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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56971 | 16 February 2024 11:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Automatic Actuation of Auxiliary Feedwater System | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 56852 | 14 November 2023 16:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip | 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 56329 | 1 February 2023 15:56:00 | 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 and Automatic Actuation of Auxiliary Feedwater System | The following information was provided by the licensee via email: At 0956 CST with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The trip was not complex with all safety related systems responding normally post-trip. During the trip, the non safety related '1A' 4160V bus lost power resulting in the loss of one Reactor Coolant Pump (RCP-1A). Operations responded and stabilized the plant. The '1A' 4160V bus was re-energized at 1031 CST. Decay heat is being removed by steam dumps to the main condenser. Farley Unit 2 is not affected. An automatic actuation of (Auxiliary Feedwater) AFW also occurred, which is an expected response from the reactor 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 report 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. | Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56028 | 3 August 2022 17:58:00 | 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 Loss of Power to START-UP Transformer | The following information was provided by the licensee via email: At 1258 CDT on August 3, 2022, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to the supply breakers of the 1B startup transformer opening. The fast dead bus transfer for the reactor coolant pumps did not occur during the event. Currently the plant is in Mode 3 on natural circulation. Operations responded and stabilized the plant. Decay heat is being removed by steaming with atmospheric relief valves. Unit 2 is not affected. An automatic actuation of the 1B diesel occurred because of the power loss to the 1G 4160V bus. Additionally, the actuation of motor driven and turbine driven auxiliary feedwater pumps (AFW) also occurred. AFW auto-start is an expected response from this reactor 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). 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 1B diesel and 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. | Reactor Protection System Auxiliary Feedwater | |
| ENS 54287 | 21 September 2019 13:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Trip Due to Reactor Coolant Pump Vibration Alarm | At 0800 (CDT), with Unit 2 in Mode 1 at 100 percent (power), the reactor was manually tripped due to elevated vibration indication on the 2C reactor coolant pump exceeding annunciator response procedure trip criteria. The trip was not complex, with all systems responding normally post trip. Auxiliary Feedwater (AFW) auto actuated as expected following the manual reactor trip. Operations responded and stabilized the plant. Decay heat is being removed via the use of AFW and subsequent steaming of the steam generators to the main condenser. 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). In addition, this event report is being reported as an eight-hour non-emergency notification per 10 CFR50.72(B)(3)(iv)(A) for a 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. Farley reported that there was no increase in containment unidentified leakage or fluctuations with RCP seal flow during this event. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 54040 | 1 May 2019 21:43:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | En Revision Imported Date 5/9/2019 | EN Revision Text: MANUAL REACTOR TRIP DUE TO MISALIGNED CONTROL ROD At 1643 (CDT), with Unit 2 in Mode 2 during low power physics testing, the reactor was manually tripped per procedure due to a misaligned control rod. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the atmosphere using the atmospheric relief valves. Unit 1 is not affected. Due to the Reactor Protection System actuation, 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.
This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). The reactor was tripped during low power physics testing. The misaligned rod was encountered during rod group insertion and the affected bank had been inserted to the extent that the reactor was subcritical when the operators tripped the reactor. The licensee notified the NRC Resident Inspector. Notified R2DO (Lopez) | Reactor Protection System Control Rod | |
