Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57033 | 17 March 2024 20:15:00 | 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 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. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 57024 | 12 March 2024 13:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 56580 | 16 June 2023 23:32:00 | 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 LO-LO Steam Generator Level | The following information was provided by the licensee via email: (On June 16, 2023,) at 1832 CDT, Unit 1 reactor automatically tripped on lo-lo level in the '1-04' steam generator (SG). Prior to the trip, the 1B (main feedwater pump) (MFP) tripped due to speed oscillations and a runback to 700MW was in progress. Both motor driven auxiliary feedwater pumps started due to the lo-lo level in SG '1-04'. Unit 1 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 licensee notified the NRC Resident Inspector. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 is in a normal post-trip electrical line-up. There was no effect on Unit 2 due to the Unit 1 trip. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 56091 | 6 September 2022 04:45:00 | 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 Turbine Trip | The following information was provided by the licensee via email: At 2345 CDT, Unit 1 Reactor tripped due to a turbine trip. All auxiliary feedwater pumps started due to steam generator Lo Lo levels. Unit 1 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007A. The Emergency Response Guideline procedure has been exited. Decay heat is being rejected to the main condenser via steam dump valves. The cause of the Turbine Trip is currently under investigation. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: NRC Resident Inspector has been notified. Unit 2 is unaffected by this event. | Steam Generator Auxiliary Feedwater Main Condenser | |
| ENS 55295 | 7 June 2021 20:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip on Turbine Trip Due to Fault and Fire in the Main Transformer | At 1527 (Central Standard Time) Unit 2 Reactor tripped caused by a turbine trip due to a fault and fire on Unit 2 Main Transformer #1. All Aux Feedwater Pumps started due to steam generator Lo-Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B. The Emergency Response Guideline Network has been exited. Decay heat is being rejected to the Main Condenser via the steam dump valves. Fire was extinguished at 1546 without offsite assistance. No major injuries reported and no personnel transported offsite for medical attention. Cause of the fault and fire are under investigation. NRC Resident Inspector has been notified. All rods inserted into the core during the trip. There were no relief valves or safety valves lifted during the transient. The plant is stable in its normal shutdown electrical lineup via the auxiliary transformer with all safety equipment available. Unit 1 was not affected by the transient. | Steam Generator Feedwater Main Transformer Main Condenser | |
| ENS 54460 | 1 January 2020 18:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip and Specified System Actuation | At 1253 (CST) Unit 2 Reactor was manually tripped due to a trip of both Main Feed Pumps. During an evolution to secure Circulating Water Pump 2-01, vacuum was reduced in the auxiliary condensers for both Main Feed Pumps causing both Main Feed Pumps to trip. Both Motor Driven and the Turbine Driven Auxiliary Feedwater Pumps started due to steam generator Lo-Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-0078. The Emergency Response Guideline Network has been exited. Decay heat is being rejected to the Main Condenser via the Steam Dump Valves. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Main Condenser | |
| ENS 54064 | 15 May 2019 02:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Voltage Transient Due to Loss of Offsite Switchyard | At 2151 CDT, on 14 May 2019, Comanche Peak Nuclear Power Plant (CPNPP) experienced a voltage transient within the onsite 138kV switchyard due to the loss of one of the offsite switchyards supplying power to the CPNPP 138kV switchyard. The reduction in safeguards bus voltage due to the transient caused the Unit 2 safeguard busses to load shed and perform a slow transfer to power supplied from 345kV transformer XST2A. Unit 2 was subjected to actuation of both blackout sequencers causing an automatic start of both motor driven Auxiliary Feedwater (AFW) pumps as well as the turbine-driven AFW pump. No emergency diesel generators started by design. All AFW pumps have been returned to standby status. All other safety systems functioned as designed. Unit 1 is currently defueled, and was unaffected by this event. The licensee has notified the NRC resident inspector. