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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 53287 | 25 March 2018 20:16:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Control Rod Drive Piping Potentially Inoperable | On March 25, 2018 at 1616 hours (EDT), with the reactor in cold shutdown condition, two control rod drive piping lines were determined to be potentially inoperable in the event of a design basis earthquake due to support defects. The control rod drive piping forms a portion of the reactor coolant pressure boundary and primary containment boundary. The supports will be repaired prior to plant startup. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the Commonwealth of Massachusetts.
The purpose of the notification is to retract ENS notification 53287 made on 03/25/18 for Pilgrim Nuclear Power Station. The previous notification reported that control rod drive (CRD) piping could be potentially inoperable in the event of a design basis earthquake, at the time of discovery, due to piping support defects. Subsequent evaluation has demonstrated that the piping was not inoperable. Specifically, after an engineering evaluation, it has been determined that the CRD Hydraulic System operability was never lost and the system was operable, although non-conforming, based on the support configuration not conforming to the pipe support drawings. The affected pipe supports have been restored or reworked to the proper design condition in accordance with the design drawings. The CRD System has subsequently been restored to a fully operable status. Notified R1DO (Jackson) and IRD MOC (Pham). | Primary containment Control Rod | |
| ENS 52744 | 10 May 2017 18:11:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Appendix J Local Leakage Exceeded Acceptance Criteria | On Wednesday May 10, 2017, at 1411 EDT, with the reactor at 0 percent core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) was in a Refueling Outage, performing a review of Local Leak Rate Testing results, when it was concluded that PNPS had exceeded its Title 10 Code of Federal Regulations Part 50, Appendix J, Option B, Type B and C Local Leak Rate Test (LLRT) leakage criteria. Previously, on April 22, 2017, when PNPS was performing LLRT of the High Pressure Coolant Injection (HPCI) steam exhaust line check valves, both valves failed to meet their LLRT acceptance criteria specified in plant procedures. Neither of the check valves seated acceptably. Based on the ongoing evaluation of these test exceedances, it was concluded that these test results cause the plant to exceed the overall as-found minimum path Appendix J acceptance criteria of 0.6 La (126.3 SLM (Standard Liters per Minute)). Further investigation is ongoing. This event has no impact on the health and safety of the public. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), any event or condition that resulted in the condition of the nuclear power plant, including its principle safety barriers, being seriously degraded. The licensee will notify the Commonwealth of Massachusetts Emergency Management Agency. | High Pressure Coolant Injection | |
| ENS 49061 | 23 May 2013 08:55:00 | 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 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Primary Containment Declared Inoperable During Hpci Testing | At 0455 hours on Thursday, May 23, 2013, with Pilgrim Station in the Startup/Hot Standby Mode and reactor coolant pressure approximately 550 psig, primary containment was declared inoperable due to a leak on the High Pressure Coolant Injection system (HPCI) turbine exhaust line while performing the HPCI system flow rate test. Power ascension was suspended pending investigation and repair. Repair plans to restore system integrity are in progress. The plant is in a safe condition and plant personnel are investigating the cause. The Resident Inspector has been informed of this notification. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(C) and (D). The licensee will notify the Massachusetts Emergency Management Agency (MEMA). The licensee has entered Technical Specification 3.7.A.2 to be in cold shutdown within 24 hours. | High Pressure Coolant Injection Primary containment | 05000293/LER-2013-005 |
| ENS 43322 | 26 April 2007 23:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition During Weld Deposit Overlay Repair | On April 26, 2007, at approximately 1930 it was reported that the N2K recirculation system inlet nozzle had slight water seepage. The seepage developed during welding operations being performed to install a full structural weld overlay over the nozzle and stopped almost immediately during the welding. The weld overlay was being installed as a conservative measure following UT inspections that had been performed earlier during the outage. This event is being reported in accordance with 10 CFR 50.72(b)(3)(ii)A. The plant is in stable condition. The licensee notified the NRC Resident Inspector. | ||
| ENS 40211 | 1 October 2003 08:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Pressure Vessel Boundry Leakage Found During Inspection | At approximately 0430 on October 1, 2003 with the reactor Mode Switch in Refuel, reactor water level 210" above Top of Active Fuel, and reactor water temperature less than 110F, investigations and walkdowns were being performed to identify potential sources of drywell leakage. Drywell leakage was well within Technical Specification limits. During these walkdowns a small leak in a cap of a nozzle of the reactor pressure vessel boundary, nozzle N10, located at reactor vessel height 440" was identified. This location is 84" above Top of Active Fuel. This is a cut and capped 4" Control Rod Drive return line. The leak appears to be on the cap weld. This notification is being made per 10 CFR 50.72(b)3(ii)(A). All required ECCS systems are presently operable. Investigation is continuing. The (NRC) resident inspector has been notified. | Reactor Pressure Vessel Control Rod |