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| Entered date | Event description | |
|---|---|---|
| ENS 53778 | 5 December 2018 17:06:00 | At 1010 (EST) on December 5, 2018, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge. This condition existed for approximately 3 minutes before the differential pressure was restored to normal when the Standby Gas Treatment system was manually initiated. This event was caused by a trip of the service air compressor 39AC-2A. The loss of instrument air pressure caused Reactor Building ventilation to isolate and raise Secondary Containment differential pressure. The instrument air pressure was restored when 39AC-2A was isolated and the two backup air compressors started. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector. |
| ENS 52480 | 10 January 2017 22:45:00 | This notification is a 10 CFR 21.21(a)(2) interim report for power supply model N-2ARPS-A6. Two instrument power supplies for the 'B' Residual Heat Removal (RHR) system were being bench tested prior to installation when it was discovered that they failed to meet Vendor Technical Manual specifications for voltage stability for varying loads. The deviation was a voltage drop of approximately 300mV. This did not meet the specification of less than 150mV when varying current from 5 amps (full load) to 2.5 amps. A second replacement power supply exhibited a similar 300 mV drop. James A. FitzPatrick (JAF) reviewed the work order instructions to determine if there was a deviation from the recommendations in the Foxboro technical manual F180-0309 Spec 200 Multinest Power Supply 2ARPS Series calibration. Since as-found voltage readings were within the required tolerance of the RHR instrument loops, the power supplies appear to have been capable to perform their intended function. However, this evaluation did not troubleshoot why the power supplies failed to meet the calibration requirements. The power supplies were sent to a repair vendor. The input from this vendor is expected to allow JAF to complete the evaluation per 10 CFR 21.21(a)(1) by March 21, 2017, and a notification for failure to comply or defect per 10 CFR 21.21(d)(3)(i) is expected by March 24, 2017, if necessary. This notification is being submitted as an interim report per 10 CFR 21.21(a)(2). The licensee notified the NRC Resident Inspector. |
| ENS 52122 | 25 July 2016 13:43:00 | A non-licensed supervisory employee had a confirmed positive test for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been suspended. The licensee notified the NRC Resident Inspector. |
| ENS 52042 | 24 June 2016 16:06:00 | At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York. |
| ENS 51242 | 20 July 2015 14:27:00 | On the morning of July 20, 2015 at 0740 EDT, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the Secondary Containment differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR-3.6.4.1.1) value of greater than or equal to 0.25 inch of vacuum water gauge. Both trains of the Standby Gas Treatment System were placed in service and the Reactor Building was isolated. The decrease in Secondary Containment differential pressure was caused by Reactor Building roof maintenance creating multiple openings. Maintenance workers were immediately ordered to stop work and address the condition. Secondary Containment differential pressure was restored to within the TS SR value at 0915 EDT, and remains greater than 0.25 inch of vacuum water gauge. The secondary containment is a structure that surrounds the primary containment and is designed to provide secondary containment for postulated loss-of-coolant accidents inside the primary containment. To prevent exfiltration the secondary containment requires the control volume pressure at less than the external pressure. The differential pressure requirement of TS SR-3.6.4.1.1 ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration. During this period there were no unmonitored radioactive releases; however, this event could have prevented the fulfillment of a safety function to control the release of radioactive material and it is reported pursuant to 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been informed. |
| ENS 49982 | 1 April 2014 14:02:00 | At 0645 EDT on the morning of April 1, 2014, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the Control Room received an alarm associated with the ventilation system for the 'B' division of the Residual Heat Removal Service Water (RHRSW) and Emergency Service Water (ESW) pump room. Investigation identified that the ventilation exhaust fan (73FN-3B) associated with this pump room had tripped due to thermal overload. The overload relay was reset at 0704 EDT and the fan automatically started; the fan is currently operable. During this period, the fan would not have automatically started. The ventilation systems for the RHRSW and ESW pump rooms are not included in the JAF Technical Specifications (TS), nor are they in the JAF Technical Requirements Manual (TRM). The ambient temperature limit in the RHRSW and ESW pump room was never challenged. However, with 73FN-3B non-functional, it is procedurally required to declare 10P-1B (RHRSW Pump B), 10P-1D (RHRSW Pump D) and 46P-2B (ESW Pump B) inoperable. The 'B' ESW pump cools the 'B' EDG subsystem, which would therefore also be inoperable. During this period, the 'A' EDG subsystem was inoperable for an emergent issue. Because the 'A' and 'B' EDG subsystems were concurrently inoperable for a period of approximately 45 minutes, this condition resulted in a loss of safety function for the Emergency Diesel Generators, which is reportable pursuant to 10 CFR 50.72(b)(3)(v)(A). Both affected emergency diesel generators were in the same division with another redundant division operable. The licensee notified the NRC Resident Inspector and will be notifying the State of New York.
The statement "Both affected emergency diesel generators were in the same division with another redundant division operable" is incorrect. Both divisions of emergency diesel generators were inoperable, since one of two available emergency diesel generators per division were inoperable at the time of this event. The Licensee notified the NRC Resident Inspector. Notified R1DO (Cahill). |
| ENS 49941 | 21 March 2014 12:53:00 | A significant Fitness for Duty (FFD) programmatic vulnerability was discovered during an NRC inspection. A potential exists that some members of the FFD site's random drug test pool could control and predict the date and time that the random list is run; thus mitigating the effectiveness of the 'random aspect' for the staff. Per 10 CFR 26.31(d)(2)(i), 'Random testing. Random testing must - Be administered in a manner that provides reasonable assurance that individuals are unable to predict the time periods during which specimens will be collected.' Since, these individuals may be able to predict the timeliness of random drug test events this condition is reportable per 10 CFR 26.719(b)(4) as a discovered programmatic vulnerability of the FFD program that may permit undetected drug or alcohol use by individuals who are assigned to perform duties that require them to be subject to the FFD program. The licensee has notified the NRC Resident Inspector. |
| ENS 49794 | 5 February 2014 15:15:00 | At 1130 (EST) on February 5, 2014, with the James A. FitzPatrick Nuclear Power Plant (JAF) at 100% reactor power, Oswego County Emergency Management notified JAF that the tone alert radios had been out of service since 1000. 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 criterion 10 CFR 50.72(b)(3)(xiii). 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. At 1328 on February 5, 2014, JAF was notified by Oswego County Emergency Management that the tone alert radios had been returned to service at 1325. The licensee notified the NRC Resident Inspector, the State of New York and local authorities of the outage. |
| ENS 48410 | 16 October 2012 13:16:00 | On October 16, 2012, at 0529 EDT, with the reactor in mode 4 and one operable channel for the Residual Heat Removal (RHR) shutdown cooling isolation function, as permitted by TS 3.3.6.1, Operators identified a rising reactor water level indication on instruments associated with the 3A reactor water level reference leg while actual level remained constant. At 0700 EDT, Operators determined this condition rendered the RHR shutdown cooling isolation function inoperable; specifically, an isolation signal for RHR shutdown cooling based on a low reactor water level. In accordance with Tech Spec 3.3.6.1 Required Action J.1, immediate action was initiated to restore isolation capability. RHR shutdown cooling isolation capability was restored at 1040 EDT by changing in-service RHR shutdown cooling systems. This event could have prevented the fulfillment of a safety function, 10 CFR 50.72(b)(3)(v), at time of discovery to mitigate the consequences of an accident (D). The licensee has notified the NRC Resident Inspector. |