Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 40077 | 15 August 2003 16:20:00 | 10 CFR 73.71(b)(1), Safeguards Event | Discovered by Security a Vulnerability That Could Allow Unauthorized/Undetected Access to a Vital Area. Compensatory Measures Have Been Taken. | THE NRC RESIDENT INSPECTOR WAS NOTIFIED OF THIS BY THE LICENSEE. | ||
| ENS 40091 | 20 August 2003 15:45:00 | 10 CFR 73.71(b)(1), Safeguards Event | Unauthorized Access | Actual individual had been granted unescorted access to a vital area. Compensatory measures immediately taken upon discovery. NRC Senior Resident Inspector was notified of the event notification by the licensee.
Based on subsequent review, the licensee has determined that there was no security violation. This event has been retracted. For additional information, contact the Headquarters Operation Officer. R3DO (R. Gardner) & IAT (A. Davis) notified. NRC Resident Inspector was notified of this event by the licensee. | ||
| ENS 40391 | 12 December 2003 18:24:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Both Trains of Component Cooling System Inoperable. | On 12/12/2003 , at 1224 both trains of Component Cooling System were declared inoperable. The governance by the KNPP Technical Specification (3.3.d) requires that actions commence within an hour to place the plant in Hot Standby within 6 hours, Hot Shutdown within the following 6 hours, and to achieve the RCS Tavg less than 350 degrees F by use of alternate heat removal methods within an additional 36 hours. The issue stemmed from the discovery of liquid on R-17 (Component Cooling Radiation Detector) detector housing at the location of a air port. Liquid from this port was chemically tested to contain chemicals (molybdates and sulfites) currently found in the Component Cooling system. It was determined that this leak violated guidance in Generic Letter 90-05 and thus rendered both trains of component cooling inoperable due to this being in a common header. A Technical Specifications guided background has commenced at 1315 on 12/12/2003. At the time of the call, reactor power was 84% and decreasing. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if need. The electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Emergency Diesel Generator Emergency Core Cooling System | |
| ENS 40452 | 16 January 2004 06:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Both Trains of Safety Injection Declared Inoperable | The following information was obtained from the licensee via facsimile: Inspections of the 'A' SI (Safety Injection) Pump lube oil cooler today per PMP (Plant Maintenance Procedure) 33-01 revealed silt and lake weed accumulation at tube pass inlets. Calculation C11423 Rev. 0, Addendum A was recently performed to determine service water flow and temperature requirements for the safety injection pump lube oil coolers. The calculation provides the required service water flow rate based on number of tubes blocked and SW (Service Water) temperature. At 1640 (hrs)(CST), 1/15/04, it was reported that a visual inspection was performed on the 'A' SI Pump HX (Heat Exchanger) tube inlet and 17 of 20 tubes were found to be blocked. The flow for the 'A' HX was 3 - 3.8 gpm and after cleaning elevated to 5.95 - 6.05 gpm. This concern prompted an investigation into 'B' SI Pump HX and a flow test was performed at 1951 (hrs) on 1/15/04. The results from this test was no flow from 17 of the 20 tubes as seen from the outlet of the HX and a similar flow rate as seen in HX 'A'. The determination was made that this had potentially made both trains of SI Pump HX inoperable and that this needed to be reported under GNP 11.08.04 -'Reportability Determinations'. The Calculation (C11423) used data that was contradictory to current Surveillance Procedure acceptance criteria and used values that may not be indicative of post accident conditions. In a teleconference with Senior Plant Management, it was determined that future operability of the SI Pump lube oil HX cannot be verified and that both trains would be declared inoperable at time 0020 (hrs) (on) 1/16/04. This is in contradiction with the plant Technical Specification,3.3.b Emergency Core Cooling, and placed the plant in the standard shutdown sequence. The plant shutdown will commence at 0120 hrs CST. The NRC Resident Inspector has been notified by the licensee. | Service water | |
| ENS 40489 | 30 January 2004 17:30:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event | At 1151 CST, the licensee declared an Unusual Event per chart P. At approximately 1115 CST, while filling the CARDOX carbon dioxide (CO2) storage tank, the tank had reached the full mark and the operator requested the trucker to stop filling the tank. The operator verified tank filling had stopped and isolated the fill line at the tank. The operator then isolated the fill line at the truck bay. The fill line relief valve lifted, causing the vent line to bleed CO2 through a weep hole in the tank room. At 1140 CST, the shift manager was informed that CO2 levels in the tank room were at life threatening levels (40,000 ppm at floor level, 30,000 ppm at waist level). At 1155 CST, the shift manager was informed that CO2 concentration had decreased below life threatening levels. At 1215 CST, no further leakage of CO2 through the weep hole into the room was observed. At 1245 CST, CO2 levels in the CARDOX room and adjacent spaces had returned to at or near normal levels. At 1343 EST, the NRC decided not to enter Monitoring mode. The licensee notified the NRC senior resident inspector, along with state and local authorities.
