford 3 Steam Electric Station was operating at 100 percent power when a manual reactor trip was initiated due to main feed isolation valve (FW-184B) and main steam isolation valve (MS-124B) going closed unexpectedly. Emergency feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat is being removed through the turbine bypass valves and the atmospheric dump valve on loop '2'. There is no primary to secondary system leakage. The cause of the isolations is still being investigated. | |
| ENS 57024 | 12 March 2024 13:16:00 | Comanche Peak | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via phone and email: On March 12, 2024, at 0816 CDT, Comanche Peak Unit 2 reactor automatically tripped on lo-lo level in the 2-03 steam generator (SG). Prior to the trip, main feedwater pump (MFP) 2A speed reduced and a manual runback to 700 MW (60 percent) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs. Concurrent with the loss of speed on MFP 2A, a servo filter swap was in progress on MFP 2A. Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedure IPO-007A. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the loss of the MFP is under investigation. Unit 1 was unaffected. | |
| ENS 57021 | 11 March 2024 17:37:00 | Hatch | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Protection System Emergency Diesel Generator Reactor Core Isolation Cooling Emergency Core Cooling System Main Condenser | The following information was provided by the licensee via phone and email: On March 11, 2024, at 1337 EDT, with Unit 1 in Mode 1 at 35 percent power performing power ascension activities, the reactor was manually tripped due to the 'A' reactor feed pump (RFP) tripping on low suction pressure. Due to the power level at the time, the 'B' RFP had not been placed in service. Closure of containment isolation valves (CIVs) in multiple systems and 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. The 'B' RFP was placed in service and is controlling reactor water level. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 2 is 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 NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'A' RFP is under investigation. The reactor electric plant remains in a normal lineup with both emergency diesel generators available. There were no temperature or pressure technical specification limits approached. | |
| ENS 57006 | 5 March 2024 06:32:00 | Watts Bar | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via email: At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. 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). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. | |
| ENS 57004 | 4 March 2024 00:42:00 | Nine Mile Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Shutdown Cooling Decay Heat Removal Control Rod | The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration. | |
| ENS 57003 | 3 March 2024 17:42:00 | Prairie Island | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater | The following information was provided by the licensee via email: At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs (power operated relief valves). Unit 1 remains at 100 percent power and is unaffected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation. | |
| ENS 56995 | 28 February 2024 14:39:00 | Monticello | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Main Condenser | The following information was provided by the licensee via fax and email: At approximately 0839 (CST) with Unit 1 in Mode 1 at 100 percent power, the reactor automatically scrammed due to the depressurization of the SCRAM air header caused by an invalid signal that (occurred) during system testing. The SCRAM was uncomplicated with all systems responding as expected. The cause and details of the event are under investigation. Containment isolation valves actuated and closed on a valid Group 2 signal. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 isolation signal. Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. State as well as Wright and Sherburne Counties will be notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Anticipated Transient Without Scram (ATWS) circuit was being tested when an invalid signal was sent to depressurize the SCRAM air header. | |
| ENS 56991 | 24 February 2024 20:46:00 | Calvert Cliffs | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via email: At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56971 | 16 February 2024 11:34:00 | Farley | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater | The following information was provided by the licensee via email: At 0048 CST on February 16, 2024, with Unit 2 in mode 1 at 100 percent power, the reactor was manually tripped due to a loss of 2A 125V DC distribution panel. The trip was complex due to the loss of components associated with A-train DC power. Operations responded and stabilized the plant. Decay heat is being removed by the atmospheric relief valves. Unit 1 is not affected. An automatic actuation of the auxiliary feedwater system (AFW) occurred due to low-low steam generator levels. The AFW auto-start is an expected response with low-low steam generator levels from the reactor trip. AFW is still currently controlling steam generator levels. 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). This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56968 | 15 February 2024 08:47:00 | Callaway | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Auxiliary Feedwater Control Rod | The following information was provided by the licensee via email: At 0247 CST on 2/15/2024, Callaway Plant was in mode 1 at approximately 100 percent power when a turbine trip and reactor trip occurred. All safety systems responded as expected with the exception of an indication issue on the feedwater isolation valves, which were confirmed closed. A valid feedwater isolation signal and auxiliary feedwater actuation signal were also received as a result of the reactor trip. The plant is being maintained stable in mode 3. All control rods fully inserted from the reactor trip signal and decay heat is being removed via the auxiliary feedwater system and steam dumps. The NRC Resident Inspector was notified. | |
