|Entered date||Site||Region||Reactor type||Event description|
|ENS 47992||25 February 2020 07:37:00||Fort Calhoun||NRC Region 4|
During a review of plant installed instrumentation racks inside containment, two instrument racks were identified that were over the analyzed weight for the seismic analysis. The instruments on these racks are used for reactor coolant pressure transmitters that are part of the Reactor Coolant System (RCS) pressure boundary, as they are connected via instrument lines to the RCS with no remote closure capabilities. A failure of these racks during a seismic event due to the excessive weight could result in an unisolable leak from the RCS within containment based on engineering judgment. This results in the RCS principal safety boundary being in a degraded nonconforming condition as the Updated Safety Analysis Report (USAR) specified Class 1 requirement is not being met for the current seismic design. Further engineering analysis is in progress to address the weight issue for these racks and mounting requirements. The plant is shutdown and in Mode 5 with the reactor vessel head removed, so RCS is not intact and not required to be for current plant conditions. This report is being made in accordance with 10CFR72(b)(3)(ii)(A) for a degraded condition. The licensee has notified the NRC Resident Inspector.
On June 4, 2012, at 2059, Fort Calhoun Station made an 8-hour non-emergency notification for a degraded condition. Subsequent internal review has determined that the initial reporting criterion for degraded condition, 10 CFR 50.72(b)(3)(ii)(A), was incorrect. The instrument racks were identified as being over the analyzed weight for the seismic analysis. This is an unanalyzed condition, not a degraded condition. The report made on June 4, 2012 should have been made under 10 CFR 50.72(b)(3)(ii)(B), unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R4DO (Werner).
|ENS 47953||25 February 2020 07:36:00||Fort Calhoun||NRC Region 4||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 47696||25 February 2020 07:36:00||Fort Calhoun||NRC Region 4|
Communications has been lost to 21 sirens out of 101. The loss of communications does not allow the activation of the sirens. Almost all of Harrison County and Pottawattamie County in Iowa are without communications to the sirens. There are compensatory measures in place to ensure notification by local law enforcement in case of an actual emergency to inform the public in these areas. This is being reported per 10CFR50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, off site response capability, or communications capability'. An attempt is being made to reboot the siren's communication system in order to restore the sirens. During the reboot, sirens in Washington County, Nebraska, will also lose communications and therefore will not be functional. Local Law Enforcement has been notified in Washington County to perform compensatory measures in case of an emergency. The NRC Resident Inspector has been informed.
As of 0518 CST, communications has been reestablished and all sirens were returned to service. The licensee notified the NRC Resident inspector. Notified R4DO (Deese).
Following an investigation of the siren failure, it was determined that all sirens were lost for a period of time from approximately 1809 CST February 23, 2012 until 0518 CST February 24, 2012. The control room was notified at 0215 CST February 24, 2012 and compensatory measures were established for all affected counties in Iowa and Nebraska. No new compensatory measures or actions are or were required. The licensee has notified the NRC Resident Inspector. Notified R4DO (Haire).
|ENS 47088||25 February 2020 07:02:00||Fort Calhoun||NRC Region 4|
Both Fire Suppression Pumps are not operable because the required monthly surveillance tests will not be completed for June and July. The surveillance tests will be completed when flood waters recede to below 1004 feet MSL. The current river level is 1006.3 feet. Both fire pumps, FP-1A and FP-1B, are available and lined up for use. Other options are also available to provide a means of backup fire water supply that include: - Water Plant Pumps DW-8A and DW-8B aligned to the Fire Protection (FP) system. - Temporary connection to the fire protection water distribution system by the Fort Calhoun Fire Truck that is staged on site or any other fire pumper truck via fire hydrant FP-3G. - Admin Building/Training Center fire hydrant via fire hoses or water truck. This supply is from Blair water system and FP storage tank west of Highway 75. - Drafting from the Missouri River via temporary pumps. The licensee notified the NRC Resident Inspector.
Further review of the plant design and licensing basis determined that the plant is adequately analyzed for the reported situation and that it does not constitute an unanalyzed condition significantly degrading plant safety as originally reported. Therefore, this event is being retracted. The licensee has notified the NRC Resident Inspector. Notified R4DO (Proulx).
