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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57026 | 13 March 2024 01:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Power to Containment Radiation Monitors | The following information was provided by the licensee via phone and email: On March 12, 2024, at 2111 EDT, a valid containment ventilation isolation train 'A' and 'B' signal was received due to a spurious loss of power to 1EMF-38 (containment particulate radiation monitor) and 1EMF-39 (containment gas radiation monitor). The power to 1EMF-38 and 1EMF-39 was restored. 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: There were no plant evolutions ongoing at the time of the event and the cause of the loss of power is under investigation. There was no impact to Unit 2.
After further review of the event, it was determined the actuation of the associated containment ventilation isolation train 'A' and 'B' was not valid. This is due to the loss of power being associated with the control room modules for 1EMF-38 and 1EMF-39, and not a result of an actual sensed parameter or plant condition. Therefore, this event notification is being retracted. The NRC Resident Inspector has been notified. Notified R2DO (Miller) | ||
| ENS 56177 | 24 October 2022 12:57:00 | 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 Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: On 10/24/2022 at 0857 EDT, with Unit 2 in Mode 1 at 7 percent power, the reactor was manually tripped due to a 2B train main feedwater pump trip. The trip was not complex, with all systems responding normally post-trip. The auxiliary feedwater (AFW) system started automatically as expected. Operations responded and stabilized the plant. Decay heat is being removed by the steam generators and discharging to the condenser. 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 automatic start of the auxiliary feedwater system is being reported as an eight-hour, non-emergency notification per 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 55856 | 23 April 2022 06:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip During Control Rod Testing | The following information was provided by the licensee via fax or email: On April 23, 2022, at 0224 (EDT) hours, with Unit 2 in Mode 1 at 100 percent power, two control rods dropped during control rod testing resulting in misalignment, which required a manual reactor trip in accordance with plant procedure. All safety systems functioned as expected. The Auxiliary Feedwater system actuated as designed to provide makeup flow to the steam generators. Operations responded and stabilized the plant. Decay heat is being removed by the steam generator power operated relief valves. 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 automatic start of the Auxiliary Feedwater system is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). The cause of the dropped rods is being investigated. Unit 1 is not affected. 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: Electrical power is in normal off-site arrangement. All emergency electrical supplies are available. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 55549 | 27 October 2021 18:29:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the 1B EDG Load Sequencer on Bus Undervoltage Condition | At 1429 EDT on October 27, 2021 with Unit 1 in Mode 6 at 0 percent power, the 1B Emergency Diesel Generator (EDG) Load Sequencer was actuated by a valid undervoltage condition on the 1B 4160V Essential Bus that occurred during 1B Sequencer calibration activities. Valid signals were sent to both the 1B EDG and Unit 1 Auxiliary Feedwater (CA) systems. Neither system automatically started as they were both removed from service for maintenance activities at the time. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the 1B EDG and Unit 1 CA systems. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 55229 | 1 May 2021 11:55:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip and Actuation of the Auxiliary Feedwater System | At 0755 EDT, on May 1, 2021, with Unit 2 in Mode 3 at 0 percent (not critical) power, the reactor trip breakers opened during heat-up activities. The trip was not complex, with all systems responding normally post-trip. At 1013 EDT, on May 1, 2021, with Unit 2 in Mode 3 at 0 percent power, an actuation of the Auxiliary Feedwater (AFW) System occurred. The loss of both main feedwater pump turbines caused an AFW auto-start. The 2A and 2B motor driven auxiliary feedwater (MDAFW) pumps automatically started as designed when the loss of both main feedwater pumps signal was received. The cause of the actuation is still being evaluated. Operations responded and stabilized the plant. Decay heat is being removed by the steam generators and discharging steam to the condenser. Unit 1 is not affected. Due to the Reactor Protection System (RPS) actuation while not critical and the actuation of the AFW system, this event is being reported as an eight-hour, non-emergency notification per 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 54307 | 2 October 2019 08:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Containment Spray System | On October 2, 2019, at 0415 EDT, with Unit 2 in Mode 5 at 0 percent power, an actuation of the Unit 2 containment spray system occurred during valve strokes of the 2A train containment spray header isolations while the 2B train containment spray pump was in recirculation. The reason for the containment spray actuation was due to a conflicting procedural alignment with the 2B containment spray recirculation procedure. The containment spray system does not have an automatic function, and only receives manual actuation. The Unit 2 containment spray actuation was secured at 0416. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the containment spray system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Containment Spray | |
| ENS 53271 | 19 March 2018 07:29:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Actuation of Auxiliary Feedwater System During Maintenance | Automatic actuation of the 2B CA (Auxiliary Feedwater) System. At 0329 (EDT on) 03-19-18, with Unit 2 in Mode 5, an actuation of the 2B CA (Auxiliary Feedwater) system occurred during maintenance associated with the 2B Main Feedwater Pump. The cause of the CA (Auxiliary Feedwater) system automatic start was an unanticipated reset and subsequent trip signal of the 2B Main Feedwater pump during maintenance activities. The 2B CA Pump/Train automatically started as designed when the 'Loss of Both Main Feedwater Pumps' signal was received. The system operated as required. This event is being reported in accordance with 10CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the CA System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. | Feedwater Auxiliary Feedwater | |