| ENS 52395 | 27 November 2016 06:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip with Automatic Auxiliary Feedwater System Actuation Due to Voltage Oscillations | At 0026 (CST) on November 27, 2016, Farley Unit 1 was manually tripped from 100% reactor power due to voltage swings suspected to be caused by the Auto Voltage Regulator. All control rods fully inserted and Auxiliary Feedwater (AFW) auto-started as expected. All systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The cause of the main generator voltage oscillations is under investigation. The NRC Resident Inspector has been notified. The trip was uncomplicated. Decay heat is being removed via the steam dumps to condenser. The plant is at normal operating pressure and temperature with auxiliary feedwater supplying the steam generators. The electrical grid is stable and supplying plant loads. All safety equipment is available, if needed. Unit 2 was unaffected by the event and remains at 100% power. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 52356 | 8 November 2016 06:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Steam Generator Reduced Feedflow | At 1331 (CST) on November 8, 2016, Farley Nuclear Plant Unit 1 manually tripped from 32% reactor power. The plant was ramping down to remove the main generator from service due to an unrelated issue. 1A SGFP did not respond to control Steam Generator (SG) level as expected when the miniflow was opened per procedure. SG levels lowered due to lower feed flow and the reactor was manually tripped in accordance with plant procedures. All control rods fully inserted and Auxiliary feedwater (AFW) auto started as expected. The Main Steam Isolation Valves were closed to minimize the cool-down. Decay heat is being removed through the Atmospheric Relief Valves. All other systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The failure of the 1A SGFP control is under investigation. Unit 2 was not affected. The NRC Resident Inspector has been notified. There is no primary to secondary leakage. | Steam Generator Main Steam Isolation Valve Auxiliary Feedwater Control Rod | |
| ENS 52274 | 1 October 2016 10:12:00 | 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)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip Due to Inadvertent Closure of Main Steam Isolation Valve | At 0512 (CDT) on October 1, 2016, Farley Nuclear Plant Unit 1 automatically tripped from 99 percent reactor power due to the inadvertent closure of a main steam isolation valve (MSIV). The closure of the MSIV caused a turbine trip resulting in an automatic reactor trip. Concurrent with the reactor trip, a safety injection (SI) occurred. The plant is stable in Mode 3 (Hot Standby) and auxiliary feedwater (AFW) autostarted as expected. The cause of MSIV closure and SI actuation is under investigation. Cooldown will continue to Mode 5 (Cold Shutdown) as planned for entry into a scheduled refueling outage. Restart is not planned until the completion of the refueling outage. Unit 2 was not affected. The NRC Resident Inspector has been notified. The MSIVs are open with the steam generators discharging steam to the main condenser using the turbine bypass valves. SI was from high head injection which has been secured. | Steam Generator Main Steam Isolation Valve Auxiliary Feedwater Main Condenser | |
| ENS 51918 | 11 May 2016 11:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Farley Unit 2 Trip on Hi Hi Steam Generator Level | At 0653 (CDT) on 5/11/16, Farley Unit 2 reactor was manually tripped from 29 (percent) power. The initiating event was hi-hi Steam Generator level. Steam Generator levels began to rise following the start of a second condensate pump. The hi-hi steam generator level setpoint was reached causing the only running main feedwater pump to trip, a main feedwater isolation, and an automatic turbine trip. Auxiliary feedwater automatically started as expected. The reactor was manually tripped per procedure. All other systems responded properly for the event and there were no complications. The plant is currently stable in Mode 3. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 51043 | 5 May 2015 09:22:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater Pump Auto Start Signal Initiated During Startup | At 0422 CDT on 5/5/2015, with Farley Nuclear Plant Unit 1 in Mode 2, and the 1A Steam Generator Feedwater Pump (SGFP) in the tripped condition, the 1B SGFP was manually tripped during troubleshooting. The trip of the second SGFP initiated the auto start signal for the MDAFWPs (motor driven aux feedwater pumps) due to the auto start signal not being defeated. Both MDAFW pumps were in service supplying AFW to the steam generators (SG) when the actuation signal was received. The effects of the auto start signal were to fully open the AFW Flow Control Valves and isolate SG blowdown and SG blowdown sample valves. These actions occurred successfully and the auto start signal was reset. There was no adverse impact to the plant and decay heat continued to be removed through the condenser throughout the event. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 50618 | 15 November 2014 09:53:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of the Reactor Protection System for Abnormal Rod Position Indication | This notification is being made as required by 10CFR 50.72(b)(3)(iv)(A) due to control room operators opening the Farley Nuclear Plant Unit 2 reactor trip breakers via the main control board hand switch during reactor startup procedures. This is a valid actuation of the reactor protection system. The reactor was not critical at the time the reactor trip breakers were opened. While withdrawing control rods for reactor startup and low power physics testing with control bank C at approximately 50 steps, Digital Rod Position indication for one of the control rods (M12) changed to 90 steps. The control room operators stopped withdrawing rods and entered FNP-2-AOP-19 (Malfunction of the Rod Control System). The reactor trip breakers were opened at 0353 CST. All rods inserted as expected. All other systems functioned normally. The plant is stable at normal operating pressure and temperature. The NRC Resident Inspector has been notified. | Reactor Protection System Control Rod | |