|ENS 44655||25 February 2020 04:51:00||Fort Calhoun||NRC Region 4||CE||On November 13, 2008, at 0020 CST, the State of Nebraska, Department of Environmental Quality and National Response Center were notified by Omaha Public Power District's Fort Calhoun Nuclear Station of an oil spill due to a crack in upper pump bearing sight glass line for 'C' Circulating Water Pump, CW-1C. The spill involved a small quantity of oil estimated to be less than one gallon to the intake Structure sump with trace amounts discharged to the Missouri River. Per Fort Calhoun Nuclear Stations' National Pollutant Discharge Elimination System (NPDES) permit, the notifications were made and samples were taken for offsite analysis to determine if the quantity discharged involved an actual violation of the discharge permit. This notification is being made in accordance with 10 CFR 72.75 (b) (2) (xi), 4 hour non-emergency notification due to a notification being made to a Government Agency (State of Nebraska, Department of Environmental Quality and National Response Center). The licensee notified NRC Resident Inspector. The licensee also notified the Nebraska Department of Environmental Quality and the National Response Center.|
|ENS 53125||18 December 2017 10:44:00||Fort Calhoun||NRC Region 4||On December 18, 2017 at 0710 CST, five (5) sirens in the Ft. Calhoun Station Emergency Planning Zone for Pottawattamie County, Iowa, were inadvertently activated for less than one (1) minute. There was no emergency. The plant is permanently offline and undergoing decommissioning. This siren actuation was a result of a Pottawattamie County testing error and was cancelled. The licensee notified the State of Iowa, Pottawattamie County, the National Weather Service, local radio station and informed NRC Region 4 (Browder).|
|ENS 52188||17 August 2016 21:30:00||Fort Calhoun||NRC Region 4||At 1826 (CDT), Operators identified that off-site 161 kV power source predicted post-trip voltage was below the operability limit of 161.3 kV. (Operators) entered AOP-31, 161 kV Grid Malfunctions, Section 1, 161 kV Grid Instability, and declared House Service Transformers T1A-3 and T1A-4 inoperable. (Operators) entered Technical Specification 2.7(2)c. Per Technical Specification 2.7(2)c, 'Both house service transformers T1A-3 and T1A-4 (4.16kV) may be inoperable for up to 72 hours. The loss of the 161kV incoming line renders both transformers inoperable. The NRC Operations Center shall be notified by telephone within 4 hours after inoperability of both transformers.' Per OPPD (Omaha Public Power District) Transmission, grid conditions are currently stable. OPPD Transmission has successfully raised predicted post-trip 161 kV voltage with all predicted voltages meeting or exceeding operability requirement of 161.3 kV as of 1834. Current post-trip predicted voltage is 162.2 kV as of 2029. Lowest observed actual voltage was 163.7 kV. The 161 kV line and transformers T1A-3 and T1A-4 remained available at all times. The licensee notified the NRC Resident Inspector.|
|ENS 52033||22 June 2016 13:02:00||Fort Calhoun||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 51807||18 March 2016 18:43:00||Fort Calhoun||NRC Region 4|
During a scheduled surveillance test on 3/18/2016 at 1128 (CDT), Fort Calhoun ultrasonic testing technicians discovered a void on the common shutdown cooling heat exchanger discharge piping. This piping is normally isolated during power operation, and the void does not adversely affect the Containment Spray function, Low Pressure Safety Injection function, or High Pressure Safety Injection function.
This isolated piping with the void is placed in service only during shutdown cooling operation. The fluid height measured was 10.8 inches, compared to the required height of 11.7 inches for the surveillance test. The void could potentially complicate the initiation of shutdown cooling in the required mode of operation. This piping was last tested satisfactory on 12/31/2015. The source of the void is still under investigation. Fort Calhoun maintenance was successful in venting the void on 3/18/2016 at 1704 CDT. The NRC Resident Inspector has been notified.
Following the 8-hour 10 CFR 50.72 notification made on 3/18/16 (EN 51807), further engineering analysis has determined that the ensuing water hammer transient would not have prevented the shutdown cooling system from performing its required safety functions. Specifically, it was found that the resulting system pressure transient would not cause any relief valves to lift and that piping and supports would not be significantly challenged. Therefore, the common shutdown cooling heat exchanger discharge piping remained operable by the detailed analysis. As such, the safety function was not lost and the event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(B). Notified the R4DO (G Miller).
|ENS 51487||21 October 2015 18:33:00||Fort Calhoun||NRC Region 4||The isolation function when transferring control from the Main Control Room to the Alternate Shutdown Panel (AI-185) for Pressurizer Heater Bank No. 4 (including Groups No. 10, No. 11, and No. 12) has been identified as a potential circuit failure. Identification of the potential circuit failure vulnerability is for Pressurizer Heater Bank No. 4 when isolated from Alternate Shutdown Panel (AI-185) and operated locally from Motor Control Center (MCC-4C1) for Alternate Shutdown Fire Areas 41 (Cable Spreading Room) and 42 (Main Control Room). The vulnerability involves an external hot short affecting the conductor connecting to the control room switch which may keep the 94/10 relay energized and defeat MCC control of the heaters. In a postulated event, a fire in the control room could prevent the heaters from energizing when demanded, or cause the heaters to unintentionally energize. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Per National Fire Protection Association (NFPA) 805, fire watch is an adequate compensatory measure. Therefore, this vulnerability has been added to the existing NFPA 805 Fire Protection compensatory measure for Fire Area 41. For Fire Area 42, the Main Control Room is continuously staffed, which has been credited as the compensatory measure. The NRC Resident Inspector has been notified.|
|ENS 51439||2 October 2015 10:27:00||Fort Calhoun||NRC Region 4|
At 0905 CDT on 10/02/15, Ft. Calhoun Station declared a Notification of Unusual Event based on criteria in the site security plan. The licensee notified State and local agencies and the NRC Resident Inspector. Notified DHS SWO, DHS NICC, FEMA, and Nuclear SSA via email.
