Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 47892 | 2 May 2012 19:02:00 | While investigating operating experience from another station it was determined that Fort Calhoun Station (FCS) is subject to similar conditions. The operating experience involved setpoint drift of safety related pressure switches beyond what had been accounted for in the station's safety analyses. Following investigation and evaluation, it was determined that pressure switches that provide safety related signals for high containment pressure to the reactor protection system (RPS) and engineered safeguards actuation circuitry may be similarly affected at FCS. The impact of the potential drift was evaluated, and it was determined that neither RPS nor the engineered safeguard circuitry may actuate at the required containment pressure of 5 psig. An evaluation determined that the actuation may not occur until slightly higher than the required pressure. Other systems are currently being evaluated to see if this same condition applies. The station is in MODE 5, refueling shutdown condition, and there is no immediate safety concern. The pressure instruments are located in the penetration area which is subject to elevated temperatures. The licensee notified the NRC Resident Inspector.
The condition was initially determined to be reportable under 10CFR50.72(b)(3)(ii)(B), plant in unanalyzed condition, based on a conservative assumption that the error introduced violated not only the Technical Specification limit (5.0 psig) but also the safety analysis limit of 5.4 psig, USAR Table 14.1-1. Subsequent evaluation of actual data concluded that the safety analysis limit was not exceeded and therefore not reportable under 10 CFR 50.72(b)(3)(ii)(B). LER 2012-004-1 reported this condition under 10CFR50.73(a)(2)(i)(B), 10CFR50.73(a)(2)(ix)(A), and 10CFR50.73(a)(2)(v)(A,B,C,D). Revision 2 of the LER will correct the reporting criteria. The NRC Resident Inspector was notified by the licensee. Notified the R4DO (Pick). |
| ENS 47202 | 26 August 2011 16:45:00 | During performance of IC-PM-VA-0200, Preventive Maintenance Air Flow Measurement of TSC Air Filter Unit VA-119, it was discovered that the required flow-rate of 2700 cfm and minimum overpressure of 0.1 inches of water in the TSC could not be achieved. Troubleshooting on VA-107, TSC Air Handling Unit, is in progress. This condition renders the Technical Support Center unavailable for Emergency Planning Responses. Approved compensatory actions are to relocate personnel to alternate facilities if required. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for Loss of Emergency Preparedness Capabilities. The licensee has notified the NRC Resident Inspector. |
| ENS 46989 | 26 June 2011 12:22:00 | At approximately 0125 CDT, the AquaDam providing enhanced flood protection for Fort Calhoun Station Unit 1 failed. This resulted in approximately 100 gallons of petroleum being released into the river after a protective barrier was breached and many fuel containers were washed out to the river. The fuel/oil containers were staged around the facility to supply fuel for pumps which remove water within the flood containment barriers. The spill was reported to the State of Nebraska at 10:45 AM CDT on 6/26/2011. This condition is being reported pursuant to 10 CFR 50.72(b)(2)(xi) for News Release or Notification of Other Government Agency. Applicable governmental agencies have been notified per plant procedures. The licensee notified the NRC Resident Inspector. |
| ENS 46506 | 23 December 2010 12:40:00 | At 1050 CST on 12/23/2010, Fort Calhoun Station experienced a reactor trip due to an unknown cause. No abnormal indications were present at the time of the trip. (Procedure) "EOP-00, Standard Post Trip Actions, was performed. At 1054 CST, transitioned to (procedure) EOP-01, Reactor Trip Recovery. All Safety Functions of EOP-01 are being satisfied. At 1207 CST, 345KV Backfeed was established to all 4160V buses. The plant is currently stable in Mode 3, and the station continues efforts to determine the cause of the reactor trip. All tripable control rods fully inserted during the reactor trip. No PORVs or safety valves lifted. The steam generators are being fed from the Main Feed Pumps and decay heat is being removed to the main condenser. No Emergency Safety Features actuated during the trip except for the reactor trip. The licensee notified the NRC Resident Inspector. |
| ENS 46490 | 19 December 2010 01:39:00 | At 2355 CST, on 12/18/2010, the Control Room was notified by security of the inability of offsite personnel to call into the plant. (At) 0008 CST, on 12/19/2010, the control room verified that the ENS Phone, Conference Operation Network (COP), Security Building, Training Center, and Blair Phone Lines were not functional. At 0011 CST, a Notice of Unusual Event (NOUE) was declared per IC SU6 EAL 2. This also meets the criteria for a Major Loss of Communications Capability under 10 CFR 50.72(b)(3)(xiii). At 0028 CST, the control room was notified by security that Huntel Communications was attempting to correct the communications issue. At 0030 CST, the COP phone was restored. At 0046 CST, the ENS phone was verified to be functional. At 0050 (CST), the criteria to exit the NOUE (was met) and the event was terminated. At 0059 (CST), the NRC Senior Resident Inspector was notified. The state and local agencies were notified. The NRC Operations Center conducted a satisfactory test of the ENS line. |
| ENS 46254 | 15 September 2010 21:00:00 | At 1740 CDT, VA-81A, Hydrogen Analyzer Panel, was declared not functional due to failing surveillance test OP-ST-VA-0006, Containment Hydrogen Monitor Monthly Check. VA-81B, Hydrogen Analyzer Panel, was previously not functional due to performance of surveillance test IC-ST-VA-0033, 18 Month Channel Calibration of Containment Hydrogen Analyzer, VA-81B. This results in no Hydrogen Analyzers being available to monitor containment, which prevents being able to assess for potential loss of containment barrier for Emergency Action Level purposes via the containment hydrogen greater than 3% method. USAR (Updated Safety Analysis Report) section 9.10.2.5 allows for both Hydrogen Analyzers to be out of service for up to 72 hours. The licensee informed the NRC Resident Inspector.
