Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 57017 | 8 March 2024 12:42:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | The following information was provided by the licensee via email and phone call: A non-licensed supervisor had a confirmed positive fitness for duty test. Unescorted access for the individual has been denied at all Dominion Energy sites. The NRC Senior Resident Inspector has been notified. | |
| ENS 56774 | 3 October 2023 15:54:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition | The following information was provided by the licensee via email: At 1154 EDT on 10/03/23, investigation into a boric acid indication was determined to be through a leak on a weld-o-let upstream of a pressurizer level transmitter isolation valve. Unit 2 is currently in MODE 6 with reactor coolant system (RCS) operational leakage limits not applicable. The leak is not quantifiable as it only consists of a small amount of dry boric acid at the location. The failure constitutes welding or material defects in the primary coolant system that are unacceptable under ASME Section XI. Therefore, this is a degraded condition reportable under 10 CFR 50.72(b)(3)(ii)(A). This condition does not affect the health and safety of the public or station employees. The Resident Inspector was notified. | |
| ENS 56743 | 18 September 2023 18:20:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification | The following information was provided by the licensee via email: On 09/17/2023 at 2218 (EDT), Operations identified that the bearing cooling (BC) tower basin was overflowing. Earlier in the day, the BC tower was isolated as part of a planned maintenance evolution and the overflow condition was due to isolation valve leak-by. At 2255, the leak-by was corrected and stopped the overflow. Approximately 75 gallons may have been discharged to the lake from the overflow. The BC water was sampled by Chemistry and all chemical parameters were within VP DES (Virginia Pollutant Discharge Elimination System) limits. At 1420 on 09/18/23, a 24-hour notification was made to the Virginia Department of Environmental Quality (DEQ) in accordance with the North Anna VPDES permit. This issue is being reported per 10CFR50.72 (b)(2)(xi) due to the notification of other government agency. The NRC Resident Inspector was notified. | |
| ENS 56618 | 11 July 2023 19:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification | The following information was provided by the licensee via email: At 1530 (EDT) on 7/11/2023, North Anna Power Station notified the Virginia Department of Environmental Quality (DEQ) that a small volume of filtered/purified water potentially discharged into Lake Anna from a leak from a reverse osmosis unit. The leak did not follow the normal release path for discharge through outfall 013. No environmental impact associated with this leak was observed or would be expected because the water in question is cleaner than the lake water, and would have met all discharge requirements for outfall 013. The NRC Resident Inspector was notified. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). | |
| ENS 56787 | 18 June 2023 04:00:00 | 10 CFR 50.73(a)(1), Submit an LER | 60 Day Notification for an Invalid Specified System Actuation | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid specific system actuation of the North Anna Power Station Unit 1 Emergency Core Cooling System (ECCS). On 6/18/2023, a comparator card power supply associated with 1-CH-PC-1121A, charging pressure low-standby pump start signal comparator, failed and caused the `A' and `B' charging pumps to auto-start and the previously running `C' charging pump to trip and lock-out. This event is considered an invalid system actuation because the actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. The ECCS pumps functioned as expected in response to the actuation. The `A' Charging pump was shut down in accordance with plant procedures following replacement of the comparator card. There was no impact on the health and safety of the public or plant personnel. The reportability requirement was determined beyond the 60-day notification requirement on 9/21/2023. The NRC Resident Inspector has been notified. | |
| ENS 56374 | 20 February 2023 21:18:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to State Agency | The following information was provided by the licensee via fax: At 1618 EST on February 20, 2023, North Anna Power Station notified the Virginia Department of Environmental Quality (DEQ) that water discharged into Lake Anna following hydrostatic testing of tanks associated with a new on-site sewage treatment plant had exceeded the project's general permit (VAG83) pH value. Hydrostatic test water discharge activities to Lake Anna began on February 20, 2023 at 0900 EST. A pH sample was collected at 0955 EST on February 20, 2023 and determined to have a pH of 9.1 which exceeded the maximum permit pH of 9.0. Discharge ceased after the reading was collected. Approximately 354 gallons were discharged to Lake Anna. A follow-up ambient pH sample result of 7.8 was collected on February 20, 2023 at 1401 EST from Lake Anna in the vicinity of the discharge pipe. No evidence of dead fish, foam, or other negative environmental impacts were observed. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi) for 'Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an on-site fatality or inadvertent release of radioactively contaminated materials.'