At 1136 CDT, the Unusual Event was terminated based on the fact that criteria for entry into the site security plan no longer exists. The licensee notified State and local agencies and the NRC Resident Inspector. Notified R4DO (Taylor), R4RA (Dapas), NRR ET (Dean), NSIR ET (Holian), NSIR (Lewis, Stapleton), ILTAB (Johnson), IRD MOC (Gott). Notified DHS SWO, DHS NICC, FEMA, and Nuclear SSA via email.
|ENS 51270||27 July 2015 18:01:00||Fort Calhoun||NRC Region 4||On July 8, 2015, Fort Calhoun Station was in Mode 1, 100% when personnel identified an increase in the Reactor Coolant System unidentified leakage rate. As a result, personnel performed a containment entry and identified the source coming from the Reactor Coolant Pump 3A seal area. Based on this observation, a monitoring plan was established and on July 20, 2015 the leak rate exceeded the pre-establish leak limit and operators manually shutdown the reactor. On July 22, 2015, at approximately 1330 CDT, personnel identified the source of the seal leak as a crack on the middle seal inlet line, which is part of the reactor coolant system boundary. Maintenance personnel have since repaired the seal line and it has passed post-maintenance testing. The 8-hour verbal report is being made post event due to additional review of the leakage condition identifying the leakage constituted a degraded condition due to be material defects in the primary coolant system. The leak rate was between 1 and 2 gpm. The licensee notified the NRC Resident Inspector.|
|ENS 51131||5 June 2015 19:22:00||Fort Calhoun||NRC Region 4|
The following information was provided via email and telephone. Fort Calhoun Station is currently completing a scheduled refueling outage. On June 5, 2015 at 1330 during performance of surveillance testing on the auxiliary feed water system, (Hydraulic Control Valve) HCV-1107A, Steam Generator RC-2A Auxiliary Feedwater Inlet Valve, did not open when given an open signal. HCV-1107A has been declared inoperable. HCV-1107A is required to open to meet the decay heat removal safety function for Steam Generator A. Fort Calhoun Station is in Mode 3 (Reactor Coolant System temperature is greater than 515 degrees Fahrenheit and not critical). With HCV-1107A inoperable and unable to feed the A steam generator both auxiliary feedwater trains are considered inoperable. HCV-1107A is inside the Containment Building. Fort Calhoun Station Technical Specifications 2.5(1)D. requires: With both AFW trains inoperable, then initiate actions to restore one AFW train to OPERABLE status immediately. Technical Specification (TS) 2.0.1 and all TS actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Fort Calhoun Station is evaluating the best approach to repairing HCV-1107A. The Resident Inspector has been notified
Fort Calhoun Station has determined plant cooldown required to perform repairs. Plant cooldown in progress. The licensee will notify the NRC Resident Inspector. Notified R4DO (Whitten)
|ENS 50945||1 April 2015 18:53:00||Fort Calhoun||NRC Region 4||At 1210 CDT the transformer supplying power to the Emergency Operating Facility (EOF) stopped working due to the failure of a capacitor bank. The EOF is located adjacent to OPPD's (Omaha Public Power District) North Omaha facility, approximately 17 miles south of Fort Calhoun Station. The event caused a small grass fire which was quickly extinguished. The local fire department was called. The backup emergency diesel generator for the EOF started and supplied power to the facility, as designed. With the EOF diesel operating, the facility is able to function as required during emergency conditions. At 1440 CDT the EOF emergency diesel generator stopped running. At 1545 CDT the Conference Operations (COP) network phone system failed. The COP network is the primary emergency notification system between OPPD, state and county agencies. It is used to provide initial and updated notifications and for general information flow between these agencies. Alternate means of communication have been established (commercial lines) and a dedicated communicator is stationed in the control room to ensure that we can facilitate communication should the need arise. Power to the EOF was restored at 1713 CDT. At time 1720 CDT the COP tested as normal. The licensee notified the NRC Resident Inspector.|
|ENS 50800||10 February 2015 01:25:00||Fort Calhoun||NRC Region 4|
On October 14, 2013 a calculation for the containment internal structural analysis was revised and accepted by the station. This calculation limited the Safety injection tank level to 74%. On October 16, 2013 Safety injection tank level was raised to 100% for approximately 13 hours in preparations for plant start-up. While the plant was safely in a cold shutdown condition, this represents a reportable unanalyzed condition. This issue is of a historical nature and does not question the current operability of any plant systems or structures. This was self identified during a Fort Calhoun calculation review. The licensee notified the NRC Resident Inspector.