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 53906 | 2 March 2019 09:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Feedwater Isolation Valve Closure | At 0317 CST, the Unit 2 Reactor tripped due to Feedwater Isolation Valve (FWIV) 2-04 going closed. All Auxiliary Feedwater Pumps started due to steam generator Lo-Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B. The Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via the Steam Dump Valves. The cause of the FWIV going closed is currently under investigation. All control rods fully inserted and the reactor trip was uncomplicated. Unit 2 is in a normal post-trip electrical line-up. There was no impact on Unit 1 due to the Unit 2 reactor trip. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 53767 | 3 December 2018 06:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Blackout Sequencers Due to Loss of One Offsite Power Source | At 0315 (CST) on 12/3/18, the Comanche Peak Nuclear Power Plant experienced a loss of 138 KV transformer XST1. Unit 1 is currently at 100% power. Unit 2 was subjected to actuation of both blackout sequencers causing an automatic start of both motor driven Auxiliary Feedwater (AFW) pumps as well as the turbine driven AFW pump. No emergency diesel generators started as per design. Train A and B motor driven and the turbine driven AFW pumps have been returned to automatic. All other safety systems functioned per design. The loss of power to 138 KV transformer XST1 resulted in loss of power to both safeguards busses on Unit 2. The busses performed a load-shed and slow transfer to power supplied from 345 KV transformer XST2A as designed and were re-energized and loads sequenced back onto the busses. The emergency diesel generators are not required to start unless the busses are not re-energized by the alternate offsite transformer. All electrical power related actuations functioned as designed. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 53550 | 13 August 2018 05:00:00 | 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 Generator Lockout | At 23:58 (Central Daylight Time) Unit 2 Reactor Tripped (automatic reactor trip) due to a Turbine Trip/ Generator Lock Out. All Auxiliary Feedwater Pumps started due to steam generator Lo Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IP0-007B. The Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via Steam Dump Valves. The cause of the Generator Lockout is currently under investigation. All control rods fully inserted in response to the automatic reactor trip. The licensee notified the NRC resident. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 53091 | 26 November 2017 02:25:00 | 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 Main Feed Water | At time 2025 (CST) on 11/25/17, Unit 2 reactor was manually tripped due to a loss of all Main Feedwater. Operators observed both Main Feed Pumps tripped and SG (Steam Generator) levels decreasing, resulting in the direction for a manual reactor trip. The reactor trip actuated a turbine trip, both Motor Driven Auxiliary Feedwater Pumps started on the loss of both Main Feed Pumps, and Steam Generator Lo Lo levels started the Turbine Driven Auxiliary Feedwater Pump. All systems responded as expected. There was no work in progress at the time of the incident. Currently Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IP0-0078 and the Emergency Response Guideline Procedure Network has been exited. Decay Heat is being rejected to the Main Condenser via Steam Dump Valves. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 52945 | 2 September 2017 02:40:00 | 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 Two Dropped Rods | At time 2140 (CDT) on September 1, 2017, CPNPP (Comanche Peak Nuclear Power Plant) Unit 2 experienced two (2) dropped rods, one control, one shutdown. The reactor was then manually tripped. This event is being reported in accordance with 10CFR50.72(b)(2)(iv)(B) for reactor trip and 10CFR 50.72(b)(3)(iv)(A) for an actuation of auxiliary feedwater. Currently Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IP0-007B, Emergency Response Guideline Procedure Network has been exited. Decay Heat is being rejected to the Main Condenser via Steam Dump Valves (Turbine Bypass Valves). The NRC Resident Inspector has been notified. All rods inserted into the core during the trip. No relief or safety valves actuated during the plant transient. The electrical grid is stable and supplying plant loads. Unit 1 was not affected by the transient. | Auxiliary Feedwater Main Condenser | |