| ENS 56936 | 29 January 2024 17:02:00 | Peach Bottom | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | High Pressure Coolant Injection Primary Containment Isolation System Reactor Core Isolation Cooling Residual Heat Removal Main Condenser Control Rod | The following information was provided by the licensee via email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed 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: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified. | |
| ENS 56935 | 28 January 2024 02:41:00 | Watts Bar | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via email: At 2141 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. 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). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. The licensee notified the NRC Resident Inspector. | |
| ENS 56894 | 16 December 2023 09:50:00 | Grand Gulf | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Control Rod | The following information was provided by the licensee via email: On December 16, 2023, at 0350 CST, Grand Gulf Nuclear Station was operating in mode 1 at 81 percent power when an automatic scram occurred due to a turbine trip signal. Before the scram the unit was performing a rod sequence exchange, and no critical work was underway. The cause of the turbine trip signal is not known at this time and is being investigated. All control rods fully inserted, there were no complications, and all plant systems responded as designed. Reactor water level is being maintained by main feedwater and condensate. Reactor pressure is being maintained with main turbine bypass valves. No radiological releases have occurred due to this event. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of the reactor protection system when the reactor is critical and specified system actuation due to expected reactor water level 3 isolation signals on a reactor scram. The NRC Senior Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Group 2 and Group 3 isolations occurred on the Level 3 isolation signal. | |
| ENS 56889 | 15 December 2023 00:39:00 | Hope Creek | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Feedwater Main Condenser Control Rod | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. | |
| ENS 56887 | 13 December 2023 07:02:00 | River Bend | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Reactor Core Isolation Cooling | The following information was provided by the licensee via phone and email: At 0102 CST, while operating at 100 percent (reactor) power, River Bend Station experienced an automatic reactor scram caused by a turbine trip signal. The cause of the turbine trip signal is not known at this time and is being investigated. At 0108, reactor core isolation cooling (RCIC) was initiated due to a loss of reactor feed pumps following feedwater heater string isolation. At 0114, reactor water level control was transferred back to feedwater and RCIC was secured. Reactor water level is being maintained by feedwater pumps and reactor pressure is being maintained by turbine bypass valves. The scram was uncomplicated and all other plant systems responded as designed. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) specified system actuation as result of expected post scram (reactor water) level 3 isolations and manual initiation of RCIC. No radiological releases have occurred due to this event from the unit. The NRC Senior Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the turbine trip, while still under investigation, was likely due to an electrical transient involving the main generator. Walkdowns in the switchyard post-scram identified damage to one of the output breaker disconnects. | |
| ENS 56863 | 18 November 2023 05:55:00 | River Bend | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Emergency Diesel Generator Control Rod | The following information was provided by the licensee via phone and email: On November 17, 2023, at 2215 CST, River Bend Station (RBS) was operating at 30 percent reactor power performing plant startup activities when an isolation of low-pressure feedwater string `A' occurred. The team entered applicable alternate operating procedures and inserted control rods to exit the restricted region of the power to flow map. Feedwater temperature continued to lower until it challenged the prohibited region of the AOP-0007 graph requiring a reactor scram. The team inserted a manual reactor scram at 2355 from 24 percent reactor power. All control rods fully inserted and there were no complications. All systems responded as designed. Currently RBS Unit 1 is stable with reactor level being maintained 10 to 51 inches with feed and condensate, and pressure being maintained 500 to 1090 psig using steam drains. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations. The NRC Senior Resident inspector has been notified. No radiological releases have occurred due to this event from the unit. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The electric plant is in a normal lineup for current plant conditions with all emergency diesel generators available. The cause of the initial isolation of low-pressure feedwater string "A" is still under investigation. | |
