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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55320 | 21 June 2021 05:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip on Generator Lockout Relay Trip | At 0051 CDT Braidwood Unit 1 experienced an automatic reactor trip due to a generator lockout relay trip and subsequent turbine trip and reactor trip. The cause of the generator lockout relay trip is unknown at this time and is under investigation. Numerous lightning strikes were present in the area during the time of the generator lockout relay trip. Both trains of auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected with the exception of failure of source range nuclear instruments to automatically re-energize following the reactor trip. Both source range nuclear instruments were manually energized in accordance with station procedures. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the 1B Diesel Generator in standby. 1A Diesel Generator is out of service for planned maintenance. All other safety systems are available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4 hr. notification, and per 10 CFR 50.72(b)(3)(iv)(A) for an automatic actuation of the Auxiliary Feedwater system, 8 hr. notification. The NRC Resident Inspector and Illinois Emergency Management Agency have been informed. | Steam Generator Auxiliary Feedwater Main Condenser Main Steam | |
| ENS 54289 | 23 September 2019 16:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Lowering Steam Generator Levels | At 1106 CDT Braidwood Unit 1 experienced an automatic reactor trip due to lowering steam generator water levels following closure of the 1B steam generator feed water regulating valve. The cause of the 1B steam generator feedwater regulating valve failing closed is unknown at this time and is under investigation. Both trains of auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected with the exception of intermediate range nuclear instrument N-36 which was identified as being undercompensated following the reactor trip. Both source range nuclear instruments were manually energized in accordance with station procedures. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in stand by and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4 hour notification, and per 10 CFR 50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8 hour notification. The NRC Resident Inspector has been informed. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Main Steam | |
| ENS 53443 | 4 June 2018 14:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip on Lowering Steam Generator Water Level | At 0920 CDT, Braidwood Unit 1 reactor was manually tripped due to lowering steam generator water levels following a trip of the 1C main feedwater pump. The cause of the 1C main feedwater pump trip is unknown at this time and is under investigation. Both trains of Braidwood Unit 1 auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8-hr. notification. All rods inserted into the core during the trip. Concerning the relief valves momentarily lifting and reseating, there is no known primary-to-secondary leakage. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Main Steam | |
| ENS 53371 | 30 April 2018 16:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Following Turbine Trip | At 1124 CDT, Braidwood Unit 1 experienced an automatic Reactor Trip. The cause of the Reactor Trip was a Turbine Trip with reactor power greater than P-8. The turbine trip was actuated as a result of a Turbine Motoring Generator Trip. The cause of the generator trip is unknown at this time and is under investigation. After the Reactor Trip occurred, the 1A Auxiliary Feedwater pump was manually started to provide feedwater flow to all four steam generators. The 1A Auxiliary Feedwater pump was subsequently secured and placed in standby when the Startup Feedwater pump was placed in service. Train A Main Control Room Ventilation Filtration system shifted to Makeup Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. The Main Steam dump valves are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the Diesel Generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr notification, and per 10 CFR 50.72(b)(3)(iv)(A) for a manual actuation of the Auxiliary Feedwater system, 8-hr notification. The licensee notified the NRC Resident Inspector and Illinois Emergency Management Agency. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Main Steam | |