The following is a summary of information provided by the licensee via telephone: The licensee called to correct the pH sample results from a pH of 9.1 to a pH of 9.93. Notified R2DO (Miller) via phone | |
| ENS 56088 | 4 September 2022 23:39:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Security Event - Unauthorized Individual in Owner Controlled Area | At 1939 EDT, the North Anna Power Station Units 1/2 declared a Notice of Unusual Event (NOUE) under emergency declaration HU1.1 confirmed security event. Both units were unaffected by the event. The licensee exited the NOUE at 2036 EDT. | |
| ENS 56073 | 27 August 2022 14:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Turbine Trip Due to Main Transformer Fire | The following information was provided by the licensee via email: At 0810 EDT on August 27, 2022, with Unit 2 at 27% power, the operating crew received an annunciator for a Turbine Trip Without Reactor Trip. At 0812 EDT, a report came in from the field of a fire in the north yard due to an "A" Main Transformer upper bushing failure. The station fire brigade was dispatched and offsite assistance was requested. However, at 0842 EDT the fire was put out, prior to needing the offsite assistance. No Emergency Action Level threshold was exceeded for this event. The switchyard is in a normal alignment for providing offsite power to Unit 2. At 1015 EDT, the Virginia Department of Emergency Management (VDEM) was notified of the event. Additionally, a notification to the Virginia Department of Environmental Quality will be made due to approximately 100 gallons of oil reaching the ground. As such, this issue is being reported per 10 CFR 50.72(b)(2)(xi) for "'Any event or situation for which a news release is planned or notification to other government agencies has been or will be made.' The NRC Resident Inspector was notified. | |
| ENS 55873 | 3 May 2022 12:19:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report | The following information was provided by the licensee via email: A non-licensed Dominion Energy supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | |
| 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. | |
| ENS 55570 | 10 November 2021 20:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Pcb Oil Spill | At 1515 EST on 11/10/21, approximately 89 gallons of PCB oil is unaccounted for from the Substation 'N' Transformer, located in the Owner Controlled Area. Transformer nameplate oil capacity is 569 gallons. Prior to removal of the original Substation 'N' Transformer, approximately 475 gallons of 10-CA-OIL (PCB Oil) was evacuated and stored by HEPACO (a licensee vendor). Approximately 5 gallons of oil is inaccessible to evacuate and remains in the original transformer. Below the transformer was evidence of oil leakage to the ground. The leakage appears to have been occurring over time, not as a result of a catastrophic failure. This condition is reportable to the Virginia Department of Environmental Quality (VA DEQ). The VA DEQ was notified of this condition at 1815 on 11/10/21. Cleanup activities are on-going. This event is reportable in accordance with 10CFR50.72(b)(2)(xi) for 'Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an on-site fatality or inadvertent release of radioactively contaminated materials.' The licensee will be notifying the Louisa County administrator and has notified the NRC Resident Inspector.
Initial, unofficial, field testing performed by HEPACO indicated the oil released to the ground was PCB-Contaminated Oil. The official test results from the lab indicated that the oil is in fact not classified as PCB-Contaminated Oil. Therefore, this update is being made to EN55570 to clarify that the oil released to the environment was not PCB-Contaminated Oil. The licensee notified the NRC Resident Inspector and the VA DEQ of this update. Notified R2DO (Miller). | |
| ENS 55457 | 12 September 2021 21:28:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Identified While Unit Shutdown | On September 12, 2021, at 1728 EDT, with Unit 1 in Mode 5 (Cold Shutdown) while performing inspections of the North Anna Power Station Unit 1 reactor vessel head flange area, a weld leak was identified on the reactor vessel flange leak-off line that connects to the flange between the inner and outer head o-rings. Entered TRM 3.4.6 Condition B for ASME Code Class 1,2, and 3 components. With known leakage past the inner head o-ring, this condition is reported since the fault in the tubing is considered pressure boundary (Reactor Coolant System) leakage. This event is reportable in accordance with 10 CFR 50.72(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 NRC Resident has been notified.
