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| Entered date | Event description | |
|---|---|---|
| ENS 50214 | 20 June 2014 00:29:00 | The National Weather Service predicts that the Missouri River level at Fort Calhoun Station will exceed 1004 feet above mean sea level on 6/20/14 at approximately 2300 CDT. Fort Calhoun Station will begin a ramp down in power to satisfy technical specification 2.16 which states, 'When the Missouri River level reaches elevation 1004 feet mean sea level, the reactor shall be in a HOT SHUTDOWN condition (Mode 3) and in Cold Shutdown (Mode 4) within 36 hours following entry into Hot Shutdown.' The river level is currently 998 feet 3 inches and rising approximately 0.5 inches per hour. At time 0001 CDT 6/20/14 Fort Calhoun station will initiate a plant shutdown to Hot Standby and will proceed to a Cold Shutdown condition within 36 hours following entry into Hot Shutdown, as required. The licensee notified the NRC Resident Inspector.
EN 50214 is being retracted. At 2300 (CDT) on 6/21/14, actual Missouri River level at Fort Calhoun Station peaked at 1001' 2" mean sea level (msl) and did not reach TS 2.16 (1) shutdown criteria of 1004' msl. Based on current Army Corps of Engineers and National Weather Service published Missouri River levels, the river upstream of Fort Calhoun Station has peaked and is trending down following the recent storms. River level is predicted to continue trending downward to normal summer navigation season levels. Plans to shutdown have been terminated. Fort Calhoun Station will continue to monitor weather and river levels. Current reactor power is 66% and Fort Calhoun Station is raising power to 100%. The lowest reactor power achieved was 30 percent at 1200 CDT on 06/20/2014. The licensee notified the NRC Resident Inspector. Notified R4DO (Hay) |
| ENS 49452 | 18 October 2013 19:38:00 | A postulated High Energy Line Break (HELB) between the Letdown Heat Exchanger and its containment penetration has been identified in the Fort Calhoun Station calculation FC07885, Rev. 0, Stress Analysis of Small Bore Piping on Isometric CH-4106 High Energy Line Break Assessment. Based on this, there is a potential for a HELB on the inlet of the Letdown Heat Exchanger that may adversely impact the upstream piping and the outboard containment isolation valve. Assuming a single failure of the inboard isolation, this condition has the potential to degrade a principal safety barrier by bypassing the containment building. The licensee has notified the NRC Resident Inspector. |
| ENS 49378 | 23 September 2013 18:37:00 | At 1340 CDT, on 09/23/2013, as part of a vendor analysis for the high energy line break reconstitution project, it was determined that Room 81 and 82 epoxy floor coatings do not meet the design basis requirements for a high energy line break barrier. This is an unanalyzed condition based on 10 CFR 50.72(b)(3) as loss of the floor coating could affect multiple redundant trains of safety-related equipment during a design basis event. The plant is currently in a cold shutdown condition. The licensee has notified the NRC Resident Inspector. |
| ENS 48094 | 11 July 2012 18:10:00 | Fort Calhoun Station is making an 8-hour verbal report per 10CFR50.72(b)(3)(ii)(B) Unanalyzed Condition. An internal containment support beam (B-22) has been identified by the station as not passing the required load combination as stated in the USAR for at power conditions. Beam B-22 is the designation for the two beams that directly support Safety Injection Tanks 6B and 6D. This beam was also identified as having potential loading conditions outside the allowable limits for the load combination for shutdown conditions. Specifically, it was determined that in order to bring the beam loading to within acceptable levels, the allowable floor live load would need to be reduced from the current designated load distribution of 200 pounds per square foot (psf) to 140 psf. A walkdown of the area by Design Engineering estimates the current floor live load is approximately 100 psf. Compensatory actions are being established to remove any equipment that is contributing to current live loading of the support beam and to isolate and post the affected area to ensure no equipment is stored without engineering analysis. The licensee has notified the NRC Resident Inspector. |
| ENS 47953 | 23 May 2012 19:54:00 | During inspections to determine the physical integrity of a failed pressurizer heater it was determined that the heater (number 26) was cracked. Due to the location of the pressurizer heater crack, this is considered a degradation of the RCS Barrier. The initial visual inspection of heater 26 in November of 2011, did not identify the cracking. During efforts to remove the heater, a crack was observed on May 21, 2012. The crack is above and below the heater support plate. The crack is an axial crack showing some branching. The crack is about an inch above and inch below the heater support plate. These inspections were being performed as a result of the operating experience at the Sizewell B reactor in the United Kingdom. The licensee has notified the NRC Resident Inspector. |
| ENS 47774 | 27 March 2012 14:06:00 | At 1145 CDT on 3/27/12, OPPD (Omaha Public Power District) was notified by Harrison County Emergency Management that the sirens in Harrison County Iowa were inadvertently activated at 0950 CDT on 3/27/12. This activation occurred during an exercise at Fort Calhoun Station which included FEMA evaluation of local State and County participation. The sirens were activated by Iowa Emergency personnel for less than 5 seconds. An OPPD and State of Iowa joint media press release will be conducted following the termination of the Emergency Planning exercise. The licensee notified the State of Iowa EMA and the Harrison County Sherriff's Dispatch. The licensee also notified the NRC Resident Inspector. |