| ENS 46262 | 20 September 2010 22:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 1704 CDT, Braidwood Unit 1 experienced an automatic reactor trip. The reactor trip red first out was Over Temperature Delta Temperature (OTDT). At the time of the reactor trip, the Instrument Maintenance Department was performing a calibration of Power Range Channel N-43 and a calibration of the 1C S/G Narrow Range Level Channel 1L-0538. The cause of the trip is unknown at this time. After the reactor trip occurred, all four Steam Generators reached their Low-2 reactor trip setpoint and Pressurizer pressure reached its low pressure reactor trip setpoint which is an expected response on a trip from full power. Steam Generator levels and Pressurizer pressure have been restored. Both the 1A and 1B Auxiliary Feedwater pumps auto started on the Low-2 Steam Generator levels as expected. All control rods fully inserted into the core. Train B Main Control Room Filtration system shifted to makeup mode and the Train B Fuel Handling Building ventilation shifted to Emergency Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam (was) released as a result of the reactor trip. The Main Steam Dumps are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10CFR50.72(b)(2)(iv)(B) for RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater system, 8-hr. notification. AC power is being provided by offsite power with the Diesel Generators in standby and all safety systems available. There is no Unit 2 impact. The licensee notified the NRC Resident Inspector. The licensee also anticipates that there will be a press release issued regarding this event. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 46178 | 16 August 2010 07:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trips at Both Units | Braidwood Unit 2 automatically tripped at 0206 (CST) due to a turbine generator trip due to generator lockout relay actuation. All systems responded as expected, with the auxiliary feed water pumps starting on Low-2 Steam Generator level. The Unit is stable in Mode 3, all primary systems are stable with the secondary heat sink being maintained via aux feed water and the steam dumps. Offsite power is supplying Unit 2, and both emergency diesel generators are available. Cause of generator lockout is under investigation. Braidwood Unit 1 automatically tripped at 0219 (CST) on a turbine trip caused by a loss of condenser vacuum. All systems responded as expected, with the auxiliary feed water pumps supplying steam generator levels. Secondary heat sink is steam generator PORVs. One steam generator safety valve is not fully seated. No steam generator tube leakage. Cause of the loss of vacuum is under investigation. For both Units all control rods fully inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The licensee notified the NRC resident inspector. Braidwood Unit 1's loss of condenser vacuum was caused by the loss of an electrical bus supplying the circ water pumps. At the time of this report, both plants were in a normal shutdown electrical lineup with the exception of the deenergized bus supplying power to the circ water pumps on Unit 1. The steam generator safety valve that has not fully seated was characterized as weeping a small amount of steam. The licensee is uncertain if the Unit 1 trip is related to the Unit 2 trip. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 45238 | 31 July 2009 02:08:00 | 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 | Unusual Event Declared Due to a Loss of Offsite Power for Greater than 15 Minutes | Unit 2 automatically tripped from 100% reactor power as a result of the over-current trip of the 2C Reactor Coolant Pump. Both station auxiliary transformers on Unit 2 subsequently tripped offline. All control rods fully inserted on the trip. Auxiliary feedwater auto-started and maintained Steam Generator water level. The unit is stable in Mode 3. The Emergency Diesel Generators auto started and loaded supplying both emergency busses with power. All systems functioned as required. There was no affect on Unit 1. The licensee notified the NRC Resident Inspector.
At 2059 on July 30, 2009, a reactor trip of Unit 2 at Braidwood occurred. A loss of offsite power occurred and an Unusual Event was declared at 2108. NRC Headquarter Operations was notified at 2155 (ENS call # 45238). Power from System Auxiliary Transformer (SAT) (credited offsite power supply) 242-2 was restored to buses 241 and 242 (safety related buses) at 0036 on August 2, 2009. The Unusual Event was terminated at 0036 on August 2, 2009. This call is being made due to the termination of the Unusual Event declared on July 30, 2009. An Event Summary Report is required by Exelon procedures within 24 hours of termination of the Unusual Event and will be communicated to the Headquarter Operations later today. The initial event was the result of the actuation of the SAT sudden pressure relay. When the transformer tripped, a slow automatic bus transfer resulted. When the RCPs (Reactor Coolant Pump) and condensate pumps were reenergized, they tripped on overcurrent causing the reactor trip. The sudden pressure relay has subsequently tripped during testing and may have caused the initial event. The licensee reported no damage to the plant. The licensee notified the NRC Resident Inspector. Notified the R3DO (Daley), IRD (McDermott), NRR (Howe), DHS (An), and FEMA (Biscoe).
