Semantic search
| Start date | Title | Description | Topic | |
|---|---|---|---|---|
| ENS 55457 | 12 September 2021 21:28:00 | 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) | Pressure Boundary Leakage |
| ENS 51557 | 20 November 2015 13:23:00 | 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). | Grab sample Offsite Dose Calculation Manual |
| ENS 48940 | 17 April 2013 20:00:00 | 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). | Boric Acid Pressure Boundary Leakage Through-Wall Leak |
| ENS 45013 | 22 April 2009 09:00:00 | Discovery of After-The-Fact Emergency Condition - Alert - Due to Fire Damage to Safety-Related Breaker | At 1410 hours on April 23, 2009, it was identified that an ALERT classification had not been declared on April 22, 2009 as required by EPIP-1.01. North Anna Emergency Plan, Emergency Action Level H2.1 requires the declaration of an ALERT for a fire or explosion in any safe shutdown area and either plant personnel report visible damage to any safety-related structure, system or component within the area or affected system parameter indications show degraded performance. A description of the event is provided below. On April 22, 2009, at approximately 0500 hours, Operations personnel identified a strong odor in the North Anna Unit 1 Cable Vault area. Subsequent investigation identified that the odor was coming from circuit breaker 01-EE-BKR-1J1-2S-J1 associated with the "D" Control Rod Drive Mechanism (CRDM) Fan (1-HV-F-37D). Operations personnel locally opened the circuit breaker to place it in a safe condition. 1-HV-F-37D had tripped approximately 30 minutes prior to the event. 1-HV-F-37D is not safety-related and not required for safe shutdown however; the supply breaker is safety-related since it is located on an emergency bus. Operation personnel then opened the circuit breaker cabinet and a small (6-inch) flame was observed. Operations personnel used a CO2 extinguisher on the internals of the circuit breaker to quickly extinguish the small fire. Appropriate levels of management were informed. The breaker has been quarantined. The cause of the circuit breaker failure has not been identified. A Root Cause Evaluation is in progress. There were no injuries. The plant continues to operate at full power. As a result of identifying that the criterion for the EAL was exceeded and no longer exists, a notification is being made to the NRC Operations Center in accordance with 10CFR50.72(a)(1)(i). The NRC Resident Inspector has been notified and the State and local governments will be notified.
On April 22, 2009, at approximately 0500 hours, operations personnel identified a strong odor in the North Anna Unit 1 Cable Vault area. Subsequent investigation identified that the odor was coming from circuit breaker 01-EE-BKR-1J1-2S-J1 associated with the 'D' Control Rod Drive Mechanism (CRDM) Fan (1-HV-F-37D). Operations personnel locally opened the circuit breaker to place it in a safe condition. 1-HV-F-37D had tripped approximately 30 minutes prior to the event. 1-HV-F-37D is not safety-related and not required for safe shutdown however; the supply breaker is safety-related since it is located on an emergency bus. Operation personnel then opened the circuit breaker cabinet and a small (6-inch) flame was observed. Operations personnel used a C02 extinguisher on the internals of the circuit breaker to quickly extinguish the small fire. A root cause evaluation is in progress. At 1447 hours on April 23, 2009, a one hour notification was made to the NRC Operations Center in accordance with 10CFR50.72(a)(1)(i), which identified that the criterion for a ALERT EAL was exceeded due to the small fire in the circuit breaker and subsequent damage to the breaker internals. The notification also stated that the condition no longer exists. Subsequent reviews have determined that the 'Initiating Condition' for the Emergency Action Level was not met and the event was not required to be classified as an ALERT. The initiating condition states -Fire or explosion affecting the operability of plant safety-related structures, systems or components required to establish or maintain safe shutdown. The 'D' Control Rod Drive Mechanism is not required to establish or maintain safe shutdown and the emergency bus remained operable during the event. The notification made to the NRC on April 23, 2009 is being retracted. The NRC Resident Inspector has been notified. Notified the R2DO (Nease). | Safe Shutdown |
| ENS 41322 | 10 January 2005 18:45:00 | Offsite Notification to Federal Energy Regulatory Commission | At 1345 on 1-10-2005, the Federal Energy Regulatory Commission (FERC) Local Engineer was notified by voice mail message of a failure of the North Anna Spillway Emergency Diesel (1-EE-EG-4) during a surveillance PT. 1-EE-EG-4's voltage regulator would not control voltage. The cause of the failure is unknown at this time. This failure only affects the hydro project and there is no impact on station operation. The FERC notification was made per 18CFR12.10(a). The licensee has notified the NRC Resident Inspector.
The following information was provided by licensee via facsimile: At 1457 hours on January 10, 2005, a 4 hour non-emergency notification was made to the NRC Operation Center as a result of an Offsite Notification to other government agency (i.e- FERC due to an inoperable spillway diesel) in accordance with 10 CFR 50.72 (b)(2)(xi). NRC guidance in NUREG -1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, provides discussion and notification examples intended to set a threshold that ensures reporting when an issue is safety significant. Some notifications to other government agencies may have little or no significance and may not be reportable to the NRC. Although the project safety device was inoperable and therefore reportable to FERC there was no safety significance since normal power was available throughout the time the device was out of service. Based on the discussions in NUREG - 1022, Section 3.2.12 for the notification in question, the North Anna Station Nuclear Safety Operating Committee has determined a NRC Notification was not required because there was no safety significance and is therefore being retracted. The licensee notified the NRC Resident Inspector. Notified R2RDO (Charlie Payne). |