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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57148 | 29 May 2024 10:24:00 | 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 email and phone: On May 29, 2024, at 0624 EDT, Unit 1 automatically tripped from 100 percent power due to a negative rate trip. The unit has been stabilized in mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater pumps actuated as designed because of the reactor trip and is reportable per 10 CFR 50.72(b)(3)(iv)(A) for a valid engineered safety feature (ESF) actuation. Decay heat is being removed by the condenser steam dump system and Unit 1 is in a normal shutdown electrical lineup. Unit 2 was not affected by this event. The NRC Resident has been notified. | Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 55843 | 14 April 2022 13:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Control Rod Testing | The following information was provided by the licensee via email: On April 14, 2022, at 0928 (EDT) hours, Unit 1 automatically tripped from 100 percent power during the control rod operability periodic test. The reactor trip occurred during the manipulation of the rod control mode selector switch as part of the rod operability testing. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated, and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed because of the reactor trip and provide makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10 CFR 50.72(b)(3)(iv) (A) for a valid actuation of an ESF (Engineered Safety Features) system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. An investigation into the cause of the reactor trip is underway. 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 was no affect to Unit 2. Unit 2 is operating at 100 percent power. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 53579 | 3 September 2018 04:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of the Emergency Diesel Generator | At 1045 (EDT) on 9/3/18, with Unit 1 and Unit 2 at 100% power, off-site power feed to the 'A' Reserve Station Transformer was lost which resulted in a loss of power to Unit 1'J' Emergency Bus. As a result of the power loss, the 1'J' Emergency Diesel Generator started as designed and restored power to the Emergency Bus. During this event, the Unit 1 'A' Charging Pump, 1-CH-P-1A automatically started as designed due to the loss of power event. The valid actuation of these ESF (Engineered Safety Features) components due to the loss of power is reportable per 10 CFR 50.72 (b)(3)(iv)(A). The Unit 1 'J' Emergency bus off-site power source was restored via the Unit 2 'B' Station Service bus and the 1 'J' Emergency Diesel was secured and returned to Automatic. The Unit 1 'A' Charging pump has been stopped and returned to Automatic. Both Units are in a stable condition. The apparent cause for the loss of power appears to be a bird strike to the 'A' RSST (Reserve Station Service Transformer) Overhead Cable. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 51892 | 30 April 2016 02:14:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Offsite Power to the a Reserve Station Service Transformer | At 2214 (EDT) on 4/29/16, with Unit 1 and Unit 2 operating at 100 (percent) power, the North Anna 34.5 kV Bus 5, off site power feed to the 'A' Reserve Station Service Transformer, was lost which resulted in the loss of power to the Unit 1 'J' Emergency Bus. Loss of Bus 5 is still undergoing investigation. As a result of the power loss, the 1J Emergency Diesel Generator automatically started as designed and restored power to the 1J Emergency bus. During the event, the Unit 1 'A' Charging Pump (1-CH-P-1A) automatically started as designed due to the loss of power event. The valid actuation of these ESF components due to the loss of electrical power is reportable per 10 CFR 50.72(b)(3)(iv)(A). The Unit 1 'J' Emergency Bus off-site power source was restored to service and the 1J Emergency Diesel Generator was secured and returned to automatic. Restoration of offsite power to Operable is complete. The Unit 1 'A' Charging Pump has been secured and returned to automatic. Both units are currently stable. An investigation is underway to determine the cause of the Bus 5 loss of power. The NRC Resident Inspector was notified. | Emergency Diesel Generator | |
| ENS 51678 | 23 January 2016 22:03:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Partial Loss of Power Results in Emergency Diesel Generators Starting | At 1703 (EST) on 1/23/16, with Unit 1 and Unit 2 operating at 100% power, the North Anna 34.5 kv Bus 3, off-site power feed to the 'C' Reserve Station Service Transformer, was lost which resulted in the loss of power to the Unit 1 'H' Emergency Bus and the Unit 2 'J' Emergency Bus. Loss of 34.5kV Bus 3 resulted from feeder breaker L102 opening. As a result of the power loss, the 1H Emergency Diesel Generator and the 2J Emergency Diesel Generator automatically started as designed and restored power to the associated emergency bus. During this event, the Unit 1 'B' Charging Pump, 1-CH-P-18 automatically started as designed due to the loss of power event. The valid actuation of these ESF components due to the loss of electrical power is reportable per 10 CFR 50.72 (b)(3)(iv)(A). The Unit 1 'H' Emergency Bus off-site power source was restored to service and the 1H Emergency Diesel Generator was secured and returned to Automatic. The Unit 2 'J' Emergency Bus power feed continues to be from the 2J Emergency Diesel Generator. Restoration of offsite power to operable status is currently being pursued. The Unit 1 'B' Charging Pump has been secured and returned to automatic. Both units are in a stable condition. An investigation is underway to determine the cause of the L102 feeder breaker opening resulting in the 34.5 kv Bus 3 loss of power. The licensee notified the NRC Resident Inspector | Emergency Diesel Generator | |