Following review of the reported event, attendant calculations and associated documentation, engineering personnel determined that the condition described in event notification EN50800 did not place the plant in an unanalyzed condition. Revision 1 of a calculation for the containment internal structural analysis demonstrated that when the safety injection tanks 'B' and 'D' are 100% filled in an outage condition, approximately a 10% safety margin is maintained. This revision was the calculation of record at the time the safety injection tank levels were raised above 74%, in October, 2013. Revision 2 of the calculation was completed to remove excess conservatism and to provide a closer representation of available margin. In addition, margin was also improved by limiting tank level to 74%. However, improving margin by limiting tank level to 74% does not result in an unanalyzed condition when tank level is 100%, as adequate margin remains. Therefore this event is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Okeefe).
|ENS 50688||17 December 2014 12:56:00||Fort Calhoun||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 50214||20 June 2014 00:29:00||Fort Calhoun||NRC Region 4|
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 50038||15 April 2014 19:16:00||Fort Calhoun||NRC Region 4|
At 1505 CDT the 'A' control room air conditioner (VA-46A) trouble alarm annunciated in the control room. The unit was confirmed to be not functioning properly and was declared inoperable at time 1515 (CDT). The 'B' control room air conditioner (VA-46B) was previously declared inoperable due to maintenance. With both control room air conditioners inoperable the plant entered technical specification 2.0.1, a 6 hour shutdown action statement. Repairs to the 'B' had been previously planned and are in progress to allow the unit to be returned to service as soon as possible. Troubleshooting and subsequent repairs to the 'A' unit are in progress. At 1812 (CDT), the station commenced a shutdown to comply with the required action statement. The NRC Resident Inspector has been informed.
At time 2050 (CDT), VA-46A was declared operable based on installation of an emergency temporary modification. TS 2.0.1 has been exited. Shutdown has been secured and FCS is stable at a nominal 33% power. The licensee notified the NRC Resident Inspector. Notified the R4DO (Gaddy).
|ENS 49926||17 March 2014 15:55:00||Fort Calhoun||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 49710||10 January 2014 06:39:00||Fort Calhoun||NRC Region 4|
During an engineering walk down inside containment a minor leak from a pipe fitting was discovered on the primary sampling piping connected to penetration M-45. There was previously identified minor leakage from a fitting on this same penetration outside of containment in the auxiliary building. The containment isolation valves are currently tagged closed. The licensee notified the NRC Resident Inspector.
Additional testing was performed on penetration M-45 which determined that the penetration met the required leakage limits. Therefore this issue does not represent a degraded condition for the containment and is not reportable. The previous notification is being retracted. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Kellar).
|ENS 49703||9 January 2014 06:42:00||Fort Calhoun||NRC Region 4||At 2230 CST on 1/8/14 during operator rounds it was self identified there was a block of ice formed on the shaft and top of one of the intake structure sluice gates. This has bent the sluice gate operating shaft. At 0315 CST on 1/9/14 it was verified this gate could not be closed. There are six intake sluice gates that are required to be able to close to act as flood barriers. The other 5 sluice gates are not affected by this condition. The licensee informed the NRC Resident Inspector.|
|ENS 49704||9 January 2014 06:42:00||Fort Calhoun||NRC Region 4|
At 0315 CST T.S. 2.0.1 was entered for all four Raw Water pumps being declared inoperable. The pumps were declared inoperable due to inability to close one of the sluice gates. There are six sluice gates and one is not functional. At 0518 the technical specification required shutdown commenced. The licensee notified the NRC Resident Inspector.
At 0900 CST 1/9/14 Fort Calhoun Station Unit 1 was manually tripped and entered Mode 3. Reactor Coolant System (RCS) cooldown to less than 300 deg F was commenced at time 1030 CST 1/9/14. The RCS temperature was less than 300 deg F at time 1433 CST. A press release has been issued. The licensee informed the NRC Resident Inspector. Notified R4DO (Hagar)
|ENS 49478||28 October 2013 18:02:00||Fort Calhoun||NRC Region 4||A review of industry operating experience regarding the impact of unfused direct current (DC) ammeter circuits in the control room has determined that the condition described below applies to Fort Calhoun Nuclear Station. This could result in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The control room ampere indications for the Class 1E batteries and their chargers do not include over-current protection to limit fault current. In a postulated event, a fire in the control room could cause one of the ammeter wires to hot-short to the ground plane. Simultaneously, the event could cause another DC wire from the opposite polarity on the same battery to hot-short to the ground plane. This would cause a ground-loop through unprotected ammeter wiring. Since this circuit is not protected (not fused), this event could result in excessive current flow in the ammeter wiring to the point of causing a secondary fire in the associated raceway system. This could potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified.|
|ENS 49452||18 October 2013 19:38:00||Fort Calhoun||NRC Region 4||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||Fort Calhoun||NRC Region 4||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 49324||5 September 2013 17:31:00||Fort Calhoun||NRC Region 4|
Current design basis calculations indicate the Low Pressure Safety Injection (LPSI) pumps could potentially operate in a run-out condition under certain worst case design basis conditions. The LPSI pumps could operate in a run-out condition beyond the analyzed time by 20 minutes. Current design basis calculation assumes LPSI Pump would be shutdown by (the) RAS (Recirculation Actuation Signal) in less than one hour, however due to past changes to Containment Spray Pump Start Logic, the time was lengthened to 80 minutes which is beyond the one hour analyzed. This represents a reportable unanalyzed condition. The licensee notified the NRC Resident Inspector.
