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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 51807 | 18 March 2016 16:28:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling Pipe Void | 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). | Shutdown Cooling Containment Spray | |
| ENS 51131 | 5 June 2015 18:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Loss of Decay Heat Removal to the 'A' Steam Generator | 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) | Steam Generator Reactor Coolant System Auxiliary Feedwater Decay Heat Removal | |
| ENS 49703 | 9 January 2014 09:15:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Intake Structure Sluice Gate Inoperable | 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. | 05000285/LER-2014-001 | |
| ENS 48551 | 2 December 2012 21:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Inadequate Raw Water Pump Anchor Embedment | 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 47473 | 22 November 2011 23:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Temporary Loss of Shutdown Cooling | During walkdown of scheduled work it was discovered that HCV-335 (Shutdown Cooling Heat Exchanger Inlet Header Isolation Valve) would not be able to be manually positioned open due to a missing idler gear key. Upon a loss of instrument air, HCV-335 would have failed closed, interrupting shutdown cooling flow with no ability to open HCV-335 manually. Alternate shutdown cooling pump and paths were available at the time of discovery. No loss of instrument air or interruption in shutdown cooling flow occurred while preparing to align alternate shutdown cooling. An 8 hour LCO under Technical Specification 2.8.1(3)2 was entered at 1700 CST. Alternate shutdown cooling was established on a containment spray pump as allowed by procedure. The 8 hour LCO was exited at 2306 CST. A replacement idler key has been fabricated for HCV-335. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM ERICK MATZKE TO PETE SNYDER AT 1654 EST ON 12/16/11 * * * Additional analysis has determined that the shutdown cooling system was capable of performing its design safety functions during the time that the idler key was missing. Therefore this event is being retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Walker). | Shutdown Cooling Containment Spray | |
| ENS 47359 | 20 October 2011 15:50:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Flood Barrier Penetrations Not Water Tight | During follow-up inspections of flood barrier penetrations into two rooms in the plant it was determined that some of the water tight conduit fittings were not filled with the material required to make them water tight. Inspection caps were removed from the fittings to perform the inspections. Three fittings into room 19 (auxiliary feedwater and plant air compressors) and fittings into room 56E (electrical switchgear) were found to contain no filling material. One additional fitting into room 56E that was thought to be capped was found to be open with a sheet metal box covering the inside access thereby obscuring inspections. All of the affected penetrations have modifications in progress to assure that they are modified and qualified for design basis flood levels. Of the 16 penetrations 6 have been verified to be made water tight by other means, specifically fire foam barrier installed in the conduit from the room 56E side. The remaining 10 penetrations will leak with a 1014 flood, although the plate will restrict flow to some degree. The stations auxiliary feedwater and safety related electrical switch gear could be affected. This eight-hour notification is being made pursuant to 10CFR50.72 (b)(3)(v). The licensee notified the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 46723 | 4 April 2011 20:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Scaffolding Affecting Safety Related Equipment | At 1500 (CDT), a concern was raised with regard to scaffolding that had been constructed around safety related equipment in the Auxiliary Building which contains both trains of safety injection and containment spray. As a result T.S. 2.0.1 was entered (which is the Fort Calhoun equivalent to standard T.S. 3.0.3). The scaffolding in question was removed and the equipment was returned to operable status and T.S. 2.0.1 was exited at 1726 (CDT). The NRC Resident Inspector has been notified.