At 1346 the Unusual Event was terminated by the licensee. CO2 levels in all areas of the plant have returned to normal working levels. The CO2 release was terminated at 1239 CST. Notified P. Louden, T. McGinty, T. Reis, FEMA, DHS (D. Lewis) | ||
| ENS 40950 | 13 August 2004 15:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Zone Boundary Outside Design Basis | On 08/13/2004 at 1030 CDT the Kewaunee Shift Manager was notified the control room emergency zone (CREZ) boundary was outside assumed design basis for an unknown amount of time. This is an eight hour non-emergency reportable under 10 CFR 50.72(b)(3)(v)(D), Accident Mitigation. On 08/12/2004 at 1915 (CDT), the local operator reported to the control room that he found a control room emergency zone (CREZ) boundary door in the unlatched open position. The local operator immediately ensured the door was secured in the latched closed position. Per Kewaunee's barrier control procedure the Shift Manager requested an evaluation be performed to determine the impact of potential air in leakage on the control room post accident dose consequences. On 08/13/2004 at 1030 CDT, the engineering staff reported that based on engineering judgment the unfiltered in leakage that could have entered the CREZ would likely exceed that assumed in the post accident dose analysis. Based on this judgment, the in leakage would have prevented the Control Room Post Accident Recirculation System from mitigating the dose consequences to the control room staff to values less than those determined in accident analysis. At the time of discovery the local operator immediately secured the boundary door rendering the CREZ boundary operable. This event has been entered in the Kewaunee corrective action process to investigate the cause of the door being in the open position. No other safety related equipment was out of service at the time of discovery. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41120 | 14 October 2004 18:44:00 | 10 CFR 50.72(b)(3)(xii), Transport of a Contaminated Person Offsite | Offsite Medical | At 1330 EDT on 10/14/04, the control room was notified that a worker in the Containment Building had a potential neck injury. An ambulance was requested to transport the worker. The individual was taken directly from the Containment building to the ambulance and a HP Technician with a radiation detector accompanied the individual to the hospital. A contamination survey was not completed on site. At 1344, the HP Tech and worker left the site for the hospital. The Shift Manager contacted the hospital to alert them of the potential for contamination on the injured person. At 1430, HP Technician confirmed that there was no contamination present. The licensee notified the NRC Resident Inspector. | ||
| ENS 41152 | 15 September 2004 14:11:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Containment Isolation Valves | In accordance with the requirements and guidance of 10 CFR 50.73(a)(1), a telephonic report of an invalid signal affecting containment isolation valves in more than one system is being submitted. This report is being made under 10 CFR 50.73(a)(2)(iv)(A), automatic actuation of a system listed in paragraph (a)(2)(iv)(B), in lieu of a written LER. At 0911 hours, on 09/15/2004, while operating at full power, the plant experienced a loss of power to eight process radiation monitors. The loss of power was caused by the failure of an internal radiation monitor system power supply, PS-100. The affected radiation monitors actuated, where applicable, their associated safety and non-safety related components. The safety related components affected included eight containment isolation valves and the plant's auxiliary building special ventilation system equipment. The containment isolation valves were: - 4 steam generator blowdown system isolation valves (inside and outside containment isolation valves for each of the plant's two, 'A' and 'B' steam generators). - 4 steam generator sampling system isolation valves (inside and outside containment isolation valves for each of the plant's steam generators' sample system lines). All the components affected operated as designed. The effect on plant operation was not significant. The power supply was replaced and the ventilation systems, sampling and blowdown piping systems were realigned to normal. The power supply failure and subsequent closure signal to the noted containment isolation valves is an example of an invalid signal. No full or partial containment isolation signal was generated. None of the valve closures were needed to mitigate the consequences of the power supply failure. The investigation to determine the cause of the power supply failure continues. The licensee has informed the NRC Resident Inspector. | Steam Generator | |
