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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| 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 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 |