| ENS 46780 | 23 April 2011 13:21:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Auxiliary Feedwater Pump Start While Cooling Down Reactor | At 0921 (EDT) on 4/23/11, CNS Unit-1 received an auxiliary feedwater pump auto start due to low-low level in the 1B steam generator. The 1A auxiliary feedwater pump started and restored steam generator level to the normal operating band. A cooldown was in progress at the time of the event and the tagout isolated feedwater flow to the 1B steam generator. The licensee notified the NRC Resident Inspector. The states of North and South Carolina, and York, Gaston, and Mecklenburg Counties will be notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 43597 | 25 August 2007 14:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Autostart of All Emergency Diesel Generators Due to a Grid Disturbance | A switchyard grid disturbance caused undervoltage relays to start both unit's Emergency Diesel Generators (EDGs). 1A, 1B, 2A, and 2B (EDGs) started which met ESF actuation (conditions). The (EDGs) did not supply the emergency (buses) due to voltage returning to normal immediately. The grid fault was isolated and the (EDGs) were secured. The EDGs are now available to respond if needed. The licensee informed the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 41246 | 6 December 2004 02:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip/Reactor Trip Due to Moisture Separator Reheater (Msr) High Level Trip Signal. | Reactor Trip caused by Turbine trip. Turbine trip caused by a Moisture Separator Reheater 1B high level. Maintenance personnel were working in vicinity of level switches at time of trip. Root cause investigation in progress. All systems operated as designed on Reactor trip. Maintenance personnel working in the vicinity of Moisture Separator Reheater 1B were setting up radiography equipment when the MSR 1B high level trip was generated. All rods fully inserted. Both Motor Driven Auxiliary Feedwater pumps actuated as expected. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. Catawba Unit 2 is operating at 100% power and the electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System | |
| ENS 41154 | 28 October 2004 04:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After Shutdown Bank Dropped Into Core | At 0052 on 10/28/04 the Unit 2 reactor was manually tripped after an electrical fault caused Shutdown Bank D control rods to drop into the core. An automatic start of the Auxiliary Feedwater system occurred when water levels in all four Steam Generators reduced to the Lo-Lo level setpoint. A letdown isolation occurred on Lo Pressurizer level resulting from the cooldown of the primary system after the Auxiliary Feedwater system start. Reactor Coolant system temperature has been recovered and the plant has been stabilized at Hot Shutdown conditions. The licensee stated that there were no complications during the trip. All systems functioned as required. No significant safety systems were out of service when the trip occurred. No primary or secondary relief valves lifted during the transient. Decay heat is currently being removed using aux feedwater to the steam generators steaming to the main condenser. Tave dropped to a low point of 540 degrees F during the transient. The reactor was manually tripped in approximately 18 seconds after the shutdown bank dropped. The licensee plans to remain in mode 3. The licensee will notify the NRC Resident Inspector. The licensee will also notify state and local authorities in North and South Carolina. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 40538 | 22 February 2004 22:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manually Tripped After the "B" Sg Feedwater Isolation Valve Failed Closed | The 1CF-42B (1B Steam Generator Feedwater Isolation Valve) failed closed with Unit 1 at 100% power. Operators performed a manual reactor trip as required by 'Loss of Normal Steam Generator Feedwater Supply', AP-06. Emergency procedures were utilized to respond and stabilize Unit 1 in mode 3. The plant is currently stable in mode 3 at normal operating pressure and temperature. Currently, steam generator levels are being maintained using the auxiliary feedwater system. Other than the failure of 1CF-42B (Initiating Event), all plant safety systems operated as required. No primary or secondary safeties or reliefs lifted during the transient. All control rods fully inserted. The main condenser is in-service to remove decay heat. The licensee had identified a problem with the hydraulic motor associated with 1CF-42B and was in the process of developing corrective actions before the failure. The licensee will inform both state/local agencies (courtesy notification) and has informed the NRC resident inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 40114 | 29 August 2003 06:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Catawba Unit 1 Automatic Reactor Trip Due to Reactor Protection System Actuation | Unit 1 tripped on Over Temperature - Delta Temperature (OTDT) protection. Channel 1 OTDT was in a trip condition due to a reactor coolant hot leg temperature problem. Per Technical Specifications, this required Channel 1 OTDT to be placed in the tripped condition. Channel 2 Pressurizer pressure failed low at 0203 EDT. This input (Channel 2 Pressurizer Low Pressure) caused Channel 2 OTDT to enter a trip condition. With 2 of 4 OTDT Channels in the trip condition, a reactor trip signal was generated by the reactor protection system resulting in an automatic reactor trip. After the reactor trip occurred, RCS (Reactor Coolant System) letdown was lost on the chemical and volume control system due to low Pressurizer level. 1NV-1A (Reactor Coolant Letdown Isolation Valve) did not close as required. This necessitated the alignment of the excess letdown flowpath to maintain the desired Pressurizer level. ESF (Engineered Safety Features) operated as designed. Auxiliary feedwater automatically started and main feedwater isolated as designed. All control rods fully inserted on the reactor trip. Station service electrical systems, emergency diesel generators, and ESF equipment are available. Catawba Unit 2 was unaffected by the trip of Unit 1. The licensee has notified the NRC Resident Inspector and will be notifying state and local officials. | Reactor Coolant System Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Control Rod |