Following the initial report, Fort Calhoun performed a seismic analysis of the impact of the scaffolding previously reported to determine if the equipment in the room would be capable of performing its required safety functions. The evaluation determined that the safety related function of the affected equipment would be able to be accomplished. Therefore, this event is being retracted. Notified R4DO (Haire). The licensee has notified the NRC Resident Inspector of this retraction. | Containment Spray | |
| ENS 46595 | 6 February 2011 04:17:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Both Trains of Auxiliary Feed Water Disabled | On February 5, 2011, both trains of the Auxiliary Feed Water system were disabled while transitioning from Auxiliary Feed Water to Main Feed Water during plant start up. While performing OI-AFW-4 attachment 3, FW-6 Electric Driven AFW Pump Operations, both Steam Generator Auxiliary Feed Water inlet valves (HCV-1107A and HCV-1108A) control switches were placed in 'CLOSE'. This action defeated the Auxiliary Feedwater Actuation Signal (AFAS) ability to open the valves, rendering both trains of Auxiliary Feed Water to the Steam Generators inoperable. The condition was subsequently recognized and the control switches were placed in 'AUTO' restoring both trains to operable. The duration of the condition was 3 minutes from start to finish. Reference Fort Calhoun Station Condition Report 2011-0839. This eight-hour notification is being made pursuant to 10 CFR 50.72(b)(3)(v)(B). The NRC Resident Inspector has been notified. | Steam Generator Auxiliary Feedwater | |
| ENS 45790 | 25 March 2010 15:28:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | High Pressure Safety Injection Inoperable Due to Voids Identified in Suction Piping | At 1028 CDT today 3/25/2010, a rejectable void was found in one of the suction lines (cooled suction line) to the B HPSI (High Pressure Safety Injection) pump. The main suction line is water filled. The piping was declared inoperable and the appropriate technical specification was entered. At 1357 CDT the void was cleared from the cooled suction line and the piping was declared operable. Subsequently, at 1409 CDT a similar rejectable void was discovered in the other cooled suction line for HPSI pumps A and C. The piping was declared inoperable. The appropriate technical specification was entered. At 1447 CDT the void was cleared and the piping was declared operable. Due to the close proximity of these occurrences the station is conservatively reporting this as a safety system functional failure as it appears that both trains of HPSI cooled suction were inoperable at the same time. The licensee notified the NRC Resident Inspector.
Subsequent to making this event notification a detailed analysis of the event was performed by FAUSKE / Westinghouse Engineering. This evaluation determined the quantity of gas voiding in the piping, the duration of time the pump(s) would be subjected to gas voiding and potential effects on the piping support design loading. The hydraulic data was then reviewed by the HPSI pump manufacturer (Sulzer) to determine the effect on pump performance. Based on these evaluations, it has been determined that the HPSI pumps were not degraded, and that they were capable of performing their design basis function at all times and were operable. Therefore, this event does not meet the 10 CFR 50.72 reporting criteria and the notification is being retracted. The licensee notified the NRC Resident Inspector. Notified the R4DO (Shannon). | ||
| ENS 45724 | 24 February 2010 17:40:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Auxiliary Feedwater Pumps May Experience Runout at Low Steam Pressures | At 1140 CST, both of the stations safety related auxiliary feedwater (AFW) pumps were declared inoperable. The AFW pumps were declared inoperable due to an evaluation that determined that the pumps may experience runout at low steam generator pressures during some design basis events. The stations Technical Specifications require that: '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 change are suspended until one AFW train is restored to OPERABLE status. The station is aggressively working to develop a solution to return at least one train of the AFW system to operability. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM DAVID SPARGO TO PETE SNYDER AT 2323 ON 2/24/10 * * * At 2133 CST, Steam Driven Auxiliary Feedwater Pump, FW-10, was declared operable based on an engineering evaluation that determined FW-10 will perform as required under all design basis conditions. Motor Driven Auxiliary Feedwater Pump, FW-6, remains inoperable. (The licensee) exited Technical Specification 2.0.1 and 2.5(1)D, and entered TS 2.5(1)B, 24 hour LCO, effective as of 1140 CST. The licensee notified the NRC Resident Inspector. Notified R4DO (Lantz).
At 0238 CST, Motor Driven Auxiliary Feedwater Pump, FW-6, was declared operable per an Engineering Operability Evaluation. (The licensee) exited Technical Specification 2.5(1)B. The licensee notified the NRC Resident Inspector. Notified R4DO (Lantz).
