Semantic search
| Start date | Report date | Site | Reporting criterion | System | Event description | |
|---|---|---|---|---|---|---|
| ENS 57177 | 18 June 2024 07:17:00 | Saint Lucie | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Auxiliary Feedwater | The following information was provided by the licensee by email: On 06/18/2024 at 0317 EDT with Unit 2 at 18 percent power, the reactor was manually tripped due to elevated secondary chemistry levels (sodium and chlorides). The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in Mode 3. Decay heat is being removed by auxiliary feedwater and atmospheric dump valves. St. Lucie Unit 1 was unaffected and remains at 100 percent power. This event is being reported pursuant 10CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. An investigation is underway to determine the root cause of the elevated chemistry levels.
The following retraction was provided by the licensee via email: The purpose of this notification is to retract a previous report made on 06/18/2024 at 0652 (EDT) (EN# 57177). Notification of the event to the NRC was initially made because of inserting a manual reactor trip due to elevated secondary chemistry levels (sodium and chlorides). After the initial report, Florida Power and Light has concluded that the event did not meet the reporting requirements on 10 CFR 50.72(b)(2)(iv)(B) since it was part of a normal plant shutdown. Therefore, this event is not considered an unplanned scram and is not reportable to the NRC as a Licensee Event Report per 10 CFR 50.73. The NRC Senior Resident Inspector has been notified. | |
| ENS 57175 | 16 June 2024 17:33:00 | Waterford | 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 Decay Heat Removal Main Condenser Control Rod | The following information was provided by the licensee via email: On June 16, 2024, at 1233 CDT, Waterford Steam Electric Station Unit 3 was operating at 93 percent power when an automatic reactor trip occurred. Immediately following the reactor trip, emergency feedwater (EFW) actuated automatically. The unit is currently in Mode 3. All control rods fully inserted. Decay heat removal is via the main condenser. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip, except steam generator (SG) feedwater pump 'A' tripped and SG '1' main feed regulatory controller went to manual. Steam generator water levels are being controlled with the SG feedwater pump 'B'. The cause of the trip is currently being investigated. 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 when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as a valid actuation of the EFW system. The NRC Resident Inspector has been notified. | |
| ENS 57164 | 5 June 2024 20:48:00 | 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 1648 EDT on 06/05/2024, with Unit 4 in Mode 1 at approximately 35 percent power, the reactor was manually tripped due to unexpected response of the turbine run back circuitry following a trip of main feedwater pump `C. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam through the steam dumps to the main condenser and main feedwater is supplying the steam generators. Units 1, 2 and 3 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 57161 | 4 June 2024 17:52:00 | Saint Lucie | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Main Condenser | The following information was provided by the licensee via email: At 1352 EDT, on June 4, 2024 with Unit 2 at 92 percent power, the reactor was manually tripped due to a loss of condenser vacuum resulting from a circulating water pump trip. The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat is being removed by steam discharge to the main condenser using the turbine bypass valves. St. Lucie Unit 1 was not affected and remains at 100 percent power. This event is being reported pursuant to 10 CFR 50.72 (b)(2)(iv)(B). The NRC Resident Inspector has been notified. | |
| ENS 57158 | 2 June 2024 21:08:00 | Surry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Emergency Core Cooling System Control Rod Main Steam | The following information was provided by the licensee via phone and email: Surry Unit 1 reactor automatically tripped at 1708 EDT on 6/2/24 due to a turbine trip (with the reactor operating at greater than 10 percent reactor power). The turbine control system is currently under investigation. Reactor coolant temperature is being maintained via main steam dumps with main feedwater supplying the steam generators. All systems operated as required. The trip was uncomplicated and all control rods fully inserted into the core. There was no emergency core cooling system (ECCS) or auxiliary feedwater system actuation. Offsite power remains available. There is no impact to Surry Unit 2. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of reactor protection system activation. The NRC Resident Inspector has been notified. There were no radiation releases, personnel injuries, or contamination events due to this event. | |
| ENS 57155 | 30 May 2024 23:49:00 | Harris | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Protection System Control Rod | The following information was provided by the licensee via email and phone: On May 30, 2024, at 1949 EDT, Unit 1 automatically tripped from 100 percent power due to an electrical fault on the B unit auxiliary transformer. The unit has been stabilized in mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). Decay heat is being removed by the condenser steam dump system and Unit 1 is in a normal shutdown electrical lineup. There was no impact on the health and safety of the public or personnel. The NRC Resident Inspector has been notified. | |
