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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| 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. | Reactor Coolant System | |
| 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) | Reactor Coolant System | |
| 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 | Reactor Coolant System | |
| 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. | Reactor Coolant System | |
| 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 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). | Reactor Coolant System | |
| 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 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. | Steam Generator Reactor Coolant System | |
| ENS 46094 | 14 July 2010 23:34:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec Shutdown Due to Through Wall Flaws in S/G Sample Line, Secondary Side | 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.
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). | Steam Generator | |
| ENS 45460 | 23 October 2009 20:34:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Plant Shutdown Due to Excess Letdown Heat Exchanger Tube Leak and After-The-Fact Unusual Event | On 10/23/09 at 1633, North Anna Unit 1 was placing Excess Letdown in service per 1-OP-8.5 due to a small unquantifiable throughwall leak on 1-CH-TV-1204B. At 1634 (the licensee) entered action of TS 3.4.13 due to what appears to be an Excess Letdown Heat Exchanger tube leak. The Component Cooling Water Head Tank (level) increased from 59% to 79% and the VCT (Volume Control Tank) level dropped 20 % indicating approximately 260 gallons of RCS had flowed into the Component Cooling Water System. At 1638 Excess Letdown was removed from service and the leak was terminated. At 1718 (the licensee) commenced ramping Unit 1 from service (TS Required Shutdown) to comply with TS 3.6.1, 'Containment Integrity' due to the throughwall leak on 1-CH-TV-1204B. The licensee has placed the normal letdown system back in service while the plant is being shut down. The throughwall leak on 1-CH-TV-1204B is relatively small and unquantifiable compare to the tube leak on the excess letdown heat exchanger. The licensee plans to proceed to Mode 5 to repair both the Letdown valve and the Excess Letdown Heat Exchanger tube leak. The licensee will notify the NRC Resident Inspector.
The licensee determined that it exceeded EAL SU6.1, Unidentified or Pressure Boundary Leakage greater than 10 gpm, for 4 minutes but currently does not meet the EAL Criteria. This requires the licensee to make a 1-hour notification that it has classified the event after-the-fact as an unusual event but did not actually declare the unusual event . The licensee will notify the NRC Resident Inspector and has notified the Virginia Department of Emergency Management. | ||
| ENS 43195 | 28 February 2007 00:47:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Failed Surveillance Testing | Both Trains of Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS) were declared inoperable at 1620 when dampers 2-HV-AOD-228-1 and -2, Safeguards Area Exhaust Bypass Dampers, failed surveillance testing. A ramp down was initiated at 1947 as required by Technical Specification 3.0.3. Temporary repairs to the dampers were completed at approximately 2125 and the ramp down was terminated. Temporary repair was made to the bypass damper seating surface. The licensee notified the NRC Resident Inspector. | Emergency Core Cooling System |