| ENS 41163 | 30 October 2004 19:43:00 | 10 CFR 50.72(b)(3)(xii), Transport of a Contaminated Person Offsite | Contaminated, Injured Person Transported Offsite | The following information was obtained from the licensee via facsimile: At 1405 (hrs. CDT), Shift Manager was notified that a contract individual in containment had collapsed and needed EMT (Emergency Medical Technician). At 1410, Shift Manager called for an ambulance and individual was transported out of containment. At 1443, he (the injured person) was transported offsite to the hospital. Shift manager notified hospital of individual being contaminated. HP (Health Physics technician) identified minor contamination on the hair on the back of his head, ~200 cpm (counts per minute). Shift Manager also contacted State Department of Health and Family Services (Radiation Protection Section). The patient's status is unknown at this time but appears to be heat stress related. The contamination appears to be the result of a co-worker catching the patient when he was falling down. Plant Health Physics personnel attempted to decontaminate the patient but were unsuccessful. The patient was transported to Two Rivers Hospital in Two Rivers, WI. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41200 | 16 November 2004 13:40:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | on 11/16/04, at Approximately 0740 (Cst) the Knp (Kewaunee Nuclear Plant) Shift Manager Was Informed That the Kewaunee/Point Beach External Communications Manager Received a Phone Call from the Milwaukee Journal/Sentinel Inquiring About an Individual That Was Taken to the Two Rivers Aurora Hospital for a Heat Stress Related Condition. This Was Previously Reported on 10/30/04 Under Event #41163. the Communications Manager Provided Information to the Individual from the Milwaukee Journal/Sentinel. | This notification is being made to inform the NRC of information provided to the media. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41207 | 18 November 2004 22:30:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Both Trains of Radiation Monitoring Automatic Isolation Features Disabled During Refueling Operations | The following information was obtained from the licensee via facsimile: At 1630 (hrs CST) the control room was performing head lift and pre-rod drive unlatching checkout and discovered R-12 (Containment Radiation Monitor) would not isolate reactor ventilation isolation. Investigation found that work on Engineered Safeguards Features pre-startup logic test had defeated both Train A and B automatic isolation signals to (Reactor Building Ventilation). This test was started and signals defeated at 0803 (hrs.) and 0804 (hrs.). Refueling evolutions, upper internals lift and rod latching were performed. Technical Specifications require auto-isolation capability for (Reactor Building Ventilation) during refueling ops. Both R-12 and R-21 would respond to high radiation and manual isolation of Reactor Building Ventilation was still available. Currently, refueling operations are stopped until auto-isolation of Reactor Building Ventilation is restored. The licensee has determined that no increased gaseous or particulate radiation levels existed while the auto-isolation features were defeated. The licensee has notified the NRC Resident Inspector. | Reactor Building Ventilation | |
| ENS 41226 | 26 November 2004 06:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Safety Injection Accumulator Isolation Valves Found Closed and Their Breakers Locked Off. | During plant startup following a Refueling Outage, the Reactor Coolant System was pressurized greater than 1000 psig with the Safety Injection Accumulator Isolation Valves (SI-20A and SI-20B) closed and their breakers locked off. This is contrary to the plant Technical Specification requirement to open the valves and lock out their breakers prior to the Reactor Coolant System exceeding 1000 psig. The Safety Injection Accumulators are required to inject into the Reactor Coolant System to mitigate the consequences to a LOCA. This is conservatively being reported under 10CFR50.72(b)(3)(v)(D) as "Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. At the time of discovery, Reactor Coolant System pressure was approximately 1090 psig and Reactor Coolant System temperature was approximately 440 deg. Fahrenheit. Approximately three minutes after the condition was discovered, the SI Accumulator Isolation Valves were opened and their power breakers were locked out. The STA discovered the problem while reviewing Technical Specification's and Plant Conditions. The NRC Resident Inspector was notified of this event by the licensee. | Reactor Coolant System | |
| ENS 41309 | 6 January 2005 13:00:00 | Information Only 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Modification to Process Computer That Will Impact Erds and Spds | The Plant Process Computer System (PPCS) will be taken out of service for an approximate 2 week period to implement a planned modification. The current PPCS is being replaced and the computer outage is required to allow cutover to the new PCCS. During this time period ERDS and SPDS will not be available. Also, a small portion of the plant annunciator system (Trouble Light Alarms) will not be available. Regulatory Guide 1.97 Category 2D Containment Temperature and Category 3D Component Cooling Water Temperature indicators will not be available in the Control Room, although other instrumentation is available to monitor these parameters. This is an 8-hour reportable event per 10 CFR50.72(b)(3)(xiii) Major Loss of Assessment Capability. The operation of plant systems will not be affected due to this planned action. ERDS and SPDS parameters will be monitored by control board indications. Compensatory actions have been developed. The PPCS outage is expected to commence around 0700 CST on 1/6/05. The licensee has informed the NRC Resident Inspector.