| ENS 57148 | 29 May 2024 10:24:00 | North Anna | 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 and phone: On May 29, 2024, at 0624 EDT, Unit 1 automatically tripped from 100 percent power due to a negative rate trip. The unit has been stabilized in mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater pumps actuated as designed because of the reactor trip and is reportable per 10 CFR 50.72(b)(3)(iv)(A) for a valid engineered safety feature (ESF) actuation. Decay heat is being removed by the condenser steam dump system and Unit 1 is in a normal shutdown electrical lineup. Unit 2 was not affected by this event. The NRC Resident has been notified. | |
| ENS 57141 | 23 May 2024 16:46:00 | 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 email: At 2223 CDT on May 23, 2024, with Quad Cities Unit 2 at 38 percent power, the reactor automatically tripped due to a turbine trip signal resulting in main stop valve closure, creating a valid reactor protection system signal. Reactor vessel level reached the low-level set-point following the scram, resulting in valid Group II and Group III containment actuation signals. The trip was not complex with all systems responding as expected 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) and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group II and Group III isolation. Operations responded using their emergency operating procedures 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 remains at 100 percent power. 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: Unit 2 was at a reduced power for maintenance. | |
| ENS 57132 | 19 May 2024 04:30:00 | Beaver Valley | 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 Control Rod | The following information was provided by the licensee via email: At 0030 (EDT) on 5/19/24, with Beaver Valley Unit 1 in mode 1 at 14 percent power, the reactor was manually tripped due to inability to control the A steam generator water level. The trip was not complex, with all systems responding normally post-trip. The turbine driven auxiliary feedwater pump automatically started on a valid actuation signal. All control rods inserted into the core. Operations responded and stabilized the plant. Decay heat is being removed by the feedwater system and the main condenser. Beaver Valley Unit 2 is unaffected. Due to the reactor protection system 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 emergency safety feature system actuation (automatic start of the turbine driven auxiliary feedwater pump) while critical, this event is being reported as an eight-hour, non-emergency notification per 10CFR 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 verbally notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 is stable on off-site power, normal configuration. All emergency systems are available. | |
| ENS 57128 | 15 May 2024 08:27:00 | Cook | 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 Turbine | The following information was provided by the licensee via email: On May 15, 2024 at 0427 EDT, DC Cooks Unit 2 reactor was manually tripped due to difficulty maintaining steam generator water levels. DC Cook Unit 2 had removed the main turbine from service at approximately 0354 EDT during a planned down-power to repair a steam leak on the high pressure turbine right outer steam/stop control valve upstream drip pot. Stable steam generator water levels were unable to be maintained. As a result, DC Cook Unit 2 was manually tripped with reactor power stabilizing at approximately 20 percent. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System actuation as a four hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight hour report. The reactor trip was not complicated and all plant systems functioned normally. The DC Cook NRC Resident Inspector was notified. | |
| ENS 57124 | 12 May 2024 21:41:00 | South Texas | 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 Emergency Diesel Generator Main Turbine Control Rod | The following information was provided by the licensee via email and phone: At 1641 CDT on May 12, 2024, with Unit 2 in Mode 1 at 15 percent power, the reactor automatically tripped due to a unit auxiliary transformer lockout. During the trip, all control rods fully inserted. The cause of the transformer lockout is currently unknown. Emergency diesel generator (EDG) 21 and 23 actuated and all three engineered safety feature (ESF) busses were energized. All equipment responded as expected except for steam generator power operated relief valve (PORV) 2C which failed to open when required in automatic, and the load center (LC) E2A output breaker which failed to close automatically but was closed manually. Steam generator PORV 2C did open when placed in manual, although it subsequently failed to full open and was then closed. Primary system temperature and pressure are currently being maintained at 567 degrees/2235 psig following start of reactor coolant pumps 2A and 2D. Due to the reactor protection system actuation (RPS) 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 per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the emergency diesel generators. There was no impact to 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: South Texas Project Unit 2 was in Mode 1 at 15 percent power due to performance of testing and analysis on the main turbine prior to the RPS actuation.