The condition identified in EN 55457, pursuant to 10 CFR 50.72 (b)(3)(ii)(A) has been evaluated, and has been determined not to be Reactor Coolant System (RCS) pressure boundary leakage. As such, the 8-hour report is being retracted, as it is not an event or condition that results in, 'the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The leakage was subsequently determined to be in a tubing connection downstream of the reactor vessel inner O-ring. Leakage past a seal or gasket is not considered to be pressure boundary leakage, as defined by Technical Specifications. The NRC Resident Inspector has been notified. Notified R2DO (Miller) | |
| ENS 55375 | 22 July 2021 21:51:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition of Fire Safe Shutdown Equipment | On July 20, 2021, at 1707 EDT, an apparent non-compliance with 10 CFR 50, Appendix R, section III.G.2 (separation of redundant fire safe shutdown equipment) was identified. This issue was initially categorized as not affecting train separation or the ability of the equipment to perform their Design Basis functions. The original concern was entered into the licensee's Corrective Action Program as CR1177199. Subsequently, on July 22, 2021, at 1751 EDT, a further review of the affected control circuits for the Unit 1 and Unit 2 Emergency Diesel Generator (EDG) output breakers and emergency bus feeder breakers identified a concern that breaker position interlocks routed to or through non-safety related components or spaces may affect the ability to provide emergency power on the affected unit due to impacts on the control power circuits during an Appendix R fire associated with a loss of offsite power. The following are the affected fire areas: - Unit 1 and Unit 2 Turbine Buildings - Unit 1 and Unit 2 Cable Spreading Rooms - Unit 1 and Unit 2 Normal (307) Switchgear Rooms This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60-day written report pursuant to 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as CR 1177399. The NRC Resident Inspector has been notified of this event. | |
| ENS 55239 | 6 May 2021 16:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | On May 6, 2021 at 1223 (EDT), Unit 1 was manually tripped from 60 percent power due to degrading main condenser vacuum. Unit 1 was in the process of decreasing power due to increased secondary sodium levels identified earlier in the day. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. This reactor protection actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). 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 pumps is reportable per 10 CFR 50,72(b)(3)(iv)(A) for a valid actuation of an ESF (Engineered Safeguards Features) system. Decay heat is being removed by the condenser steam dump system. The electrical system is in normal lineup for shutdown conditions. There was no effect on Unit 2 operation. The NRC resident inspector has been notified. | |
| ENS 54921 | 28 September 2020 20:00:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report | A non-licensed Dominion Energy supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | |
| ENS 54652 | 9 April 2020 05:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Reactor Coolant System Pressure Boundary Leakage | On April 9, 2020 at 0100 EDT, while performing a containment walkdown due to a small increased Reactor Coolant System (RCS) unidentified leakage, a leak was identified on the 'A' Reactor Coolant Pump (RCP) seal injection piping. The source of the leakage cannot be isolated and is considered RCS pressure boundary leakage. At that time, Condition B of Technical Specification (TS) LCO 3.4.13, 'RCS Operational Leakage' was entered due to pressure boundary leakage. TS 3.4.4 'RCS Loops - Mode 1 and 2' and Technical Requirement (TR) 3.4.6 'ASME Code Class 1, 2, and 3 Components' are also applicable. Unit 2 is projected to be taken to Mode 5 for repairs. This event is reportable in accordance with 10 CFR 50.72(b)(2) for 'Initiation of plant shutdown required by Technical Specifications' and 10 CFR 50.72(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. There is no effect on Unit 1 | |
| ENS 53909 | 3 March 2019 14:16:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notice of Unusual Event Due to Unconfirmed Fire Alarm | At 0916 EST on March 3, 2019, North Anna Unit 2 declared a Notice of Unusual Event under Emergency Action Level HU 2.1 (fire in/or restricting access to any table H-1 area not extinguished within 15 minutes of control room notification or verification of a control room alarm). At 0906 the control room received a heat sensor alarm for the Unit 2, Reactor Coolant Pump motor cube. The fire brigade was dispatched to the scene where they found no indication of fire, no smoke and no fire damage. There were no actuations associated with the alarm and no redundant indications of fire. There was no effect on plant equipment and no indications of RCS leaks. The site determined that the alarm was invalid and terminated the NOUE. Unit 2 is in a stable condition and in a normal electrical lineup. Offsite support was not requested. The NRC Resident Inspector, State, and local authorities have been notified by the licensee. Notified R2RA (Haney), DNRR (Evans), IRD MOC (Grant), R4RDO (Rose), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). | |