The Plant Process Computer System was restored at 0005 CST on 01/20/05. ERDS and SPDS are now available. The licensee notified the NRC Resident Inspector. Notified R3DO (A.M. Stone). | ||
| ENS 41398 | 23 January 2005 06:00:00 | Other Unspec Reqmnt | Discovery of After-The-Fact Emergency Condition (Unusual Event) | On 2/10/05 at 13:35, the Shift Manager became aware of a condition that previously existed which met the Emergency Plan criteria for declaration of an Unusual Event. On 1/23/05, a Radiation Protection Technician released nitrogen into the RAF Countroom as part of an experiment. During the experiment, the atmospheric monitors in the room detected life-threatening levels of nitrogen (oxygen deficient atmosphere). This has been determined to meet the criteria for an Unusual Event per Emergency Classification Level Chart P for a release of toxic or flammable gas on site and portable monitors indicate concentrations at life threatening levels. Management became aware of the unauthorized experiment conducted by the Technician during a management review. There were no personnel injuries as a result of this incident. The licensee will inform both state/local agencies and the NRC Resident Inspector.
On 2/24/05 at approximately 1720 hours, it was discovered that a second incident had occurred prior to the previously reported occurrence (EN#41398) that met the Emergency Plan criteria for an Unusual Event declaration. Based on the investigation of the 1/23/05 incident, the same Radiation Protection Technician had released liquid nitrogen in the Auxiliary Building elevator, as part of the same data collection activity. The release may have been at the levels of 'Immediate Danger to Life and Health' in this separate case. This previous incident occurred on 1/22/05 at approximately 1200 hours. The licensee will inform both state/local agencies and the NRC Resident Inspector. R3DO(Lanksbury), NRR EO (Carpenter), and IRD (Kennedy) notified. | ||
| ENS 41406 | 12 February 2005 04:26:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Loss of Afw Pumps Following Tornado Damage to Condenstate Storage Tank | Plant engineering staff discovered that the auxiliary feedwater (AFW) system discharge pressure trip protective equipment may not operate to protect the AFW pumps from condensate storage tanks (CST) failing from a tornado event. The AFW system protection design uses discharge pressure switches to protect the AFW pumps from a loss of suction pressure. The AFW system is required equipment to support a plant shutdown which is also assumed to occur as a result of a tornado. Current analysis predicts substantial damage to the CSTs caused by a tornado. The CSTs are aligned as the AFW pumps' initial/preferred water source. The tank damage is predicted to result in a rapid, near complete, loss of water inventory due to a tornado. The CST volume loss quickly causes air to enter the pump suction piping to all three AFW pumps. Under the postulated conditions, the current discharge pressure switch design does not act fast enough to trip the pumps before significant pump damage can occur, it appears that the original design analyses failed to consider the consequences of a loss of suction source causing air to enter the pump suction. The design basis for AFW pump discharge pressure switches, in part, is assumed to protect the AFW pumps from damage due to a tornado. Given the event postulated as described, the pressure switches may not meet this design basis assumption. Therefore, all three AFW pump discharge pressure switches were declared inoperable at 2226 on 2/11/2005. To ensure the AFW Pumps are protected in the event of a tornado, compensatory actions were incorporated in Operating Procedure E-0-05 Response to Natural Events. At the earliest indication of a tornado threat (i.e. Tornado Watch, Warning or Strike) operators are directed to align Service Water to the AFW Pump Suction to ensure adequate NPSH In the event of a loss of the CST supply. These actions do not replace the automatic function but ensure the pumps are protected against loss of NPSH. Based on the proceduralized compensatory actions put in place, the AFW Pump Low discharge Pressure Trips were declared Operable but Non-Conforming at 0056 on 2/12/2005. The licensee has notified the NRC Resident Inspector. | Service water Auxiliary Feedwater | |
| ENS 41423 | 20 February 2005 01:10:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specifications Required Shutdown Due to Inoperable Auxiliary Feedwater Pumps | During continuing evaluation of the operability of the Auxiliary Feedwater (AFW) Pump discharge pressure switches, engineering determined that a high energy line break had the potential to affect the AFW Pump Suction line from the Condensate Storage Tank (CST) due to the inability of the discharge pressure switches to protect the AFW pumps from a loss of suction from the CST. At 1910 CST on 02/19/2005 it was determined that all three AFW Pumps were inoperable as a result of the condition discovered by engineering. Due to the high energy line break, there is the potential for damage to the CST supply line to the AFW Pumps (due to pipe whip resulting from a feedwater line circumferential break). Damage to the CST supply line may result in air entrainment in the AFW Pump supply and potential AFW pump damage following an automatic AFW Pump Start. Technical Specification 3.4.b.7 allows AFW Pumps to be placed in "pull-out" at less than 15% power because analysis shows that there is at least 10 minutes available for an operator to manually initiate AFW flow if needed. At 2003 CST a power reduction to <15% power was initiated to restore the operability of an AFW Pump. When power is less than 15%, the Turbine Driven AFW Pump will be placed in "pull-out" and Service Water will be aligned to the suction of the Turbine Driven AFW Pump to restore Operability of the Turbine Driven AFW Pump. When operability is restored to one AFW pump, the plant will enter a 4 hour LCO as a result of two AFW Pumps remaining out of service. Since operability of the two motor driven pumps will not be restored within the 4 hour LCO time, plant cooldown to less than 350 deg F will continue in accordance with Technical Specification 3.4.b.6. The licensee notified the NRC Resident Inspector. | Feedwater Service water Auxiliary Feedwater | |