Fort Calhoun completed additional analysis which verified that the LPSI pumps will not go into run-out as previously reported. Therefore Fort Calhoun is withdrawing the event notification. The licensee will notify the NRC Resident Inspector. Notified R4DO (Drake).
|ENS 49119||14 June 2013 18:28:00||Fort Calhoun||NRC Region 4||While revising calculations for the station analyses for potential high-energy line breaks outside of containment, the station determined that the conditions required to validate the exclusion from analyzing for a break in some small-bore (1- to 4-inch diameter) piping could not be validated. The piping is contained within the station's auxiliary building. In the unlikely event of a break of one of these lines during power operations, the plant may not have been able to respond as expected. The plant is currently in cold shutdown, with the fuel removed from the core. The licensee notified the NRC Resident Inspector.|
|ENS 49112||13 June 2013 16:55:00||Fort Calhoun||NRC Region 4||The station is reporting an unanalyzed condition involving the steam driven auxiliary feedwater pump. A postulated high energy line break in the room containing the pump could result in steam communicating with equipment in the safety related switchgear and battery rooms which are immediately above the room. The plant is currently in cold shutdown with the fuel removed from the core. The licensee stated that in the event of a postulated high energy line break, steam could possibly enter the switchgear and battery rooms via a stairwell and ventilation ductwork. The licensee notified the NRC Resident Inspector.|
|ENS 49056||21 May 2013 22:15:00||Fort Calhoun||NRC Region 4||A non-licensed, contract supervisory employee failed a random fitness-for-duty test. The employee's access to the plant has been revoked. The licensee informed the NRC Resident Inspector.|
|ENS 49050||17 May 2013 17:28:00||Fort Calhoun||NRC Region 4|
It has been determined that some instrument racks in the Containment and Auxiliary buildings do not meet their design basis capacity due to inadequate embedment depth of the seismic anchors. Assumptions made about embedment depth for a previous event were determined to be incorrect; therefore, the design basis capacity cannot be assured. This report is being made under 10 CFR 50.72(b)(3)(ii)(B), 'Unanalyzed condition'. The licensee has notified the NRC Resident Inspector.
Additional evaluation has determined that the instrument racks are adequately anchored. Therefore, this event is not reportable. The licensee notified the NRC Resident Inspector. Notified the R4DO (Gepford).
|ENS 49027||13 May 2013 11:22:00||Fort Calhoun||NRC Region 4|
Planned intermittent outages of all FCS sirens will occur the week of 5/13/13-5/17/13 due to scheduled upgrades to the radio system. Based on the planned maintenance, all sirens for the Alert Notification System within the Emergency Planning Zone (EPZ) will be nonfunctional for various amounts of time. Prior notifications and coordination with Local Law Enforcement will be completed with compensatory measures established prior to work each day to support notification of the public in case of an actual emergency during the scheduled maintenance. Updates will be made to the NRC on 5/13/13 when the work starts, and upon completion of the work not to exceed 5/17/13. Work is currently scheduled to be complete 5/17/13. Also, contingencies have been established to back out if required in support of the plant or Law Enforcement activities. Work is scheduled to commence today, 5/13/13, at 10:10 AM. This is being reported per 10CFR50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, off site response capability, or communications capability.' The licensee notified the NRC Resident Inspector.
The work was completed on 5/16/2013 at 1700 CDT. The licensee notified the NRC Resident Inspector. Notified the R4DO (Walker).
|ENS 48994||3 May 2013 06:20:00||Fort Calhoun||NRC Region 4|
A planned outage of all FCS (Ft. Calhoun Station) sirens will occur today at 0530 CDT to transfer in-service zone controllers. During the planned maintenance, all sirens for the Alert Notification System within the Emergency Planning Zone (EPZ) are nonfunctional. Prior notifications and coordination with Local Law Enforcement have been completed with compensatory measures established to support notification of the public in case of an actual emergency during the scheduled maintenance. The licensee has notified the NRC Resident Inspector, Washington, Harrison, and Pottawattamie counties.