| ENS 53908 | 3 March 2019 03:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Degrading Main Condenser Vacuum | On March 2, 2019 at 2237 EST, North Anna Unit 2 reactor was manually tripped, while operating at approximately 12 percent power, due to degrading vacuum in the main condenser. The unit was in the process of a planned shutdown for refueling when condenser vacuum degraded to greater than 3.5 inches of mercury absolute. 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. The reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). There were no ESF system actuations. Decay heat is being removed by the Steam Generator Pressure Operated Relief valves. Unit 2 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified. | |
| 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. | |
| ENS 53268 | 18 March 2018 00:07:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty Report Involving Discovery of Kombucha Tea Inside the Protected Area | At 2007 (EDT) on 3/17/18, a security Officer reported finding a container of herbal tea (Kombucha) on a platform in the Unit 1 Emergency Switchgear Room, which is located inside the Protected Area. Kombucha tea is a fermented tea containing trace amounts of alcohol, and is legally sold without restrictions. Dominion had previously notified its workforce that Kombucha tea was prohibited from being consumed or carried on-site. This is considered an alcoholic beverage and is being reported under the requirements of 10 CFR 26.719. The individual who brought the beverage on-site was identified and escorted out of the Protected Area. The NRC Resident Inspector has been notified. The licensee will also be contacting the County Administrator for Louisa County, Virginia. | |
| ENS 52438 | 15 December 2016 16:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Made for Potential Fuel Oil Leakage | On 12/14/16 at 1206 (EST), 0-PT-89.9K (underground fuel-oil piping pressure test), was performed UNSAT after a failed attempt to maintain pressure in the supply line to the Unit 2 'H' Emergency Diesel Generator (EDG) Day Tank. The exact source of the leakage is unknown at this time but is reasonable to believe some fuel oil was released underground. The associated Fuel Oil line is currently tagged out and isolated. This condition is reportable to the Virginia Department of Environmental Quality (VA DEQ) as part of the Underground Storage Tank Program. Pressure tasting of the other EDG fuel oil supply lines has been previously completed satisfactorily. The 2H EDG remains Operable as the redundant fuel oil transfer pump and its fuel oil piping are Operable and capable of maintaining adequate day tank level. The VA DEQ was notified of this condition at 1130 (EST) on 12/15/16. This event is reportable in accordance with 10CFR50.72(b)(2)(xi) for 'Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the NRC Resident Inspector and will notify Louisa County.
This is a follow-up report to Event Number 52438, made on 12/15/2016, regarding offsite notification to the Virginia Department of Environmental Quality for potential fuel oil leakage from the supply line to the Unit 2H Emergency Diesel Generator (EDG) Day Tank. During investigation of the potential fuel oil leak from the 2HB EDG Day Tank supply line (as previously reported in EN 52438), the 2HA fuel oil line, which runs close to the 2HB line, was disturbed and began to leak a mist of fuel oil. The fuel oil was contained in the area and was being cleaned via vacuum truck as it was leaking. Personnel at the scene noted the soil was not contaminated with fuel oil initially and saw the leak start on 2HA line. It has been estimated that less than one gallon of fuel oil was released to the surrounding soil during troubleshooting of the leak. The fuel oil was immediately vacuumed. The 2HA line was isolated and the leakage was stopped. The 2HA line is to be repaired and tested. The 2H EDG remains available, however, it is considered inoperable at this time. Investigation of the 2HB fuel oil line continues. The condition of the 2HA fuel oil line was reported to the Virginia Department of Environmental Quality (VA DEQ) as part of the Underground Storage Tank Program on 12/28/16. Therefore, this event is reportable in accordance with 10CFR50.72(b)(2)(xi) for 'Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the NRC Resident Inspector and will notify Louisa County. Notified R2DO (Rose). | |