| ENS 50533 | 14 October 2014 08:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of a Start-Up Transformer | This notification is being made as required by 10 CFR 50.72(b)(2)(iv)(B) due to a Farley Nuclear Plant Unit 2 manual reactor trip. The trip was initiated when the in service train of CCW cooling to the Reactor Coolant Pumps was lost due to a loss of the 2B Start Up Transformer (SUT). The control room team manually tripped the reactor then tripped all three reactor coolant pumps as required by station procedure. There was a line of severe thunderstorms with lightning passing through the plant site at the time of loss of the 2B Start Up Transformer. The 2B emergency diesel generator was out of service for maintenance therefore there was a loss of the 'B' train emergency power 4160V electrical bus ('B' train LOSP (Loss of Offsite Power)). 'A' train emergency power remained energized from offsite sources. The plant is stable at normal operating pressure and temperature. At 0433 (CDT), 2B Reactor Coolant Pump was re-started when support conditions were re-established. Heat sink is adequate using the 2A Motor Driven Auxiliary Feedwater Pump. Unit 2 'B' train power was restored by starting the 2C emergency diesel generator at 0523 (CDT). This restored power to the Digital Rod Position Indication system, and control rod K-8 in control bank 'C' indicated full out, and all other control rods fully inserted. An emergency boration is in progress to compensate for the stuck rod. Additionally, the reactor trip resulted in a valid actuation of the Aux Feedwater system which is an eight hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A). During the transient, one primary PORV momentarily opened, then reseated. Decay heat is being directed to the atmospheric relief valves with no indicated primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector.
Digital rod position indication troubleshooting was conducted on 10/14/2014 and confirmed all control rods, including control rod K-8, fully inserted following the reactor trip. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ayres), NRR EO (Davis) and IRD (Gott). | Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 49106 | 12 June 2013 02:05:00 | 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 | Unit 1 Automatic Reactor Trip Due to the Loss of a Start-Up Transformer | This is a report of an automatic RPS actuation and automatic ESF actuation per 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). Additionally, this is to report intentions for a press release per 10CFR50.72(b)(2)(xi). At 2105 CDT on 6/11/13, Farley Unit 1 experienced an automatic reactor trip from 100% power. The initiating event was the loss of the 1B Start up Transformer which resulted in de-energization of the B-Train ESF 4KV buses and the 1B and 1C Reactor Coolant Pump Buses. The 1B Emergency Diesel Generator auto started and tied to the B-Train 4KV Emergency buses. Both MDAFW (Motor Driven Auxiliary Feedwater) Pumps and the TDAFW (Turbine Driven Auxiliary Feedwater) Pump auto-started and are supplying AFW flow to the steam generators. Decay heat removal is via the steam dumps to the main condenser. The cause of the loss of the 1B Start-up Transformer is unknown and is currently under investigation. All other systems functioned as expected in response to the loss of the 1B Start-up Transformer and reactor trip. The NRC Senior Resident Inspector has been notified. A press release is planned. All control rods fully inserted. There is no impact on Unit 2. Currently the licensee does not plan to restart the 1B and 1C Reactor Coolant Pumps. Pressurizer spray has been isolated from the 1B loop per procedure. Main Condenser vacuum is adequate for decay heat removal. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 47813 | 6 April 2012 19:44:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Load Shed Signal Occurred Due to Loss of the 1B Startup Transformer | At 1444 (CDT) on April 6, 2012, during a planned refueling outage on Unit 1, maintenance activities in the high voltage switchyard caused feeder breaker 820 to inadvertently trip. With the second feeder breaker, 924, already out of service, power was lost to the 1B startup transformer. An undervoltage condition was then experienced on the 1G 4160 V emergency bus. As a result, the B1G Sequencer initiated a valid load shed of the 1G 4160 V emergency bus. Due to outage conditions, the B-Train, 1B Emergency Diesel Generator (EDG) was tagged out and did not automatically start but did receive a valid start signal. None of the ESF loads supplied by the 1G bus started automatically since the 1B EDG was out of service. With a B-Train equipment outage in progress, the 1A RHR pump (A-Train) remained in service for shutdown cooling throughout the event. Although the bus safety function was not needed for plant conditions a valid load shed signal occurred and therefore this event is considered reportable. The 1G 4160 V emergency bus was restored to service at 1542 on April 6, 2012. Investigation revealed a technical inaccuracy in the instructions used during the maintenance activity in the high voltage switchyard that caused feeder breaker 820 to trip. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Shutdown Cooling | |
