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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55871 | 29 April 2022 16:51:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection (HPCI) Inoperable | The following information was provided by the licensee via email: At 1251 EDT on April 29, 2022, while troubleshooting the failure of the High Pressure Coolant Injection (HPCI) Exhaust Drain Pot High Level Alarm to clear, it was discovered that the High Pressure Coolant Injection exhaust line condensate drain system was not functioning as designed to support removal of condensate from the turbine exhaust. This resulted in some water accumulation in the turbine casing. Subsequently, the High Pressure Coolant Injection System was declared inoperable. As a result, this condition is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented fulfillment of the safety function at the time of discovery.
A technical evaluation of this event was performed and concluded that the HPCI system would have been operable with this condition. If HPCI turbine actuated with the estimated amount of condensate accumulated in the casing and connecting piping, it would have performed its safety function; the HPCI Turbine Exhaust Rupture Disc would not have been challenged by calculated peak pressures; and calculated water hammer loads were within specified load capacities of the turbine flange, downstream piping, struts, snubber, and spring hanger. Based on this, the condition reported in EN 55871 is being retracted. Notified R1DO (Bickett) | High Pressure Coolant Injection | |
| ENS 55593 | 18 November 2021 22:02:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Inoperable Due to Isolation Valve Failure to Automatically Open | On November 18, 2021, during the performance of High Pressure Coolant Injection (HPCI) surveillance testing, 23MOV-19 (HPCI PUMP DISCH TO REACTOR INBD ISOL VALVE) did not go open as expected while performing the sensed low water level portion of the test. The ability to manually open 23MOV-19 from the control room was unaffected as such, the HPCI system remained available for use. Failure of 23MOV-19 to open automatically prevents the HPCI system from performing its safety function as such this condition renders HPCI inoperable but available and is being reported as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(D). HPCI inoperable placed the licensee in a 14-day limiting condition for operation for Tech Spec 3.5.1.c. The NRC Resident Inspector was notified. | High Pressure Coolant Injection | |
| ENS 54657 | 10 April 2020 07:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Injection System Declared Inoperable | On April 10, 2020, at 0300 (EDT), an oil leak from 23PCV-12, HPCI (High Pressure Core Injection) Trip System Pressure Control Valve (PCV), resulted in the system being declared inoperable. This condition is being reported as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. | ||
| ENS 52664 | 4 April 2017 11:35:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Inadvertently Isolated During Conduct of Maintenance Surveillance | On April 4, 2017, at 0735 (EDT), the HPCI System was inadvertently isolated during the performance of l&C (Instrument and Control) testing. Technicians were in the process of performing instrument surveillance tests for the HPCI (high pressure coolant injection) System (using Allowed Out of Service Times) when a trip signal was applied to the incorrect instrument. This caused a HPCI System isolation signal on High Area Temperature, resulting in the closure of the HPCI steam isolation valves and rendering the system inoperable and unavailable. RCIC was immediately verified to be operable. The surveillance testing was aborted and system restoration is in progress. This condition is being reported as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(D). This placed the plant in a 14-day LCO action statement under Technical Specification 3.5.1. The licensee has notified the NRC Resident Inspector. | High Pressure Coolant Injection | |
| ENS 51613 | 18 December 2015 22:22:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Trains of Containment Atmosphere Dilution System Inoperable | On December 18, 2015 at 1722 EST, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, JAF received a notification pursuit to 10 CFR 21.21(d)(3)(ii) related to Moore Industries RTD temperature transmitters. Specifically, wire insulation in T2 transformer was damaged during assembly which reduced the insulation resistance and dielectric breakdown between the windings of the transformer. This equipment is in both redundant trains (A and B) of the Containment Atmosphere Dilution (CAD) System. Preliminary review by Operations and Engineering, which was completed on 12/18/15 at 2100 EST, determined the Part 21 results in both trains of CAD being inoperable and the applicable Technical Specification (TS) for both redundant trains of CAD being inoperable was entered. Per TS 3.6.3.2 Condition B, this places the unit in a 7-day shutdown LCO, provided the hydrogen control function is maintained. Per the TS Bases, the alternate hydrogen control capabilities are provided by the Primary Containment lnerting System, which is unaffected. The event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D), as an event or condition that could prevent fulfillment of a safety function. The licensee notified the NRC Resident Inspector. | Primary containment | 05000333/LER-2015-008 |
| ENS 50979 | 12 April 2015 22:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Temporary Loss of Control Room Envelope Boundary | The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(3)(v)(D) to notify the NRC of a temporary loss of the Control Room Envelope (CRE) boundary. The safety function of the CRE boundary is to ensure the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants. In addition to an intact CRE boundary maintaining CRE occupant dose from a large radioactive release below the calculated dose in the licensing basis consequence analysis for DBAs, it also ensures the occupants are protected from hazardous chemicals and smoke. The loss of the CRE boundary was due to a failed latching mechanism for a CRE boundary door used for normal passage of personnel into and out of the CRE. The failure of the door to latch as designed is considered a condition that could have prevented the fulfillment of a safety function at the time of discovery, and is therefore reportable as required by paragraph 50.72(b)(3), 'Eight-hour reports.' Procedural controls have restored the safety function of the CRE boundary by mechanically locking the subject door in the closed position through the use of a specifically designed mechanical strong-back until a permanent repair is made. The NRC Resident Inspector has been notified.