| ENS 52137 | 30 July 2016 15:52:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Unisolable Reactor Coolant System Boundary Leakage | On July 30, 2016 at 1152 hours (EDT) following a containment walkdown to investigate an increase in RCS unidentified leakage to 0.15 gpm, a leak was identified on the seal return line from 2-RC-P-1C, 'C' Reactor Coolant Pump. The source of the leakage cannot be isolated and is considered RCS pressure boundary leakage. (Technical Specification) LCO 3.4.13, RCS Operational Leakage, Condition B for the existence of pressure boundary leakage was entered. Technical Requirement TR 3.4.6, ASME Code Class 1, 2, and 3 Components is also applicable. Unit 2 is projected to be taken to Mode 5 for repair. This event is reportable in accordance with 10 CFR 50.72(b)(2) for 'the initiation of any nuclear plant shutdown required by the plant's Technical Specifications' and 10 CFR 50.72(b)(3)(ii)(A) for 'any event or condition that results in the condition of the nuclear plant including its principal safety barriers, being seriously degraded.' The licensee will be notifying the Louisa County Administrator and has notified the NRC Resident Inspector. | |
| 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. | |
| 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 | |
| ENS 51557 | 20 November 2015 13:23:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Inoperable Vent Stack Radiation Monitor | At 0823 EST on 11/17/2015, the Unit 1 'A' Vent Stack radiation monitor, 1-VG-RI-179, was declared non-functional due to a faulty heat trace circuit. Compensatory measure to perform grab samples every 6 hours was implemented. At 0823 EST on 11/20/2015, the Unit 1 'A' Vent Stack radiation monitor, 1-VG-RI-179, had been out-of-service for 72 hours. The loss of 1-VG-RI-179 is being reported per 10 CFR 50.72(b)(3)(xiii) as a loss of emergency assessment capability. Corrective actions continue to be pursued to restore 1-VG-Ri-179 to functional status. The NRC Senior Resident Inspector has been notified by the licensee.
The Unit 1 'A' Vent Stack radiation monitor, 1-VG-RI-179, was returned to service at 1440 EST on 11/21/15. The licensee has notified the NRC Resident Inspector. Notified R2DO (Ernstes).
The purpose of this report is to retract the event notification report made in accordance with 10 CFR 50.72(b)(3)(xiii) on November 20, 2015 at 1534 EST (EN# 51557). After further review it has been determined that the performance of grab samples is an approved back-up method for radiological assessment capabilities as described in the North Anna Emergency Plan implementing procedure EPIP-4.24 Gaseous Effluent Sampling During Emergency. During non-emergencies, VPAP-2103N Offsite Dose Calculation Manual governs grab sampling and is tracked by Operations using 1-LOG-14 Non-Routine Surveillance Log. As such, a loss of radiological assessment capability did not exist and the ability to assess EAL RU1.4 was not affected. This is consistent with NUREG 1022, Rev.3, Supplement 1 and NEI 13-01, Rev. 0. The action was cleared at 1440 hours on 11/21/15 and the Unit 1 'A' Vent Stack radiation monitor, 1-VG-RI-179, was returned to functional status. The NRC Senior Resident lnspector has been informed of this event notification retraction. Notified R2DO (Masters). | |
| ENS 51460 | 8 October 2015 01:47:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | High Energy Line Break Door Found Unlatched | At approximately 2147 EDT on October 7, 2015, a high energy line break (HELB) door between the Turbine Building (TB) and the safety related Emergency Switchgear Room (ESGR) was determined to be unlatched. The door was immediately closed (latched). Investigation determined the door was unlatched for approximately 47 minutes. At 1617 EDT on October 8, 2015, it was determined the Unit 2 ESGR was outside of the design analysis for a Unit 1 HELB. A high energy line break in the TB with the door open could result in equipment in the Unit 2 ESGR experiencing high temperature, pressure, or humidity beyond conditions analyzed for equipment qualification which has the potential to render redundant safety-related equipment inoperable. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition and in accordance with 10 CFR 50.72(b)(3)(v)(A) & (B) & (D) as a condition that could have prevented the fulfillment of safety functions to shutdown the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident. The NRC Senior Resident Inspector has been notified. | |
| 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. | |
| ENS 50884 | 13 March 2015 13:28:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to Osha Due to Onsite Plant Employee Passing Away | On 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 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. | |