| ENS 47809 | 5 April 2012 17:20:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Systems Actuations While Performing a Surveillance Test | Unit 1 was performing a planned refueling outage surveillance test, FNP-1-STP-40.0, 'Safety Injection with Loss of Off-Site Power (LOSP).' The systems were being returned to normal following the actuation portion of the test. When the B1F Sequencer Test Trip Override Switch was taken to the 'ON' position, the 1-2A Diesel Generator output breaker opened, which caused a loss of power to the 'A' Train 4 kV busses. Prior to the event, the 1-2A Diesel Generator was running at normal speed and voltage carrying the 'A' Train 4kV busses. When the diesel generator output breaker opened, it then reclosed upon receipt of the LOSP signal causing the LOSP sequencer loads to automatically start. This included the 1C Component Cooling Water Pump, the 1A High Head Safety Injection Pump (discharge isolation was closed prior to the event), and the 1A and 1B SW pumps. Therefore, during the test, the system actuated in a way that was not part of the planned surveillance testing. The 1A RHR pump was in shutdown cooling mode at the time of the event and was load shed. RHR was restarted manually by the operating crew approximately 1 minute later (no auto start (signal) present due to a loss of site power - LOSP signal without a safety injection signal present). The investigation revealed that a step in the procedure sequence was not performed during the restoration portion of the test. The operator did not parallel the diesel with off-site power prior to operating the B1F Sequencer Test Trip Override Switch which opened the diesel output breaker without off-site power aligned to the 'A' Train 4kV busses. The 1-2A Diesel Generator was subsequently paralleled to the grid and properly shutdown per the test procedure restoration. The licensee notified the NRC Resident Inspector. | Shutdown Cooling | 05000348/LER-2012-003 |
| ENS 46661 | 7 March 2011 07:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertant Auto-Start Signal to the 1B Diesel Generator During Testing | Farley Unit One was conducting FNP-1-STP-80.8, (test procedure for) 1B DG (Diesel Generator) 1000 KW load rejection. After successfully completing the load rejection portion of the procedure, the control room staff was restoring the 1B diesel to a normal auto start alignment. With the 1B diesel running, the plant operator was required to reset the 1B DG loading sequencer. He incorrectly pressed the Emergency Start reset push-button instead of the Sequencer reset push-button. As a result, the Emergency Diesel generator stop light illuminated for a brief few seconds and then extinguished. Subsequently due to the test configuration, the 1B diesel received an auto-start signal and returned to the running condition prior to the Emergency Start reset. Although further investigation is continuing, this report is being made due to an apparent valid actuation of ESF equipment. This event had no impact on other equipment or the plant electrical alignment. The Sequencer reset push-button and the Emergency Start reset push-button are not in close proximity to each other. The plant operator was assessed for fatigue and it was determined that fatigue was not a factor. The plant operator was removed from duties pending remedial training and assessment. The licensee has notified the NRC Resident Inspector.
The 8-hour non-emergency report (EN #46661) per 10CFR50.72(b)(3)(iv)(A) was conservatively reported based on the potential for a valid actuation of an Emergency Diesel Generator (EDG) during a 1000 KW load rejection surveillance test. During the restoration phase of the load rejection test to align the 1B EDG to a normal shutdown configuration, a plant operator incorrectly pressed the Emergency Start Reset (ESR) push-button instead of the Sequencer Reset push-button. As a result, the 1B EDG stop light illuminated momentarily and then extinguished. The 1B EDG received a momentary shutdown signal, but remained in a running condition. Upon completion of the 1B EDG circuit analysis, It was determined that the 1B EDG did not receive a valid actuation of the EDG safety function. Depressing the ESR push-button caused the emergency start relays to deenergize and remain de-energized. The emergency start relays energize on receipt of valid signals in response to actual plant conditions or parameters satisfying the requirements for the initiation of the safety function of the EDG. Therefore per section 3.2.6 of NUREG-1022, the 1B EDG did not receive a valid actuation signal. The NRC Resident Inspector has been notified. Notified R2DO(Musser) | Emergency Diesel Generator | |