| ENS 56856 | 16 November 2023 07:27:00 | Calvert Cliffs | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report 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: All rods fully inserted. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). | |
| ENS 56852 | 14 November 2023 16:41:00 | Farley | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | The following information was provided by the licensee via phone and email: At 1041 CST on 11/14/23 with Farley Unit 2 in Mode 1 at 10 percent power, the reactor was manually tripped due to rising steam generator levels. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Auxiliary feedwater (AFW) was manually initiated in accordance with plant procedures and is feeding the steam generators. Heat removal is being provided via the atmospheric relief 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). This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. The licensee attempted to take manual control of the feedwater control valves to lower steam generator level but, due to reaching a steam generator level that requires a manual trip, the licensee manually tripped the reactor. | |
| ENS 56846 | 10 November 2023 08:14:00 | Susquehanna | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Main Condenser | 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 56839 | 7 November 2023 21:17:00 | Calvert Cliffs | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. | |
| ENS 56826 | 1 November 2023 10:48:00 | Hatch | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Emergency Core Cooling System Main Condenser | 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 56813 | 25 October 2023 01:59:00 | Turkey Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Auxiliary Feed Water | The following information was provided by the licensee via email: At 2159 on 10/24/2023, with Unit 3 in Mode 1 at 100 percent power, the reactor was automatically tripped due to an actuation signal into the Unit 3 reactor protection system protection rack during maintenance. The trip was uncomplicated with all systems responding normally post trip. Decay heat is being removed via auxiliary feed water system and the steam dump system. Unit 4 is not affected. This event is being reported pursuant to 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The cause of the automatic reactor trip will be investigated by the licensee. | |
| ENS 56803 | 19 October 2023 16:10:00 | Prairie Island | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via email: On 10/19/2023, at approximately 1110 (CST), with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The cause of the trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generator through the steam generator power operated relief valve. The trip was complex as non-safety related power was lost to both Unit 1 and Unit 2. Unit 1 is currently in mode 3 and on natural recirculation as both reactor coolant pumps are without power. Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was lost for approximately 70 minutes. No impacts to the SFP temperature were observed. 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). Due to the actuation of the auxiliary feedwater system following the reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The second paragraph of the original report is amended as follows to correct information regarding the spent fuel pool for Unit 2: Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was maintained at all times with one train of SFP cooling. The second train lost power and was restarted approximately 70 minutes (after power was lost). No impacts to the SFP temperature were observed. Notified R3DO (Orth) and IR MOC (Crouch) and NRR EO (Felts) via email | |
| ENS 56790 | 13 October 2023 01:27:00 | Ginna | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Control Rod Auxiliary Feed Water Main Steam | The 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. | |
| ENS 56769 | 1 October 2023 03:14:00 | Diablo Canyon | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater | The following information was provided by the licensee via email: At 2014 (PDT) on 09/30/2023, with (Diablo Canyon) Unit 1 in Mode 1 at 11 percent reactor power in preparation for a pre-planned manual reactor trip into a scheduled refueling outage, the reactor was manually tripped due to a failed secondary system dump valve. Auxiliary feedwater was manually started in accordance with plant procedures. This event is being reported in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). There was no plant or public safety impact. 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: Diablo Canyon Unit 2 was unaffected. | |
| ENS 56759 | 27 September 2023 15:41:00 | Monticello | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Steam Isolation Valve Primary containment Reactor Pressure Vessel Main Steam Line | The following information was provided by the licensee via fax: (On 09/27/2023) at 1041 CDT, with the plant at 75 percent power and main turbine control valve testing in progress, a reactor pressure transient resulted in a reactor steam dome high pressure scram and subsequent group 1 primary containment isolation of the main steam lines (MSL). All main steam isolation valves closed as a result of the group 1 isolation signal. Additionally, a group 2 containment isolation signal was received due to reactor pressure vessel (RPV) level less than plus 9 inches during the transient. Operations personnel responded and stabilized the plant. The high-pressure coolant injection (HPCI) system was placed in service to control RPV pressure. HPCI did not inject into the RPV and was not needed to control RPV water level. The cause of the initial pressure transient is under investigation. The NRC Resident Inspector has been notified.