| ENS 50946 | 2 April 2015 08:26:00 | 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 Generator Voltage Regulator Failure | On April 2, 2015 at 0426 EDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a failure of the main generator voltage regulator. This also resulted in a turbine trip. The operations crew entered the reactor trip procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated as designed and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified. There was no effect on Unit 2 as a result of this trip. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 50851 | 26 February 2015 20:11:00 | 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 Feedwater Regulating Valve Failing Closed | On February 26, 2015, at 1511 EST, with Unit 1 operating at 95% power in an end of cycle coastdown, the 'B' Main Feedwater Reg Valve failed closed which resulted in a Unit 1 automatic reactor trip due to 'B' Steam Generator low/low level. The operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10CFR50.72(b)(3)(iv)(A) for the valid actuation of an ESF system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified and are in the Control Room. The Louisa County Administrator will be notified. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 50116 | 15 May 2014 23:20:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 and Unit 2 Emergency Diesel Generators Start and Load Due to Loss of One 34.5Kv Line | On 5-15-2014 at 1920 hours (EDT), with Unit 1 & 2 operating at 100% power, the North Anna 34.5 kV Bus 5, offsite power feed to the 'C' Reserve Station Service Transformer, was lost which resulted in the loss of power to the Unit 1 'H' Emergency Bus and the Unit 2 'J' Emergency Bus. As a result of the power loss, the 1H Emergency Diesel Generator and the 2J Emergency Diesel Generator automatically started as designed and restored power to the associated emergency bus. During this event, the Unit 2 'A' Charging Pump, 2-CH-P-1A, automatically started as designed due to the loss of power event. The valid actuation of these ESF (Engineered Safety Feature) components due to the loss of electrical power is reportable per 10 CFR 50.72 (b)(3)(iv)(A). The Unit 1 'H' Emergency Bus off-site power source was restored to service and the 1H Emergency Diesel Generator was secured and returned to automatic. The Unit 1 Action Statement of Technical Specification 3.8.1 was cleared at 2115 hours on 5-15-2014. The Unit 2 'J' Emergency Bus power feed continues to be from the 2J Emergency Diesel Generator and this line-up will remain until the off-site power source can be restored to operable status. The Unit 2 'A' Charging Pump has been secured and returned to automatic. Both Units are in a stable condition. An investigation is underway to determine the cause of the 34.5 kV Bus 5 loss of power. Power was returned to the Unit 1 'H' Emergency Bus via the Unit 1 'B' Reserve Station Service Transformer. The licensee will be notifying local Louisa County officials and has notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 49784 | 2 February 2014 13:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Loss of the "a" Main Feedwater Pump | On 2-2-2014 at 0859 (EST), with Unit 2 operating at 100% power, a manual reactor trip was initiated by the control room staff following a trip of the 'A' main feedwater pump and automatic start of the 'C' feedwater pump due to crew concerns that both motors of the 'C' feedwater pump had not actuated. When the 'C' feedwater pump auto started, the running indicator light for one of the 'C' feedwater pump motors failed to illuminate. Both motors of the 'C' feedwater pump had started as designed. Following the reactor trip, all control rods fully inserted into the core and Unit 2 was stabilized in Mode 3 at normal reactor coolant system temperature and pressure. Decay heat is being removed using the normal condenser steam dump system. Unit 2 is in a normal shutdown electrical alignment with power being supplied from the Reserve Station Service Transformers. This event is reportable per 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system. Following the reactor trip, the auxiliary feedwater pumps automatically started as designed and provided makeup flow to the steam generators. The steam generator levels were returned to normal operating level and the auxiliary feedwater pumps were returned to the normal standby automatic alignment. This event is reportable per 10CFR50.72(b)(3)(iv)(A) for actuation of an ESF system. Unit 1 is operating at 100% power and was not affected by the event. The licensee informed the NRC Resident Inspector and will inform the Louisa County Administrator. