Semantic search
Start date | Reporting criterion | Title | Event description | System | LER | |
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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 53358 | 22 April 2018 21:46:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Undervoltage Actuation of the Engineered Safety Feature Bus | On Sunday, April 22, 2018 at 1646 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned 1A Diesel Generator (DG) Emergency Core Cooling System (ECCS) Actuation Surveillance, initiating the 1A DG to emergency start and sequence loads on a safety injection signal. Following the 1A DG solely supplying electrical power to Bus 141, the 1A DG lost voltage, resulting in an unplanned UV actuation of ESF Bus 141. The 1A DG output breaker was manually opened and local emergency stop of the 1A DG was attempted. The 1A DG continued to run at idle. Fuel supply was secured to the 1A DG and the engine stopped. Subsequently, operators restored power to ESF Bus 141 from the Unit 1 Offsite Power Source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. The licensee notified the NRC resident inspector. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal Emergency Core Cooling System | |
ENS 53347 | 19 April 2018 16:52:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Undervoltage Actuation of the Engineered Safety Feature Bus | On Thursday, April 19, 2018 at 1152 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned Bus 141 Undervoltage Actuation Surveillance, initiating the 1A Emergency Diesel Generator (EDG) to emergency start and sequence loads on the UV signal. Following the 1A EDG solely supplying electrical power to Bus 141, the EDG lost voltage resulting in an unplanned UV actuation of the ESF Bus 141. Subsequently, operators restored power to ESF Bus 141 via crosstie of the Unit 2 offsite power source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal | |
ENS 51450 | 5 October 2015 06:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuations | Braidwood Unit 2 was performing a planned plant shutdown for refueling outage A2R18. In accordance with plant shutdown procedures while in Mode 1 (Power Operations) at approximately 15% power, operators attempted to start the Start Up Feedwater (SFWP) pump and the pump immediately tripped on Phase A Overcurrent. The 2A Motor Driven Feedwater pump (MDFWP) was manually started to maintain Steam Generator Water Level during the shutdown and subsequent plant cooldown. While in Mode 3 (Hot Standby) at (550 Degree-F), the 2A MDFWP was manually secured due to pump inboard journal bearing temperature exceeding its (200 Degree-F) operating limit. At 0105 (CDT) an anticipated automatic Auxiliary Feedwater actuation signal was generated on low Steam Generator level (36.3%) and both the 2A and 2B Auxiliary Feedwater pumps (AFP) auto-started. Also at 0105 (CDT) a Reactor Protection System (RPS) Reactor trip signal was received due to low Steam Generator level (36.3%) with the reactor not critical. Both Auxiliary Feedwater trains operated as designed with the Main Steam Dumps in service and the Main Condenser providing the heat sink. All systems operated as designed with the exception of the SFWP and the MDFWP described above. The plant is currently stable in Mode 5 with both AFPs secured. This report is being made per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the (1) RPS Reactor Trip with the reactor not critical and (6) Auxiliary Feedwater System, 8 hour notification. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser Main Steam | 05000457/LER-2015-002 |
ENS 50301 | 24 July 2014 00:30:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Based on Shots Fired in the Owner Controlled Area | At 1943 CDT on 7/23/14, Braidwood Station declared Unusual Event HU1 due to gunshots fired within the Owner Controlled area (OCA). This was a security condition without a Hostile Action. Both Units remain in Mode 1, 100% reactor power throughout the event. Local Law enforcement was contacted and investigated. Security stood down from the Security Condition at 1956 CDT. The Control Room was informed at 1930 CDT by Security of fourteen (14) gunshots heard inside the OCA. The licensee informed State and local agencies and the NRC Resident Inspector. Notified other FEDS (FEMA Ops Center, DHS NICC Watch Officer, DHS SWO) and (Nuclear SSA, FEMA NWC) via email.
Braidwood Station Terminated an Unusual Event, HU1 at 2134 CDT. Local Law Enforcement Agency called an 'AII Clear' and station security restored the normal security posture at 1956 CDT. The licensee informed State and local agencies and the NRC Resident Inspector. Notified R3DO (Orth), NRR (Lund) and IRD (Gott) via email. Notified other FEDS (FEMA Ops Center, DHS NICC Watch Officer, DHS SWO, Nuclear SSA, FEMA NWC) via email. | ||
ENS 46936 | 8 June 2011 15:26:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to Flooding That Potentially Affected Safety Related Equipment | Flushing activities were in progress on the 2A Auxiliary Feedwater (AF) Pump suction line from the Essential Service Water System (SX). At 1011 (CDT), during this flushing activity, the flushing hose ruptured and caused flooding in the Auxiliary Building that had the potential to affect safety related equipment needed for the current operating mode. This was due to the flood waters contacting the motors of the 1A and 2A AF Pumps. At 1026, an Unusual Event was declared by the Shift Emergency Director because the conditions for EAL entry were met for EAL HU5. Specifically the EAL conditions were 'Flooding in the Auxiliary Building that has the potential to affect safety related equipment needed for the current operating mode.' Auxiliary Feedwater is required to be operable in Mode 1 for each unit. The leak was immediately isolated (within 45 seconds of hose rupture) and remains isolated. Maintenance personnel are in the process of testing the 1A and 2A AF Pump motor to determine operability. The appropriate Tech Spec Action Statements have been entered on each Unit for the AF Pumps. The leak also caused the wetting of MCC (Motor Control Center) 132X1, which feeds the 1B AF Pump control power. The Unit 0 (Common) Component Cooling Pump feed breaker cubicle was also wetted. Investigation into these components are also in progress. The licensee has entered Technical Specification LCO 3.7.5 Condition A which requires restoration of the one of the AF pumps within 72 hrs. There were no personnel injuries. The licensee has notified the NRC Resident Inspector.