| ENS 50700 | 23 December 2014 03:30:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unit 1 Technical Specification Shutdown Due to Rcs Pressure Boundary Leakage | On December 22, 2014 at 2230 hours (EST) while performing a containment walkdown due to increased RCS (Reactor Coolant System) unidentified leakage, a leak was identified upstream of 1-RC-68, B Loop Cold Leg Drain Isolation Valve. The source of this leakage cannot be isolated and is considered RCS pressure boundary leakage. (Unit 1) Entered TS LCO 3.4.13 RCS Operational Leakage, Condition B for the existence of pressure boundary leakage. TS 3.4.4 RCS Loops - Modes 1 and 2 condition A, TR 3.4.6 ASME Code Class 1, 2, and 3 components, Condition B. Unit 1 will be taken to Mode 5 for repair. This event is reportable in accordance with 10 CFR 50.72(b)(2) for "initiation of plant shutdown required by Technical Specifications" and 10 CFR 50.72(b )(3)(ii)(A) for "any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The RCS leakage has been quantified as 0.053 gallons per minute from a containment sump in-leakage calculation. The exact location of the leak has not been identified due to the installation of lagging on the RCS components. The licensee anticipates entering Mode 3 (Hot Standby) within the next 30 minutes. There is no safety-related equipment out-of-service at this time. The licensee will inform Louisa County and has informed the NRC Resident Inspector. | |
| ENS 50666 | 10 December 2014 18:44:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Inadvertent Loss of Vital Indication During Maintenance | On December 10, 2014, at 1344 (EST), channel 1 of Refueling Water Storage Tank (RWST) level indication failed low during maintenance activities on channel 2. At that time, operators entered abnormal procedure 1-AP-3, 'Loss of Vital Indication, on Unit 1.' Additionally, Technical Specification (TS) 3.0.3 was entered due to 2 channels inoperable that affect Recirculation Spray (RS) pump auto-start capability. Had a Containment Depressurization Actuation (CDA) occurred during the time that both channels were inoperable, accident mitigation would have been adversely impacted. At 1356, both level indications returned to normal. At 1417, channel 2 was declared Operable and TS 3.0.3 was cleared. At 1439, channel 1 was declared Operable. Therefore, this 8-hour report is being made per 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function due to the RS pump auto-start concern. Technicians inadvertently went to the incorrect channel (Channel 1) during planned maintenance activity of Channel 2, causing the loss of both channels simultaneously for a short period of time. The licensee has notified the NRC Resident Inspector and will be notifying the Louisa County Administrator. | |
| ENS 50457 | 15 September 2014 13:00:00 | 10 CFR 20.2201(a)(1)(ii) 10 CFR 74.11(a) 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Failed Fuel Assembly Identified During Core Off-Load | 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.
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 50448 | 11 September 2014 18:00:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Programmatic Failure | On 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 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. | |
| 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. | |
| ENS 49672 | 23 December 2013 13:23:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Communications with Virginia State Eoc Due to Power Outage | At 0828 EST on 12/23/2013, the North Anna control room was notified that the Virginia State Emergency Operations Center (EOC) had lost all power and all offsite communications. This included commercial phone lines, automatic ring downs, and the State ring down loop (lnsta-phone). At 0900 EST, all communications were restored once emergency generators were placed in service at Virginia State EOC. The Virginia State EOC initially lost offsite power only but all communications were supplied power by an Uninterruptable Power Supply (UPS). Subsequently, the UPS batteries depleted and all communications were lost. At that time (0828 EST) Virginia State EOC personnel notified the North Anna control room and supplied individual cellular phone numbers as an alternative method of contacting the Virginia State EOC. This report is reportable in accordance with 10CFR50.72(b)(3)(xiii), any event that results in a major loss of offsite communications capability, (e.g. Emergency Notification System). The licensee has notified the NRC Resident Inspector and Louisa County. | |