Further review has determined that the auxiliary feedwater pumps are and have been fully capable of performing their required safety function. Therefore, this notification is being retracted. The licensee notified the NRC Resident Inspector. Notified R4DO(Lantz) | Steam Generator Auxiliary Feedwater | |
| ENS 44228 | 21 May 2008 00:56:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Decay Heat Removal Cooling Interupted During Core Reload | At 1956, during reactor core reload with a full refueling cavity, power was lost to the #2 non-vital instrument bus. This power loss resulted in closure of the shutdown cooling temperature control valve, HCV-341. The closure of HCV-341 interrupted the cooling capability of the in service shutdown cooling loop. While cycling a condenser motor operated valve a 480 volt ground occurred which resulted in tripping the feeder breaker to motor control center MCC-4B2. MCC-4B2 was supplying power to Instrument Bus 2 via the Inverter 2 test transformer. The test transformer was powering Instrument Bus 2 due to Inverter 2 replacement per plant modification. The loss of the #2 Instrument Bus resulted in HCV-341, the shutdown cooling heat exchangers temperature control valve, failing closed. HCV-341 was manually opened to restore cooling at 2019. At 2049 power was restored to Instrument Bus 2 and the shutdown cooling system was returned to automatic operation. Flow through the core was maintained throughout the event, as the shutdown cooling heat exchanger bypass valve responded by opening to maintain flow. At the time shutdown cooling was lost, 44 of 133 assemblies had been loaded into the vessel and shutdown cooling temperature was approximately 88 degrees F. Time to boil was conservatively calculated to be 22.5 hours per plant procedures which assume decay heat from all 133 assemblies. Shutdown cooling temperature rose approximately one degree during this event. Technical Specification 2.8.1(3)(1) was entered due to no Shutdown Cooling loop in Operation. No reactor coolant boron reductions were in progress. Irradiated fuel assembly loading into the reactor core was secured and actions to restore a Shutdown Cooling loop were being initiated. (Ft. Calhoun) entered) a 4 hour LCO to close all containment penetrations providing direct access from the containment to the outside atmosphere. The licensee notified the NRC Resident Inspector.
Following a detailed review of the event of May 20, 2008, Fort Calhoun station determined that the safety function of removing residual heat from the reactor coolant system was available throughout the entire event. The system is largely manual and the manual functions were not affected by the event. At the time of the loss of power the core was being reloaded. The heat load was very small and the temperature control valve (HCV-341) was closed to allow the system to increase in temperature. When control power to HCV-341 was lost the valve did not change position. Since the ability of the shutdown cooling system to remove residual heat was not impacted by the loss of power and plant procedures have provisions to control the system locally, the safety function of removing decay heat was not lost. Therefore this notification is being retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (William Jones) | Reactor Coolant System Shutdown Cooling Decay Heat Removal | |
| ENS 43886 | 8 January 2008 17:42:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Potential Loca Injection Flow Rate Issue | At 11:42 CST, a condition report was initiated that questioned the specified flow path for simultaneous hot and cold leg injection following a large break loss of coolant accident (LOCA). When an unisolated LOCA event exists, simultaneous hot leg and cold leg injection should be implemented if the plant cannot be placed on shutdown cooling within six hours of the LOCA initiation and RCS pressure is less than 120 psia. The procedure is implemented at five and one-half hours to provide adequate time to align simultaneous hot/cold leg injection before the six hour time limit. Injecting to each side of the reactor vessel at an injection rate greater than 165 gpm, ensures that fluid from the reactor vessel (where the boric acid is being concentrated) flows out of the break regardless of the break location and is replenished with a dilute solution of borated water from the other side of the reactor vessel. The action is taken between 5.5 and 6 hours after the LOCA in order to ensure that the buildup of boric acid is terminated well before the potential for boric acid precipitation occurs which could restrict coolant flow through the core. Once the RCS is refilled, the boric acid is dispersed throughout the RCS via natural circulation. If entry into shutdown cooling system operation is anticipated before the 5.5 hour limit, then the realignment to hot/cold leg injection is unnecessary. The potential concern is associated with a charging line thermal relief valve CH-202 bypassing flow from hot leg injection and preventing the required flow rate needed to prevent boron precipitation from occurring. A minimum injection rate of 147 gpm to the cold legs and 159 gpm to the hot legs is required to prevent boric acid precipitation. Total hot leg injection flow is measured at FIA-236. Cold leg injection flow is the total of the four HPSI flow instruments, FI-313, FI-316, FI-319, and FI-322 with 50 gpm the minimum flow indication. A total cold leg injection flow of at least 200 gpm ensures at least 150 pm flow into the core, assuming 25% spillage out the break. This meets the required minimum of 147 gpm. It could not be determined through a review of the design basis documents and associated calculations what, if any, bypass flow is assumed through CH-202. Current procedural guidance in the emergency operating procedures is to align a high pressure safety injection pump to the charging header and provide hot leg injection from auxiliary pressurizer spray valves attached to the charging headers through the pressurizer and into the hot leg. The current procedural guidance does not isolate CH-202 and due to the location of flow instrument FIA-236, it cannot be guaranteed that all the flow through the charging system is being injected into the hot leg or being diverted through the normal charging line. As a result the potential existed which could have prevented the fulfillment of the safety function of a system needed to remove residual heat. Therefore this report is being made in reference to 10 CFR 50.72 (b) (3) (v) (B). Efforts are continuing to review design basis documents and calculations to determine if bypass flow was assumed past CH-202 when determining the minimum hot leg injection rate. As a compensatory measure, Operations management has restricted the use of hot and cold leg injection via the charging header until the design basis review confirms the adequacy of the current procedural guidance or the procedural guidance is revised. Pre-approved alternative methods will be utilized via the emergency operating procedures to perform simultaneous hot and cold leg injection if required. No LCO condition exists. The licensee notified the NRC Resident Inspector.
