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 Entered dateEvent description
ENS 509462 April 2015 06:55:00On 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.
ENS 5088413 March 2015 12:48:00On March 13, 2015 at 0928 EDT, a notification to OSHA (Occupational Safety and Health Administration) was initiated due to an employee experiencing a non-work related medical event that resulted in the employee passing. When the issue was identified, the station first aid team responded to administer first aid. Subsequent to the employee passing, a report was made to OSHA in accordance with federal requirements. This event is reportable to the NRC per 10 CFR 50.72(b)(2)(xi) since another governmental agency was notified of this employer referral medical event. The plant employee was in a building within the protected area and was not contaminated. The licensee notified the NRC Resident Inspector and will notify the local government of the event.
ENS 5085126 February 2015 16:39:00On 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.
ENS 5045715 September 2014 14:54:00

With North Anna Unit 2 in Mode 6 during a scheduled refueling outage, discharged assembly 4Z9 was identified as a failed fuel assembly by In-Mast Sipping. The fuel assembly was located in core location B11. Initial inspection of the fuel assembly identified two (2) visibly split fuel pins of eight (8) to ten (10) inches long with visible damage to the top of the pins. The internals of the affected pins are visible and the springs from the top of each pellet stack are touching the top nozzle. The fuel assembly has been placed into its designated location in the Spent Fuel Pool. No abnormal increase was noted on any radiation monitor either after or during fuel assembly movement. This fuel assembly had been used during three (3) previous operating cycles and is not scheduled for reuse. On September 15, 2014, at 0900 (EDT), subsequent video inspection of the fuel assembly identified that the top springs of the two (2) fuel pins were dislodged. Video inspection of the reactor vessel identified debris that has the potential to be fragments of fuel pellets resting on the core plate. Additional investigations are in progress. Due to the fact that the failure exceeded expected conditions, this event is being reported per 10 CFR 50.72(b)(3)(ii)(A), as any event or condition that results in the condition of the nuclear plant, including its principle safety barriers, being seriously degraded. The licensee has notified the NRC Resident Inspector and will notify local county authorities.

  • * * UPDATE FROM PAGE KEMP TO HOWIE CROUCH AT 1227 EDT ON 9/30/14 * * *

Event Notification #50457 was provided on September 15, 2014, at 1454 hours, pursuant to 10 CFR 50.72(b)(3)(ii)(A), to provide notification that North Anna Unit 2 discharged assembly 4Z9 had two visibly split fuel pins and debris on the core plate that had the potential to be fuel pellet fragments. Detailed video inspections estimated that fifteen (15) fuel pellets were dislodged from fuel assembly 4Z9. For reference, the reactor core contains approximately 15 million fuel pellets. Efforts to identify and recover the fuel pellets were performed. Debris fragments, estimated to represent five (5) fuel pellets, were located within the damaged fuel assembly that is currently in the spent fuel pool. In addition, an estimated three (3) pellets worth of material was retrieved by the foreign object search and retrieval (FOSAR) efforts in the reactor vessel. The remaining seven (7) fuel pellets have already or are expected to granulate into fine particles that will remain in low flow areas of the primary plant systems or be removed by normal purification processes. However, since the specific location of the seven (7) fuel pellets is undesignated, a report is being made pursuant to 10 CFR 74.11(a) for the loss of special nuclear material. The seven (7) fuel pellets contain licensed material in a quantity greater than 10 times the quantity specified in Appendix C of 10 CFR 20; therefore a report is also being made pursuant to 10 CFR 20.2201(a)(ii). The cause of the fuel clad degradation is understood and is being addressed. It has been evaluated that the dispersion of fuel pellet material will pose no threat to the integrity or operation of the reactor fuel and primary system components. Reactor Coolant System activity will remain below Technical Specification limits during power operation. In addition, there are no adverse radiological consequences to the public as a result of this issue. The licensee will be notifying the state of Virginia, local authorities in Louisa County and has notified the NRC Resident Inspector. Notified R2DO (Vias) and IRD (Stapleton).