| ENS 45946 | 22 May 2010 21:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Steam Generator Water Level Control | Unit 2 was operating at 100% power in the normal operating procedure, FNP-2-UOP-3.1, Power Operation, when multiple alarms were received associated with 2C Steam Generator (S/G) level, and a process cabinet failure. The control room team noticed there was no power or control capability on the 2C S/G Feedwater Regulating Control Valve (FRV), and 2C S/G level was decreasing. The control room team attempted to take manual control of the 2C FRV, which did not respond. The reactor was manually tripped when 2C S/G narrow range level reached 40%. The automatic trip set point for S/G level is 28%. All systems responded properly for the reactor trip and there were no complications. The investigation indicates there was an Nuclear Controller Driver (NCD) card failure in Process Control Cabinet 8. The controller card controls the 2C S/G FRV controller, which prevented any automatic, or manual control of the 2C S/G FRV, or 2C S/G level. There were no safety or relief valves that lifted and decay heat is being removed via steam dump control valves. Auxiliary feedwater pumps are maintaining level in the steam generators. Electrical lineup is normal. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | 05000364/LER-2010-002 |
| ENS 45889 | 1 May 2010 02:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent System Actuation Occurred During a Surveillance Test | Unit 2 was performing a planned refueling outage surveillance test, FNP-2-STP-40.0, Safety Injection with Loss of Off-Site Power. The system was being returned to normal following the actuation portion of the test. When the B2F Sequencer was reset, a loss of off-site power (LOSP) occurred, which caused a loss of the 'A' Train 4kV busses, and an LOSP signal was generated. The 1-2A Diesel Generator was already running at normal speed, and voltage. Therefore, the diesel generator output breaker opened, and then reclosed which then allowed the LOSP loads to automatically start. This included the 2A Motor Driven Auxiliary Feedwater Pump, and the 2A High Head Safety Injection Pump. Therefore, during the test, the system actuated in a way that was not part of the planned evolution. The investigation indicated that a recent design change on the diesel generator output breaker circuitry had not been fully incorporated into the test procedure. The test procedure currently in progress was revised to provide guidance for operating the B2G Sequencer Test Trip Override switch. The restoration section for the 'B' Train was completed with no further complications when the B2G Sequencer Test Trip Override switch was operated before resetting the B2G Sequencer. The licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater | 05000364/LER-2010-001 |
| ENS 44855 | 16 February 2009 22:55:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent Relay Actuation Results in Auto Start of 1C Emergency Diesel Generator | No release of radiation has occurred as a result of this event. At 1655 CST, the Unit One 'H' 4160V bus phase two differential relay actuated when the relay panel was inadvertently bumped. This resulted in the de-energization and lockout of the 1H 4160V bus and the auto start of the 1C Emergency Diesel Generator (EDG). Recovery actions have been successful in restoring power to the affected switchgear. All equipment functioned as expected. Normal equipment lineup has been restored. Unit One remained at full power during this event. The 1C EDG never picked up any loads because of the lockout on the 1H bus. Some loads were lost including an air compressor and some oil pumps but this equipment was either re-started or backups were started within about two hours. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | 05000348/LER-2009-001 |
| ENS 44666 | 19 November 2008 10:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On 11/19/08 at 0425 CST was operating at 100 power when the reactor tripped with no complications. All safety systems operated properly with the plant in Hot Standby. The cause of the reactor trip is under investigation. There is no radioactive release from the site. Uncomplicated trip event. Farley Unit 1 is shutdown with all rods in. No relief valves lifted. All other safety related systems are operable. There are no EDG's running. Reactor Pressure is 2247 psig; reactor temperature is 551 degrees Fahrenheit. Decay heat path is via turbine bypass valves to the condenser. AFW is feeding the S/G's. No ECCS systems injected. There is no affect on Unit 2. Licensee is investigating a potential fault in the switchyard. The licensee will notify the NRC Resident Inspector. | ||