The maintenance has been completed and the EPZ sirens have been returned to service. Local Law Enforcement has been notified that the scheduled maintenance is complete and the primary method of alerting the public with sirens is restored. The licensee has notified the NRC Resident Inspector. Notified R4DO (Haire).
|ENS 48806||4 March 2013 18:32:00||Fort Calhoun||NRC Region 4||It has been determined that the mechanical seals used in two Low Pressure Safety Injection Pumps and three Containment Spray Pumps are made of a material that may not maintain the designed integrity of the systems under certain accident conditions. These seals have been installed since original plant construction. This issue was discovered by plant personnel while researching requirements for the replacement parts during scheduled outage activities. The licensee notified the NRC Resident Inspector.|
|ENS 48792||28 February 2013 15:34:00||Fort Calhoun||NRC Region 4|
Per 10 CFR 26.719 section (c)(2) the Omaha Public Power District (OPPD) is making this notification of an apparent false positive error that has occurred on a blind performance test sample submitted to the Health and Human Services (HHS)-certified laboratory used for drug testing. On February 27, 2013, OPPD was notified that two positive drug samples that were part of a blind performance test package provided by Professional Toxicology and submitted to Clinical Reference Laboratory tested negative. Currently Professional Toxicology indicates that the two positive samples were provided with the NRC required positive levels for the drug. OPPD will investigate the issue and report to the NRC as required by part 26.719(c)(1). Professional Toxicology and Clinical Reference Laboratory are contracted by OPPD as required by NRC regulations to provide fitness for duty services. The licensee notified the NRC Resident Inspector.
Additional investigation has determined that the samples submitted by Professional Toxicology did not contain the required levels of the drug being tested. The analysis performed by Clinical Reference Laboratory was correct. There was not a reportable condition per 10 CFR 26.719. Therefore, this report is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Powers).
|ENS 48787||27 February 2013 00:55:00||Fort Calhoun||NRC Region 4||During a follow-up review of off-site testing of a sample of General Electric model HFA relays, it was discovered that some of these relays did not pass testing for full qualification in their as-found condition. Additional torquing of the relay backing plate mounting screws was required to fully meet the required qualification. Further investigation into the as-found condition of these relays installed in the plant continues at this time. The relays in question are installed in Engineered Safeguards Features, Auxiliary Feed Water, and 4160 volt systems and are used in protective and actuation functions. The licensee has notified the NRC Resident Inspector.|
|ENS 48781||25 February 2013 16:01:00||Fort Calhoun||NRC Region 4||During a review of the plant inverters, it has been determined that the inverters may not have been operable. The inverters were replaced during the 2008 refueling outage. It appears that either the modification did not recognize that the diesel frequency range is wider than the new inverters, or did not recognize its consequence. Consequently, when the diesel is supplying power to the buses and loads are being sequenced onto the bus, the bus frequency exceeds the inverter frequency range. This causes inverter voltage transients. Operation of the inverters has been modified to improve plant reliability. This issue was discovered during scheduled plant testing of the electrical system. The licensee informed the NRC Resident Inspector.|
|ENS 48668||11 January 2013 17:12:00||Fort Calhoun||NRC Region 4||During a random screening, a non-licensed contract supervisor tested positive for a controlled substance. The individual's unescorted access has been terminated. The Licensee has notified the NRC Resident Inspector.|
|ENS 48628||27 December 2012 15:46:00||Fort Calhoun||NRC Region 4|
Fort Calhoun Station (FCS) will be implementing a scheduled modification to renovate and upgrade the interior configuration of the site's Operational Support Center (OSC). The OSC is located within the Technical Support Center (TSC). The TSC will remain fully operable. Due to the construction, beginning on December 27, 2012 with a planned completion date of February 27, 2013, the OSC will be inoperable. An alternate OSC has been established on-site and is fully operable. The licensee notified the NRC Resident Inspector.
This is an update to Event Notification 48628, dated December 27, 2012. On March 8th, 2013, Fort Calhoun Station (FCS) will be entering a new phase of scheduled modifications to renovate and alter the interior configuration of its Technical Support Center (TSC). The TSC ventilation system will not be functional during this phase of modification. Combined with renovations, the primary TSC and OSC will be inoperable. An alternate OSC and TSC have been established. If the Emergency Response Organization is activated the alternate emergency facilities will be available for emergency responders per existing Emergency Plan procedures. The project is scheduled to be completed April 12, 2013. An update to this report will be provided when the TSC renovation is complete. The licensee informed the NRC Resident Inspector. Notified R4DO (Werner).