| ENS 52898 | 11 August 2017 16:24:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Comanche Peak Automatic Turbine Trip from 10 Percent Power | At 1124 CDT on 11 August 2017, CPNPP (Comanche Peak Nuclear Power Plant) Unit 2 experienced an automatic turbine trip and trip of both main feedwater pumps on high steam generator water level (P-14, 81.5 percent level) in steam generator 2-02. Following the turbine trip, the auxiliary feedwater system actuated as required. The plant was stabilized at 2-3 percent reactor power with auxiliary feedwater feeding all steam generators with all levels within their normal bands. The cause of the high steam generator level appears to be a mechanical malfunction of steam generator 2-02 flow control valve bypass valve 2-LV-2163 (SG 2-02 FW BYP CTRL VLV) to close when demanded. Troubleshooting and repair of 2-LV-2163 is in progress. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) for an actuation of auxiliary feedwater. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 51444 | 3 October 2015 15:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip During Shutdown for a Refueling Outage | During the scheduled Unit 2 down power to start 2RF15, Steam Generator 3 Feed Control Valve failed to control in Automatic or Manual resulting in excessive feedwater flow. Control Room Operators manually tripped the reactor and aligned Auxiliary Feedwater. No automatic RPS or ESF actuations occurred. Plant conditions are stable. All control rods fully inserted on the trip and no safety or relief valves lifted. The plant is supplying water to the steam generators with auxiliary feedwater and removing decay heat through the main condenser. The unit is in its normal shutdown electrical lineup. There was no effect on unit-1. The licensee informed the NRC Resident Inspector.
The licensee added 10 CFR 50.72(b)(3)(iv)(A) to the 10 CFR codes reporting section. No other text was modified. Notified the R4DO (Taylor). | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 49743 | 18 January 2014 14:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Turbine Trip | At 0840 (CST), Comanche Peak NPP (CPNPP) Unit 1 turbine tripped causing a reactor trip. All rods inserted. No evolutions were in progress at time of trip. The cause of the trip is under investigation. A Generator Negative Sequence Current High Alarm was received immediately prior to turbine trip. Both Motor Driven Auxiliary Feed Water Pumps and the Turbine Driven Auxiliary Feed Water Pump started in response to Steam Generator Lo-Lo Levels. Currently, Unit 1 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007A and the Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via Steam Dump Valves (Turbine Bypass Valves). Normal plant response was observed and all systems actuated as expected. No radiological release is in progress. There was no impact to CPNPP Unit 2, and off-site power remains available. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Main Condenser | |
| ENS 49606 | 4 December 2013 19:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event - Loss of Startup Power Due to a Cut Cable | At 1343 CST on 12/04/2013, Comanche Peak Nuclear Power Plant experienced a loss of safeguards electrical power when 345KV transformer XST2 lost power from offsite while 138KV transformer XST1 was out of service for maintenance. Non-Safeguards electrical power remained energized. Both Units remain at power in Mode 1. Both units Emergency Diesel Generators automatically started and all safeguards busses were re-energized from the Emergency Diesel Generators. All ESF systems functioned as expected and the Turbine Driven Auxiliary Feedwater Pump automatically started as designed. Operators reduced reactor power to maintain reactor power less than 100 percent rated thermal power. At 1350 CST, the licensee declared an Unusual Event due to losing both startup transformers. The cause of the loss of the second start up transformer was a cut cable. No one was injured during this event. The licensee informed State and Local Government Agencies and the NRC Resident Inspector. Notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and Nuclear SSA via email).
Correction to initial event report description; actual loss of electrical power time was 1341 CST. Include 8-HR Non-Emergency report for Specified System Actuation for automatic start of Auxiliary Feedwater and Emergency Diesel Generators. Notified the R4DO (Vasquez).