| ENS 49629 | 13 December 2013 21:00:00 | 10 CFR 26.719, FFD Reporting requirements | Eleven Individuals Not Properly Subjected to the Random Fitness for Duty Program | In 2010, changes within the Dominion FFD (Fitness for Duty) program resulted in currently 11 individuals not being subjected to random FFD testing although required. Two individuals are actively badged at Surry Power Station and North Anna Power Station (NAPS). Both individuals accessed the protected area at NAPS recently. The other nine individuals were not badged but perform duties that require them to be subject to the FFD program. The affected individuals are now within the random testing program. This event is reportable 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 a protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program.'" A work review will be conducted for the two actively badged individuals. The program error has been corrected. The licensee has notified the NRC Resident Inspector and the Louisa County Administrator. See similar Surry report EN #49630. | |
| ENS 49571 | 21 November 2013 21:41:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Low Casing Cooling System Tank Level Potentially Causes Loss of Npsh to Outside Recirculation Spray Pumps | The Casing Cooling system at North Anna Power Station (NAPS) Units 1 and 2 provides cold, borated water to the suction of the Outside Recirculation Spray (ORS) pumps to increase net positive suction head (NPSH) following the initiation of a Containment Depressurization Actuation (CDA). As Casing Cooling tank level decreases to the isolation setpoint, it has been determined that vortexing/air entrainment may occur. This air would then enter the suction of the ORS pumps and potentially cause degradation in design flow and/or loss of NPSH. As a result, this constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable per 50.72(b)(3)(v). A prompt operability determination is in progress that should restore the function of the recirculation spray system. The licensee has notified the NRC Resident Inspector. The licensee will be notifying the Louisa County Administration in the morning. | |
| 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. | |
| 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. | |
| ENS 49072 | 27 May 2013 19:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Drowning at Lake Anna | At 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 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. | |
| ENS 48940 | 17 April 2013 20:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Suspected Valve Body Leakage | On April 17, 2013 at 1600 (EDT), while performing a valve inspection/repair of the Unit 2 'A' Reactor Coolant Loop Fill Valve (2- RC-HCV-2556A), the as-found inspection results identified evidence of a suspected flaw causing leakage from the valve body to the threads of a stud housing of the valve. This valve is a 2 (inch) 316 SS (Stainless Steel) cast ASME XI (Class 1) 1500 psi valve body of a globe style design. Due to this design and the installed orientation, the RCS pressure medium fills the upper portion of the valve bonnet where the leak is located during normal plant operations. Therefore, this leakage would be considered pressure boundary leakage. 2-RC-HCV-2556A is currently isolated from the Reactor Vessel and is at atmospheric pressure. This inspection was performed in response to dry discolored boric acid identified during the normal operating pressure boric acid accumulation inspection procedure during the Spring 2013 Unit 2 refueling outage shutdown. An engineering evaluation of the suspected defect 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 principal safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector and local County Commissioners.
Event Number 48940 was made on April 17, 2013 in accordance with 10CFR50.72(b)(3)(ii)(A) to document a suspected flaw resulting in RCS pressure boundary leakage on Unit 2 'A' Reactor Coolant Loop Fill Valve (2-RC-HCV-2556A). North Anna Power Station is retracting this notification following completion of a cause analysis and metallurgical examination. The analysis determined that the valve leakage was due to the body-to-bonnet gasket joint. The original valve body was especially susceptible to gasket creep, which lead to a loss of sufficient sealing stress. This resulted in body-to-bonnet leakage, not a through-wall leak. Based on this analysis, the reporting requirements of 10CFR50.72(b)(3)(ii)(A) are not met and this event report is being retracted. The licensee has notified the NRC Resident Inspector. Notified R2DO (Ehrhardt). | |
| 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). | |
| ENS 47769 | 26 March 2012 03:36:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Seismic Activity | On March 25, 2012 at 2336 EDT, an Unusual Event was declared due to an earthquake felt on site. The site entered EAL HU1.1. No plant systems were affected. The National Earthquake Information Center reported a magnitude 3.1 seismic event 6 miles south-south west of Mineral, Virginia. A plant inspection is on-going to determine any plant issues related to the seismic event. Unit 1 is in a refueling outage and containment integrity was maintained. Unit 2 continues in full power operation. The licensee notified the NRC Resident Inspector, State and local agencies. Notified DHS SWO, FEMA, NICC and Nuclear SSA via email.