This is an update to Event Notification 48628, dated December 27, 2012. As of April 12th, 2013 Fort Calhoun Station has completed modifications to the Technical Support Center (TSC) and Operations Support Center (OSC). The TSC ventilation system has been proved functional through required testing. The primary TSC/OSC is now fully operational. The licensee notified the NRC Resident Inspector. Notified R4DO (Deese).
|ENS 48551||2 December 2012 17:28:00||Fort Calhoun||NRC Region 4||The raw water pumps (AC-10A/B/C/D) base plate support anchors were discovered by Fort Calhoun Station personnel to have inadequate embedment to support existing analysis. Plant drawing specify a j-bolt type of anchor with a required 16 inch embedment. Actual plant configuration was found to be a j-bolt type anchor with a 9 inch embedment. Plant design analysis requirements are not being met for the existing configuration. Existing analysis requires a minimum embedment of 60 inch for a j-bolt type anchor. There are a total of 4 anchors for each raw water pump, totaling 16 anchors. The as found condition renders all four raw water pumps inoperable. In the current plant Mode 5 (De-fueled), Shutdown Condition, the raw water pumps are considered available per the station's Shutdown Operations Protection Plan. Raw water pumps AC-10B and AC - 10D are in service providing cooling to the Component Cooling Water System. The core is offloaded and the Component Cooling Water System is maintaining Spent Fuel Pool temperature. The licensee notified the NRC Resident Inspector.|
|ENS 48313||17 September 2012 15:23:00||Fort Calhoun||NRC Region 4||At 0858 hrs. (CDT), Corporate Communications notified the control room that there was a communication issue with the emergency siren router. Based on the report, all sirens for the Alert Notification System within the Emergency Planning Zone (EPZ) were declared nonfunctional and notifications were completed. Local law enforcement has been notified in the required surrounding counties and compensatory measures are in place to ensure notification of the public in case of an actual emergency. Troubleshooting of the siren's communication system revealed that a peripheral router power supply had failed. The power supply was replaced at 0935 and router restarted. At 0955 sirens were restored to the counties of Harrison and Pottawattamie in Iowa. Communications were restored to all but one siren, in Washington County Nebraska by 1031, with that one siren (Siren 35) restored at 1103. All repairs completed and retested satisfactorily with proper communications confirmed with each siren. The power supply failure resulted in 2.1 hours with the sirens being unavailable. Notifications have been completed with compensatory actions by local law enforcement secured. This is being reported per 10CFR50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, off site response capability, or communications capability.' The licensee has notified the NRC Resident Inspector.|
|ENS 48148||1 August 2012 13:24:00||Fort Calhoun||NRC Region 4||At 0327 CDT, a call was received from the NRC Operations Center for Emergency Notification System (ENS) phone status check on the commercial phone line. The ENS phone system had no incoming rings, nor a dial-tone. The NRC Headquarters Operations Officer (HOO) was to submit a trouble ticket. At 0949 CDT, Emergency Planning personnel were working with Verizon Communications to determine the cause of the loss of the ENS phone. Verizon had isolated the problem to the American Broadband network and was working to resolve the problem. At 1010 CDT, Emergency Planning personnel identified that the Health Physics Network (HPN) line, the Protective Measures Counterpart (PMC) and Reactor Safety Counterpart (RSC) lines with the NRC had also been lost. The Conference Operations Network (COP) and all other commercial phone lines were functional and remained available. All areas were reachable by commercial phone lines. At 1022 CDT, the FCS Control Room was informed by Emergency Planning personnel of the communication means that were lost. The Emergency Plan per EPIP-OSC-1 was reviewed and verified no initiating conditions applied. At 1034 CDT, Emergency Planning personnel verified the communications problem with ENS, HPN, the Protective Measures Counterpart and Reactor Safety Counterpart lines was limited to the Fort Calhoun site. The Emergency Operations Facility (EOF) was not affected. At 1054 CDT, the ENS phone was restored to the FCS Control Room. This was verified by FCS placing a call to the NRC Operations Center. The NRC Operations Center then returned the call. All communications were satisfactorily restored. At 1132 CDT, the Protective Measures Counterpart line was restored. At 1142 CDT, the HPN and Reactor Safety Counterpart lines were restored. This is being reported per 10CFR50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, off site response capability, or communications capability.' The licensee notified the NRC Resident Inspector.|
|ENS 48111||17 July 2012 15:04:00||Fort Calhoun||NRC Region 4||Results of a thermal fatigue analysis on the Chemical and Volume Control System (CVCS) charging line concluded that the socket weld fittings above the RCS piping cannot be qualified. As an interim action, shut down cooling purification has been secured and charging has been isolated to the RCS. The plant is shutdown and in Mode 5 with the reactor vessel head removed, so RCS is not intact and not required to be for current plant conditions. This report is being made in accordance with 10CFR50.72(b)(3)(ii)(A) for a degraded condition. The licensee has notified the NRC Resident Inspector.|
|ENS 48094||11 July 2012 18:10:00||Fort Calhoun||NRC Region 4||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 47900||4 May 2012 16:20:00||Fort Calhoun||NRC Region 4||While evaluating the station environmental equipment qualification for equipment inside the containment it was determined that a number of different pieces of equipment were not analyzed for the environmental conditions associated with the current analysis of record. The equipment is subject to adverse conditions for a time frame longer than currently accounted for (220 verses 60 seconds). In addition, the equipment is subject to potentially different temperature effect than that to which it is currently analyzed. This affects a variety of equipment in containment. The licensee notified the NRC Resident Inspector.|
|ENS 47892||2 May 2012 19:02:00||Fort Calhoun||NRC Region 4|
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 47884||1 May 2012 17:46:00||Fort Calhoun||NRC Region 4|
During a review of environmental qualification records for reactor containment building electrical penetrations, six penetrations were identified that may not provide an adequate seal during worst case (Design Basis Accident (DBA)) conditions as required. These penetrations are through wall from the containment into the auxiliary building. The conditions that could cause degradation of the electrical penetration seals are not applicable to this operating mode. The station is currently in a refueling mode. This event was identified on March 2, 2012. The reportability was confirmed on May 1, 2012 at 1502 CDT. The current penetration configuration has existed since the plant was built. The area of concern is that the Teflon connections may degrade under conditions of high radiation and high temperature during a DBA event. The licensee is investigating the extent of the condition and repair techniques. The licensee notified the NRC Resident Inspector.