Comanche Peak terminated from the Unusual Event at 1717 CST on 12/05/13. Offsite power has been restored to both safeguard busses on both units via the transformer XST2. Work continues on transformer XST1. Comanche Peak Units 1 and 2 continue to operate at 100% power. The licensee has notified the NRC Resident Inspector. Notified R4DO (Vasquez), NRR EO (Bahadur), IRD MOC (Gott), and DHS SWO, FEMA Ops, DHS NICC, and Nuclear SSA via email. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 49497 | 2 November 2013 02:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Solid State Protection System Testing | At 2147 CDT on 11/1/2013, Unit 2 Reactor tripped during Solid State Protection System Slave Relay Testing. This test utilizes a blocking circuit to verify the operability of the slave relay which trips the Main Turbine and both Main Feedwater pump turbines, on a Hi-Hi Steam Generator level or Safety Injection. No valve actuation is expected to occur. While positioning the Slave Relay switch in a testing lineup, the relay actutated. The Unit 2 Turbine tripped as well as both Main Feedwater Pumps. The Turbine Trip actuated the Reactor Trip since power was above 50%. The trip of both Main Feedwater Pumps started both Motor Driven Auxiliary Feedwater Pumps. The Steam Generator Lo Lo Levels started the Turbine Driven Auxiliary Feed Water Pump. All systems responded as expected. Currently, Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B and the Emergency Response Guideline Procedure Network has been exited. Decay Heat is being rejected to the Main Condenser via Steam Dump Valves (Turbine Bypass Valves). The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine Main Condenser | |
| ENS 48514 | 17 November 2012 16:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Steam Generator Water Level | At 1023 CDT Comanche Peak Unit 2 automatically tripped due to steam generator low level. The trip was uncomplicated. All control and shutdown rods fully inserted. Neither emergency diesel generator started. All safety systems functioned as designed. Both motor driven and the turbine driven AFW pumps started as required to restore steam generator level as a result of the trip. The turbine driven AFW pump was returned to auto-start status. The turbine driven AFW pump has been placed in pull-out as per procedure after verifying normal motor driven AFW pump operation. Currently Unit 2 is in Mode 3, (no-load Tave 557 - degrees Fahrenheit) with temperature being maintained with the AFW and steam dumps to the main condenser. Recovery actions will be planned. All electrical busses are powered from offsite power and grid conditions are stable. The initiating event appears to be a sheared instrument air tubing line that feeds the common heater drain pump discharge valve which will fail close resulting in loss of all heater drain flow. The licensee has notified the NRC Resident Inspector. | Steam Generator Emergency Diesel Generator Main Condenser | |
| ENS 48475 | 2 November 2012 06:42:00 | 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 High Reactor Coolant Pump Motor Bearing Temperature | At 0142 CDT, Comanche Peak Nuclear Power Plant Unit 1 was manually tripped due to high temperature indications on the 1-04 reactor coolant pump motor radial bearing concurrent with a high / low oil reservoir level alarm. The trip was uncomplicated; all control rods and shutdown rods fully inserted; neither emergency diesel generator started; all safety systems functioned as designed. Both motor driven and the turbine driven auxiliary feed water pumps started as required to restore steam generator levels as a result of the trip. The turbine driven AFW pump was stopped per procedure and returned to auto-start status. Currently, Unit 1 is in Mode 3, no load Tave of 557 degrees F, with temperature being maintained with AFW and steam dumps to the main condenser. All electrical busses are powered from offsite and grid conditions are stable. The licensee has notified the NRC Resident Inspector. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 46863 | 19 May 2011 19:07:00 | 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 a Main Condenser Tube Leak | At 1407 CDT CPNPP (Comanche Peak Nuclear Power Plant) Unit-2 was manually tripped due to indication of a main condenser tube failure. The trip was uncomplicated, all control and shutdown banks fully inserted, neither emergency diesel generator started, all safety systems functioned as designed. Both motor driven and the turbine driven AFW (auxiliary feedwater) pumps started as required to restore steam generator level as a result of the trip. The turbine driven AFW pumps has been returned to auto start status. Currently Unit-2 is in mode 3, no-load Tave 557 degrees F with temperature being maintained with AFW and steam dumps to the main condenser. Recovery actions are currently being planned. All electrical busses are powered from offsite power and grid conditions are stable. The reactor was manually tripped because of rising sodium concentration in the main condenser/feedwater. The licensee notified the NRC Resident Inspector, State of Texas, and local government. | Steam Generator Emergency Diesel Generator Main Condenser | |