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 49429 | 11 October 2013 17:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine and Reactor Trip Due to Station Service Transformer Lockout | At 1317 hours on 10/11/2013, Unit 1 experienced an automatic turbine and reactor trip from 48% power. Unit 1 was in the process of increasing power level following a refueling outage when the 1C Station Service Transformer Lockout Relay actuated as the 'C' Condensate Pump was started. The 1C Station Service Transformer Lockout resulted in the turbine trip which subsequently tripped the reactor. All three station service electrical buses transferred to the Reserve Station Service Transformers. The 1C Station Service Transformer does not have any visible exterior damage. All control rods fully inserted into the core following the reactor trip. The actuation of the Reactor Protection System is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater Pumps actuated as designed following the trip and provided makeup flow to the steam generators. The steam generator levels were returned to normal operating level and the Auxiliary Feedwater Pumps were returned to automatic. The actuation of the Auxiliary Feedwater Pumps is reportable per 10CFR50.72(b)(3)(iv)(A). Due to low decay heat loads, the Main Steam Trip Valves were closed as the Reactor Coolant Tavg temperature decreased, as directed by the reactor trip response procedure and decay heat is being removed using the atmospheric steam dumps. Decay heat control will be transferred to the main condenser steam dump system. Unit 1 is stable in Mode 3 at normal Reactor Coolant System temperature and pressure. Unit 2 is operating at 100% power and was not affected by this event. The licensee has notified the NRC Resident Inspector and the local government. | Steam Generator Reactor Coolant System Reactor Protection System Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 49075 | 28 May 2013 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 Due to a Feedwater Transient | On May 28, 2013, at 1507 (EDT), Unit 2 was manually tripped from approximately 98 percent power due to decreasing steam generator levels as a result of a main feedwater system transient. The main feedwater system transient was initiated when the 'C' Main Feedwater Pump Discharge Motor-Operated Valve, 2-FW-MOV-250C, spuriously closed. The cause of the spurious closure of 2-FW-MOV-250C is unknown at this time. The Operations crew entered the reactor trip procedure and stabilized Unit 2 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. The reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater (AFW) pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the AFW system is reportable per 10CFR50.72 (b)(3)(iv)(A) for a valid actuation of an ESF system. The AFW pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 2 is in the normal shutdown electrical line-up. Unit 1 was not affected by this event. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 49020 | 10 May 2013 10:12:00 | 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 Turbine Bearing Vibration | On May 10, 2013 at 0612 hours (EDT), Unit 2 was manually tripped from 60% power due to increased vibrations and a report of arcing on bearing #9 of the main turbine. Unit 2 was in the process of increasing power following a refueling outage when this event occurred. The Operations crew entered the reactor trip procedure and stabilized Unit 2 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10CFR50.72(b)(3)(iv)(A) for a valid actuation of an ESF system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 2 is in a normal shutdown electrical lineup. The #9 bearing is on the main generator exciter. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector and will be notifying local government agencies. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Turbine Control Rod | |
| ENS 48436 | 24 October 2012 05:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low Steam Generator Water Level | On 10/24/12 at 0147, North Anna Unit 2 reactor tripped automatically. The reactor first out is the 'C' steam generator lo-lo level. The turbine first out is reactor tripped, turbine trip. The event was apparently initiated by a loss of load on the secondary side. The cause of the loss of load is still being investigated. All systems responded as expected. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) due to actuation of the Reactor Protection System. The Auxiliary Feedwater pumps received an automatic start signal due to low-low level in all steam generators at the time of the trip, Steam generator levels have been restored to normal operating level. The Auxiliary Feedwater System operated as designed with no abnormalities noted. This event is reportable per 10 CFR 50.72(b)(3)(iv)(A) due to actuation of an ESF system. All control rods inserted into the core at the time of the trip and decay heat is being removed via the main condenser steam dumps. Several secondary (feedwater) relief valves lifted and reseated during the event. North Anna Unit 2 is currently stable at no load temperature and pressure in mode 3. At 0147 EDT, the Unit 2 Pressurizer Power Operated Relief Valve (PORV) , 2-RC-PCV-2455C, opened during an automatic reactor trip of Unit 2. The valve indicated open for less than 1 second. During this time, the identified leakage threshold for EAL SU6.1 (25 gpm) was exceeded. The cause of the loss of secondary load, which is believed to have caused the low steam generator water level and the lifting of the pressurizer PORV, is still under investigation. The licensee is focusing on the high pressure to low pressure turbine intercept valves or reheat valves going shut for reasons unknown at this time. The licensee's data shows that a pressurizer PORV opened momentarily. The instantaneous leak rate exceeded the unusual event threshold leak rate of 25 gpm. The PORV reseated and no ongoing leakage occurred during the transient. The rest of the transient was characterized as uncomplicated. The unit is in a normal post-trip electrical configuration. All systems functioned as required. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector.