At 1122 (CDT) on 6/8/11, the Unusual Event was terminated due to the flooding conditions no longer existing. Evaluation of wetted components are still in progress. The NRC Resident Inspector has been notified of event. Notified IRD (Gott), R3DO (Lara), NRR EO (Thorp), FEMA (Blankenship), and DHS (Flinter). | Service water Auxiliary Feedwater | |
ENS 46694 | 24 March 2011 15:18:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unexpected Loss of Annuciators During Planned Maintenance | During a planned maintenance activity on the Unit 2 main control room alarm cabinets, it was identified that all Unit 2 safety system annunciators were lost. This was identified at 1006 (CDT). Main Control Board indicators remained functional. At 1018 (CDT), the Shift Manager declared an Unusual Event under Emergency Action Level MU6. This was due to an unplanned loss of most (approximately 75%) safety system annunciators for > 15 minutes. The planned maintenance activity was not expected to affect the amount of annunciators that were lost. At 1030 (CST), the (planned maintenance) clearance order was cleared and power was restored to the Unit 2 annunciators. There was not transient on other plant equipment and the plant remained stable before and after this event. The cause for the unexpected loss of annunciators is not clearly understood and is still under investigation. The licensee notified the NRC Resident Inspector.
The Unusual Event was terminated at 1047 CDT on 03/24/11. All annunciators have been restored and an investigation will be conducted to determine the cause. Notified R3DO (Cameron) | ||
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 45618 | 10 January 2010 01:51:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Declared Due to a Fire in the Auxiliary Building Lasting Greater than 15 Minutes | At 1925 (hrs. CST), the Main Control Room received a notification of smoke in the Auxiliary Building Ventilation Supply Plenum. Fire Brigade and the Incident Commander were dispatched. Once (they) arrived at the scene, they noted smoke in the area of the 0C VA Supply Fan and requested the fan be shutdown. The 0C VA supply fan was shutdown at 1933 hrs. The operators reported smoke and a small fire coming from the inboard bearing of the 0C VA supply fan. A CO2 fire extinguisher was used to put the fire out and cool the bearing. A total of 2 extinguishers were used. The fire was declared out at 1941 hrs. Both Units remained stable and at full power during the entire event. There were no injuries, and no off site assistance was required. The licensee notified state and local authorities and the NRC Resident Inspector.
At 2111 CST (on) 1-9-2010, the Unusual Event was terminated. The fire is out. Both Units are stable. A fire watch is established. There are no signs of reflash. (Braidwood) no longer (meets the) Unusual Event action level threshold. The licensee has notified the NRC Resident Inspector. Notified IRD (Grant), R3DO (Stone), NRR EO (Giiter), DHS (Inzler) and FEMA (Casto). | ||
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 41534 | 28 March 2005 19:59:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Hydrogen Leak | The licensee faxed the following information: Following the Unit 2 trip, (due to a malfunction of the generator protection circuitry), a hydrogen leak was identified on the Unit 2 main generator. The leakage was sufficient enough to cause a flammable gas release affecting normal plant operation. The Unusual Event was declared under HU6 - Hazards and Other Conditions. The State and local authorities were notified at 14:09 CST. There is NO fire, it is a hydrogen release only. The licensee said the hydrogen totalizer, which indicated a flow rate of 100 cfm, may be a probable area of leakage. The hydrogen leaked directly into the turbine building. Air samples indicated personnel breathing apparatus was not required. The licensee expects to exit from the UE once the hydrogen system was purged with CO2. Estimated timeframe is 3-6 hours. The licensee notified the NRC Resident Inspector.
The licensee terminated their Unusual Event at 16:40 CST, and the plant status is 0% power / Mode 3. The hydrogen leak was determined to be from a bushing on the main generator. The hydrogen leakage is believed to have lasted only 5 minutes. The licensee notified the State and NRC Resident Inspector. Notified R3DO (Riemer), IRD Manager (Crlenjak), NRR EO (Berkow), FEMA and DHS.
Received Event Summary Report from Braidwood station. Corrected Unusual Event termination time from 16:40 to 16:23 CST. Notified R3DO (Riemer) and NRR EO (Brenner) | ||
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 | |
ENS 40298 | 5 November 2003 03:32:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Braidwood 2 Afw Support System Actuation During Outage | Unit 2 auxiliary feedwater support systems actuated during scheduled ATWS testing when an unrelated clearance order placement de-energized two 6.9 kv busses (256 and 258). The 2 of 4 6.9 kv bus undervoltage coincidence initiated a valid auto-start signal causing lube oil pumps 2AF01PA-A, 2AF01PB-A and 2AF01PB-C to start. Auxiliary Feedwater Pump AOV discharge valves 2AF004A and 2AF004B auto opened. 2AF01PA 4kv breaker, which was in the equipment test position. Neither auxiliary feedwater pump started and no water transferred to the steam generators. The licensee notified the NRC resident inspector. | Steam Generator Auxiliary Feedwater |