The Event Summary Report was received and documented the following technical conclusions: The Unusual Event declaration was caused by a sudden pressure relay on SAT 242-1 causing a lockout of both SATs followed by a trip of Unit 2 due to the 2C RCP tripping during the automatic bus transfer for bus 258. This led to a loss of offsite power to Unit 2. It is currently unknown why the sudden pressure relay on SAT 242-1 actuated. Troubleshooting on the sudden pressure relay is in progress. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Daley). Notified the IRD (McDermott) and NRR (Howe) via e-mail. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 45017 | 24 April 2009 16:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Instrument Calibration | At 1141 CT, Braidwood Unit 2 experienced an automatic Reactor Trip. The Reactor Trip red first out annunciator was Over Temperature Delta Temperature (OTDT). At the time of the Reactor Trip the Instrument Maintenance Department was performing a scheduled calibration of a Pressurizer Pressure channel (2PT-456) which is in the B loop of reactor protection. During the calibration a spike occurred on the D loop of reactor protection. Specifically, the RCS (Reactor Coolant System) temperature for the D loop. This caused a Reactor Trip on a 2 of 4 coincidence. After the reactor trip occurred, all four steam generators reached their low-2 Reactor Trip setpoints and pressurizer pressure reached its low pressure Reactor Trip setpoint all of which is an expected response on a trip from full power. Steam Generator levels and Pressurizer pressure have been restored. Both the 2A and 2B Auxiliary Feedwater pumps auto started on the low-2 steam generator levels as expected. All control rods fully inserted into the core. No secondary relief valves lifted and no secondary steam released as a result of the Reactor Trip. Steam Generators are now being filled by the 2A Main Feedwater pump and the Auxiliary Feedwater pumps have been placed in standby. The main steam dumps are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, 4 hour notification, and per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater System, 8 hour notification. The electrical line up transferred to the normal shutdown configuration with the standby diesel generators and safety systems available. There is no Unit 1 impact. The licensee plans on issuing a press release and has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 44743 | 27 December 2008 20:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip as a Result of a Generator Trip | At 1418 on 12-27-08 Braidwood Unit 2 experienced an automatic Reactor Trip. The Reactor Trip red first out annunciator was Turb(ine) Trip above P8 Rx Trip. At the time of the trip the Unit Aux Transformer (UAT) 241-1 sudden pressure relay actuated causing a main generator trip which resulted in a main turbine trip which resulted in a Reactor Trip. Also at the same time as the Reactor Trip, the 2C Heater Drain Pump tripped on phase A over current. Damage was subsequently noted on the pump motor terminal box. No fire or smoke was observed at UAT 241-1 or the 2C Heater Drain Pump. After the Reactor Trip occurred, all four steam generators reached their low-2 Reactor Trip setpoints and the pressurizer reached its low pressure Reactor Trip setpoint all of which is an expected response on a trip from full power. Steam generator levels and pressurizer pressure have been restored. Both the 2A and the 2B Auxiliary Feedwater Pumps auto started on the low-2 steam generator levels as expected. All control rods fully inserted into the core. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. Steam generators are now being filled by the Startup Feedwater Pump and the Auxiliary Feedwater Pumps have been placed in standby. The main steam dumps are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10CFR 50.72(b)(2)(iv)(B) for RPS actuation, 4 hr (notification), and per 10CFR 50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater System, 8 hr (notification). The electrical line up transferred to the normal shutdown configuration with standby diesel generators and safety systems available. There was no impact on Unit 1. The licensee plans on issuing a press release and has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 43590 | 23 August 2007 20:30:00 | 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)(3)(xiii), Loss of Emergency Preparedness | Manual Reactor Trip Because of Lowering Condenser Vacuum | At 1530 hours on 8/23/07, Braidwood Station Unit 2 was manually tripped due to lowering condenser vacuum. The lowering condenser vacuum resulted from the trip of two circulating water pumps. The cause of the two circulating water pump tripping is under investigation. All control rods inserted and there were no complications during the trip and all systems functions as required. Following the unit trip, the Auxiliary Feedwater System actuated as expected to maintain steam generator level. At the time of the unit trip, the Braidwood Station area was experiencing severe thunderstorms. Additionally, at 1604 hours, 19 of 70 emergency sirens for the Braidwood Station were declared inoperable due to a loss of power from storms in the area. As of 1704 hours, 19 sirens (greater than 25%) remain inoperable. This event is considered a major loss of offsite response capability and applies to both Braidwood Station Unit 1 and Unit 2. These events are is being reported under: (1) 10 CFR 50.72(b)(2)(iv)(B) as an event that results in the actuation of the reactor protection system (RPS) when the reactor is critical, (2) 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PWR auxiliary feedwater system. (3) 10 CFR 50.72(b)(3)(xiii) as a major loss of offsite response capability. All safety buses remained powered by offsite power throughout this event. Emergency diesel generators are available if needed. No steam generator PORV's lifted as a result of the trip. Decay heat is being discharged to the condenser via the steam dumps. The licensee informed the NRC Resident Inspector. | Steam Generator Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 43449 | 27 June 2007 14:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Off-Site Power Fluctuation | Switchyard line 2001 tripped and re-energized during a thunderstorm. This caused main generator output breaker, ACB 3-4, to trip open. At this time '1D' reactor coolant pump (RCP) tripped and caused a reactor trip. Cause for the '1D' RCP trip is under investigation. The heat sink is being provided by Aux Feedwater and the use of Steam Dumps. Electrical power is being provided by offsite power. All rods inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The electrical transient had no impact on Unit 2. The licensee notified the NRC Resident Inspector.