| ENS 43687 | 3 October 2007 19:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip on Loss of Forced Flow to Rcs Loop "a | An emergency has not been declared and there has not been a release of radioactivity to the environment. Farley Unit 2 tripped from 100% power due to a loss of forced flow in the 'A' Reactor Coolant System loop. The loss of flow resulted from de-energization of the 2A 4160V Reactor Coolant Pump Bus that provides power to the 2A Reactor Coolant Pump following the de-energization of the 2B Start Up Transformer. The loss of the 2B Start Up Transformer appears to have been the result of relay testing on Farley Unit 1. Farley Unit 1 is currently in Mode 6 for a refueling outage. The loss of the 2B Start Up Transformer also de-energized the 2G 4160V Engineered Safeguards Feature Bus. This bus has been re-energized from the 2B Emergency Diesel Generator. All rods fully inserted and all systems functioned as required following the trip. Decay heat removal is being provided by both Motor Driven Auxiliary Feedwater Pumps supplying flow to all three Steam Generators. Cause of event is under investigation. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal | 05000529/LER-2007-003 |
| ENS 41175 | 5 November 2004 15:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent Loss of "B" Train Site Power During Testing | At 09:11 a.m. CST on November 5, 2004, an inadvertent "B" Train Loss of Site Power occurred when the normal power supply breaker to the 1G 4160V bus opened during testing. The 1B Emergency Diesel Generator started and powered up the 1G 4160V Bus as required and the Loss of Site Power Sequencer started all "B" Train LOSP Loads as required. Abnormal Operating Procedures for the LOSP and Loss of Train "A" or "B" Train RHR were completed as required for the event. The "A" Train RHR pump was manually started to provide core cooling and circulation flow as required for Mode 6. Core alterations were in progress for reload which were suspended immediately upon the event Systems which did not function as required: Upon the auto start signal for the 1B Motor Driven Auxiliary Feedwater Pump, the 4160V power supply breaker tripped open. The flag on the breaker indicated it was due to a time delay over current on one of the phases. Investigation into this problem is in progress. Spent Fuel Pool cooling was not lost because it was being powered from "A" Train. Site power was restored to "B" Train 4160V bus at 0959 CST. An event investigation is being performed by the licensee. The NRC Resident Inspector has been informed of this event by the licensee. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 40667 | 12 April 2004 08:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 2 Experienced a Reactor Trip During Startup Due to an Invalid Sr Trip Signal | Unit 2 reactor tripped during low power physics testing. Trip appeared to be from an invalid source range (SR) trip signal in one train of solid state protection system. All systems responded properly. The Unit is currently stable in mode 3. The licensee informed the NRC Resident Inspector. | ||
| ENS 40666 | 11 April 2004 16:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip for Unknown Reasons During Plant Startup | At 1247 EDT on 04/11/04, the licensee reported that at 1105 CDT on 04/11/04, control room operators were performing low power physics testing in accordance with FNP-2-STP-101 during startup of Unit 2 following a refueling outage. With reactor power at 10E-8 amps in the intermediate range in Mode 2, the 'B' reactor trip breaker opened for unknown reasons. All control rods inserted completely. The licensee is investigating the cause. The licensee notified the NRC Resident Inspector. | Control Rod | |
| ENS 40558 | 1 March 2004 11:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Experienced a Reactor Trip on Turbine Trip Due to a Feedwater System Problem | Reactor trip Unit 1 as a result of Turbine Trip from Hi-Hi Steam Generator level '1C' Steam Generator. Steam Generator Feed pump suction pressure initially dropped, deviation alarms were received on 'A' and 'C' Steam Generator levels. Third condensate (pump) was started, increasing feed to S/Gs. '1C' S/G went high and caused turbine trip/reactor trip. Autostart of 'A' and 'B' motor driven feed pumps occurred following the trip. All rods fully inserted following the reactor trip. Both motor-driven auxiliary feedwater pumps are currently supplying the steam generators with the steam dump system in-service to remove decay heat via the main condenser. Offsite power is stable with the EDGs in standby, if needed. All systems functioned as required. The licensee is conducting an investigation to determine the root cause. The licensee will inform the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 40309 | 10 November 2003 16:36:00 | 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 False Rcp Breaker Open Signal | Reactor Protection System actuation in response to an indicated (not an actual) 2A RCP (Reactor Coolant Pump) breaker open position signal. All reactor protection and support systems operated as expected. The Aux Feedwater System started as required in response to the tripping of both Steam Generator Feed Pumps. All 3 RCPs are running; none have tripped. Not understood is the indication of the 2A RCP breaker open when the breaker has remained closed. The licensee reported that all control rods fully inserted; decay heat is being rejected to the condenser via the steam dumps; a steam generator atmospheric relief may have momentarily lifted during the transient; and that the electrical grid is stable. The licensee will be notifying the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Control Rod |