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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 45957 | 26 May 2010 19:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram While Increasing Power | At 1526 on 5/26/2010, while operators were increasing power with reactor recirculation flow, an RPS (Reactor Protection System) actuation occurred in both channels and all control rods inserted. RPV (Reactor Pressure Vessel) level decreased to 114.5 inches (Low level setpoint is less than 127 inches). Following the scram, the PCIS (Primary Containment Isolation System) groups 2, 3, 4 and 5 received actuation signals and all open valves isolated. Both trains of standby gas treatment system actuated. Plant actions taken included entering procedures OT-3100, Reactor Scram on RPS Actuation and EOP-1, RPV Control on Low Level Signal. The EOP-1 was exited per shift manager direction because of no emergency. The operators stabilized the plant and reset both RPS and PCIS. An investigation into the cause of the scram is continuing. Electrical power is being supplied from offsite sources through the startup transformers. The licensee notified the NRC Resident Inspector. | Reactor Protection System Primary Containment Isolation System Reactor Pressure Vessel Standby Gas Treatment System Control Rod | |
| ENS 43610 | 30 August 2007 19:13:00 | 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 Turbine Stop Valve Closure | Event Description: Reactor scram (4 hr notification) automatic scram Primary containment isolation of Groups 2,3,4, and 5 due to RPV Level < low level setpoint (<127") due to reactor scram. (8 hr notification). Actions Taken (reference applicable Technical Specifications): Implemented OT 3100 (Reactor Scram Procedure) EOP-1 (RPV Control). Placed the plant in a stable condition and implemented OP 0109, Plant Restoration. The NRC Resident Inspector was notified of this event by the licensee. Reactor was initially at approximately 63% power due to cooling tower damage which occurred more than a week ago. License was performing a surveillance test of the # 2 turbine stop valve. The valve was shut per the surveillance test procedure but they were unable to open the valve. Personnel were in the heater bay and mechanical assistance was applied to open the valve. The valve opened quickly at which point the licensee received a turbine stop valve closure signal which generated an automatic reactor scram. All rods fully inserted into the core. Reactor vessel water level decreased below 127 inches, due to the reactor scram, which caused primary containment isolation of groups 2,3,4 and 5. Reactor vessel water began to increase because Reactor feedwater pumps "A" & "B" were still operating. Reactor feedwater pump "B" was secured. When reactor vessel water increased to 173 inches, high level alarm, reactor feedwater pump "A" automatically tripped. Highest reactor vessel water level increased to was approximately 179 inches. No SRV's opened. All Emergency Core Cooling Systems, EDGs are fully operable if needed and the electrical grid is stable. Reactor vessel water level is being maintained using a reactor feedwater pump. Only other anomaly was that for some unknown reason automatic pressure control went to mechanical pressure control during the transient. Licensee is investigating the event. | Feedwater Primary containment Emergency Core Cooling System | |
| ENS 41868 | 25 July 2005 04:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip Caused by Failure in Switchyard | At 1525 the plant experienced a load reject generator trip due to a catastrophic failure in the 345 Kv switchyard. A reactor scram occurred as a result. The degraded AC power system prevented a fast transfer from occurring. Degraded bus voltage caused the emergency diesel generators (EDGs) to start. A residual bus transfer restored power to the 4 Kv busses. The (main steam isolation valves) (MSIVs) closed on a low-low reactor water level of 82.5 inches. (Reactor Core Isolation Cooling) and (High Pressure Coolant Injection) (HPCI) also started on the low-low reactor vessel water level. The (Safety Relief Valves) were cycled twice for pressure control. OT 3100 Reactor Scram procedure was executed. EOP-3 was entered due to elevated torus water temperature and both loops of (Residual Heat Removal) (RHR) are in torus cooling. Water level has restored and is being maintained by feedwater. The MSIVs have been reopened and the scram reset. EDGs were secured. The plant is currently shutdown and stable with all control rods fully inserted. Decay heat removal is being accomplished with HPCI in pressure control mode. The licensee is transitioning to feeding with normal feedwater and steam exhausting through drains. Both trains of RHR are providing torus cooling. Electric power is being provided by offsite power. The licensee is currently investigating the event in the switchyard. The licensee notified the NRC Resident Inspector and will issue a press release.
Corrected incorrect entry for Scram Code from N (N/A) to A/R (Automatic/with Rod Motion). R1DO (Glenn Meyer) notified. | Feedwater Emergency Diesel Generator Decay Heat Removal Control Rod | 05000271/LER-2005-001 |