| ENS 41425 | 20 February 2005 17:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation on Steam Generator Low Level | The following information was reported by fax: During a plant shutdown from 100% power due to the condition as discussed in previous event notice EN #41423, a valid actuation of the reactor protection system occurred. The plant was in the intermediate shutdown mode at approximately 510�F, using the S/G's to cool the Reactor Coolant System, S/G water level was being controlled with one Auxiliary Feedwater Pump and manual control of its outlet valve. Following the initiation of RCS cooldown, Steam Generator levels began to decrease. The operator began to throttle the operating AFW pump's outlet valve open, however at 1158 on 2/20/05, the Control Room received the Annunciator for S/G B Low-Low Water Reactor Trip, which has a setpoint of 2/3 S/G level channels <17% narrow range level. This caused a Reactor Trip signal to be initiated by the Reactor Protection System. The trip signal was generated, however the plant was already in a shutdown condition with the reactor trip breakers open." All systems were functioning normally. The operator started an additional AFW pump and, at 1206, normal level was restored to S/G B and the automatic Reactor Trip signal cleared. The lowest level during this transient in Steam Generator B was 14.5% narrow range level. The licensee notified the NRC Resident Inspector. | Steam Generator | |
| ENS 41496 | 15 March 2005 22:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Kewaunee Plant Design for Flooding Events May Not Mitigate the Consequences of Piping System Failures | The following was provided by the licensee: While reviewing Nuclear Regulatory Commission's (NRC) memorandum regarding Task Interface Agreement (TIA), TIA 2001-02,'Design Basis Assumptions For Non-Seismic Piping Failures at Prairie Island Plant,' Kewaunee staff determined that the Kewaunee plant design for flooding events may not mitigate the consequences of piping system failures. As a minimum, and as a consequence of assuming failure of non-seismically qualified piping systems as prescribed in the TIA, water has been assumed to collect in the turbine building from a circulating water system piping failure that would result in substantial damage to Engineered Safeguards (ESF) and Safe Shutdown (SS) plant equipment, most notably electrical equipment. As a consequence of high water level in the turbine building, water could flow into the ESF equipment rooms that contain the Auxiliary Feedwater pumps, Emergency Diesel Generators and both the 480 volt and 4160 volt electrical switchgear. Water is assumed to flow into the equipment rooms by way of leakage past non-water-tight doors and the plant's unchecked floor-drain system. The expected water levels In the safeguards and electrical equipment rooms are assumed to increase to the point of causing multiple trains of both ESF and SS equipment to be unavailable to safely shutdown the plant. Kewaunee's primary mitigation strategy to combat flooding events is to recognize the event and initiate manual actions to open doors/ barriers. Opening the barriers to flooding directs the water out of the turbine building through the safeguards equipment rooms and returns it to the lake. Normally the manual actions would be expected to be performed before water level accumulates to a point of causing equipment damage. However, under the seismic failure assumptions, water levels are assumed to accumulate faster than the plant's ability to identify and react in order to assure protection of equipment required to initiate and complete a safe plant shutdown. Coincidental to the condition being reported, the plant had recently implemented additional precautionary measures to combat internal flooding events that lesson the significance of the condition being reported. Temporary pumping equipment, temporary sandbag barriers and additional personnel have been staged to minimize the consequences of previously questioned flooding events. Furthermore, a number of plant equipment design changes are being processed to further improve Kewaunee's defenses against internal flooding events. However, given the event being reported, the full scope of any additional actions is still to be determined. The NRC Resident Inspector was notified. | Emergency Diesel Generator Auxiliary Feedwater Circulating Water System | |