This is a revision to a previously transmitted ENS call on 6/27/07 EN# 43449. Braidwood Unit 1 Reactor automatically tripped on a loss of '1D' RCP greater than P-8 setpoint. The cause of the RCP trip was an electrical disturbance during a thunderstorm. The reactor trip automatically caused a main turbine and generator trip. Auxiliary feedwater system automatically started on the low-2 S/G water level that is expected from a full power reactor trip. Auxiliary feedwater and main steam dumps are providing a heat sink. A switchyard line also tripped during this electrical transient causing multiple switchyard breakers to open. Electrical power is being provided by offsite power. All control rods fully inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The electrical disturbance had no impact on Unit 2. The licensee notified the NRC resident inspector." R3DO (Louden) notified. | Feedwater Auxiliary Feedwater Main Turbine Control Rod Main Steam | |
| ENS 41535 | 28 March 2005 18:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Generator Protection Circuitry | The licensee faxed the following: Unit 2 reactor trip due to generator protection circuitry. Auxiliary feedwater actuated as expected. There were no additional malfunctions or unexpected plant response. The cause of the generator protection circuitry induced trip is still under investigation. This is a 4 hour notification of an RPS actuation per 10CFR 50.72(b)(2)(iv)(B). The 8 hour notification of an auxiliary feedwater system actuation per 10CFR 50.72(b)(3)(iv)(A) is being made under the same telephone call. See Event 41534. The licensee notified the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 41280 | 22 December 2004 19:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation Due to Steam Generator Low Level Signal | Unit 2 reactor trip due to 2C steam generator LO-2 reactor protection signal. Auxiliary Feed Water actuated as expected. No additional malfunctions or unexpected plant response. Cause of LO-2 steam generator reactor protection signal under investigation. This is a 4 hour notification of an RPS actuation per 10 CFR 50.72(b)(2)(iv)(B). The 8 hour notification of Auxiliary Feed Water system actuation per 10 CFR 50.72 (b)(3)(iv)(B) is being made under this same telephone call. The licensee notified the NRC Resident Inspector. All control rods fully inserted. Decay heat is being removed to the main condenser via the turbine by-pass valves. The electrical grid is stable.
The licensee has determined that the RPS and Auxiliary Feed Water actuations were the result of an actual low level in the 2C steam generator. The cause of the low level is under investigation. The NRC Resident Inspector was informed. Notified R3 DO (L. Kozak). | Steam Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 40370 | 3 December 2003 09:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Plant Had an Auto Reactor Trip from 100% Power Due to Steam Generator Low Level | The "2 D" steam generator Lo-2 level was caused by the loss of the "2C" feedwater pump while performing the "2 BWOS" feedwater weekly surveillance of the HP stop valve. Both trains of the aux feed actuated as expected on the "2D" Lo-2 s/g level signal. The plant is currently in mode 3 with all rods fully inserted. No ECCS or safety relief valves actuated. Licensee notified the NRC Resident Inspector | Steam Generator Feedwater |