The following information was provided by the licensee via email and phone: South Texas Project is submitting the following correction to the event notification: The steam generator (SG) power operated relief valve (PORV) 2C did not fail to open automatically. System pressure during this event did not reach the automatic setpoint for the PORV (1225 psi), and there was no demand for it to open automatically. During the event, SG PORV 2C was taken to manual and it went full open when the up button was pushed slightly. It went closed when the down button was pressed to close it manually. In addition, the load center E2A output breaker initially failed to close automatically, however, after operations placed it in pull-to-lock and returned the hand switch to automatic, it closed automatically. Notified R4DO (Dixon). | |
| ENS 57107 | 5 May 2024 08:38:00 | Braidwood | 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 and phone: At 0338 CDT, with the unit 1 in mode 1 at 6 percent power, the reactor automatically tripped due to lowering steam generator water level. 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) and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for an actuation of the auxiliary feedwater system. Operations responded using procedure 1BwEP-0 and stabilized the plant in mode 3. Decay heat is removed by steam dumps via the main condenser. 1A and 1B auxiliary feedwater pumps were actuated manually prior to the reactor trip in an attempt to restore steam generator water level. Unit 2 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 57090 | 25 April 2024 03:15:00 | Browns Ferry | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Coolant System Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Emergency Core Cooling System | The following information was provided by the licensee via email: On 4/24/2024 at 2215 CDT, Browns Ferry Unit 1 experienced an automatic reactor scram. The cause of the scram is currently under investigation. The main steam isolation valves (MSIVs) remain open with the main turbine bypass valves controlling reactor pressure. The reactor feedwater pumps are in service to control reactor water level. Primary containment isolation systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all components actuated as required. Following the reactor scram, due to reactor water level reaching minus 45 inches, both high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) initiation signals were received, and both initiated as designed. All safety systems operated as expected. This event requires a 4-hour report per 10 CFR 50.72(b)(2)(iv)(A), `Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event requires a 4-hour report 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. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A), `Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B): 1) Reactor protection system (RPS) including: reactor scram or reactor trip. 2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). 4) ECCS for boiling water reactors (BWRs) - high-pressure coolant injection (HPCI). 5) BWR reactor core isolation cooling system (RCIC). All safety systems operated as expected. At no time was public health and safety at risk. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Units 2 and 3 were not affected. | |
| ENS 57075 | 13 April 2024 04:35:00 | Beaver Valley | 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 Control Rod | The following information was provided by the licensee via phone and email: At 0035 EDT on April 13, 2024, with Unit 1 at 97 percent power, the reactor automatically tripped due to 1 of 3 reactor coolant pump (RCP) low flow reactor trip (signal) associated with a loss of the A and B 4160 volt normal buses. Auxiliary feedwater and the 1-1 emergency diesel generator (EDG) automatically started on valid actuation signals. The 1-1 EDG sequenced on to supply all required loads per plant design. All control rods fully inserted and the trip was not complex with all systems responding normally post-trip. Operators have responded and stabilized the unit in Mode 3 (Hot Standby). Decay heat is being removed by discharging steam to the main condenser via the condenser steam dump system with steam generators being supplied by the main feedwater system. Unit 2 is not affected by the event. 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 valid actuations of auxiliary feedwater and the 1-1 EDG, this event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. 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: Power for the A-E Bus is on the 1-1 EDG. The D-F Bus is on offsite power. One electrical train of offsite power is down. | |