| ENS 41528 | 24 March 2005 21:08:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Relating to the Integrity of the Diesel Generator Exhaust Stacks During a Tornado | The following information was obtained from the licensee via facsimile (licensee text in quotes): While evaluating the capability of the plant to withstand a tornado it was identified that the Diesel Generator Exhaust Stacks may not maintain their design structural integrity. This would place the plant in an unanalyzed condition, however expert engineering judgment is that the damage to the Emergency Diesel Generator Exhausts would not reduce the capacity below that required to ensure decay heat removal. Currently the plant is in refueling shutdown and the Emergency Diesel Generators are not required to be operable per Technical Specifications. This condition will be corrected prior to plant startup. Power is still being supplied to the plant from the Reserve and Tertiary Auxiliary Transformers. This event was determined to be reportable per 10CFR50.73(a)(2)(ii)(B) and further review of the reporting criteria identified that if there is any doubt, this condition should be reported under 10CFR50.72(b)(3)(ii)(B). At the request of the Plant Manager at approximately 2100 on 3/25/05, the Shift Manger and Shift Technical Advisor independently reviewed the reporting criteria. Based on the statement in NUREG 1022, Rev 2, when applying engineering judgment and there is doubt regarding whether to report or not, the commission's policy is that the licensees should make the report, it was decided to report per 10CFR50.72. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator Decay Heat Removal | |
| ENS 41539 | 28 March 2005 21:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Historical Report - After-The-Fact Operability Assessment | The licensee provided the following information: This event notification reports a previously unrecognized reportable event that occurred in 2002. This past event is being reported under the criterion of 10CFR50.72(b)(3)(ii)(B), an event or condition that results in the plant being in an unanalyzed condition. On 1/23/2002, at 1815 hours, while the plant was operating at full power, the plant entered a 72 hour Technical Specifications (TS) limiting conditions for operation (LCO) for one train of the Residual Heat Removal (RHR) system. The RHR system LCO was entered when the plant discovered a potential for reaching pump runout under specific post-accident operating conditions for single pump operation of the Component Cooling Water (CCW) system. The CCW system provides cooling support to the RHR system for post-accident long term recirculation core and containment cooling operation. The RHR train was returned to service at 2002 hours on 01/25/2002, within the TS allowed 72 hours, after a flow limiting device was installed on a non-critical CCW system flaw control air operated valve. An after-the-fact past operability assessment of the condition found that, compared to corrective actions taken to remove the potential for runout, the condition potentially existed since original plant design. As a minimum, it existed for a period longer than the TS allowed LCO for the RHR system. Therefore, an unanalyzed condition did exist and as such this is a late report. The licensee notified the NRC Resident Inspector. | Residual Heat Removal | 05000305/LER-2005-007 |
| ENS 41605 | 15 April 2005 13:35:00 | 10 CFR 26.73, Applicability | Fitness for Duty | A non-licensed employee supervisor tested positive for alcohol during a for-cause fitness for duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details. | ||
| ENS 41752 | 7 June 2005 16:25:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Potential Impact on Emergency Diesel Generator Operability | At 1125 on 6/7/2005 it was determined that the Emergency Diesel Generators A and B were out of service due to the possibility of Tornado Missiles potentially collapsing the D/G Fuel Oil Tank Vents. The Emergency Diesel Generators are required as a support system for RHR Decay Heat Removal and RHR was also declared inoperable at the same time. Technical Specification requirements for RHR Decay Heat Removal are, if less than the required number of heat sinks are operable, then corrective action shall be taken immediately to restore the minimum number to operable status. Actions are being taken to restore full operability of the Emergency Diesel Generators A and B. Currently RHR is operating and providing decay heat removal and Emergency Diesel Generators are available as a support system for RHR. Event Report # 41528 had similar issues associated with the Emergency Diesel Generators exhaust ducts and their ability to withstand tornado forces. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM G. RISTE TO P. SNYDER ON 7/26/05 AT 1541 * * * Event Notice #41752 was initiated on 6/7/2005 to report an unanalyzed condition with the Emergency Diesel Generators A and B. The initial analysis for tornado missile strike probability results for the Emergency Diesel Generator fuel oil tank vent lines indicated they could be damaged by a tornado missile to the point they would potentially adversely affect Diesel operability. Additional analysis was performed and it was determined that the original fuel tank oil vent line configuration was acceptable. The licensee notified the NRC Resident Inspector. Notified R3DO (Mark Ring). | Emergency Diesel Generator Decay Heat Removal | |
| ENS 42152 | 17 November 2005 22:51:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Both Trains of Shield Building Ventilation Declared Inoperable Due to Missing Clamps | At 1651 on 11/17/2005 both trains of Shield Building Ventilation were declared out-of-service because it was identified that Shield Building penetrations 31 and 36NW flexible boot seals were not clamped per the design drawing. Since there is no documentation to show that Shield Building Ventilation would be able to perform its function during a design basis event with no clamps installed, Shield Building Vent was declared inoperable. Clamps have been installed and the penetrations were returned to the design configuration. At 1810 CST Shield Building Ventilation was declared Operable. The licensee notified the NRC Resident Inspector.