On January 8, 2008, (Event Number 43886) Fort Calhoun Station reported that there could be a potential reduction of injection flow to the hot leg during Long Term Core Cooling (LTCC) simultaneous hot and cold leg injection. The charging line thermal relief valve/check valve CH-202 could potentially divert flow from hot leg injection and reduce hot leg flow below the required flow rate needed to prevent boron precipitation from occurring. On January 8, 2008 it could not be determined through a review of design basis documents and associated calculations if bypass flow has been assumed through CH-202. Divergence of flow through CH-202 would result if a valve failure occurred. Assuming flow is diverted through CH-202, the operators would not realize that flow was going through the wrong flow path (cold leg) as their flow indication (FE-326) is located upstream of where the flow path to the hot leg and cold legs branch off. Therefore, there was nothing to alert the operator to isolate CH-202 or go to alternate hot leg injection. Previous procedural guidance was not adequate to address this condition. Current procedural guidance is adequate to address this condition as the procedures now require isolating CH-202 for LTCC. A reanalysis was performed to evaluate the required flow rate needed to prevent boron precipitation and ensure adequate LTCC. Calculations performed assumed full flow (failure) through CH-202. Under postulated design scenarios it was determined that adequate flow would have been provided to the hot legs during simultaneous hot and cold leg injection during LTCC. The calculations determined that under the evaluated scenarios, divergence of flow through CH-202 was acceptable, and that the requirements to maintain adequate flow to the core for LTCC decay heat removal and boron flushing would have been met. As a result of the analysis that were performed, it has been determined that the system was capable of performing its design function even under bypass flow conditions through CH-202. Therefore, this event is NOT reportable under 10 CFR 50.72( b) (3) (v) (B) as previously reported. The licensee notified the NRC Resident Inspector. Notified R4 DO (Miller) | Shutdown Cooling Decay Heat Removal | |
| ENS 43157 | 8 February 2007 19:42:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Raw Water Pumps Declared Inoperable Due to a Potential Failure of the Pump Breaker Linkages | On February 8, 2007 at 10:22 CST, Raw Water Pump AC-10C was declared inoperable when the pump breaker experienced a plunger linkage failure during a scheduled pump start for a scheduled rotation. Due to a similar failure of Raw Water Pump AC-10B two weeks earlier, a condition report was initiated for engineering to review the potential for a common failure mode. Engineering reviewed the number of cycles of the breakers for AC-10A and AC-10D both have greater than 1600 cycles. The breaker for AC-10B had over 1200 cycles at the time of its failure and AC-10C was over 1700 cycles. Based on the similar numbers of cycles, engineering has determined that continued operation of AC-10A and AC-10D without breaker failure could not he assured. All other safeguards associated breakers have been verified to have less than 1000 cycles and are not considered to be susceptible to breaker failure at this time. After discussion with engineering, Raw Water Pumps AC-10A and AC-10D were declared inoperable on February 8, 2007 at 13:42 CST based on the number of breaker cycles for these two components in relation to the two breakers that failed. With three Raw Water Pumps inoperable, Fort Calhoun Station entered Technical Specification 2.0.1 which requires the plant be placed in hot shutdown within six (6) hours. The potential for a common mode failure of the linkage in the 4160 VAC circuit breakers (ABB Combustion Engineering Model No. 5VKBR-250) could have prevented operation of the Raw Water Pumps to fulfill the required design function to remove residual heat during a design basis accident. At 15:22 CST, repairs were completed on Raw Water Pump AC-10C breaker and the pump was declared operable. At 15:34 CST, Technical Specification 2.0.1 was exited and Technical Specification 2.4(1)c, 7-day LCO was entered for the inoperability of Raw Water Pumps AC-10A and AC-10D. The repair of the breakers for AC-10A and AC-10D are scheduled for February 9, 2007. The licensee informed the NRC Resident Inspector.