| ENS 57046 | 23 March 2024 04:04:00 | Fermi | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Primary containment Reactor Pressure Vessel Main Condenser Control Rod Main Steam | The following information was provided by the licensee via email: At 0004 EDT on March 23, 2024, with the unit in Mode 1 at 23 percent power, the reactor automatically scrammed due to high reactor pressure vessel pressure when the turbine bypass valves unexpectedly closed while attempting to lower generator MW to 55 MWe to support shutdown for a refueling outage. The scram was not complex, with systems responding normally post-scram, with the exception of the pressure control system. The transient occurred while lowering on turbine speed/load demand which caused a rise in pressure and power until the reactor protection system setpoint for reactor pressure high was exceeded and resulted in an automatic reactor scram. The plant was preparing to shut down for a refueling outage when the trip occurred. Operations responded and stabilized the plant. Reactor water level is being maintained at normal level. Decay heat is being removed by the main steam system to the main condenser using manual operation of the turbine bypass valves. All control rods inserted into the core. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CPR 50.72(b)(2)(iv)(B). Additionally, received expected (primary containment) isolations for Level 3: Group 13 drywell sumps, Group 15 (traverse in-core probe) TlPs (which was already isolated) and Group 4 (residual heat removal - shutdown cooling) RHR-SDC (which was already isolated). The primary containment isolation event is being reported under 10 CFR 50.72(b)(3)(iv)(A). Also, due to the main turbine bypass valves unexpectedly closing, this is also being reported under 10 CFR 50.72(b)(3)(v)(D). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract the 10 CFR 50.72(b)(3)(v)(D) reporting criteria of event notification 57046 reported on March 23,2024. Based on further evaluation, Fermi 2 has concluded that there was no event or condition that could have prevented fulfillment of a safety function that was needed to mitigate the consequence of an accident. Although discussed in Chapter 15 of the UFSAR, the turbine bypass valves do not provide a safety related function and are not credited safety related components for accident mitigation. Therefore, Fermi 2 is retracting the 10 CFR 50.72(b)(3)(v)(D) reporting criteria that was included on the March 23, 2024 event notification. Notified R3DO (Betancourt-Roldan) | |
| ENS 57042 | 22 March 2024 04:37:00 | Waterford | 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 | Feedwater Reactor Protection System Emergency Diesel Generator Main Transformer Emergency Core Cooling System Decay Heat Removal Main Condenser Control Rod | The following information was provided by the licensee: A Notification of Unusual Event, HU4.4 (see note below) was declared based a fire in the protected area requiring off site assistance to extinguish. The fire was in the main transformer yard. The fire was detected at 2328 CDT on March 21, 2024, and the fire was declared out at 0009 CDT on at March 22, 2024. An automatic reactor trip was initiated due to a loss of offsite power to the "B" train and a failure to automatically transfer from unit auxiliary transformer "B" to startup transformer "B. The licensee notified State and local authorities and the NRC Resident Inspector. The NRC remained in Normal. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). NOTE: Due to a typographical error initiating condition HU4.1 was initially recorded for the event. The correct initiating condition is HU4.4 as now shown.