| ENS 46748 | 13 April 2011 03:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Esf Actuation Following Loss of a Safeguard Bus During Edg Testing | The following event occurred with the unit in a "No Mode" defueled condition. On the 12th of April at 2215 CDT Comanche Peak Nuclear Power Plant Unit 2 had a Train B ESF Actuation (Black Out) reportable per 10CFR50.72(b)(3)(iv) for (the 6.9KV Train B) Emergency AC Electrical power systems including Emergency Diesel Generator (EDG). At the time of the event, Unit 2 was in 'No Mode' - the core was off loaded and post maintenance testing was being performed on the 2-02 Emergency Diesel Generator. The scope of the EDG testing was to verify proper operation of the EDG when it was the sole supply to the Unit 2 Train B 6.9KV safeguards buses. Unit 2 Train B 6.9 KV buses have 3 AC sources - XST1 (138KV) preferred source, XST2 (345KV) alternate source and the 2-02 Emergency Diesel Generator. The 2-02 EDG was started normally and was operating in parallel with XST1 (preferred AC source) just before the event. The event was triggered when the preferred AC supply breaker to the Train B 6.9KV Safeguard bus was opened per plan. The opening of the preferred feeder breaker resulted in the 2-02 EDG being the sole AC source to the Train B 6.9KV bus. Immediately following this breaker operation, the EDG tripped on an Auto Voltage Regulator failure. This caused the Train B 6.9KV safeguard bus to de-energize which resulted in the automatic closure of the alternate AC (XST2) feeder breaker. This in turn triggered the Train B Solid State Sequencer to sequence on Train B Black Out loads - as designed. The Train B Black Out Solid State Sequencer operated as designed. No abnormalities were noted. All available equipment operated as designed. The 2-02 EDG is currently tagged out and the problem associated with the (Auto Voltage Regulator) failure is being investigated. (The) NRC Resident Inspector has been informed and is currently located on site. No fuel movement was in progress during this event. Train A safeguard power was unaffected and available throughout. Non-safety related power was also not affected. The only operational impact was the loss of a Train B spent fuel pool cooling pump for approximately 10 minutes. The spent fuel pool temperature increased approximately 1/2 degree during before cooling was restored.
Further investigation of this event has determined that the 2-02 Emergency Diesel Generator (EDG) auto voltage regulator functioned as designed. The EDG did not receive an emergency start signal, which was consistent with the design for the plant conditions present. In addition, Unit 2 was defueled and in a 'No Mode' condition, and there was no actuation of any of the systems listed in paragraph 50.72 (b)(3)(iv)(B) by the sequencer operation. Therefore, this event is not reportable because it did not meet the criteria for a system actuation per 50.72(b)(3)(iv)(A) and the guidance in NUREG-1022. The licensee will be notifying the NRC Resident Inspector. Notified R4DO (Miller). | Emergency Diesel Generator | |
| ENS 45833 | 12 April 2010 12:21:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Blackout Sequencers Due to Loss of One Offsite Power Source | At 0721 (CDT) on 4/12/10 the Comanche Peak Nuclear Power Plant experienced a loss of 138 KV transformer XST1 due to a failed transmission line insulator. Unit 1 is currently defueled per 1RF14 refueling outage schedule. Unit 2 was subjected to actuation of both blackout sequencers causing an automatic start of both motor driven Aux Feedwater (AFW) pumps as well as the turbine driven AFW pump. No emergency diesel generator started as per design. Train B and the turbine driven AFW pumps have been returned to automatic. Train A AFW pump has been removed from service due to damaged pump packing. Unit 2 is currently operating in Technical Specification (TS) 3.8.1, Condition A, for one required source of offsite power inoperable and in TS 3.7.5, Condition B, for one train of AFW inoperable. With the exception of Train A AFW on Unit 2, all other safety systems functioned per design. The licensee has notified the NRC Resident Inspector. The loss of power to 138 KV transformer XST1 resulted in loss of power to both safeguards busses on Unit 2. The busses performed a load-shed and slow transfer to power supplied from 138 KV transformer XST2 as designed and were re-energized and loads sequenced back onto the busses. The emergency diesel generators are not required to start unless the busses are not re-energized by the alternate offsite transformer. All electrical power related actuations functioned as designed. During this event, the A train AFW pump on Unit 2 started as expected but had to be removed from service due to a pump packing leak. There was no impact on Unit 1 because of its defueled condition.
