|Entered date||Site||Scram||Region||Reactor type||Event description|
|ENS 52033||22 June 2016 13:02:00||Fort Calhoun||Automatic Scram||NRC Region 4||At 0841 (CDT) an automatic turbine trip occurred, resulting in an automatic reactor protective system (RPS) actuation due to loss of turbine load. The source of the turbine trip was from the distributed control system (DCS) and is being investigated via a root cause analysis. This was an uncomplicated trip, all systems responded as expected post trip, and the reactor trip recovery procedure was entered at 0852 (CDT). The plant is stable in Mode 3 with a normal electrical line up and decay heat removal via steam dumps to the condenser. The NRC Resident Inspector has been notified.|
|ENS 50688||17 December 2014 12:56:00||Fort Calhoun||Automatic Scram||NRC Region 4||At 1014 (CST) on 12/17/2014, the reactor automatically tripped due to loss of load on RPS (Reactor Protection System). Preliminary information indicates turbine-generator trip caused an uncomplicated reactor trip . All safety functions are met. Currently maintaining Mode 3, Hot Shutdown. At time of trip, non-safety related house transformers, T1A1 and T1A2 became inoperable as expected due to generator trip. Entered Technical Specification 2.7 (2)a, 72 hr LCO (for the inoperable T1A1 and T1A2 transformers). All control rods fully inserted. All busses are energized via offsite power. Decay heat is being released via AFW and the condenser bypass valves. The unit is stable in Mode 3. Cause of the loss of load is being investigated. The NRC Resident Inspector has been informed.|
|ENS 49926||17 March 2014 15:55:00||Fort Calhoun||Automatic Scram||NRC Region 4||Ft. Calhoun station automatically tripped due to a loss of turbine load. The turbine tripped due to loss of stator cooling water. Maintenance was in progress on the stator cooling system when inventory was lost and low pump discharge pressure caused an automatic turbine trip and reactor trip. All systems operated as expected. Ft. Calhoun station is shutdown and stable in mode 3 at this time. All control rods fully inserted into the core and decay heat is being removed using the normal condenser steam dump system. The licensee has notified the NRC Resident Inspector.|
|ENS 49717||12 January 2014 06:44:00||Fort Calhoun||Manual Scram||NRC Region 4|
After achieving criticality a deviation between control rods was observed by plant personnel. When attempting to level the control rods, one rod could not insert to the level of the rest of the group. A manual reactor trip was initiated by the operating crew. All tripable control rods fully inserted into the core. The trip was uncomplicated and the licensee is investigating the cause of the control rod position deviation. The licensee has notified the NRC Resident Inspector.
Report was updated to indicate 8-hour reportable criteria for Valid Specified System Actuation (Reactor Protection System). The licensee will notify the NRC Resident Inspector. Notified the R4DO (Kellar).
|ENS 45828||8 April 2010 20:34:00||Fort Calhoun||Manual Scram||NRC Region 4||CE|
At 1622 hours CDT, an electrical ground on 480 Volt Bus 1B3A was determined to be from a supply cable to Motor Control Center (MCC)-3A1. Isolating loads on this MCC required securing power to HCV-1385, Steam Generator RC-2B Inlet Isolation Valve. This condition results in the valve being unable to close on a Steam Generator Isolation Signal (SGIS), which requires entry into Technical Specification 2.0.1(1). This Technical Specification requires the plant to be placed in a Hot Shutdown condition within 6 hours. A plant shutdown to Mode 3 was commenced at 1740 hours CDT. The licensee notified the NRC Resident Inspector.
At 2123 hours CDT, the Reactor was manually tripped from 22% Reactor Power in order to meet the requirements of Technical Specification 2.0.1(1) and have the reactor in a hot shutdown condition. All systems functioned properly. At 2124 CDT, the plant entered Mode 3 Shutdown Condition. AT 2235 CDT, HCV-1385, Steam Generator RC-2B Inlet Isolation Valve has been manually closed. Technical Specification 2.0.1(1) has been exited. The licensee is reporting the manual scram under 10CFR50.72(b)(2)(iv)(B). The licensee notified the NRC Resident Inspector.
|ENS 40156||13 September 2003 06:10:00||Fort Calhoun||Manual Scram|
|NRC Region 4||CE||This report is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to manual initiation of the RPS, although briefed, was not part of any written pre-planned sequence. The unanticipated negative ASI response that began to approach RPS APD trip settings drove the decision to manually initiate an RPS trip. Fort Calhoun Station Unit 1 was being shutdown for a planned refueling outage. During scheduled power shutdown at 3% per hour, ASI was trending more negative (power shifting to top of core) than expected. Reactor Engineering was contacted about the trend at 67% reactor power on September 11, 2003. New guidance was issued, however negative ASI trend continued. RPS Axial Power Distribution (APD) pretrip/trip setpoints were -0.55/-0.58 ASI Units. Below 15% power APD trips are bypassed. Operations briefed the continued power descension with manual trip criteria being the receipt of any 1 of 4 APD pretrip setpoints. At 15.8% delta T power it became apparent that with ASI approaching APD pretrips (-0.5305 ASIU recorded), an Automatic Trip could be initiated if the downpower continued. At 2055 on September 12, 2003, with power at 15.8% delta T, a briefing was conducted and an manual reactor trip was initiated by tripping the Primary RPS Trip pushbutton on Control Board-4. Operations entered EOP-00, "Standard Post Trip Actions" and transitioned to EOP-01, "Reactor Trip Recovery" with all safety functions verified. EOP-01 was exited at 2200 hours and the plant entered OP-3A, "Plant Trip". The plant is currently stable in Mode 3 with RCS Tave at 535 degrees F. The only abnormal plant response noted was a feedwater side relief valve lifted and failed to fully reseat on FW-16A, High Pressure Feedwater Heater. The relief valve was cycled by maintenance and was reseated approximately 10 minutes after opening. The NRC Resident Inspector was notified of this event by the licensee.|