The main control room corridor fire door (76FDR-A-300-10) was found to not be able to latch. The latch was stuck in the latch mechanism because the latch bolt was bent. The latch was replaced on 4/15/2015. The Control Room Emergency Ventilation Air Supply System (CREVAS) provides a protected environment from which occupants can control the plant following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. The Control Room Envelope (CRE) is the physical boundary around the CREVAS environment. The Operability of the CRE boundary depends on its ability to minimize in-leakage of unfiltered air such that after a design bases accident a habitable environment can be maintained for 31 days without exceeding 5 rem whole body dose or its equivalent to any part of the body. The control room is normally pressurized greater than the 0.125 inches water gauge. This causes air to leak out rather than allowing infiltration of air from surrounding areas into the CRE boundary. The pressurized control room pushes this door (76FDR-A-300-10) outward, toward the open direction; however, even though the latch to the door did not work the door was still able to close. The closed door minimized in-leakage and a positive differential pressure was maintained in the control room during this event. These doors are kept closed against the door seals primarily by the closure mechanism. The latch is a secondary means of ensuring that the doors remain closed as well as a means to control personnel access to the control room. The Control Room Envelope (CRE) remained Operable with this deficiency and there was no loss of safety function per 10 CFR 50.72(b)(3)(v)(D). The original notification may be retracted. The licensee has notified the NRC Resident Inspector. Notified the R1DO (Powell). | Control Room Emergency Ventilation | |
| ENS 50532 | 13 October 2014 23:35:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Injection Degraded Accident Mitigation Capability | 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 50210 | 18 June 2014 19:45:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of Hpci Room Cooling | At 1545 (EDT), while testing of the Emergency Service Water system (ST-8Q) was being performed at the James A. FitzPatrick Nuclear Power Plant (JAF), two of five unit coolers (66UC-22H and 66UC-22K) in the East Crescent were found with indicated flow of 0 gpm. The other three unit coolers in the East Crescent Area were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation Subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation Subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from the areas, in the event that ECCS and RCIC were used to mitigate the consequences of an accident. The West Crescent Area Ventilation Subsystem remained functional. The accident mitigating function of the division of ECCS and RCIC located in the West Crescent Area were unaffected by this condition. However, this condition could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition could have prevented fulfillment of the safety function of HPCI and it is being reported under 10 CFR 50.72(b)(3)(v)(D). As part of the testing, the throttle valves to the unit coolers (66UC-22H and 66UC-22K) were cycled and normal flow was restored. This condition no longer exists. The licensee is investigating the loss of flow to the "H" and "K" unit coolers and the restoration of flow by cycling the unit cooler supply throttle valves. The licensee will be notifying the NRC Resident Inspector.
FitzPatrick is retracting EN # 50210 made on June 18, 2014 at 2120 EDT. The plant was at 86% power at the time. The ENS notification was an 8-Hr non-emergency notification to 10 CFR 50.72(b)(3)(v)(D) when it was discovered that two of five unit coolers in the East Crescent (66UC-22H and 66UC-22K) were found with indicated flow of 0 gpm while testing. The other three unit coolers in the East Crescent (66UC-22B, 66UC-22D, 66UC-22F) were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from these areas in the event that ECCS and RCIC are used to mitigate the consequences of an accident. As part of testing, throttle valves to unit coolers 66UC-22H and 66UC-22K were cycled and normal flow was restored. The West Crescent Area Ventilation subsystem remained functional. The accident mitigating function of the division of the ECCS and RCIC located in the West Crescent Area were unaffected by this condition. Initial review of this condition determined that it could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition was initially reported under 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented fulfillment of the Safety function of HPCI. This EN# 50210 is being retracted based upon a subsequent engineering analysis that determined that there is reasonable assurance that the three unit coolers with sufficient flow (66UC-22B, 66UC-22D, and 66UC-22F) would have been capable of removing accident heat loads as a function of time to maintain East Crescent area temperatures at a value which ensures operability of supported equipment. The analysis considered unit cooler heat transfer capability at the modified design condition flow of 22 gpm for historically observed lake temperatures and for flow at tested conditions. Additional margin in flow at the tested condition provided increased heat removal capability and provided added assurance that accident heat load would have been removed. The East Crescent Area Ventilation subsystem was, therefore, functional with three unit coolers (functionality never was lost) and the supported ECCS remained Operable. The Operability determination for the condition has subsequently been revised based upon the engineering analysis, to state the condition was not immediately reportable per 10 CFR 50.72. The licensee has notified the NRC Resident Inspector Notified R1DO (Kennedy) | Service water High Pressure Coolant Injection Reactor Core Isolation Cooling Emergency Core Cooling System | |