ENS 5044811 September 2014 18:00:00On September 11, 2014, at 1400 hours, it was determined that four (4) personnel in the Emergency Response Organization (ERO) were not subject to random testing requirements of the Fitness for Duty (FFD) Program. The personnel involved do not have unescorted access to the Protected Area, but they do respond and perform duties as a member of the ERO. The affected individuals are now included in the random FFD testing pool. 10CFR26.4(c) requires all persons who are required by a licensee in 10CFR26.3(a) and, as applicable, (c) to physically report to the licensee's Technical Support Center or Emergency Operations Facility by licensee emergency plans and procedures shall be subject to an FFD program that meets all of the requirements of this part. This event is a 24-hour reportable event per 10CFR26.719(b)(4) - Any programmatic failure, degradation, or discovered vulnerability of the FFD program that may permit undetected drug or alcohol use or abuse by individuals within the protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program.
ENS 5011615 May 2014 23:10:00On 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.
ENS 497842 February 2014 11:01:00On 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.
ENS 4942911 October 2013 15:07:00At 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.
ENS 4907528 May 2013 18:09:00On 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.
ENS 4907227 May 2013 18:49:00At 1545 hours on 05/27/2013, the North Anna Control Room was notified by local authorities that a potential drowning had taken place at the number 3 Dike in Lake Anna. This incident has been reported to the FERC (Federal Energy Regulatory Commission) Regional Engineer under FERC requirements. Therefore, this is reportable to the NRC under 10CFR50.72(b)(2)(xi). In addition, this incident has received significant media interest. The identity of the victim is not known at this time. The licensee notified the NRC Resident Inspector and the Louisa County Administrator.
ENS 4902010 May 2013 08:20:00On 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.
ENS 4776725 March 2012 01:44:00On March 24, 2012, at 1855 (EDT) during the performance of work activities to support Alloy 600 dissimilar metal weld overlay work on the 'B' Reactor Coolant loop hot leg to the 'B' Steam Generator nozzle weld, two through-wall defects were identified. The workers noted a small amount of water seeping from the indications in the nozzle weld area. The indications are in the area of excavation that was being performed for the weld overlay project. Approximately 1 (inch) of weld material had been removed prior to the seepage being identified. Entered Technical Requirement 3.4 .6, 'ASME Code Class 1, 2 and 3 Components' and immediately initiated actions to isolate the 'B' Reactor Coolant loop. The 'B' Reactor Coolant loop stop valves were closed at 2312 hours on March 24, 2012, which isolated the defects from the reactor coolant system . An engineering evaluation of the defects will be performed and corrective actions implemented. This event is reportable in accordance to 10CFR50.72(b)(3)(ii)(A) for 'any event or condition that results in the condition of the nuclear power plant, including its principle safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector and will notify Louisa County.
ENS 4720126 August 2011 16:23:00

On August 23, 2011 at 1351 hours, North Anna Power Station experienced a seismic activity event which resulted in a loss of offsite power and automatic reactor trip of both units. At 1403 hours, an Alert was declared, based on Shift Manager judgment, due to significant seismic activity on the site. Subsequent to the earthquake, both units were stabilized and offsite power was restored. Following the event, seismic data was retrieved from the installed monitoring system and shipped to the vendor to determine the response spectrum for the event. On August 26, 2011 at 1340 hours, initial reviews of the data determined that the seismic activity potentially exceeded the Design Basis Earthquake magnitude value above 5 Hz. Therefore, this is reportable per 10CFR50.72(b)(3)(ii) (B) for the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. North Anna Unit 1 is currently in Cold Shutdown with the Residual Heat Removal System providing core cooling. North Anna Unit 2 is currently in Hot Shutdown and will be taken to Cold Shutdown with the Residual Heat Removal System providing core cooling. No significant equipment damage to Safety Related system (including Class 1 Structures) has been identified through site walk-downs nor has equipment degradation been detected through plant performance and surveillance testing following the earthquake. Therefore, there is reasonable assurance that the Safety Related systems are fully functional. The Spent Fuel Pit cooling system also remains fully functional and the temperature of the Spent Fuel Pit remained unchanged during the event. The vendor will complete the analysis of the seismic data and this information will be utilized to address the long term actions following the earthquake. The licensee notified the NRC Resident Inspector. 

* * * UPDATE FROM DON TAYLOR TO PETE SNYDER AT 1739 EDT ON 9/9/11 * * *

This is an update to EN 47201 reported on 8/26/2011 where It was reported that North Anna potentially exceeded the Design Basis Earthquake (DBE) magnitude value above 5 Hz. The vibratory motion from the 5.8 magnitude earthquake were recorded in all three orientations at several locations in the plant using two types of instruments: the Engdahl scratch plates that record 12 discrete spectral accelerations between 2 and 25.4 Hz, and the Kinemetrics analog recorders that recorded time histories of the accelerations. Based on evaluation of recorded plant data, it is concluded that the Central Virginia earthquake of 8/23/2011 exceeded the spectral accelerations for the Operational Basis Earthquake (OBE) and DBE of North Anna Plant. Extensive actions are underway to inspect. evaluate, test, and repair if necessary. systems and components to ensure they are capable of performing their required functions. To date, no significant damage to safety related structures, systems or components (SSC) has been identified. The licensee notified the NRC Resident Inspector. Notified R2DO (Rich).