The licensee terminated the Notification of Unusual Event at 0221 CDT on 3/22/24. The licensee notified the NRC Resident Inspector. Notified R4DO (Gepford), IR-MOC (Grant), NRR-EO (Felts), DHS-SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
RPS ACTUATION The following information was provided by the licensee via email: On March 21, 2024, at 2328 CDT, Waterford 3 Steam Electric Station, Unit 3 was operating at 98 percent power when an automatic reactor trip was initiated due to a loss of offsite power to the B train and a failure to automatically transfer from unit auxiliary transformer B to startup transformer B. Emergency feedwater actuation signal 2 (EFAS), safety injection actuation signal (ECCS), containment isolation actuation signal and emergency diesel generators automatically actuated. The unit is currently stable in Mode 3. All control rods fully inserted and all other plant equipment functioned as expected. Forced circulation remains with one reactor coolant pump per loop running. Decay heat removal is via the main condenser. A train safety bus is being supplied by off-site power, and B train safety bus is being supplied by emergency diesel generator B. Following the loss of offsite power to the B train, it was reported that main transformer B and startup transformer B were both on fire. The Emergency Director declared an Unusual Event at time 2337 CDT. The fire was reported extinguished at 0009 CDT on March 22, 2024, and the Unusual Event was terminated at 0221 CDT on March 22, 2024. Offsite assistance was requested. The local fire department responded to the site but the fire was extinguished by the on-shift fire brigade. NRC Region IV management was contacted regarding the emergency plan entry at 0030 CDT on March 22, 2024. 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 non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system, ECCS, Containment Isolation and Emergency Diesel Generators. The NRC Resident Inspector has been notified. Notified R4DO (Gepford)
The initial notification in event notice #57042 by Waterford Steam Electric Station, Unit 3, reported a Notice Of Unusual Event (NOUE) emergency declaration due to a fire in the protected area requiring off site support to extinguish. The basis for retraction of the initial emergency notification is that this event did not meet the definition of a fire in the protected area that requires off site support to extinguish. Guidance provided in Nuclear Energy Institute (NEI) 99-01, Rev. 6 and implemented in Waterfords Emergency Plan procedure, initiating Condition HU4.4 states, The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. (NOTE: The Initial Notification Form sent from the Control Room at 2341 CDT on March 21, 2024, requested by and provided to the NRC Headquarter Operations Center via e-mail at 0302 CDT on March 22, 2024, stated that the Emergency Classification had been made on Initiating Condition HU4.4 rather than HU4.1)" When the event occurred on March 21, 2024, the Emergency Director declared an Unusual Event at 2337 CDT and requested offsite support based on the information available at that time including the initial assessment by the fire brigade leader and expected need for offsite support to extinguish the fire. As reported in the 0420 EDT update on March 22, 2024, the fire was reported extinguished at 0009 CDT on March 22, 2024, by the Waterford Fire Brigade without the need of offsite support." Notified R4DO (Kellar).
The initial notification in EN 57042 by Waterford Steam Electric Station, Unit 3, reported an emergency declaration of an Unusual Event due to a fire in the protected area requiring off site support to extinguish. The basis for the update to the initial notification is that this event did not meet the definition of a Fire in the Protected Area that requires offsite support to extinguish. As provided in NEI 99-01, Rev. 6 and implemented in Waterfords emergency plan procedure, initiating condition HU4.4 states, The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Additionally, EAL 4.1 for a fire not extinguished within 15 minutes of detection in any Table H-1 fire area was not applicable because the fire did not occur in a Table H-1 fire area. When the event occurred on March 21, 2024, the Emergency Director declared an Unusual Event at 2337 CDT and requested offsite support based on the information available at that time including the initial assessment by the fire brigade leader and expected need for offsite support to extinguish the fire. As reported in the 0420 EDT update on March 22, 2024, the fire was reported extinguished at 0009 CDT on March 22, 2024, by the Waterford Fire Brigade without the need of offsite support. (NOTE: The Initial Notification Form sent from the Control Room at 2341 CDT on March 21, 2024, requested by and provided to the Headquarters Operation Center via e-mail at 0302 CDT on March 22, 2024, stated that the emergency classification had been made on initiating condition HU4.4 rather than HU4.1) In accordance with NRC Approved guidance in FAQ 21-02 (ML21117A104), Waterford 3 is retracting the initial event notification made at 0117 EDT on March 22, 2024. The remaining events that were reported in EN 57042 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 non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW (emergency feedwater) system, ECCS (emergency core cooling system), containment isolation and emergency diesel generators are still applicable and require no additional update at this time. The licensee also provided a site map. Notified R4DO (Kellar) | |
| 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 | 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. |