| | Site | Start date | Title | Description |
|---|
| ENS 57033 | Comanche Peak | 17 March 2024 20:15:00 | Manual Reactor Trip Due to Main Feedwater Pump Trip | The following information was provided by the licensee via phone and email:
On March 17, 2024, at 1515 CDT, the Comanche Peak Unit 2 reactor was manually tripped due to an anticipated automatic trip due to lo-lo steam generator (SG) water levels. Prior to the trip, main feedwater pump '2B' tripped and an auto runback to 700 MW (60 percent power) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs.
Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedures IPO-007B. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
The cause of the '2B' main feed pump trip was due to loss of primary and redundant power to the servo control valve. The loss of power to the servo control valve is under investigation. |
| ENS 57024 | Comanche Peak | 12 March 2024 13:16:00 | Automatic Reactor Trip | The following information was provided by the licensee via phone and email:
On March 12, 2024, at 0816 CDT, Comanche Peak Unit 2 reactor automatically tripped on lo-lo level in the 2-03 steam generator (SG). Prior to the trip, main feedwater pump (MFP) 2A speed reduced and a manual runback to 700 MW (60 percent) was in progress. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started due to lo-lo level in all SGs.
Concurrent with the loss of speed on MFP 2A, a servo filter swap was in progress on MFP 2A.
Unit 2 is being maintained in hot standby (Mode 3) in accordance with integrated plant operating procedure IPO-007A. The emergency response guideline network has been exited. Decay heat is being rejected to the main condenser via the steam dump valves.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
The cause of the loss of the MFP is under investigation. Unit 1 was unaffected. |
| ENS 57006 | Watts Bar | 5 March 2024 06:32:00 | Automatic Reactor Trip | The following information was provided by the licensee via email:
At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected.
Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A).
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. |
| ENS 57003 | Prairie Island | 3 March 2024 17:42:00 | Automatic Reactor Trip Due to Loss of Main Feedwater Pump Suction | The following information was provided by the licensee via email:
At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs (power operated relief valves). Unit 1 remains at 100 percent power and is unaffected.
This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation. |
| ENS 56991 | Calvert Cliffs | 24 February 2024 20:46:00 | Manual Reactor Trip Due to Steam Generator 22 Feed Pump Trip | The following information was provided by the licensee via email:
At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B).
Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected.
ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56971 | Farley | 16 February 2024 11:34:00 | 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. |
| ENS 56968 | Callaway | 15 February 2024 08:47:00 | Automatic Turbine Trip/Reactor Trip | The following information was provided by the licensee via email:
At 0247 CST on 2/15/2024, Callaway Plant was in mode 1 at approximately 100 percent power when a turbine trip and reactor trip occurred. All safety systems responded as expected with the exception of an indication issue on the feedwater isolation valves, which were confirmed closed. A valid feedwater isolation signal and auxiliary feedwater actuation signal were also received as a result of the reactor trip. The plant is being maintained stable in mode 3.
All control rods fully inserted from the reactor trip signal and decay heat is being removed via the auxiliary feedwater system and steam dumps.
The NRC Resident Inspector was notified. |
| IR 05000317/2023004 | Calvert Cliffs | 1 February 2024 | Integrated Inspection Report 05000317/2023004 and 05000318/2023004 | |
| ENS 56935 | Watts Bar | 28 January 2024 02:41:00 | Automatic Reactor Trip | The following information was provided by the licensee via email:
At 2141 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip.
Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected.
Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A).
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
All control rods are fully inserted. The cause of the turbine trip is being investigated.
The licensee notified the NRC Resident Inspector. |
| ML24026A011 | Ginna | 26 January 2024 | R. E. Ginna Nuclear Power Plant, Relief Request Associated with Inservice Testing of B Auxiliary Feedwater Pump | |