The utility notified the State of Minnesota and Wright and Sherburne counties. Notified R3DO (Orlikowski) | |
| ENS 56755 | 22 September 2023 22:19:00 | Turkey Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feed Water | The following information was provided by the licensee via email: At 1819 EDT on 9/22/2023, with Unit 3 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main generator lockout. The probable cause of the main generator lockout was from a lightning strike. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in Mode 3. Decay heat is being removed by the steam dump system. Unit 4 is not affected. Auxiliary feed water was actuated as expected as a result of the reactor trip. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. | |
| ENS 56731 | 9 September 2023 15:43:00 | Ginna | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Control Rod Auxiliary Feed Water Main Steam | The 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. | |
| ENS 56710 | 2 September 2023 10:32:00 | Nine Mile Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Feedwater Reactor Protection System Reactor Core Isolation Cooling High Pressure Core Spray Reactor Water Cleanup Control Rod | The 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 56692 | 22 August 2023 21:24:00 | Vogtle 1/2 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser Main Steam | The following information was provided by the licensee via email: At 1724 EDT, on August 22, 2023, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to a failure of the non-safety heater drain pump 'B' and the failure of the non-safety condensate pump 'A' to automatically or manually start. At 1735 EDT, a fire was identified on heater drain pump 'B' and was extinguished by the onsite fire brigade at 1807 EDT. Operations responded and stabilized the plant. The trip was not complex, with all safety systems responding normally post-trip. Decay heat is being removed by the main steam system to the main condenser using the steam dumps. There was no impact to Units 2, 3, or 4. An automatic actuation of the auxiliary feedwater system (AFW) also occurred, as expected, due to lo-lo steam generator levels resulting from the reactor trip. AFW is currently controlling all steam generator levels at their normal levels. 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). Due to the notification of another government agency, the Burke County Fire Department, this event is being reported as a four-hour, non-emergency notification under 10 CFR 50.72(b)(2)(xi). The Burke County Fire Department was not needed to extinguish the fire. This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56676 | 11 August 2023 08:29:00 | Quad Cities | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Main Condenser | The following information was provided by the licensee via phone and email: At 0329 (CDT) on August 11, 2023, with Unit 2 in Mode 1 at 90 percent power, the reactor automatically tripped due to a turbine trip. The trip was uncomplicated with all systems responding normally post-trip. The cause and details of the event are under investigation. Containment isolation valves actuated closed in multiple systems on a valid Group II signal. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group II isolation. Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56673 | 10 August 2023 04:39:00 | Perry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Feedwater Reactor Protection System Reactor Core Isolation Cooling Reactor Pressure Vessel Core Spray Emergency Core Cooling System Main Steam Line Safety Relief Valve | The following information was provided by the licensee via email: At 0039 (EDT) on 8/10/23, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped during a reactor protection system (RPS) bus shift. All systems responding normally post-trip. There was no equipment inoperable at the time of the trip. Operations responded and stabilized the plant. Reactor water level being maintained via feedwater. Decay heat is being removed by cycling safety relief valves. An actuation of high-pressure core spray, division 3 diesel generator, and reactor core isolation cooling occurred during the scram and main steam line isolation closure. The reason for the auto-start was reaching Level 2 (130 inches in the reactor pressure vessel) during the transient. The systems automatically started as designed and injected to the reactor vessel when the Level 2 signal was received. The RPS actuation is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The emergency core cooling system (ECCS) injection is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). The ECCS actuation is being reported as a eight-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56660 | 4 August 2023 21:46:00 | Watts Bar | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via phone and email: At 1746 EDT on 08/04/2023, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to number 2 steam generator low low level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by using the auxiliary feedwater and steam dump systems. 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). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the number 2 steam generator low low level is being investigated. | |
| ENS 56658 | 4 August 2023 17:20:00 | Turkey Point | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Auxiliary Feed Water | The following information was provided by the licensee via email: At 1320 (EDT) on 08/04/2023, with the Unit 3 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering level in the 3C steam generator. The trip was uncomplicated with all systems responding normally post-trip. Decay heat is being removed via the auxiliary feed water system and the atmospheric steam dumps. Unit 4 is not affected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of lowering level in the 3C steam generator was unknown at the time of the notification and will be investigated by the licensee. | |
| ENS 56645 | 30 July 2023 19:26:00 | Seabrook | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Main Turbine Decay Heat Removal Main Condenser | The following information was provided by the licensee via email: On July 30, 2023 at 1526 EDT, with unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to low main turbine electro-hydraulic control oil level. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished using the steam dumps in steam pressure mode to the main condenser. Emergency Feedwater actuated due to low-low steam generator level as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. | |