The licensee has performed an engineering review and determined that the boot seals would have been capable of performing their design basis function without the clamps installed. Consequently, this event is no longer considered reportable and is being retracted. The NRC Resident Inspector has been notified by the licensee. R3DO (O'Brien) has also been notified. | Shield Building | |
| ENS 42169 | 25 November 2005 22:24:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Dangerous Levels of Carbon Dioxide Following a Cardox Release | At 16:09 CST, the Control Room received a fire alarm on the Main Generator and actuation of the fire protection system. The Main Generator is protected by Carbon Dioxide for fire protection purposes. There was no evidence of a fire at the Main Generator. Air sampling of the Cardox Storage Tank Room indicated Carbon Dioxide levels at life threatening levels. An Unusual Event was declared at 1624 CST based on Emergency Action Level Chart P 'External Events and Chemical Spills' for a release of toxic or flammable gas on site AND portable monitors indicate toxic or explosive concentrations at life threatening levels of the gas near the spill area. There were no personnel injuries reported and the licensee is conducting a survey to account for personnel. Ventilation of the affected areas is in progress to reduce the toxic gas levels. The licensee informed state/local agencies and the NRC Resident Inspector.
(The) declaration of Unusual Event has been terminated (at 1915 CST on 11/25/05). The Control Room received a fire alarm on the Main Generator and actuation of the fire protection system. The Main Generator is protected by Carbon Dioxide for fire protection purposes. There was no evidence of a fire at the Main Generator. (The) source of Carbon Dioxide gas has been isolated. The area with (the) life threatening concentration of Carbon Dioxide has been ventilated. (The) plant no longer has any areas with (a) toxic concentration of Carbon Dioxide at the life threatening levels. Normal plant operation continues. The licensee is continuing their investigation into the cause of the Cardox discharge. The licensee informed stat/local agencies and the NRC Resident Inspector. Notified R3DO(Skokowski), NRR (Gillespie), IRD (Leach), DHS (Akers) and FEMA (Ligett). | ||
| ENS 42173 | 29 November 2005 04:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Following Main Feed Pump Trip | At 22:19 CST, Main Feedwater Pump B tripped on over current. A secondary plant runback from 100% power was automatically initiated. During the secondary plant runback, the reactor automatically tripped on Steam Generator B low-low level at 22:20 CST. All three Auxiliary Feedwater pumps automatically started due to low-low Steam Generator level. The plant has been stabilized at Hot Shutdown (RCS temperature approximately 547 degrees F, RCS pressure approximately 2235 psig). Investigation into the cause of the trip is on-going. This event is being reported under 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system (RPS) when the reactor is critical and 10CFR50.72(b)(3)(iv)(A) for valid actuation of the Auxiliary Feedwater System. All control rods fully inserted on the automatic trip. Steam generator water levels have recovered to indicate in the narrow range. The current decay heat removal path is auxiliary feedwater to the steam generators steaming through the power operated relief valves. There are no known primary to secondary leaks. All safety related buses are powered from offsite power. Emergency diesel generators are available and in standby. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 42509 | 18 April 2006 17:00:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Shield Building Ventillation System Declared Inoperable | On 4/18/2006, at 1200 hours, while the plant was operating at full power, the plant entered a 12 hour Technical Specification (TS) action statement for both trains of the Shield Building Ventilation (SBV) System being declared inoperable. The SBV System action statement was entered when the plant declared Relay Flacks RR-119 and RR-120 inoperable due to non-qualified fuses and cables being installed in six (6) of the boxes contained in these racks. The instruments associated with the six (6) boxes were not found in Technical Specification required instruments or alarms. The issue was that a downstream failure on the non-qualified Instruments may not have qualified fault protection, therefore a fault could impact safety related equipment. Other Technical Specification equipment affected by the inoperability of RR-119 and RR-120 include both trains of Inadequate Core Cooling Monitoring System (ICCMS), Reactor Vessel Level Indication (RVLIS), Pressurizer Safety Valve Outlet Temperature, and Pressurizer Power Operated Relief Valve Outlet Temperature. At 1611 actions taken by plant staff returned RR-120 to operable and the 12 hour action statement for SBV was exited. The plant remains in a 7 Day Limiting Condition of Operation (LCO) pending the return of RR-119. Due to the fact that Train B of SBV has been made operable, a plant shutdown was not commenced. This event is being reported under 10CFR50.72(b)(3)(v)(C) 'Any event that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to control the release of radioactive material'. The licensee notified the NRC Resident Inspector. | Shield Building | 05000305/LER-2006-002 |