On March 26, 2012 at 0410 EDT, the Unusual Event was terminated. The basis for the termination was that all equipment walkdowns are complete with no damage discovered. The licensee will notify the NRC Resident Inspector. Notified the R2DO (Haag), NRR EO (Brown), IRD (Grant), DHS SWO, FEMA, NICC and the Nuclear SSA via email. | |
| ENS 47767 | 24 March 2012 22:55:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degradation of Steam Generator Nozzle Weld Areas | On 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 47682 | 21 February 2012 18:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Tritium Identified Onsite Above the Voluntary Reporting Threshold | On February 17, 2012, North Anna Power Station (NAPS) was notified by its vendor laboratory that a water sample, taken from an onsite ground water sample point, was confirmed to contain tritium above the voluntary reporting threshold of 20,000 picocuries per liter(pCi/L). The water sample, measuring 53,300 pCi/L, was obtained as a part of ongoing activities to determine the source of tritium previously reported to the state and NRC on October 29, 2010 (Event Notification - 46377). Current hydrological studies have determined the ground water in the area migrates to the station power block which is in the opposite direction from the lake. The ground water at the power block is collected in building subsurface drains and transported to a clarifier for processing. Clarifier discharge is accounted for as a monitored liquid effluent release pathway under the radiological effluent control program in accordance with the station's Offsite Dose Calculation Manual. As such, there is no increase to the projected annual dose to a member of the public. There are also no sources of drinking water in this area. Sampling of eight (8) ground water sample points outside the station protected area show no detectable levels of tritium confirming there is no migration offsite. The NRC Resident Inspector has been notified. A 30 day written report will be submitted to the NRC in accordance with NEI 07-07, Industry Ground Water Protection Initiative - Final Guidance Document. The licensee will inform both state and local agencies. | |
| ENS 47665 | 15 February 2012 17:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of Sewage System Release | At 12:44 on 2/15/2012, the Virginia Department of Environmental Quality was notified of a sewage system release that had the potential to reach state waters. On 2/14/2012, it was identified that water was flowing from a manhole cover near the North Anna training building. Further review identified the training building sewage lift station had lost power and that the water line in the manhole discharges to the lift station. It was estimated that approximately 120-200 gallons of untreated water reached the ground around the manhole before power was restored to the lift station. Upon further investigation, the station could not confirm whether untreated water reached Lake Anna. The NRC Resident Inspector has been notified. | |
| ENS 47626 | 30 January 2012 23:56:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to an Earthquake of Magnitude 3.2 | At 1856 EST, North Anna, Units 1 and 2, experienced an onsite seismic event. This was confirmed by the National Earthquake Information Center to be a magnitude 3.2. Based on these 2 indications the licensee declared an Unusual Event per EAL HU 1.1. The earthquake did not impact operations or equipment, and both units remain at 100% power. To exit the Unusual Event the licensee will need to verify that no damage was sustained to any systems or equipment. Presently the licensee is conducting a walkdown to verify no damage was incurred. The licensee has notified the NRC Resident Inspector. The licensee also notified the state and other government agencies. * * * UPDATE AT 0048 EST ON 1/31/12 FROM ROGER SMUTHERS TO PETE SNYDER * * * The NOUE declared at 1856 EST on 1/30/12 has been terminated (at 0035 EST). The event has been terminated after completion of AP-36, which is the procedure used when seismic activity is experienced. The abnormal procedure included engineering and operations inspections of various systems throughout the station, all of which were completed satisfactorily with no damage found relating to the seismic event. None of the stations seismic instrumentation actuated during the event due to low seismic activity felt at the station. The newly installed Free Field Seismic Instrumentation did not actuate which has a trigger point of 0.1 g's further demonstrating the low level of the seismic activity experienced at the station. The licensee notified the NRC Resident Inspector. Notified R2DO (Bonser), NRR (Lund), IRD (Grant), DHS (Gates), and FEMA (Biscoe). | |
| ENS 47497 | 5 December 2011 13:47:00 | Other Unspec Reqmnt | Discovery of After-The-Fact Emergency Condition - Unusual Event | At approximately 0847 EST, Unit 1 Letdown Pressure Control Valve, 1-CH-PCV-1145, began acting erratically which resulted in the Letdown Relief Valve, 1-CH-RV-1203, lifting and flowing to the Pressurizer Relief Tank. At 0848 EST, the relief valve reseated and the leakage stopped. Approximately 42 gpm leakage resulted from the relief valve lifting. This 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 is troubleshooting the Letdown Pressure Control Valve. The licensee notified the NRC Resident Inspector and will notify State and local agencies. |