The licensee is updating their report to RETRACT the portion related to the after-the-fact entry into EAL SU6. At 0147 hours EDT on 10-24-12, a Unit 2 Pressurizer Power Operated Relief Valve, 2-RC-PCV-2455C, opened during automatic reactor trip. The valve indicated open for less than 1 second. 2-RC-PCV-2455C opened as designed in response to the plant trip and allowed a small amount of water to transfer to the Pressurizer Relief Tank, as designed. The Pressurizer Power Operated Relief Valve subsequently re-closed and remains available for automatic operation, if needed. Initially, this issue was reported to the NRC at 0240 hours on 10-24-12 as an After-The-Fact Unusual Event for EAL SU6.1. Subsequent review has determined that the Pressurizer Power Operated Relief Valve functioned as designed and the small amount of inventory was transferred to the Pressurizer Relief Tank as designed and therefore does not meet the criteria for an Unusual Event and this notification is being retracted. NEI 99-01, Rev. 5 provides additional guidance that relief valve normal operation should be excluded from this Initiating Condition. However, a relief valve that operates and fails to close per design should be considered applicable to this Initiating Condition if the relief valve cannot be isolated. In this case, the Pressurizer Power Operated Relief Valve operated as designed and returned to automatic operation. The licensee informed state and local agencies and the NRC Resident Inspector. Notified R2DO (Musser). | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 46020 | 16 June 2010 23:20:00 | 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 a Lightning Storm | On 6-16-2010 at 1920 hours, Unit 2 experienced an automatic reactor trip/turbine trip from 98% power. A severe lightning storm was in progress at the time of the trip and a lightning strike appears to be the cause of the event. The reactor trip was actuated from Channel 1 and Channel 2 Over Temperature Delta T. All control rods fully inserted into the core during the trip. The control room staff responded to the trip in accordance with plant procedures and the unit is stable in Mode 3. This event is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps started as designed following the reactor trip and steam generator inventory was restored to normal operating level. The Auxiliary Feedwater pumps have been secured and returned to automatic. This event is reportable per 10CFR50.72(b)(3)(iv)(A) due to the ESF actuation. Decay heat is being removed by the condenser steam dump system. The 'A' loop wide range hot and cold leg thermocouples remain failed high and the 'B' loop wide range cold leg thermocouple also failed high during the event. The plant is in a normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector and will notify the local authorities. See EN #41898 for similar occurrence. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 45960 | 28 May 2010 04:03:00 | 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 Loss of Reactor Coolant Pump | A Unit-2 reactor trip was initiated by a loss of the Unit-2 'B' station service bus. The loss of the 'B' station service bus caused a reactor trip due to the loss of flow on one-of-three loops due to the loss of the 'B' Reactor Coolant Pump. The Auxiliary Feed Water system actuated as expected due to the reactor trip. The plant was stabilized in Mode 3 using the appropriate emergency procedure. During the transient, the 'B' Reserve Station Service Transformer de-energized and the Unit-2 'H' Emergency Diesel Generator was previously tagged out for planned maintenance. This resulted in the Unit-2 'H' emergency bus being de-energized. The alternate AC diesel generator has been placed in service and is providing power to the Unit-2 'H' emergency bus. The automatic tap changer for the 'C' reserve station service transformer did not work in automatic and had to be manually adjusted to control voltage. Unit-2 'C' Reactor Coolant Pump remains in service. All control rods fully inserted on the trip and no relief valves lifted or safety valves lifted in either the primary or secondary systems. The turbine drive and 'B' motor driven Auxiliary Feed Water pumps automatically started and injected into the 'A' and 'B' steam generators on a low level signal. The 'A' motor driven Auxiliary Feed Water pump failed to start due to the loss of the 'H' emergency bus. The 'C' steam generator is being controlled with main feed water though the 'C' main feed regulating valve bypass valve. Decay heat removal is via the condenser steam dumps. The licensee has notified the NRC Resident Inspector.