| ENS 49660 | 18 December 2013 20:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Safety Function for the Hpci Suction Automatic Low Cst Level Swap Over Inoperable | During the performance of surveillance testing on 12/17/2013 for the HPCI CST low water level switch instrument functional test, the safety function for the low CST level suction swap-over was lost. While performing testing, 23LS-75B setpoint was found to be below the TS required value with 23LS-74B already declared inoperable. Tech Spec 3.3.5.1 Condition D was entered and within 1 hour of finding 23LS-75B setpoint below the TS required value, this switch was adjusted to within tolerance as allowed by procedure and tested satisfactory. Both switches were inoperable concurrently for a period of less than 1 hour from time of discovery of 23LS-75B being out of tolerance. This issue has been discussed with the NRC Resident Inspector. | 05000333/LER-2013-006 | |
| ENS 48410 | 16 October 2012 11:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Shutdown Cooling Isolation Capability Lost | 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. | Shutdown Cooling Residual Heat Removal | 05000333/LER-2012-006 |
| ENS 48270 | 3 September 2012 06:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Inoperable Due to Erroneous Indication on Flow Indicating Controller | At 0225 EDT on September 3, 2012, with the James A. Fitzpatrick Nuclear Power Plant (JAF) operating at 93% reactor power, High Pressure Coolant Injection (HPCI) was declared inoperable due to abnormal indication on the HPCI Flow Indicating Controller (FIC). The FIC was found to be indicating a HPCI System flow rate of 700 gpm while the system was in the standby lineup. Under these conditions, the capability of the system to achieve the required flow rate cannot be assured. This failure meets NRC 8 hour reporting criterion 10CFR50.72(b)(3)(v)(D). Reactor Core Isolation Cooling (RCIC) and other Emergency Core Cooling Systems (ECCS) remain operable. The NRC Resident Inspector has been notified.
The improper HPCI flow indication was determined to be due to minor air intrusion following restoration of the system after maintenance. The flow transmitter for the HPCI system was repeatedly vented with no air observed. The HPCI system has been restored to a normal standby line-up and is OPERABLE as of 9/4/2012 at 1415 EDT. The NRC Resident Inspector has been notified. R1DO (Conte) notified. | High Pressure Coolant Injection Reactor Core Isolation Cooling | |
| ENS 48258 | 30 August 2012 16:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Inoperable Due to Failed Pressure Control Valve | At 1215 EDT on August 30, 2012, with the James A. FitzPatrick Nuclear Power Plant (JAF) operating at 95% reactor power, High Pressure Coolant Injection (HPCI) was declared inoperable due to the failure of a pressure control valve on the HPCI oil cooling system. The failure of this pressure control valve results in a safety-valve lifting and releasing approximately 75 gallons per minute to the reactor building equipment drain tank. There was no release to the environment. This failure meets NRC 8 hour reporting criterion 10CFR50.72(b)(3)(v)(D). Reactor Core Isolation Cooling (RCIC) and other Emergency Core Cooling System (ECCS) systems remain operable. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Reactor Core Isolation Cooling Emergency Core Cooling System | 05000333/LER-2012-002 |
| ENS 46355 | 23 October 2010 04:45:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Declared Inoperable Due to Power Supply Failure | The High Pressure Coolant Injection (HPCI) system was declared inoperable due to an instrument power supply failure. The cause of the failure is under investigation. All other ECCS, Emergency Diesels, and Reactor Core Isolation Cooling (RCIC) are operable. The licensee has notified the NRC Resident Inspector. | High Pressure Coolant Injection Reactor Core Isolation Cooling | |
| ENS 45008 | 22 April 2009 15:21:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection (Hpci) System Inoperable | On 4/22/09 at approximately 1121, during surveillance testing on the High Pressure Coolant Injection (HPCI) System, the associated Primary Containment Isolation valves closed. This isolation rendered the HPCI system incapable of performing its safety function and is, therefore, reportable under 10 CFR 50.72(b)(3)(v). Technical Specification LCO 3.5.1 Condition C had been entered at 0837 to support testing. The cause of the isolation is unknown at this time. Troubleshooting is in progress. LCO 3.5.1 Condition C provides 14 days for restoration of the HPCI System, the plant will remain in the LCO until the cause is determined and any associated repairs have been completed. A follow-up Licensee Event Report will be filed within 60 days. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Primary containment | 05000333/LER-2009-006 |
| ENS 42964 | 4 November 2006 23:10:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Inoperable | During HPCI testing with reactor pressure at 975 psig, HPCI flow and turbine speed oscillations occurred. The HPCI turbine was run for approximately one minute and was manually tripped by the operator when system conditions did not improve. A cause investigation is underway. All other ECCS and RCIC remain operable. The unit is in a 14 day LCO for this event. The NRC Resident Inspector has been notified. |