| ENS 56632 | 21 July 2023 15:48:00 | Vogtle | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Feedwater Reactor Protection System | The following information was provided by the licensee via email: At 1148 EDT on 07/21/2023, with Unit 3 in Mode 1 at 32 percent power, the reactor automatically tripped on low reactor coolant pump (RCP) speed due to decaying RCP motor voltage during power ascension testing. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56614 | 9 July 2023 17:28:00 | Vogtle | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Feedwater Reactor Protection System | The following information was provided by the licensee via email: At 1328 EDT on 07/09/2023, with Unit 3 in Mode 1 at 45 percent power, the reactor automatically tripped during power ascension testing due to low reactor coolant flow from decaying voltage to the reactor coolant pumps. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56593 | 27 June 2023 20:26:00 | Watts Bar | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via phone and email: At 1626 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. 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). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. | |
| ENS 56585 | 22 June 2023 14:35:00 | Robinson | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Main Condenser | The following information was provided by the licensee via email: At 1035, on June 22, 2023, with Unit 2 in Mode 1 at 100% power, the reactor automatically tripped due to `A' train reactor trip breaker and `B' train reactor trip bypass breaker opening during testing. The trip was not complex, with all systems responding normally post-trip. MST-021 (Reactor Protection Logic Train `B' At Power) testing was in progress at the time of trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. 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). As a result of the reactor trip, emergency feedwater actuated; therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56580 | 16 June 2023 23:32:00 | Comanche Peak | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via email: (On June 16, 2023,) at 1832 CDT, Unit 1 reactor automatically tripped on lo-lo level in the '1-04' steam generator (SG). Prior to the trip, the 1B (main feedwater pump) (MFP) tripped due to speed oscillations and a runback to 700MW was in progress. Both motor driven auxiliary feedwater pumps started due to the lo-lo level in SG '1-04'. Unit 1 is being maintained in hot standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007A. The Emergency Response Guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves. The licensee notified the NRC Resident Inspector. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 is in a normal post-trip electrical line-up. There was no effect on Unit 2 due to the Unit 1 trip. | |
| ENS 56551 | 2 June 2023 11:05:00 | Palo Verde | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Condenser | The following information was provided by the licensee via email: The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS (Emergency Notification System) or under the reporting requirements of 10 CFR 50.73. At 0405 MDT on June 2, 2023, the Unit 2 reactor automatically tripped on low steam generator water levels due to degraded flow from the A main feedwater pump. Steam generator water levels reached the automatic Auxiliary Feedwater Actuation System (AFAS) setpoint resulting in automatic AFAS-1 and AFAS-2 actuations and subsequent start of both class auxiliary feedwater pumps. Steam Generator water levels are being restored to normal band with the class 1E powered motor driven auxiliary feedwater pump. Following the reactor trip, all control element assemblies inserted fully into the core. No emergency plan classification was required per the Emergency Plan. Safety related buses remained powered from offsite power during the event and the offsite power grid is stable. Both emergency diesel generators automatically started on the AFAS-1 and AFAS-2 actuations as designed and are currently running unloaded. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed. Unit 1 and 3 are in Mode 1 at 100 percent power. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat is being removed to main condenser via automatic steam bypass and B auxiliary feedwater pump. | |
| ENS 56544 | 30 May 2023 08:46:00 | Millstone | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Steam Generator Reactor Protection System Auxiliary Feedwater | The following information was provided by the licensee via email: At 0446 EDT on 5/30/2023, with Millstone Power Station Unit 3 operating at approximately 100 percent reactor power, an automatic reactor trip occurred due to a turbine trip caused by electrical protection. The reactor trip was uncomplicated and decay heat is being removed via steam dumps to the condenser. All systems responded as expected to the trip. Auxiliary feedwater actuated automatically as expected following the trip due to low-low levels in the steam generators. There was no risk to the public. There was no impact to Millstone Unit 2. This event is being reported as a four hour report under 10CFR50.72(b)(2)(iv)(B) as a condition that resulted in actuation of the reactor protection system while the reactor was critical, and as an eight hour report under 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B) for actuation of the auxiliary feedwater system. The NRC Resident Inspector has been notified. | |
| ENS 56543 | 27 May 2023 23:34:00 | Prairie Island | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater | The following information was provided by the licensee via email: Notification of Unusual Event, HU4.1 declared based on multiple fire alarms in the containment building not verified within 15 minutes. Turbine trip causing reactor trip due to fault on 2GT transformer. At 1845 CDT, verification of no fire in the containment building. Notified DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email).
