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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 50047 | 20 April 2014 19:42:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Emergency Core Cooling System Piping Leak While Shutdown | A leak was discovered in the area of EPV0109, COMBINED SI/RHR (Safety Injection/Residual Heat Removal) TO ACCUMULATOR D OUTLET LINE VENT. The leak rate was estimated at 2.5 gallons per hour. The plant is in Mode 5, RCS (Reactor Coolant System) depressurized. This leak is considered a material problem that causes abnormal degradation of or stress upon the reactor coolant system pressure boundary, reportable in accordance with 10CFR 50.72(b)(3)(ii)(A). Efforts are underway to characterize the leak and plan for repairs. The leak has been secured after realigning RHR cooling from the B to the A train. The Resident Inspector has been notified. The cause of the leak is being investigated. | Reactor Coolant System Emergency Core Cooling System | |
| ENS 48713 | 4 February 2013 19:40:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Pressure Boundary Leakage Identified on Seal Injection Line Drain Valve | Unit is shut down in Mode 4 and in process of cooling down to Mode 5 for a scheduled refueling outage. During a containment walkdown to inspect for boron leakage, personnel identified that Reactor Coolant Pump 'A' seal injection line drain valve has a crack in the socket weld on the upstream side of the valve with active leakage visible as a fine mist. This valve is part of the Reactor Coolant System pressure boundary. TS 3.4.13 requires that the unit be placed in Mode 5 prior to 0140 hours on 2/6/13. All Mode 4 required safety related equipment is operable. The NRC Resident Inspector has been notified. The last primary leak rate test was conducted at 100% power and determined that there was 0.114 gpm of unidentified leakage. The leak location is inside the containment bio-shield and is not accessible at power. | Reactor Coolant System | |
| ENS 42897 | 11 October 2006 17:40:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unanalyzed Condition - Indications on Pressurizer Nozzle to Safe End Welds | The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. On October 11, 2006, at approximately 12:40 Central Standard Time (CST), engineering personnel performing preplanned in-service examination of the Pressurizer nozzle to safe end welds notified the Control Room that five circumferential indications had been discovered that cannot be found acceptable under ASME Section XI. Since the indications cannot be found acceptable under ASME Section XI the condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(A). Three indications are located on the surge line, one indication is on the safety nozzle, and one indication is on the relief nozzle. The locations are part of the Reactor Coolant System (RCS) pressure boundary. There is no evidence of RCS pressure boundary leakage. WCGS is currently shutdown for its 15th refueling outage and is in Mode 6. Weld overlay of the safe ends is an activity scheduled for the current outage. An investigation of this event will be conducted in accordance with the WCGS corrective action program. The licensee notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 41602 | 15 April 2005 10:17:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Steam Generator Bowl Drain Line Weld Leak Discovered During Visual Exam | During the performance of Alloy 600 bare metal visual examinations, the presence of boric acid deposits indicated a leak in a weld in the steam generator D (EBB01D) bowl drain line. This condition is being reported as a degraded condition on the primary coolant system under 10 CFR 50.72 (B)(3)(ii)(A). During investigation the defect was characterized as a 1/16" rounded indication in the drain line to boss weld. This condition cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws' or ASME Section XI, Table IWB-3410-1, 'Acceptable Standards.' The (NRC) Resident Inspector has been notified.
On 04/20/05 at 12:10 CDT, during the performance of Alloy 600 Bare Metal Visual Examinations, the presence of boric acid deposits indicated a leak in a weld in the Steam Generator C (EBB01C) bowl drain line. The inspection of Steam Generators A & B showed no presence of boric acid deposits. The licensee notified NRC Resident Inspector. Notified the R4DO (Shannon). | Steam Generator | |
| ENS 40327 | 17 November 2003 15:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Leakage Due to a Cracked Socket Weld | During routine operations activities a leak was identified in the area of EPV0109 (combined SI/RHR to Accumulator "D" outlet line vent). At 0900 it was determined that the leak was due to a cracked weld on the socket up stream of EPV0109. The leak rate is 25-30 drops/per minute . This condition is being reported as a degraded condition in the primary coolant system under 10CFR50.72(b)(3)(ii)(A). The plant is currently in Mode 5 with RCS pressure at approximately 345 psig and temperature at approximately 125F. Shutdown cooling is being supplied by the "A" train Residual Heat Removal System. The Resident Inspector has been notified. | Reactor Coolant System Shutdown Cooling Residual Heat Removal |