ENS 4637729 October 2010 14:11:00This report is being made pursuant to 10 CFR 50.72(b)(2)(xi) for notification to other government agencies. Dominion North Anna Power Station intends to voluntarily notify state and local agencies regarding an increase in tritium levels in one (1) onsite ground water monitoring sample point. This increase in tritium levels has not exceeded any NRC regulatory dose limits nor has it exceeded the voluntary reporting limits (i.e., 20,000 picoCuries per liter) specified in NEI 07-07 Industry Ground Water Protection - Final Guidance Document. Two (2) adjacent onsite ground water monitoring sample points have not shown a similar increase. None of the eight (8) ground water monitoring sample points surrounding the station have shown any detectable levels of tritium. All indications show that the tritium in the one (1) onsite ground water monitoring sample point has not migrated to the lake or any drinking water sources. The station continues to monitor, sample and investigate the source of the tritium anomaly. This condition does not present a health hazard to station employees or the general public. Normal tritium levels at the particular sample point are 3-4000 picoCuries per liter. One sample read 16,500 picoCuries per liter. Samples afterwards have return to normal readings. The NRC Resident Inspector has been notified.
ENS 461535 August 2010 14:53:00At 1345 EDT, following maintenance on 1-CH�LCV-1115A, VCT divert valve, testing was commenced. At 1355 EDT, the valve failed to the full divert position and could not be repositioned from the normal control system, resulting in 75 gpm letdown flow being diverted to the gas stripper. At 1401 EDT, the normal letdown flowpath was isolated in accordance with plant procedures and the leakage isolated. The identified flow rate exceeded the threshold for entry into a Notice of Unusual Event under EAL tab SU6.1 due to identified leakage greater than 25 gpm. The licensee will inform State and local agencies and has notified the NRC Resident Inspector.
ENS 4609414 July 2010 19:53:00

At 1834 hours on 7/14/10, the Unit 1 'C' Reactor Coolant Loop was declared inoperable due to small unisolable leaks on the 'C' Steam Generator secondary side surface sample line. Two small through-wall flaws were identified in the piping upstream of 1 -SS-217, 'C' Steam Generator surface sample line manual isolation valve. The piping is Class 2 and the non-conforming condition could not be evaluated with the steam generator pressurized. Based on the condition of the piping and the inability to evaluate the flaw, the 'C' Steam Generator was declared inoperable per Technical Requirements Manual 3.4.6, ASME Code Class 1, 2 and 3 Components. Subsequently, Technical Specification 3.4.4 was entered to place Unit 1 in Mode 3 within 6 hours. At 1934 hours on 7/14/10, North Anna Unit 1 initiated a shutdown in accordance with Technical Specification 3.4.4. The unit will be shutdown and the line will be evaluated and repaired. The licensee is presently at 82% power and coasting down in power. All safety systems are fully operable. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM PAUL TRENT TO DONALD NORWOOD AT 0015 HRS ON 7/15/2010 * * *

North Anna Unit 1 entered mode 3 at 2353 hrs. There were no complications during shutdown. One source range monitor failed downscale low. The other source range monitor is operating correctly. The failure of this source range monitor did not affect shutdown capabilities. Notified R2DO (Seymour).

ENS 4602016 June 2010 21:08:00On 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.
ENS 4587727 April 2010 18:44:00On 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.
ENS 4562311 January 2010 16:53:00A verbal report was made to the Federal Energy Regulatory Commission (FERC) Engineer on January 11, 2010 at 1510 hours pursuant to 18CFR12.10(a) for a condition affecting the safety of a project or works. A new wet area has been identified on the slope of Dike 6, approximately 25 feet from the top of the structure. Dike 6 was designed and created during original plant construction to support electrical towers and provide isolation between Lake Anna and the proposed Unit 3 & 4 intake structures. The Unit 3 & 4 intake area is currently being used as a settling pond. The settling pond is used to process storm water and other secondary water before the liquid is discharged from the station. This minor seepage is located about half way down the slope to the settling pond. Water samples were obtained from the wet area and it was determined that the seepage was not ground water. The sample composition was similar to the water in Lake Anna. Currently, there is no threat to the public or North Anna Power Station. Engineering is investigating this issue and a corrective action plan will be developed to correct the minor seepage. This issue is being reported pursuant to 10CFR50.72(b)(2)(xi). The licensee has notified the NRC Resident Inspector and the local government.
ENS 455569 December 2009 17:15:00At 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.
ENS 4502628 April 2009 12:00:00At 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.
ENS 4501323 April 2009 14:47:00