This communication is meant to supplement the notification of February 8, 2007. The common mode failure of the linkage in the 4160V circuit breaker (ABB Combustion Engineering model number 5VKBR-250) could have prevented the operation of the raw water pumps as required by design. This could have resulted in the inability of the raw water system to remove residual heat during design basis accidents as required. The licensee informed the NRC Resident Inspector. Notified R4DO (Smith).
On February 3, 2007, (Event Number 43157) Fort Calhoun Station reported a condition that, at the time, was believed (to be) a possible common mode failure of the 4160V feeder breakers for the Raw Water (RW) System (redacted). The circuit breakers for two of the four RW pump feeder breakers had failed. Investigation had determined that the failure mechanism was the same for the two failures. On February 8, it was not known if the other two RW pump breakers could fail by this same mechanism. The component that failed in the circuit breaker was an offset rod that transmitted operation of the circuit breaker to a set of auxiliary switches associated with the circuit breakers. The offset rods of the A and D RW pumps have subsequently been tested. Non destructive testing determined that the rods had no cracks. Destructive testing determined that the rods would not have failed for several hundred operations of the circuit breakers. As a result, the system was capable of performing its design safety function. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(B) as previously reported and EN 43157 is being retracted. While no verbal report criteria is applicable to this condition it is reportable as an LER under 10 CFR 50.73(a)(2)(vii). The licensee notified the NRC Resident Inspector. R4DO ( Shannon) notified. | ||
| ENS 43013 | 27 November 2006 19:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Momentary Loss of Shutdown Cooling Due to Rhr Isolation Signal | On November 27, 2006 Fort Calhoun Station was on shutdown cooling, while in the process of moving the plant from Mode 3 to Mode 4 to fix a leak on an In-Core Instrument (ICI) Grayloc fitting. At 1327 Reactor Coolant Pump RC-3B was secured. At 1330, the last Reactor Coolant Pump (RC-3A) was secured. Once the reactor coolant pumps were secured, spray to the pressurizer was lost. This loss of pressurizer spray caused pressure in the pressurizer to rise; the systems pressure interlock caused HCV-347, Shutdown Cooling Loop 2 Outboard Isolation Valve, and HCV-348, Loop 2 to Shutdown Cooling Isolation Valve to close due to pressurizer pressure being greater than 250 psia. At 1330 the Control Room entered AOP-19 for a Loss of Shutdown Cooling. The control room operators initiated auxiliary spray, lowered pressure, reopened HCV-347/348, and started Low Pressure Safety Injection Pump SI-1A. At 1342 shutdown cooling was reestablished. At 1352 all conditions to exit AOP-19 were met and the procedure was exited. During the 12 minutes that shutdown cooling was lost the highest pressure reached in the RCS was 254 psia up from 233 psi and the highest temperature was 135 degrees, up from 134 degrees as read on the core exit thermocouples. The licensee informed the NRC Resident Inspector. | Shutdown Cooling | |
| ENS 42692 | 7 July 2006 22:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Auxiliary Feedwater Pump Recirculation Valve Inoperable | At 1700 CDT, Fort Calhoun Station determined that one of the two safety related auxiliary feedwater pumps may not provide sufficient flow to the steam generators during certain design basis events. The circuitry for the minimum flow recirculation valve for the electric motor driven auxiliary feedwater pump contains some components which were not designed as critical quality components. Since these components are not critical quality components, they are assumed to fail in a manner such that the minimum flow recirculation valve would fail open during a design basis event. If this valve should fail open, flow to the steam generators from the electric motor driven auxiliary feedwater pump may not provide adequate flow to cool the steam generators. The station has entered the appropriate technical specification action statement which requires the plant to be in Mode 2 (hot standby) within 30 hours of declaring the pump inoperable. The electric motor driven auxiliary feedwater pump was declared inoperable at 1700 CDT today. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater |