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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54503 | 31 January 2020 10:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | At 0555 (EST), on January 31, 2020, James A. FitzPatrick was at 38 percent power when an automatic scram occurred as a result of a main turbine trip on high Reactor Pressure Vessel (RPV) water level. The plant was at reduced power in preparation for maintenance activities. The 'A' Reactor Feed Pump (RFP) was being removed from service when a perturbation in reactor water level reached the high RPV water level setpoint. This resulted in a main turbine trip and 'B' RFP trip. The automatic scram inserted all control rods. A subsequent low water level resulted in a successful Group 2 isolation. The plant is stable in Mode 3 with the 'B' RFP maintaining RPV water level. The initiation of the reactor protection systems (RPS) due to the automatic scram signal 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). The licensee notified the NRC Resident Inspector, and the State and Local government for the scram. Decay heat is being removed via the main condenser. | Reactor Protection System Reactor Pressure Vessel Main Condenser Control Rod | |
| ENS 52042 | 24 June 2016 16:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pumps Degradation | 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. | Feedwater Secondary containment Reactor Protection System Emergency Diesel Generator Reactor Recirculation Pump Reactor Pressure Vessel Reactor Building Ventilation Reactor Building Closed Loop Cooling Decay Heat Removal Main Steam Line Control Rod | 05000333/LER-2016-004 |
| ENS 51680 | 24 January 2016 03:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Due to Lowering Intake Level | At 2241 (EST) on 1/23/2016, James A FitzPatrick inserted a manual scram from 89 percent power due to lowering intake level. Following the successful scram, a residual transfer occurred, resulting in a loss of the non-vital busses, loss of all Circulating Water Pumps, and a manual closure of the Main Steam Isolation Valves (MSIVs). The cause of the residual transfer is unknown. RPV (Reactor Pressure Vessel) level shrink during the scram resulted in a successful Group 2 isolation. Reactor Vessel (level) and pressure are being maintained with the High Pressure Coolant Injection System which was manually started. A cooldown is in progress. FitzPatrick will proceed to Mode 5 until the cause is identified and corrected. The Emergency Diesel generators auto started as a result of the loss of power to the non-vital busses. Offsite power remained available throughout the event. Operators are controlling pressure manually via the relief valves. FitzPatrick will notify the Public Service Commission of the event. The NRC Resident Inspector was notified. | High Pressure Coolant Injection Emergency Diesel Generator Main Steam Isolation Valve | 05000333/LER-2016-001 |
| ENS 48501 | 11 November 2012 08:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Reactor Scram Due to Turbine Trip Followed by Unusual Event Declared Due to Main Transformer and Bus Duct Fire | An unplanned, automatic reactor scram occurred at 0355 EST due to a Main Turbine trip signal. All safety systems operated and actuated as expected. Both the Main Transformer, T-1A and normal station services transformer T-4 activated their respective deluge systems. On-site fire brigade and offsite fire assistance have successfully extinguished the T-1A transformer fire. There is still an active fire in the T-1A bus ductwork. The plant will be taken to cold shutdown conditions. At 0545 EST the plant entered the emergency plan at the NUE level due to inability to successfully extinguish the fire. All control rods fully inserted following the reactor scram. MSIVs remain open with decay heat being removed via steam to the main condenser using the bypass valves. All electrical buses are powered from their normal offsite reserve source. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.
As of 0639 EST the fire in the T-1A bus ductwork has been extinguished.
Local fire department is on-site. No radiological release and no protective actions required. Plant cooldown in progress.