| ENS 47848 | 18 April 2012 18:53:00 | The Waste Disposal System (WDS) Class 1 piping requires operable seismic supports downstream of the isolation valve class break. Currently, eight (8) INC (International Nuclear Safety, Corp.) snubbers have been degraded to (Non Nuclear System) NNS Class 4 ridged struts. The snubbers original design function was to allow thermal motion but restrain seismic motion. The snubbers have been identified as potential to create an unanalyzed condition that over stresses the safety class 1 drain pipe upstream of the isolation valve if the snubbers on the drain pipe downstream of the isolation valve were in a locked condition (acting as a strut). Per NRC bulletin 81-01, these snubbers are assumed to be frozen and do not allow movement of the pipe; thus, they have been degraded to rigid struts as they are not in the snubber program and are not tested. They still provide a seismic safety function for (class) II/I issues and act as a strut to provide horizontal restraint to the WDS piping. The snubbers were removed from the piping system and tested to determine their performance and if they would have moved to allow thermal growth. Six snubbers failed the test and were either in a locked condition or their movement was dimensionally small relative to the required movement. The (Reactor Coolant System) RCS is within acceptable stress values with the snubbers removed. The 8-hour regulatory reporting time has been exceeded. An initial Reportability Evaluation was completed on March 26, 2012 and had determined the supports were operable. A second Reportability Evaluation later determined the supports have been inoperable since October 6, 2011. The WDS is used to drain the RCS. The licensee will notify the NRC Resident Inspector.
Additional review and testing demonstrated that (there was) no degradation of the RCS from thermal fatigue. The analysis demonstrates adequate past performance of the snubbers with regard to thermal fatigue. The impact of the snubber has been analyzed and determined to have not resulted in an unanalyzed condition that significantly degraded plant safety. Therefore, this event is being retracted. The failure to retract this notification in a timely fashion has been entered into the corrective action system. The licensee will notify the NRC Resident Inspector. Notified R4DO (Werner). |
| ENS 47067 | 18 July 2011 13:23:00 | At approximately 1215 CDT FCS began discharging raw sewage to the Missouri River via the Raw Water Header. Portable toilets staged inside the plant for plant flooding are being pumped for disposal. Per discussions with the State of Nebraska, this is an additional bypass of the sewage system and requires state notification. 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. NRC Resident informed. |
| ENS 46988 | 26 June 2011 07:58:00 | At approximately 0125 (hrs. CDT), the AquaDam providing enhanced flood protection for FCS (Ft. Calhoun Station) Unit 1 failed. At 0221, as a precautionary measure, DG-2 (Diesel Generator) was automatically started per plant procedure to divorce bus 1A4 from offsite power. At 0250, DG-1 was automatically started to divorce bus 1A3. Both Emergency Diesel Generators loaded on buses as designed. 345KV and 161KV offsite power remain available. This is reportable pursuant to 10CFR50.72(b)(3)(iv)(A), 'System Actuation of the Emergency Diesel Generators'. All safety related flood protection barriers remain in place. Current river level is 1006.5' MSL and stable. FCS remains in a NOUE (Notification of Unusual Event) due to high river level. NRC personnel stationed onsite for flood monitoring have been informed. |
| ENS 46893 | 26 May 2011 06:24:00 | Operations identified a potential flooding issue in the Intake Structure 1007' 6" level. The areas of concern are the holes in the floor at the 1007' 6" level where the screen wash header penetrates the ceiling of the Raw Water Vault. There are five of these penetrations of concern. Flooding through the penetrations could have impacted the ability of the station's Raw Water (RW) pumps to perform their design accident mitigation functions. This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v). A one foot sandbag berm has been placed around each penetration of concern. The licensee notified the NRC Resident Inspector.
Following additional review of the reported condition, it has been determined that the Raw Water pumps are adequately protected during flooding conditions and that the open penetrations would not impact the ability of the Raw Water pumps to perform their design accident mitigation functions. The licensee notified the NRC Resident Inspector. Notified R4DO (Haire). |
| ENS 46716 | 31 March 2011 23:32:00 | During investigations of flood barrier penetrations, two approximately 4 inch conduits have been identified that are not sealed. These conduits penetrate the south wall of the auxiliary building near the transformers into room 19. Flooding through the penetrations could have impacted the ability of the station's auxiliary feedwater (AFW) pumps to perform their design accident mitigation functions. This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v). The penetrations are at an approximate elevation of 1007 (feet). The river level has been less than 997 feet Mean Sea Level (MSL) since prior to December 1, 2010. The AFW pumps are operable. There are not any indications of conditions that might result in a flood. Actions are in progress to plug the penetrations. The NRC Resident Inspector has been notified. |