| | Site | Start date | Title | Description |
|---|
| ENS 57032 | Waterford | 16 March 2024 19:49:00 | Manual Reactor Trip Due to Main Feedwater and Main Steam Isolations | The following information was provided by the licensee via phone and email:
At 1449 CDT, Waterford 3 Steam Electric Station was operating at 100 percent power when a manual reactor trip was initiated due to main feed isolation valve (FW-184B) and main steam isolation valve (MS-124B) going closed unexpectedly.
Emergency feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected.
This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system.
The NRC Resident Inspector has been notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
Decay heat is being removed through the turbine bypass valves and the atmospheric dump valve on loop '2'. There is no primary to secondary system leakage. The cause of the isolations is still being investigated. | | ENS 57021 | Hatch | 11 March 2024 17:37:00 | Manual Reactor Trip | The following information was provided by the licensee via phone and email:
On March 11, 2024, at 1337 EDT, with Unit 1 in Mode 1 at 35 percent power performing power ascension activities, the reactor was manually tripped due to the 'A' reactor feed pump (RFP) tripping on low suction pressure. Due to the power level at the time, the 'B' RFP had not been placed in service. Closure of containment isolation valves (CIVs) in multiple systems and actuation of high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) occurred as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. The 'B' RFP was placed in service and is controlling reactor water level. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 2 is not affected.
Due to the emergency core cooling system (ECCS) discharging into the reactor, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, the Reactor Protection System actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI.
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:
The cause of the 'A' RFP is under investigation. The reactor electric plant remains in a normal lineup with both emergency diesel generators available. There were no temperature or pressure technical specification limits approached. | | 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 56995 | Monticello | 28 February 2024 14:39:00 | Automatic Reactor Scram | The following information was provided by the licensee via fax and email:
At approximately 0839 (CST) with Unit 1 in Mode 1 at 100 percent power, the reactor automatically scrammed due to the depressurization of the SCRAM air header caused by an invalid signal that (occurred) during system testing. The SCRAM was uncomplicated with all systems responding as expected. The cause and details of the event are under investigation. Containment isolation valves actuated and closed on a valid Group 2 signal.
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, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 isolation signal.
Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves.
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. State as well as Wright and Sherburne Counties will be notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
The Anticipated Transient Without Scram (ATWS) circuit was being tested when an invalid signal was sent to depressurize the SCRAM air header. | | 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 56957 | Peach Bottom | 9 February 2024 18:22:00 | Unanalyzed Condition - Inadequate Fuses for Fuel Pool Cooling | The following information was provided by the licensee via email:
On 2/9/24 at 1322 EST, it was determined that the unit was in an unanalyzed condition. A review of DC feeder circuit protection schemes identified a circuit for the fuel pool cooling system is uncoordinated due to inadequate fuse sizing. This results in a concern that postulated fire damage in one area could cause a short circuit without adequate protection, leading to the unavailability of equipment credited for in 10 CFR 50 Appendix R, Fire Safe Shutdown. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B).
The postulated event affects the following fire zones: fire areas 6S and 6N (within the Unit 2 reactor building). Compensatory actions for affected fire areas have been implemented. An extent of condition review is being performed.
The NRC Senior Resident Inspector has been notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
Fire watches have been established in the affected areas. These will be maintained until the protection scheme is revised.
- * * UPDATE ON 03/08/24 FROM PAUL BOKUS TO TOM HERRITY * * *
The following updated information was provided by the licensee via email and phone call:
On 03/08/24 at 1418, extent of condition reviews identified circuit(s) in the Units 2 and 3 Reactor Protection Systems (RPS) which are also uncoordinated due to improper fuse sizing. These circuits are not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel.
The postulated event affects the following fire areas: 32, 33, 38 and 39 (Units 2 and 3 Switchgear Rooms). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved.
The NRC Senior Resident Inspector has been notified.
Notified R1DO (Arner)
- * * UPDATE ON 3/13/2024 AT 1538 FROM TROY RALSTON TO SAM COLVARD * * *
On March 13, 2024, at 1350 EDT, extent of condition reviews identified a circuit in the Unit 2 reactor protection system (RPS) which is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50, Appendix R, Fire Safe Shutdown, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel.
The postulated event affects fire area 57 (Switchgear Corridor, common to Units 2 and 3). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved.
Additionally, it was previously reported that fire area 6N contained a circuit which was not bounded by the Fire Safe Shutdown analysis; however, after further review it has been determined that compliance is maintained in this fire area and is therefore retracted from the scope of this report.
The NRC Senior Resident Inspector has been notified.
Notified R1DO (Jackson)
- * * UPDATE ON 3/21/2024 AT 1525 FROM PAUL BOKUS TO IAN HOWARD * * *
The following information was provided by the licensee via email:
On 03/21/24 at 1211, extent of condition reviews identified an annunciator circuit for the Unit 3 emergency service water (ESW) and high pressure service water (HPSW) pump structure heating and ventilation panel that is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel.
The postulated event affects fire area 47 (Unit 3 pump structure for `B' ESW and `3A'-`3D' HPSW pumps) and the yard fire area (Manhole 026D). In order to restore immediate compliance, the cable has been de-energized to eliminate the possibility of the event of concern. This circuit will remain de-energized or other measures will be implemented until the condition is permanently resolved.
The NRC Senior Resident Inspector has been notified.
Notified R1DO (Ford) | | IR 05000317/2023004 | Calvert Cliffs | 1 February 2024 | Integrated Inspection Report 05000317/2023004 and 05000318/2023004 | | | IR 05000397/2023004 | Columbia | 29 January 2024 | Integrated Inspection and Independent Spent Fuel Storage Installation Report 05000397/2023004 and 07200035/2023001 | | | 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. |
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