Fort Calhoun Station had previously reported that the loss of both hydrogen monitors on September 15, 2010, constituted a major loss of emergency response assessment capability. Additional investigation has determined that other methods of assessment were available that would have provided sufficient indication to make the proper emergency classification. Therefore, this event is being retracted. The licensee has notified the NRC Resident Inspector. Notification was sent to R4DO (Thomas Farnholtz). |
| ENS 45898 | 4 May 2010 16:56:00 | On 5/4/2010 at 15:12 CDT when attempting to flush CH-1A charging pump, a loss of charging flow occurred. CH-193 discharge valve for CH-1A charging pump was found to be open. This resulted in a flow path to the auxiliary building sump tank when CH-356 charging pump CH-1A discharge drain valve to waste disposal system was opened. This resulted in an approximately 38 gpm leak from the reactor coolant system via letdown to waste. This leak was isolated within 1 minute. The site entered and exited the conditions for NOUE before the shift manager was able to make an E-plan call. There are currently no emergency conditions on site and an NOUE was never declared. The reactor continued to operate at 100% power throughout the event. The cause of this event is believed to be a valve line-up error. The NRC Resident Inspector has been notified. The site plans to notify the State of Nebraska. No other notifications are planned.
The declaration of Unusual Event SU 5 EAL 2, RCS leakage, on 5/4/10 is being retracted, because the declaration was inaccurate. The leak was an intersystem leak of the CVCS, not RCS leakage, and the leak was isolated as previously noted. Therefore, the leakage did not meet the initiating condition for the EAL. The licensee has notified the NRC Resident Inspector and plans to notify the State of Nebraska. Notified R4DO (Farnholtz), NRR EO (Quay), and IRD (McDermott). |
| ENS 45828 | 8 April 2010 20:34:00 | 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 45389 | 29 September 2009 20:09:00 | At 15:09 CDT, today the Technical Support Center ventilation system stopped running. The cause for the failure of the Technical Support Center ventilation system is suspected to be an interlock between the fire detection system and the ventilation unit. The cause of the this condition renders the Technical Support Center unavailable for Emergency Planning Responses. Alternate facilities are available. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for Loss of Emergency Preparedness Capabilities." The licensee has notified the NRC Resident Inspector.
Technical Support Center ventilation system is now functional and is available for Emergency Planning Responses. The licensee has notified the NRC Resident Inspector. Notified R4DO (Mike Shannon) |
| ENS 46594 | 4 February 2011 18:17:00 | On September 9, 2009, the NRC Component Design Basis Inspection (CDBI) Team identified Fire Protection penetrations on the west side of the Intake Structure were not sealed and it has been determined that the penetrations were below the USAR (Updated Safety Analysis Report) credited flood level. Flooding through the penetrations could have impacted the ability of all the station Raw Water Pumps to perform their design accident mitigation functions. Reference Fort Calhoun Station Condition Report 2009-4166. This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v). This report should have been made on September 9, 2009, and is late. Subsequent review of the issue determined this reportability. The penetrations have since been sealed. The licensee notified the NRC Resident Inspector. See EN #46590 dated 2/3/2011 for a similar event at Fort Calhoun. |