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| Entered date | Site | Region | Scram | Reactor type | Event description | |
|---|---|---|---|---|---|---|
| ENS 56971 | 16 February 2024 05:34:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0048 CST on February 16, 2024, with Unit 2 in mode 1 at 100 percent power, the reactor was manually tripped due to a loss of 2A 125V DC distribution panel. The trip was complex due to the loss of components associated with A-train DC power. Operations responded and stabilized the plant. Decay heat is being removed by the atmospheric relief valves. Unit 1 is not affected. An automatic actuation of the auxiliary feedwater system (AFW) occurred due to low-low steam generator levels. The AFW auto-start is an expected response with low-low steam generator levels from the reactor trip. AFW is still currently controlling steam generator levels. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56852 | 14 November 2023 14:36:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 1041 CST on 11/14/23 with Farley Unit 2 in Mode 1 at 10 percent power, the reactor was manually tripped due to rising steam generator levels. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Auxiliary feedwater (AFW) was manually initiated in accordance with plant procedures and is feeding the steam generators. Heat removal is being provided via the atmospheric relief valves. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. The licensee attempted to take manual control of the feedwater control valves to lower steam generator level but, due to reaching a steam generator level that requires a manual trip, the licensee manually tripped the reactor. |
| ENS 56329 | 1 February 2023 13:40:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 56028 | 3 August 2022 16:25:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 54287 | 21 September 2019 10:28:00 | Farley | NRC Region 2 | Manual Scram | 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. | |
| ENS 54040 | 1 May 2019 21:03:00 | Farley | NRC Region 2 | Manual Scram | 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) | |
| ENS 52395 | 27 November 2016 03:08:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 52356 | 8 November 2016 17:36:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 52274 | 1 October 2016 09:42:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 51918 | 11 May 2016 10:56:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 50533 | 14 October 2014 07:37:00 | Farley | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | 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). |
| ENS 49106 | 12 June 2013 01:33:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 45946 | 22 May 2010 19:10:00 | Farley | NRC Region 2 | Manual Scram Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 40666 | 11 April 2004 12:47:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |
| ENS 40309 | 10 November 2003 12:41:00 | Farley | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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. |