Semantic search
Jump to navigation
Jump to search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57103 | 3 May 2024 08:11:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary Containment Degraded | The following information was provided by the licensee via email: At 0411 EDT on 5/03/2024, it was determined that primary containment did not meet TS (Technical Specification) 4.6.1.2 (surveillance) requirement due to a primary containment leak rate test exceeding `La (allowable leakage rate). This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The final observed leak rate is still being calculated as the test is still within the stabilization period. Testing is allowed within the stabilization period for an unspecified amount of time. Short term corrective actions are to identify and repair any leak paths. No mode changes are required due to this event. | Primary containment | |
| ENS 56494 | 30 April 2023 05:00:00 | 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)(A), Seriously Degraded | Primary Containment Integrity Degraded | The following information was provided by the licensee via email: At 0100 EDT on 04/30/23, it was determined that the primary containment integrity did not meet (Technical Specification) TS 4.6.1.1.d requirement, suppression chamber in compliance with TS 3.6.2.1 due to the inability to establish test conditions for the bypass leakage test in accordance with TS 4.6.2.1.f. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D) & 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Primary containment | |
| ENS 49110 | 13 June 2013 06:52:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Discovery of Pressure Boundary Leakage | On 6/13/13 at approximately 0252 EDT, during the initial walk down of the drywell of the Hope Creek Unit 1 forced outage, water was observed leaking from a 40% circumferential crack in a weld in the RHR vent line adjacent to the outboard isolation valve (BCV-597). The estimated leakage rate through the crack is less than one gallon per minute. The RHR vent line is one-inch ASME Class 1 piping. Hope Creek Unit 1 is stable in Operational Condition 3, Hot Shutdown. This report is being made under 10CFR50.72(b)(3)(ii)(A) for a weld or material defect in the primary coolant system which cannot be found acceptable under applicable ASME standards. The NRC Resident Inspector has been notified. The only safety-related equipment out of service at the time event was the 'C' Service Water Pump, which was tagged for scheduled maintenance. No personnel injuries occurred. No radiation releases occurred. The licensee entered a 24 hr. action statement under T.S. 3.4.3.2, Condition A to be in mode 4 by 0252 EDT on 6/14/13. The licensee will be notifying the Lower Alloways Creek township. | Service water | 05000354/LER-2013-003 |
| ENS 41536 | 28 March 2005 15:35:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Rcs Through Wall Leakage Identified at Welded Piping Junction | At 1035 hours on the morning of 3/28/05, investigation into a previously reported leak (Event #41530) identified an approximate 3 inch through wall flaw at the welded junction of the 4 inch diameter decon port and the 28 inch diameter 'B' Reactor Recirculation Systems pump suction piping. The identification of the indication required the removal of insulation in order to complete the inspection. The unidentified leakage rate is below Tech Spec limits and the leak location is within the isolable boundary of the 'B' Reactor Recirculation Pump. This failure constitutes a welding or material defect in the primary coolant system that is not acceptable under ASME Section XI for ASME Code Class 1 piping. This report is being made under Hope Creek Event Classification Guide RAL# 11.2.1 in accordance with 10CFR50.72(b)(3)(ii). At the time of this notification, the Hope Creek Generating Station is in OPCON 4 at 150 degrees reactor coolant temperature to support piping repairs. The licensee informed the Lower Alloways Creek Township and NRC Resident Inspector. | Reactor Recirculation Pump |