The faulted 138 KV (line) insulator has been repaired, and the Unit 2 safeguard busses have been transferred back to their normal power supplies. The 2-01 motor driven AFW pump has been successfully surveillance tested and is available. Unit 2 remains in TS 3.7.5 condition B pending the close out of an engineering evaluation. No further update notifications are planned. The licensee has informed the NRC Resident Inspector. Notified R4DO (Spitzberg). | Feedwater Emergency Diesel Generator | |
| ENS 45617 | 9 January 2010 16:28:00 | 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 Turbine Trip on Sudden Pressure in Output Transformer | At 1028 CST on the 9th of January, Unit 1 tripped due to turbine trip greater than 50% reactor power. The turbine trip was caused by a sudden pressure fault of the Unit 1 main generator output transformer (1MT1). All systems operated as designed. All three auxiliary feedwater pumps auto-started as required. There were no complications noted on the reactor trip. Current status: Holding in mode 3 and shutting down secondary equipment. Decay heat removal is via auxiliary feed and main condenser steam dumps. All rods inserted during the trip. No relief or safety valves lifted during the transient. The electrical grid is stable and Unit 1 is in its' normal shutdown electrical lineup. Investigation of the transformer fault is underway. Unit 2 was unaffected by the transient. The licensee will be notifying the NRC Resident Inspector. | Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 44067 | 16 March 2008 16:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Broken Sensing Line | At 1137 CDT, Comanche Peak NPP, Unit 2 reactor tripped due to a broken sensing line feeding the vacuum inputs to the 2/3 low vacuum turbine trip. With the reactor power steady at 100%, the turbine trip signal generated a valid RPS reactor trip signal. All control rods fully inserted. Both motor driven AFW pumps and the turbine driven AFW pump started as a result of low steam generator level following the trip. Currently Unit 2 is in Mode 3, stable with RCS temperature at no-load program temperature of 557 degree F being maintained on the steam dumps. All 4 steam generators are at program level, being maintained by the motor driven AFW pumps. The turbine driven AFW pump has been returned to standby. The licensee notified the NRC Resident Inspector. | Steam Generator Control Rod | |
| ENS 42945 | 29 October 2006 21:20:00 | 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 Lowering Steam Generator Level | Unit 2 was operating in Mode 1 at 80% power following refueling holding for stabilization xenon in preparation for incore - excore calibration. An alarm condition for steam generator 2-03 indicating Steam - Feedwater flow mismatch was observed. Manual control of the feed regulating valve was taken, but operators were unable to control feed flow while steam generator 2-03 level continued to decrease. A manual reactor trip was initiated and the reactor was tripped. Auxiliary feedwater automatically started on low-low level in steam generators. All systems responded normally during and following the trip. Unit 2 is being maintained in Mode 3 pending the event investigation. All control rods fully inserted. RCPs are in operation transferring decay heat to the steam generators. The MSIV's are open with the steam generators discharging steam to the main condensers using the steam dumps. AFW is maintaining steam generator water levels. Unit 1 was not affected. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42937 | 27 October 2006 08:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Loss of Both Main Feed Pumps | The licensee was performing load reject testing on the Unit 2 turbine generator when oscillations were observed on the plant steam dump system. These oscillations caused a high steam generator water level that caused both Main Feed Pumps (MFP) to trip. Upon receiving this indication the licensee manually tripped the reactor. All Auxiliary Feed Water (AFW) pumps started as expected. All rods inserted as expected. Decay heat is being removed using AFW supplying both steam generators and steaming to the steam dumps. The plant is in a normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 41432 | 23 February 2005 07:53:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater Action Following Momentary Power Interruption | A momentary interruption in the preferred offsite power source to Unit 2 safeguards 6.9 KV busses occurred due to adverse weather. A blackout signal was generated as a result which in turn initiated an auxiliary feedwater signal. Operators took action to reduce turbine load to 1100 MWe to ensure reactor power would remain below 100% following the reactivity effects associated with the addition of cold auxiliary feedwater. The event was caused by both feeder breakers to the 138 KV switchyard opening. Breaker 7050 auto-reclosed but breaker 7020 needed to be manually reclosed. The cause of the breaker operation is not fully understood and is still under investigation. Auxiliary feedwater has been returned to standby condition and reactor power is being restored to 100%. The NRC Resident Inspector has been notified. | Auxiliary Feedwater | 05000446/LER-2005-002 |
| ENS 41134 | 19 October 2004 10:28:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Normal Power Supply to "2E" Switchgear Resulted in Afw Autostart | Unit 2 experienced an autostart of Auxiliary Feedwater System (AFW) and Emergency Diesel Generator 'B' following transfer of the '2E' Switchgear to Alternate Power Supply. Normal Power Supply (XST1) experienced a loss of power, causing the '2E' Switchgear to slow transfer to the -2 Breakers from the Alternate Power Supply (XST2). The AFW (MDAFWP (Motor-driven Auxiliary Feedwater Pumps) 'A' and 'B' and TDAFWP (Turbine-driven Auxiliary Feedwater Pump)) autostarted due to Sequencer Operations. The EDG 'B' autostarted on low voltage condition but did not supply power to the affected '2E' bus. The EDG 'B' cooling water was provided by the service water system. Power was restored to XST1 at 0606 CDT. The cause of the EDG 'B' autostart is under investigation. The cause of the loss of power to XST1 is under investigation. Power was reduced to 98.2% to ensure Nuclear Power <100%. All other systems functioned as required. Local NRC Resident was informed. | Service water Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 40406 | 22 December 2003 14:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Main Turbine Trip | At 0827 on 12/22/2003, Comanche Peak Unit Two (2) tripped due to a turbine trip > 50%. The turbine trip was caused by ingestion of a small metallic cover plate into the main generator exciter housing, as witnessed by maintenance personnel. A small fire ensued, which self-extinguished in five (5) minutes. An Aux Feedwater auto start (ESF actuation) occurred due to normal post-trip steam generator shrink. The turbine-driven and both motor-driven aux feedwater pumps started as designed. All control rods inserted into the core upon trip. The local electrical grid is stable. The number one (1) steam generator atmospheric relief valve cycled as designed upon the secondary pressure transient. The number three (3) Reactor Coolant Pump tripped during the transfer of power sources post-trip. Unit 1 was not affected by the trip. Unit 2 is stable in Mode 3. Damage assessment is in progress which will determine future (near-term) plant mode of operation. RCPs are in operation transferring decay heat to the steam generators. Decay heat removal is through the MSIVs to the main condenser using the turbine steam dump valves. Both motor-driven AFW pumps are running to maintain Steam Generator levels. The trip of the number (3) Reactor Coolant Pump is being investigated. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Decay Heat Removal Main Condenser Control Rod |