At 1410 hours on April 23, 2009, it was identified that an ALERT classification had not been declared on April 22, 2009 as required by EPIP-1.01. North Anna Emergency Plan, Emergency Action Level H2.1 requires the declaration of an ALERT for a fire or explosion in any safe shutdown area and either plant personnel report visible damage to any safety-related structure, system or component within the area or affected system parameter indications show degraded performance. A description of the event is provided below. On April 22, 2009, at approximately 0500 hours, Operations personnel identified a strong odor in the North Anna Unit 1 Cable Vault area. Subsequent investigation identified that the odor was coming from circuit breaker 01-EE-BKR-1J1-2S-J1 associated with the "D" Control Rod Drive Mechanism (CRDM) Fan (1-HV-F-37D). Operations personnel locally opened the circuit breaker to place it in a safe condition. 1-HV-F-37D had tripped approximately 30 minutes prior to the event. 1-HV-F-37D is not safety-related and not required for safe shutdown however; the supply breaker is safety-related since it is located on an emergency bus. Operation personnel then opened the circuit breaker cabinet and a small (6-inch) flame was observed. Operations personnel used a CO2 extinguisher on the internals of the circuit breaker to quickly extinguish the small fire. Appropriate levels of management were informed. The breaker has been quarantined. The cause of the circuit breaker failure has not been identified. A Root Cause Evaluation is in progress. There were no injuries. The plant continues to operate at full power. As a result of identifying that the criterion for the EAL was exceeded and no longer exists, a notification is being made to the NRC Operations Center in accordance with 10CFR50.72(a)(1)(i). The NRC Resident Inspector has been notified and the State and local governments will be notified.

  • * * RETRACTION FROM KEMP TO SANDIN AT 1105 ON 07/09/09 * * *

On April 22, 2009, at approximately 0500 hours, operations personnel identified a strong odor in the North Anna Unit 1 Cable Vault area. Subsequent investigation identified that the odor was coming from circuit breaker 01-EE-BKR-1J1-2S-J1 associated with the 'D' Control Rod Drive Mechanism (CRDM) Fan (1-HV-F-37D). Operations personnel locally opened the circuit breaker to place it in a safe condition. 1-HV-F-37D had tripped approximately 30 minutes prior to the event. 1-HV-F-37D is not safety-related and not required for safe shutdown however; the supply breaker is safety-related since it is located on an emergency bus. Operation personnel then opened the circuit breaker cabinet and a small (6-inch) flame was observed. Operations personnel used a C02 extinguisher on the internals of the circuit breaker to quickly extinguish the small fire. A root cause evaluation is in progress. At 1447 hours on April 23, 2009, a one hour notification was made to the NRC Operations Center in accordance with 10CFR50.72(a)(1)(i), which identified that the criterion for a ALERT EAL was exceeded due to the small fire in the circuit breaker and subsequent damage to the breaker internals. The notification also stated that the condition no longer exists. Subsequent reviews have determined that the 'Initiating Condition' for the Emergency Action Level was not met and the event was not required to be classified as an ALERT. The initiating condition states -Fire or explosion affecting the operability of plant safety-related structures, systems or components required to establish or maintain safe shutdown. The 'D' Control Rod Drive Mechanism is not required to establish or maintain safe shutdown and the emergency bus remained operable during the event. The notification made to the NRC on April 23, 2009 is being retracted. The NRC Resident Inspector has been notified. Notified the R2DO (Nease).

ENS 443937 August 2008 17:26:00At 1430 hours on 8-7-08, approximately 20 individuals from a group identified as the Southeast Convergence for Climate Action arrived at the North Anna Nuclear Information Center (NANIC) and initiated a civil disturbance within the center. (The NANIC is located just inside the Owner Controlled Area). When the NANIC closed at 1600 hours, 8 of the individuals refused to leave the building. Local Law enforcement informed the individuals that they would be placed under arrest, and 2 individuals exited the area. The remaining 6 individuals were detained by Local Law Enforcement Officers and removed from Dominion property. Licensee has notified the NRC Resident Inspector.
ENS 4386625 December 2007 23:22:00On 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.
ENS 4346229 June 2007 20:45:00At 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.
ENS 4142621 February 2005 09:37:00On February 21, 2005, at 0250 hours (EST), the Unit 1 Plant Computer System (PCS) failed rendering the Safety Parameter Display System (SPDS) inoperable. The PCS and SPDS were restored at 0650 hours. This event is a major loss of emergency assessment capability for greater than one hour and is reportable under 10 CFR 50.72(b)(3)(xiii). The licensee will notify the NRC Resident Inspector.
ENS 4078429 May 2004 09:55:00During 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.
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