The following information was provided by the licensee via email: This update is being made to report the actuation of the auxiliary feedwater system following the reactor trip at 1819 CDT. This event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). This update is also being made for the termination of the notification of unusual event at 2304 CDT on 5/27/2023. The basis for the termination was that there was no indication of a fire. Upon lockout of 2GT transformer, main to reserve power transfer did not occur on 3 of 4 non-safeguards buses. Subsequently, operator action successfully restored power to all non-safeguards buses at 1925 CDT. There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified of the update. Notified R3DO (Benjam¡n), NRR EO (Walker), IRMOC (Grant), DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email). | |
| ENS 56531 | 20 May 2023 08:15:00 | Browns Ferry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Control Rod | 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 56497 | 2 May 2023 08:23:00 | Vogtle | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Feedwater Reactor Protection System Main Condenser | The following information was provided by the licensee via email: At 0423 EDT on 05/02/2023, with Unit 3 in Mode 1 at 14 percent power, the reactor was manually tripped due to securing all main feed pumps, due to sudden high differential pressure on their suction strainers. The trip was not complex, with all safety-related systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the reactor trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dumps, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 were 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). 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 05:48:00 | Brunswick | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System Primary Containment Isolation System Shutdown Cooling Main Steam Line Safety Relief Valve Main Condenser | 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 56460 | 10 April 2023 04:48:00 | Vogtle | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Reactor Protection System | The following information was provided by the licensee via email: At 0048 EDT on 4/10/2023, with Unit 3 in Mode 1 at 18 percent power, the reactor automatically tripped due to low reactor coolant flow due to voltage decaying to the reactor coolant pumps during main generator testing activities. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to atmosphere. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56459 | 9 April 2023 04:44:00 | Palo Verde | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Main Steam Isolation Valve Auxiliary Feedwater Main Turbine Main Steam | The following information was provided by the licensee via email: The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. At 2144 MST on April 8, 2023, the Unit 1 reactor automatically tripped due to the loss of reactor coolant pumps stemming from the loss of 13.8 kV power to the pumps. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip as expected so the control room operators opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses failed to complete a fast bus transfer, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. Following the reactor trip, all control element assemblies inserted fully into the core. No automatic specified system actuation was required or occurred. No emergency plan classification was required per the Emergency Plan. Safety related buses remained powered from offsite power during the event and the offsite power grid is stable. Unit 1 is stable and in Mode 3. Decay heat is being removed by the atmospheric dump valves and the class 1E powered motor driven auxiliary feedwater pump. The loss of hydraulic pressure, the main generator output breakers failing to automatically open and the fast bus transfer not actuating are being investigated. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B). The NRC Senior Resident Inspector has been informed. Unit 2 is in a refueling outage in Mode 5 and Unit 3 is in Mode 1 at 100 percent power.
This update is being made to report the manual actuation of the B-train auxiliary feedwater pump and manual main steam isolation signal (MSIS) actuation affecting multiple main steam isolation valves (MSIVs) following the reactor trip. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Warnick)
This update is intended to clarify the initial description of the event that occurred on 4/8/2023. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip. The control room operators manually opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses de-energized. A fast bus transfer did not occur per design, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Gaddy) | |
| ENS 56453 | 5 April 2023 10:51:00 | Summer | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Reactor Protection System | The following information was provided by the licensee via email: At 0651 EDT on April 5, 2023, with Unit 1 in mode 1 at 85 percent power, the reactor was manually tripped due to loss of main feedwater pump 'C'. The trip was not complex, with all systems responding normally post-trip. Main feedwater pump 'B' had previously been removed from service in preparation for a planned shutdown as a part of refueling outage RF27. Operations responded and stabilized the plant. Decay heat is being removed by the emergency feedwater system. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) resulting from valid actuation of the reactor protection and emergency feedwater systems. There was no impact on the health and safety of the public or plant personnel. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee continues to investigate the loss of main feedwater pump 'C'. The licensee notified the NRC Resident Inspector. |