The Unusual Event (HU 6.1) has been terminated at 0801 EST. Cooldown in progress to cold condition. Reactor level at 206 inches and pressure is at 530 pounds. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified R1DO (Dentel), NRR EO (McGinty), IRD (Gott). Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email. | Main Transformer Main Condenser Control Rod | 05000333/LER-2012-008 |
| ENS 48479 | 5 November 2012 02:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram from Full Power Following Turbine Trip | The reactor was scrammed on a valid reactor protection system activation caused by a main turbine trip. The cause of the main turbine trip is under investigation. All control rods fully inserted. All isolations and initiations occurred as designed. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated as expected. RCIC injected into the reactor coolant system, HPCI did not, as expected. This scram was characterized as uncomplicated and the reactor is stable in Mode 3. The plant is in a normal post shutdown electrical lineup. All systems functioned as required. The NRC Resident Inspector has been notified. | Reactor Coolant System High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Control Rod | |
| ENS 43752 | 28 October 2007 04:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Due to Lowering Intake Level | High winds (in excess of 40 MPH) resulted in significant debris at the plant intake. Traveling screens had been placed in 'Fast' speed and continuous wash on the previous shift in anticipation of predicted high wind conditions. As the debris entered the intake, high traveling screen differential pressure and lowering intake level prompted the operating crew to enter the appropriate Abnormal Operating Procedure (AOP). Efforts to manually clean the screens were ineffective, and at 0059 hours, the crew inserted a manual scram as directed by procedure at 240 foot intake level. On scram, as expected, reactor vessel level lowered to the low level scram setpoint (177 inches above top of active fuel). At this point, as expected, an automatic Reactor Protection System (RPS) actuation, and a Group 2 Primary Containment Isolation System (PCIS) isolation occurred with no anomalies noted. During plant cooldown reactor level lowered to 177 inches above TAF. This resulted in a valid RPS actuation and a PCIS Group 2 isolation signal. All systems responded as expected. Operators were able to maintain reactor vessel level above the actuation setpoint for High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, and these systems were not required to operate. The cause of the failure of the traveling screens to maintain intake level is under investigation. All control rods are fully inserted and the plant is stable in Mode 3, Hot Shutdown. Decay heat is being removed via the turbine bypass valves to the main condenser. No SRVs lifted during the transient. The plant is in a normal shutdown electric plant lineup. The licensee notified the NRC Resident Inspector and the New York Public Service Commission. | High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Reactor Core Isolation Cooling Main Condenser Control Rod | 05000333/LER-2007-002 |
| ENS 43635 | 12 September 2007 10:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Following Intake Structure Traveling Screen Fouling | As a result of high winds and a severe storm front moving through the area a Nuclear Plant Operator (NPO) found the intake canal water murky, wind speed was noted to be 40 MPH. The Traveling Screens were placed in the continuous mode of operation and increased monitoring was put in effect. At 0520 the NPO noted that 'a lot of weeds and fish were coming in through the intake' and that a full trash basket had been changed out. At approximately 0630 the NPO reported that the traveling screens were not moving and intake level was lowering due to the influx of debris. The Shift Nuclear Operator (SNO) entered the appropriate Abnormal Operating Procedure (AOP) for lowering intake level and commenced rapid power reduction. At 0635 with intake level at 240 feet the SNO inserted a manual scram and entered the appropriate AOP and Emergency Operating Procedure (EOP). Reactor Vessel level decreased to less than 177 inches above the top of active fuel resulting in a Primary Containment Isolation System (PCIS) Group 2 isolation signal. The PCIS Group 2 isolation resulted in a valid signal to close PCIS valves in the Drywell Floor Drain System, the Drywell Equipment Drain System, the Residual Heat Removal (RHR) System, the Traversing In-Core Probe (TIP) System, the Containment Atmosphere Dilution (CAD) System, and the Reactor Water Clean-Up (RWCU) System. However, as a result of the power reduction the operators were able to maintain reactor vessel level above the actuation setpoint for the High Pressure Coolant Injection (HPCI) System and the Reactor Core Isolation Cooling (RCIC) Systems and these systems were not required to operate. At 0640 the Scram was reset, at 0710 the Group 2 PCIS Isolation was verified, and at 0729 the Group 2 PCIS Isolation was reset. The cause of the traveling screens stopping has been determined to be due to shear pins on the traveling water screen shearing due to excessive debris loading. All equipment responded as expected during the downpower and subsequent manual Scram. All rods are in and the plant is stable in MODE 3, Hot Shutdown." Decay heat is being removed by dumping steam to the condenser via the turbine bypass valves. The safety buses are being powered by offsite power. The 'B' train Residual Heat Removal pump was out of service at the time of the event for planned maintenance. The licensee notified the NRC Resident Inspector.