* * * UPDATE FROM ROBERT KROS TO PETE SNYDER AT 1523 ON 6/26/12 * * *
On review of CR 2012-01947 by a new Project Manager, who was brought in as a subject matter expert on HELB/EEQ, and issue was identified with the 530 primary containment electrical penetration feed-throughs used for non-CQE devices. The CR (Condition Report) correctly notes that under the original accident testing, the Teflon seals failed, and water was noted leaking from these penetrations. On further review, the following was noted: Due to the design of the penetration feed-throughs, when the inboard Teflon seal fails (as it is expected to, due to high level of radioactivity in the primary containment, following a Loss of Coolant Accident (LOCA)), the atmosphere of the primary containment will be introduced to the penetration assembly, first through the failed seal or seals, and then through the weep hole between the inboard and outboard seals of the feed-through. This will put the same high level of radioactivity in direct contact with the outboard seals, resulting in the failure of its Teflon Seal. This would result in approximately 530 breaches of the Primary Containment during post LOCA conditions. The existing vendor analysis does not assume any contribution to the outboard seal exposure from the mixing of containment atmosphere with the penetration air after the failure of the inboard seal. This is probable, as each feed-through has a weep hole. Once the inboard seal fails, the penetration will be filled with containment atmosphere to equalize the pressure, which will bring the associated noble gas and Iodine fraction in proportion, into the penetration. The licensee notified the NRC Resident Inspector. Notified R4DO (Clark)
During the extent of condition review for CR 2012-01655 and 2012-01947 additional penetration feed-through assemblies were identified that are subject to the same failure mechanism. These penetrations are associated with the containment sump recirculation isolation valves, and also associated with the personnel air lock. The licensee notified the NRC Resident Inspector. Notified R4DO (Walker).
|ENS 47870||27 April 2012 16:02:00||Fort Calhoun||NRC Region 4||A non-licensed supervisory employee was determined to be under the influence of illegal drugs during a random test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector.|
|ENS 47862||25 April 2012 17:22:00||Fort Calhoun||NRC Region 4||A non-conservative error was identified in the Proto-Flo input calculation FC06644 for LPSI (Low Pressure Safety Injection) flow post-RAS (recirculation actuation signal). The calculation used an incorrect (non-conservative) input for LPSI pump performance. Also, the associated procedure (EOP/AOP Attachment 11) as written does not provide adequate direction during the Alternate Hot Leg Injection mode of operation. EOP/AOP Attachment 11 (Alternate Hot Leg Injection) used 140 psia as the entry point. The LPSI pumps may not be able to meet minimum flow requirements at this pressure, affecting core cooling and possibly resulting in pump damage. Also the EOP/AOP attachment directs the operator to verify that flow is approximately 400 gpm as indicated on FIC-326. If 400 gpm cannot be achieved the contingency is to open any LPSI loop injection isolation valve. This step would not depressurize the RCS low enough to allow the 400 gpm flow rate to be achieved which would cause insufficient flow. Therefore, it is reasonable to conclude that the referenced procedural guidance may not be able to complete the safety function of providing adequate core cooling during the Alternate Hot Leg Injection mode of operation under a worst case scenario. Therefore, this condition is an unanalyzed condition and reportable under 10CFR50.72(b)(3)(ii)(B). The licensee has notified the NRC Resident Inspector.|
|ENS 47848||18 April 2012 18:53:00||Fort Calhoun||NRC Region 4|
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 47774||27 March 2012 14:06:00||Fort Calhoun||NRC Region 4||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.|