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| Entered date | Event description | |
|---|---|---|
| ENS 41530 | 28 March 2005 00:35:00 | The following information was obtained from the licensee via facsimile (licensee text in quotes): While in OPCON 2 with the Main Turbine secured, Hope Creek personnel performed a Primary Containment entry to determine the source of slightly elevated unidentified leakage. This leakage has slowly trended up since plant start-up in February to approximately 0.73 gpm. With the current unidentified leak rate at 0.436 gpm, personnel identified a steam leak from an insulated decontamination port, sealed via a bolted flange, within the isolable boundary of the 'B' Reactor Recirculation pump. This was the only source of reactor coolant system leakage identified although some minor leakage (totaling approximately 250ml/min) from two drywell recirculation fans was also identified. This is a voluntary/courtesy Emergency Notification System (ENS) notification that Hope Creek is proceeding to Cold Shutdown to precisely identify and repair the leakage source. At the time of this notification, Hope Creek Generating Station is in OPCON 3 with plant cooldown in process. The licensee reported this event under 10 CFR 50.72(b)(2) as a 4-Hour Non-Emergency Voluntary notification. The NRC Resident Inspector has been notified and the licensee plans to notify LAC Township. |
| ENS 40996 | 30 August 2004 02:54:00 | On 08/29/04 at 1928 hours (EDT), an alarm was received in the control room indicating the reactor water cleanup (RWCU) high differential flow isolation channel for the outboard RWCU supply isolation valve was inoperable. At the time, the channel functional test for the inboard RWCU high differential flow isolation channel was being performed. During the channel functional test, power to the inboard isolation valve is removed to permit continued RWCU system operation while the isolation function is tested. With the outboard isolation actuation instrumentation channel inoperable and the inboard isolation valve deenergized, in the event of a break in the RWCU system, the required isolation function may not be completed. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. In accordance with Technical Specification requirements, the RWCU system was removed from service and the inboard and outboard supply Isolation valves were closed. There is no immediate impact on water chemistry and the licensee is investigating the cause of the original failure as well as possible long term degradation of the reactor water chemistry. The licensee notified the NRC Resident Inspector and will notify the local township authorities of this incident. |
| ENS 40648 | 6 April 2004 02:17:00 | On 04/05/04 at 2030 hours, Engineering personnel informed Operations personnel of an issue affecting the ability of the offsite power sources to provide adequate 1E bus voltage consistent with the design basis. Operating procedures currently contain non-conservative values for minimum voltage on 1E 4 kV buses. In addition, the transformer auto load tap changer (LTC) is currently set to regulate at approximately 4200 VAC, a value that is below the required 1E 4 kV bus lower voltage design limit. Adequate voltage from the offsite power sources is required IAW General Design Criteria 17 to ensure that vital buses remain connected to their preferred power source and adequate terminal voltage exists at the load end device during accident conditions. The ability of the Emergency Diesel Generators to perform their design function is not affected by this condition. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function because it affects the ability of both offsite power sources to provide adequate voltage to all 1E buses to properly mitigate the consequences of an accident. Both offsite electrical power sources and all four 4 kV distribution buses have been declared inoperable and the appropriate Technical Specification Actions have been entered. The current voltage readings for all 1E 4 kV buses are between 4204 and 4263 VAC. The plant will remain in Operational Condition 4 until this condition is corrected. Engineering and Operations personnel are evaluating this condition to determine required corrective actions. All other plant systems are available to support Operational Condition change and reactor startup. The licensee notified the NRC Resident Inspector and will notify the local township. |
| ENS 40437 | 12 January 2004 12:30:00 | On 01/12/04 at 1048 hours, the Hope Creek Generating Station reactor was manually scrammed following an invalid containment isolation signal on Reactor Building High-High Radiation. The invalid signal was caused by the combination of a scheduled sensor calibration on channel 'C', coincident with an emergent failure on channel 'A.' This combination of trip signals made up the two out of three trip logic for the Reactor Building High-High Radiation containment isolation signal. While recovering from the spurious isolation signal, the operating crew observed two of the inboard MSIV's drifting closed from a loss of pneumatic pressure as a result of the isolation signal. In response to this condition, the operating crew manually scrammed the reactor. A low reactor water level scram signal was received at 12.5 inches as expected, and reactor level was subsequently returned to the normal band using the reactor feedpumps. At the time of this event, the 'A' Control Room Ventilation Train was inoperable but available pending emergent corrective maintenance. The 'C' channel Reactor Building Radiation monitor has been returned to service and is operable, and the 'A' channel remains failed in the tripped condition. All other systems functioned as expected, and a post-transient review team is being assembled to investigate the event. Decay heat is being removed via steam to the main condenser using the bypass valves. The condensate and feedwater system is in operation maintaining reactor vessel water level. No SRVs lifted during the transient and the electrical system is stable in a normal lineup. The licensee notified the NRC Resident Inspector and will be notifying the LAC Township. |
| ENS 40225 | 5 October 2003 06:26:00 | While performing common mode failure testing of the Emergency Diesel Generators (EDG), the 'C' EDG was declared INOPERABLE for planned installation of required test equipment. Concurrent with the inoperability of the 'C' EDG, the 'B' Control Room Emergency Filtration (CREF) System has been INOPERABLE for emergent corrective maintenance since 10/2/03 at 0502. Because the 'C' EDG is the emergency power supply for the 'A' CREF train, 'A' CREF was also declared INOPERABLE and Technical Specification 3.0.3 was entered as of 0300 hrs on 10/05/03. At 0430 hrs on 10/05/03, the test equipment was removed from 'C' EDG, thereby restoring it and 'A' CREF to an operable status, and Technical Specification 3.0.3 was exited. Testing did verify the absence of a common mode failure and all EDG's are operable. The Control Room Ventilation System provides heating, cooling, ventilation, and environmental control for the control room and adjacent areas. Under accident conditions, CREF ensures that the control room will remain habitable during and following all design basis accidents. Because the CREF system is required to automatically respond in the event of a design basis accident, having both trains of CREF inoperable at the same time impacted the ability to mitigate the consequences of an accident. Therefore, this event is being reported in accordance with 10CFR50.72(b)(3)(v)(D). The plant is currently in HOT SHUTDOWN for repair of an emergent turbine hydraulic fluid leak, with decay heat removal to the main condenser via turbine bypass valves. The NRC resident inspector was notified by the licensee.
At the time of the original notification, both trains of Control Room Emergency Filtration (CREF) were declared inoperable impacting the ability of CREF to mitigate the consequences of an accident. The "B" train was inoperable for emergent corrective maintenance and the "A" train was declared inoperable when test equipment was connected to the "C" emergency diesel generator (EDG). The "C" EDG provides emergency power to the "A" train of CREF. Subsequent to this event, an evaluation of the test equipment impact to the "C" EDG was performed and determined that the "C" EDG would still be capable of providing emergency power to the "A" CREF train in the event offsite power is lost. Therefore, the "A" CREF train remained available to respond to a design basis accident. Thus, the safety function would have been fulfilled." R1DO (Brain McDermott) notified. The NRC Resident Inspector will be notified of this retraction by the licensee. |