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| Entered date | Event description | |
|---|---|---|
| ENS 52503 | 22 January 2017 16:19:00 | Information from a Manual Phased Array UT (Ultrasonic Testing) examination of the 'A' RHR LPCI (Residual Heat Removal Low Pressure Coolant Injection) Injection Loop indicates an axially oriented indication 0.95 inch in length and 0.81 inch through wall. This is on weld number 24-10-130 (T to Valve dissimilar metal weld). This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A) based on the fact that the indications result in a defect in the primary coolant system which cannot be found acceptable under ASME Section XI. The licensee informed the NRC Resident Inspector. The weld is located where the RHR piping taps into the reactor vessel. The wall thickness at this location is 1.15 inches. |
| ENS 52343 | 2 November 2016 15:35:00 | During panel walkdown, it was discovered that a tag out for the 'C' Residual Heat Removal pump suction valve was active and the valve was open with its breaker open. This rendered the valve inoperable and Technical Specification 3.6.1.3 Action C for penetration with one inoperable PCIVs was entered. The action was to isolate the penetration by closing the valve within (4) hours or restore power. The event was discovered at 0845 (EDT) and the breaker was closed at 0925 (EDT). Technical Specification 3.6.1.3 Action C (Isolate penetration within 4 hrs.) was entered at 0130 (EDT) (time breaker was opened per tagout) and exited at 0925 (EDT). This condition of non-compliance existed from 0530 (EDT) on 11/02/16 until 0925 (EDT) on 11/02/16. This event is being reported under 10CFR50.72 (b)(3)(v)(C). NRC Resident has been notified.
In accordance with Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves, the TS Basis states that one or more barriers are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. When two or more barriers are provided, one of these barriers may be a closed system. During this event, one of the barriers in the penetration became inoperable: 'C' Residual Heat Removal (RHR) pump suction valve 10MOV-13C. After the initial NRC notification, it was confirmed that the RHR system piping is classified as a closed system outside containment. The integrity of the closed-loop RHR system is verified by monitoring the keep-full system. Since the piping is maintained full of water during normal and post-accident modes of operation, a barrier against post-accident, gaseous, containment leakage is provided. Therefore, the affected penetration could have performed its intended safety function since there was redundant equipment in the same system which was operable. This event is not reportable under 10 CFR 50.72(b)(3)(v)(C) and the original notification may be retracted. Finally, the primary containment penetration with 10MOV-13C is with a closed system and the completion time per TS 3.6.1.3 Required Action C is 72 hours. The valve was restored to operable prior to exceeding this time. The licensee notified the NRC Resident Inspector. Notified the R1DO (Dentel). |
| ENS 50579 | 30 October 2014 13:05:00 | On the evening of October 28, 2014 at 1708 EDT, with James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100 percent power, the Reactor Building differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR) value of at least 0.25 inches water vacuum for a period of thirty-four (34) seconds. This occurred during restoration of the Reactor Building Ventilation System (RBVS) following planned maintenance. The Reactor Building differential pressure was 0.50 inches water vacuum with the 'A' RBVS fans in-service in conjunction with the Standby Gas Treatment System (SGTS). The Reactor Building differential pressure decreased to 0.19 inches water vacuum when the SGTS was secured. The Reactor Building Vent was subsequently isolated, and the alternate 'B' RBVS fans were placed in-service; the differential pressure increased to within the required 0.25 inches water vacuum value. The JAF TS bases associated with Secondary Containment state that, 'for Secondary Containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.' Troubleshooting activities indicated that the transient was due to a non-safety related, non-TS damper downstream of one of the 'A' RBVS fans that did not fully stroke open. The subject damper is not part of Secondary Containment, and has no safety related function. This condition did not impact the leak tightness of Secondary Containment or the ability of the associated equipment to establish and maintain the required differential pressure. Secondary Containment would have fulfilled its safety function. However, because the JAF TS SR value of 0.25 inches water vacuum was not met, Secondary Containment was considered Technical Specification INOPERABLE for a period of thirty-four (34) seconds. The Secondary Containment is considered a single-train system; therefore, this condition is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector. |
| ENS 50532 | 14 October 2014 01:30:00 | During the plant response to the trip of the B Recirculating water pump, reactor water level rose to the HPCI (High Pressure Core Injection) high water level trip setpoint as indicated on the associated instrumentation. With this high water level trip actuated, the HPCI high drywell pressure initiation signal would not have allowed the HPCI system to perform its intended safety function if required. If the HPCI system received the low water level initiation signal, the system would have been able to perform Its intended safety function. This high water level signal was actuated from 1935 (EDT) until reset at 1940 (EDT). This is reportable under 50.72(b)(3)(v). The licensee notified NRC Resident Inspector.
Further review has determined that the condition was not a result of procedural errors/inadequacies, equipment failures, or design / analysis inadequacies. Plant systems responded as per design when the HPCI system high water level trip actuated when reactor vessel water level rose to the HPCI high water level trip setpoint. HPCI initiation has two logics: one for low-low vessel water level and the other for a high drywell pressure. A vessel low-low water level is an indication that reactor coolant is being lost with a need for HPCI injection for core cooling. High drywell pressure could indicate a line break in the Reactor Coolant Pressure Boundary inside the drywell. The HPCI level instrumentation is designed to shut down the HPCI system upon high water level to prevent HPCI turbine damage due to gross moisture carryover and will re-initiate HPCI if vessel water level drops to the initiation water level setpoint. A HPCI high drywell pressure initiation signal, above setpoint, would have made up the logic for HPCI initiation and as per design, HPCI would have injected at the vessel low low level setpoint without operator action to reset the trip. In this instance, the trip was reset as prescribed by station procedures. HPCI was capable of performing its safety function after the high water level trip reset either by operator action or instrumentation (low low level initiation). The licensee will be notifying the NRC Resident Inspector. Notified R1DO (Rogge). |
| ENS 48981 | 29 April 2013 17:50:00 | Offsite notification to the New York DEC (Department of Environmental Conservation) to report a Freon (R-22) release to the air of 8 lbs. 11 ozs. This release came from the cafeteria kitchen walk-in cooler. The licensee notified the NRC Resident Inspector. |
| ENS 48927 | 16 April 2013 10:17:00 | The purpose of this report is to provide a telephone notification under 10CFR50.72(b)(2)(xi) to notify the NRC of the inadvertent actuation of the Oswego County emergency notification sirens at approximately 0845 (EDT) on 4/16/13. Oswego County was performing routine weekly testing and siren #17 was inadvertently actuated for approximately 2 minutes. The Oswego County Emergency Management Office issued a News Release identifying the inadvertent actuation of the emergency siren. The NRC Resident Inspector has been notified. |
| ENS 45504 | 17 November 2009 21:35:00 | At 1819 on November 17, 2009, with James A. FitzPatrick (JAF) nuclear Power Plant operating at 100% power, Oswego County Emergency Management notified JAF that the National Weather Service had notified them that the Tone Alert Radios had been out of service since 1740. This impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) population for the Nine Mile Point and JAF Nuclear Power Plants. This failure meets NRC 8-hour reporting criteria per 10 CFR 50.72(b)(3)(xiii). The National Weather Service is currently working to determine the cause of the failure and time frame for system restoration. The County Alert Sirens which also function as part of the Public Prompt Notification System remain operable. The loss of the Tone Alert Radios constitutes a significant loss of emergency off-site communications ability. Compensatory measures have been verified to be available should the Prompt Notification System be needed. This consists of utilizing the hyper reach system which is a reverse 911 feature available from the county 911 center. Local law enforcement personnel are also available for 'Route Alerting' of the affected areas of the EPZ. The event has been entered into the corrective action program and the NRC Resident Inspector has been briefed and the state PSC will also be notified. See EN# 45503 for Nine Mile Point Nuclear Plant.
Notified by Oswego County 911 operator that the Tone Alert radios were restored at 0318 on 11/18/09. The NRC Resident Inspector was notified. Notified R1DO(McKinley). |