At 1034 with the plant shutdown and a cooldown in progress reactor vessel (RV) level lowered to less than 177 inches above TAF (Top of Active Fuel). This resulted in a valid Reactor Protection System (RPS) Scram signal and a PCIS Group 2 Isolation Signal. AP systems responded as expected. Reactor vessel level remained above the actuation setpoint for HPCI and RCIC and they were not required to operate. RV level was restored and the Scram was reset. At the time of the initial event and at this update the plant is in LCO action statement 3.5.1.A for the 'B' RHR Pump being out of service. The system is being restored at this time. The licensee notified the NRC Resident Inspector. The licensee also notified the state and is providing a press release to the media. Notified R1DO (Trapp). | High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Reactor Core Isolation Cooling Residual Heat Removal | 05000333/LER-2007-002 |
| ENS 41987 | 14 September 2005 06:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram on Low Reactor Vessel Water Level During Planned Maintenance | An automatic reactor scram on low reactor water level occurred following a momentary loss of the UPS (Uninterruptible Power Supply) system. The power loss resulted in a lock-out of the RFP (Reactor Feed Pump) controls. The HPCI system started on low reactor level but did not inject, reactor level had risen above the initiation set-point. The RCIC did not receive an initiation signal. Overall plant response was as expected. Prior to the transient the licensee was in the process of transferring UPS electrical loads to the alternate power supply. The momentary loss of UPS power locked-out the RFP controls during a downtrend in the reactor water level from the normal 201-203 inch operating band. Before Operators could establish manual control, reactor water level reached the 177 inch scram setpoint. Following the scram, reactor water level continued to decrease to the HPCI and RCIC initiation setpoint of 126 inches before recovering. HPCI received a start signal but RCIC did not for reasons under investigation. The unit will remain in mode 3 pending the results of the post-scram investigation and restart. The licensee informed the NRC Resident Inspector and is planning on issuing a press release.
On September 14, 2005 at approximately 0213. an automatic reactor scram on low reactor water level occurred following a momentary loss of the UPS (Uninterruptible Power Supply) system. The power loss resulted in a lock-out of the RFP (Reactor Feed Pump) controls. A level transient occurred causing reactor water level to lower, resulting in an automatic reactor scram on low reactor water level. The HPCI system auto initiated on low reactor water level but did not inject as reactor water level had risen above the initiation setpoint. The RCIC system auto initiated (and sealed-in) and injected into the reactor vessel. Both systems operated as designed. In addition, a Primary Containment Isolation System (PCIS) Group 2 isolation occurred, resulting in multiple system isolations. This included isolation signals to Reactor Water Cleanup, Reactor Building Ventilation, Containment Atmosphere Dilution. Torus Vent and Purge, Drywall Floor and Equipment Drain Sumps, Drywall Containment Atmospheric Monitors, Recirculation System Sample Lines, Traversing In-Core Probes, LPCI Inboard Injection Valves, Residual Heat Removal Drain to Radwaste, and auto initiation of Standby Gas Treatment. (Note that two Reactor Water Cleanup PCIS valves did not close due to their respective circuit breakers being in the open position for planned maintenance activities.) The above event meets the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation while the reactor was critical, as well as 10 CFR 50.72(b)(3)(iv)(A) for the valid actuation of systems listed in 10 CFR 50.72(b)(3)(iv)(B), including general containment isolation signals affecting containment isolation valves in more than one system, the HPCI system, and the RCIC system. The NRC Resident Inspector has been briefed. | Primary Containment Isolation System Reactor Building Ventilation Residual Heat Removal Reactor Water Cleanup | |
| ENS 40072 | 14 August 2003 20:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to a Loss of Offsite Power. | Automatic reactor scram due to a loss of offsite power. All rods fully inserted into the core. All emergency core cooling systems and the emergency diesel generators are operating properly. NRC Resident Inspector was notified of this event by the licensee.
Reactor vessel pressure is 93 psig. Plant electrical loads have been shifted back to offsite power and the EDGs were secured. The licensee exited the Unusual Event at 0039. The licensee tentatively plans to remain shutdown for a brief period to do some maintenance work. Notified R1IRC and FEMA(Hyman).
This updated is being provided to note a correction to the previously submitted time information and provide additional clarification of information that did not get captured. The automatic reactor scram occurred due to a generator load rejection. All rods fully inserted into the core, all emergency core cooling systems and the emergency diesel generators operated properly. The incorrect "event time" was communicated to the NRC during the initial notification (at 1717 hours) and subsequent updates. The event time listed was "1611" hours EDT. This was the actual time of the loss of offsite power. The declaration of the Unusual Event was at "1626" hours EDT, based on EAL 6.1.1, Loss of power to transformers T2/T3 for greater than 15 minutes (loss of offsite power). In addition, the initial notification of the Unusual Event was also used to document two reporting requirements under 10CFR50.72. Specifically, 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system (RPS) when the reactor is critical, and 10CFR50.72(b)(3)(iv)(A) for valid actuation of systems listed in paragraph (b)(3)(iv)(B). Various system actuations occurred as a result of the loss of offsite power. The NRC Resident Inspector was notified along with state and local agencies. Notified Reg 1 RDO(Meyer) | Reactor Protection System Emergency Diesel Generator Emergency Core Cooling System |