EN#45960 reported the RPS Actuation (50.72(b)(2)(iv)(B)) and AFW System Actuation (50.72(b)(3)(iv)(A). The event occurred at 0003 EDT on May 28, 2010. Technical Specification (TS) 3.0.3 was entered at 0004 hours on May 28, 2010, for inoperable offsite power sources with the 2H emergency diesel generator (EDG) being inoperable per TS 3.8.1. M. Update: At the time of the event, the station was experiencing a severe lightning storm. The Auxiliary Feedwater System was returned to auto standby at 0558 hours. At approximately 0942 hours, RCS cooldown to Mode 4 was started on Unit 2. Mode 4 was entered at 1245 hours. The 'A' and 'B' RCPs remain secured in Mode 4. Following repairs and post maintenance testing the 'C' reserve station service transformer (RSST) was declared operable at 1324 hours. This restored two (2) qualified offsite circuits for Unit 1 and one (1) qualified offsite circuit for Unit 2. TS 3.0.3 was cleared at this time on Unit 2. The 'B' RSST remains out of service (OOS) pending repairs and testing. The Unit 2 'B' station service bus remains OOS. The 2H EDG previously reported OOS for scheduled maintenance is expected to be returned to service on Monday, May 31, 2010. The alternate AC diesel generator continues to supply power to the 2H emergency bus. Limiting action remains for one (1) offsite circuit for Unit 2 being inoperable along with the 2H EDG OOS. The licensee will be notifying the NRC Resident Inspector. Notified R2DO (Haag). | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 45877 | 27 April 2010 20:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Following a Generator Lockout During Automatic Voltage Regulator Testing | On 4/27/2010 at 1637 hours, during recovery from a refueling outage when the main turbines and exciter were replaced, Unit 2 experienced a generator lockout which caused a turbine trip. The turbine trip resulted in a reactor trip when the reactor was at 74% power. The generator lockout occurred while automatic voltage regulator testing was being performed. This event is reportable per 10CFR50.72(b)(2)(iv)(B) due to actuation of the Reactor Protection System. The Auxiliary Feedwater Pumps (AFW) received an automatic start signal due to low/low level in the steam generators following the reactor trip. Steam Generator inventory was restored to normal operating level. The Steam Driven AFW pump experienced an issue with the lube oil system which resulted in some of the oil leaking onto the floor. An investigation into the oil leakage issue will be performed. The Steam Driven AFW pump was declared inoperable until this investigation is complete. The two motor driven AFW pumps automatically started and operated as designed. This event is reportable per 10CFR50.72(b)(3)(iv)(A) due to actuation of an ESF system. All control rods inserted into the core following the reactor trip and decay heat is being removed using the normal steam dump system. Several secondary (feedwater) relief valves lifted and resealed during the event. Unit 2 is stable in Mode 3 at normal operating temperature and pressure. Unit 2 is in a normal shutdown electrical lineup and there was no impact on Unit 1. The NRC Resident Inspector has been notified by the licensee. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Turbine Control Rod | |
| ENS 45556 | 9 December 2009 19:23:00 | 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 Unit 2 and Edg Start on Both Units Due to Inadvertent Loss of "C" Reserve Station Service Transformer | At 1423 hours on 12/9/2009, electrical supply breaker L102 was inadvertently opened which caused electrical Bus 3 and the 'C' Reserve Station Service Transformer to de-energize. This caused the loss of 'F' Transfer Bus which resulted in a loss of power to the 1H and 2J Emergency Busses and an automatic start of the 1H and the 2J Emergency Diesel Generators. Both emergency diesel generators started and re-energized their associated emergency bus as designed. The Unit 2 'G' Bus, which supplies power to the Unit 2 Circulating Water Pumps, did not automatically transfer to the 'B' Reserve Station Service Transformer in a sufficiently short time to prevent the loss of the Unit 2 Circulating Water pumps. The loss of the Unit 2 Circulating Water pumps resulted in an automatic low vacuum turbine trip and a subsequent (Unit 2) reactor trip due to the turbine trip. The 2 'G' Bus did automatically transfer to the 'B' Reserve Station Service Transformer and is currently energized. The Unit 2 Auxiliary Feedwater pumps automatically started and provided flow to the steam generators. There were no issues with the Auxiliary Feedwater System operation. The Unit 2 'A' Charging Pump and the Unit 2 'A' Component Cooling Water pump automatically started as designed due to the loss of power. The Unit 1 'B' Charging Pump and the Unit 1 'B' Component Cooling Water pump automatically started as designed due to the loss of power. The Unit 2 'C' Station Service Bus was lost following the trip when the electrical system automatically transferred to the Reserve Station Service transformers. With the 'C' Reserve Station Service Transformer de-energized the 'C' Station Service Bus was unable to transfer to an energized transformer. This resulted in the loss of the Unit 2 'C' Reactor Coolant Pump. The 'A' and 'B' Reactor Coolant Pumps remain in service at this time. The reactor trip is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater system, Emergency Diesel Generator system, Charging system actuations are reportable per 10CFR50.72(b)(3)(iv)(A). The electrical system is being returned to a normal lineup. The condensate and feedwater system remained in service to provide flow to the steam generators. Steam Dump operation to the condenser is not available due to low condenser vacuum, therefore steam is being released to the atmosphere from the Steam Generator Power Operated Relief Valves. The licensee suspects that switchyard maintenance activities caused the L102 trip which initiated the chain of events. All rods inserted into the core during the trip. During the transient, some secondary relief valves lifted and properly reseated. There is no known primary to secondary leakage. During the event call, the licensee reported that the 'C' Reserve Station Service Transformer was returned to service. The licensee notified the NRC Resident Inspector and will be notifying the Louisa County Administrator. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 45026 | 28 April 2009 13:53:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Auto-Start Due to Loss of the "B" Reserve Station Service Transformer | At 0953 hours on April 28, 2009, a loss of the 'B' Reserve Station Service Transformer resulted in a Degraded Voltage / Under Voltage automatic start of the Unit 2 'H' Emergency Diesel Generator. The Unit 2 'H' Emergency Diesel Generator is operating and supplying electrical power to the Unit 2 'H' 4160 Volt Bus. The Unit 2 'B' Charging Pump (2-CH-P-1B) and the Unit 2 'B' Component Cooling Water Pump (2-CC-P-1B) also auto started in response to the loss of power. (Procedure) 0-AP-10, 'Loss of Electrical Power', was entered to address the loss of the normal power source for the Unit 2 'H' Emergency Bus. Unit 2 is operating at 100% power and an investigation has been initiated to determine the cause of the event and appropriate corrective actions. The NRC Resident Inspector has been notified. There was no impact on Unit 1 operation. There was no loss of significant safety equipment as a result of the transformer loss. The licensee will continue to supply the 'H' 4160V bus with the 'H' Emergency Diesel Generator until compensatory measures are put in place. The loss of the 'H' bus places Unit 2 in a 72-hr shutdown action statement. | Emergency Diesel Generator | |
| ENS 43972 | 8 February 2008 23:12:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Control Rod Position Deviation | While withdrawing shutdown bank rods in preparation for a North Anna Unit 2 startup, a Rod Control system Urgent Failure alarm was received in the main control room. Further investigation revealed group step counters for Shutdown Bank 'A' deviated by 3 steps from demand. TRM 3.1.3 was entered and the Reactor Trip breakers were opened within the 15 minute requirement. The licensee notified the NRC Resident Inspector. | Control Rod | 05000339/LER-2008-001 |
| ENS 43866 | 26 December 2007 02:10:00 | 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 Trip of "B" Reactor Coolant Pump | On 12/25/07 at 2110 hours EST, Unit 2 tripped from 100% power due to a trip of the 'B' Reactor Coolant Pump. The reactor trip 1st out annunciator was 'Loss of flow, power >30%'. All control rods fully inserted. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater (AFW) pumps auto started due to the event and the steam driven AFW pump subsequently tripped on overspeed. The steam driven AFW pump was reset and placed in service. The ESF (Engineered Safety Function) actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). The unit is currently in mode 3 and (the licensee is) investigating the cause of the ground on the 'B' reactor coolant pump. The plant is at normal operating pressure and temperature. The electrical grid is stable and supplying plant loads through the startup transformer. Decay heat is being removed via the steam dumps to the condenser with feedwater being supplied via the normal path. The licensee has notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 43462 | 29 June 2007 21:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Trip and Spurious Safety Injection | At 1752 hours EDT Unit 2 received a 'B' train safety injection (SI). This spurious SI on 'B' train caused a trip of the main feedwater pumps and a turbine trip. The Unit 2 reactor tripped due to the turbine trip. The single train SI resulted in ECCS flow to the RCS. Both trains of SI were manually initiated per station procedures. The 'B' train of SI could not be reset and this resulted in RCS inventory increasing and lifting of the pressurizer PORV's. The pressurizer relief tank rupture disc ruptured and released water to the containment sump. SI flow to the core has been secured. Normal charging has been returned to service. This event is reportable per 10CFR50.72(b)(2)(iv)(A) for ECCS flow to the RCS. 10CFR50.72(b)(2)(iv)(B) for RCS Actuation (Rx/Turbine Trip). 10 CFR50.72(b)(3)(iv)(A) for AFW pump start, containment phase 'A' isolation, ECCS pumps actuation, and EDG starts. The AFW pump auto started during the event and operated as expected. Cause of the 'B' train SI is unknown at this time. All rods fully inserted. All systems functioned as required with the exception of the 'B' train SI which spuriously actuated and then could not be reset. All equipment started as expected from the SI actuation. AFW is still supplying cooling water to the steam generators at this time and decay heat is being discharged via steam dumps to the condenser. The licensee does not yet know how much water was discharged to the containment sump. The reactor is currently stable at no-load temperature and pressure with the level in the pressurizer a little high but tracking down to normal. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 43217 | 7 March 2007 08:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Autostart on Loss of "B" Reserve Station Service Transformer | The Unit-2 H Emergency Diesel Generator (EDG) automatically started on a Degraded Voltage/Under Voltage (DV/UV) signal due to the loss of the 'B' Reserve Station Service Transformer (RSST). The Unit-2 H Emergency Diesel Generator is supplying the Unit-2 H 4160 Volt Bus. Both Unit 1 and Unit 2 were stabilized using the appropriate abnormal procedures. During the event, the Unit 2 'B' Main Feed Water Pump motor, 2-FW-P-1B1, was noted to be running with the other motor, 2-FW-P-1B2, not running. The Unit 2 'B' Main Feed Water Pump was subsequently placed in Pull-to-Lock. Investigation continues as to the cause of the loss of the 'B' Reserve Station Service Transformer and the start of the Unit 2 'B' Main Feed Water Pump motor. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 43072 | 3 January 2007 23:03:00 | 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 a Card Failure | A Unit-1 reactor trip was initiated by a Process Rock card failure that caused the 'B' main feed regulation valve to fail closed. The closure of the 'B' main feed regulation valve caused a reactor trip due to a steam flow - feed flow mismatch with a low steam generator water level. The auxiliary feed water system actuated as expected due to the reactor trip. The plant was stabilized with no other issues using the appropriate emergency procedures. All control rods fully inserted on the trip and no relief or safety valves lifted in either the primary or secondary systems. Auxiliary feed water pumps automatically started and injected into the steam generators on a low water level signal. The operators restored normal feedwater flow to the steam generators. Decay heat is via the condenser steam dumps. The plant is aligned to the normal shut down electrical alignment. Card replacement is expected tonight and reactor startup is expected tomorrow. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 42996 | 16 November 2006 07:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip | At 0226, North Anna Power Station Unit 2 automatically tripped on steam flow greater than feed flow with low steam generator water level on 'B' Steam Generator. This was caused by a Steam Flow Channel (Channel 3) Low failure. After the reactor trip, Auxiliary Feedwater Pumps automatically started on Low-Low Steam Generator Level. All control rods fully inserted. RCPs are in operation transferring decay heat to the steam generators. The steam generators are discharging steam to the main condenser using the condenser steam dumps. Main feedwater pumps are running to maintain steam generator water levels. Unit 1 was not affected. The licensee will notify the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42483 | 8 April 2006 02:31:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Group Step Counter Failure During Reactor Startup | While in Mode 3 at 547 degrees and 2235 psig in the Reactor Coolant System, during Rod Control System rod drop testing, the group 2, 'A' shutdown bank step counter failed. The step counter is required to be operable in Modes 3, 4 and 5 per (Technical Requirements Manual requirement) 3.1.3 or the reactor trip breakers must be opened within 15 minutes. The step counter failed at 2217 and at 2231 the reactor trip breakers were opened. This was considered a valid actuation of the (Reactor Protection System) due to the (Technical Requirements Manual) requirements and due to the equipment malfunction. Shutdown margin is adequate and all emergency buses are on offsite power. Emergency Diesel Generators are available. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Emergency Diesel Generator | |
| ENS 41898 | 6 August 2005 02:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Apparently Due to Lightning Strike | The Unit 2 reactor automatically tripped due to an overpower-delta temperature (OpDeltaT) trip signal. The licensee states that an actual OpDeltaT condition did not exist at the time of the trip. The trip signal is believed to have been generated by lightning strikes from an electrical storm that was passing through the area at the time. The trip was uncomplicated and all systems functioned as required. All control rods fully inserted; no safety relief valves lifted; decay heat is being discharged to the main condenser using normal feedwater to supply the steam generators; the reactor temperature and pressure are at normal hot standby range. No obvious grid disturbance was seen during the trip and Unit 1 was not impacted (Unit 1 was being ramped offline at the time for secondary side maintenance work). The licensee noted that Auxiliary Feedwater did auto-start as expected due to a trip from full power and was subsequently secured. The licensee is still investigating the cause of the OpDeltaT trip signal but noted that other Unit 2 instrumentation was found failed after the transient including the Unit 2, A-loop, wide range T-hot and T-cold indications and the Unit 2, B-loop, T-cold indication. The licensee plans to remain in Mode 3 until the investigation is complete and instrument repairs completed. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Safety Relief Valve Main Condenser Control Rod | |
| ENS 40804 | 10 June 2004 17:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip While Performing Scheduled Solid State Protection Testing | A Unit 2 automatic reactor trip occurred while the licensee was performing planned periodic testing on train "A" solid state protection. All control rods fully inserted into the reactor core. The Auxiliary Feedwater Pumps automatically started as expected immediately following the reactor trip due to low-low level in the steam generators. The unit is being maintained stable in mode 3 and heat sink is being performed via steam dump to the condensers. All other systems functioned as required. The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector.
This is an update to event notification 40804. At 1313 hours on June 10, 2004, North Anna Unit 2 experienced an automatic trip from 100 percent during the performance of 2-PT-36.1A (Train 'A' Reactor Protection and ESF Logic Actuation Logic Test). The cause of the reactor trip, was determined to be an incorrect configuration of the cell switch (52h contract) on 'A' Reactor Bypass Breaker, 2-EP-BKR-BYA. The incorrect cell switch configuration resulted in a turbine trip signal being generated during testing which resulted in a reactor trip signal being generated in the 'B' train Reactor Protection System. The Auxiliary Feedwater System actuated in response to the event. Control room personnel responded to the event in accordance with emergency procedure E-0, Reactor Trip or Safety Injection. The control room team stabilized the plant using ES-0.1 Reactor Trip recovery. The lowest Reactor Coolant System (RCS) pressure during the event was 1988 psig and the lowest RCS temperature was 549 degrees. No human performance issues were identified during this event. A non-emergency four-hour report was made to the NRC operations center at 1611 hours pursuant to 10CFR50.72(b)(2)(iv)(B) for an actuation of the Reactor Protection System while critical. An eight-hour report was also made to the NRC in accordance with 10CFR 50.72(b)(3)(iv)(A) due to the Auxiliary Feedwater Pump starts (Engineering Safety Features Actuation). The Reactor Protection System, AMSAC (ATWAS Mitigating System Actuation Circuit), and the Auxiliary Feedwater System operated properly in response to the event. During the Unit 2 reactor trip, a blown output fuse on a logic card (that feeds the permissive for arming the Steam Dumps from loss of load) prevented the Main Steam Dump Valves from opening in Tavg Mode as expected. The Steam Generator Power Operated Relief Valves (PORVs) lifted and operated to control RCS temperature until transferring Steam Dump control to the Steam Pressure Mode. The fuse was replaced. A post trip review was conducted at 1500 hours on June 10, 2004. The cell switches on the Reactor Trip Bypass breakers have been repaired and post maintenance testing has been completed. Management approval was granted to start-up Unit 2. North Anna Unit 2 is currently in Mode 1 and is preparing to be placed on-line. The licensee notified the NRC Resident Inspector. Notified R2DO (Lesser) and NRR EO (Bateman). | Steam Generator Reactor Coolant System Reactor Protection System Auxiliary Feedwater Control Rod Main Steam | |
| ENS 40784 | 29 May 2004 06:02:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Control Rod Position Indication Failure During Hot Rod Drop Test | During performance of Hot Rod Drop Testing (and) when withdrawing 'D' Control banks, a failure of Group 1 Position Indication was identified. Entered action of Technical Requirement Manual (TRM) 3.1.3 and opened the Reactor Trip Breakers within 15 minutes per action (statement) of TRM 3.1.3. After the Reactor Trip Breakers were opened all Group 1 "D" Control Rods fully inserted into the core. There were no reactivity concerns since the reactor was borated with adequate shutdown margin. The failure of the position indicator has been identified and repaired. There were no other issues associated with this incident and the licensee will proceed with Hot Rod Drop Testing while at 0% reactor power and Mode 3. The licensee will notify the NRC Resident Inspector. | Control Rod |