| ENS 42528 | 26 April 2006 22:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Plant S/D Required by Ts Due to Leak in Service Water Supply Line to "B" Edg | At 1345 on April 26, 2006, a decision to shutdown the plant was made due to an Inoperable service water train. At 1700 on April 26, 2006, the plant shutdown was initiated. On April 25, 2006, a leak was found on the Service Water supply line to the 'B' emergency diesel generator. At 1725 on April 25, 2006, the 'B' train of service water was declared inoperable and the technical specification limiting condition for operation was entered for one train of service water being inoperable. The end state for exceeding the allowed outage time for the limiting condition for operation for service water is the reactor coolant system Tavg less than 350 degrees Fahrenheit. During investigation of the leak, an additional small leak was discovered approximately 180 degrees from the first leak on the same section of pipe. Based on this additional leak, a conservative decision was made to shutdown the unit. It is estimated that the repair to the service water piping would not be completed within the allowed outage time and therefore per 10CFFR50.72(b)(2)(i) this event is reportable as the initiation of any nuclear plant shutdown required by plant's technical specifications. All systems are currently functioning as required for the unit shutdown and the "B" EDG is operable and available. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Service water Emergency Diesel Generator | |
| ENS 42530 | 27 April 2006 01:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Failure of Rps Signal to Initiate Reactor Trip | At 2043 (CDT) on 4-26-06, during a plant shutdown with the reactor at approximately 35% power, the operating crew manually initiated a reactor trip. The operating crew had just stopped one of the two condensate pumps and then the remaining feedwater pump tripped unexpectedly. The operating crew recognized the turbine did not trip, as it is expected to automatically trip when no feedwater pumps are running. The automatic turbine trip would have automatically tripped the reactor. Therefore, the operating crew manually initiated a reactor trip. Because the reactor did not automatically trip (i.e., failure of RPS to initiate and complete a reactor trip), the Shift Manager declared an Alert, at 2049, based on Chart F of Table 2-1 EPIP-AD-02. Therefore, this is a one-hour notification in accordance with 10CFR50.72(a)(1)(i) 'The declaration of any of the emergency classes specified in the licensee's approved Emergency Plan.' The manual reactor trip is reportable (4-hour) in accordance with 10CFR50.72(b)(2)(iv)(B) 'Any event or condition that results in an actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' All systems functioned as expected following the manual reactor trip. Service Water Train B is inoperable because of a one-gallon per minute leak. All rods inserted fully. Decay heat is being removed with the steam dump and secondary PORVs. The condenser is losing vacuum due to the turbine trip. Auxiliary Feedwater Pumps started as required. The licensee notified the NRC Resident Inspector.
At 2049 on April 26, 2006, Kewaunee Power Station staff declared an Alert emergency classification (reference EN# 42530). The Kewaunee Power Station staff has assessed this event. There was no affect on the health and safety of the general public and no release of radiation. No plant personnel were injured and the only plant equipment problem was with the failure of a trip of both feedwater pumps to cause the main turbine to trip. The Kewaunee Power Plant staff has conducted a preliminary investigation of the control room indications and sequential events recorder, which indicates that before the manual reactor trip there was no automatic reactor trip signal present and a failure of the reactor trip breakers did not occur. The Alert was terminated at 0024 CDT (on 04/27/06). The unit is currently in the Hot Shutdown Mode with plans to cool the plant to less than 350 degrees Fahrenheit. The licensee notified the NRC Resident Inspector and will be notifying State and local government and issuing a press release. Notified NRR EO (MJ Ross-Lee), IRD Mgr (P. Wilson), R3DO (H. Peterson), DHS (Holz ), FEMA (Steindurf), NRC/EPA (Crews), DOE (Wyatt), USDA (Timmons), HHS (Peagler).
When performing a review of the event reported on April 26, 2006 (EN# 42530), the Kewaunee Power Station staff determined another reporting criterion was met. An eight-hour report is required to be made per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of 10 CFR 50.72 except when the actuation results from and is apart of a pre-planned sequence during testing or reactor operation.' 10 CFR 50.72(b)(3)(iv)(B)(6) is PWR auxiliary or emergency feedwater system. As described in EN# 42530, a manual reactor trip of the Kewaunee Power Station was initiated at 2043 on April 26, 2006. The manual reactor trip was initiated when the plant experienced a loss of both feedwater pumps. With a loss of both feedwater pumps and a manual reactor trip, the narrow range water level in both steam generators decreased to the actuation setpoint value for starting the Auxiliary Feedwater Pumps, causing all three Auxiliary Feedwater Pumps to start as designed. Because the steam generator water level was below the actuation setpoint, this was a valid actuation of the auxiliary feedwater system. As a valid actuation of the auxiliary feedwater system, this condition is reportable under 10 CFR 50.72(b)(3)(iv). The untimeliness of the report has been entered into the Kewaunee Power Station's corrective action program. The licensee notified the NRC Resident Inspector. Notified R3DO (H. Peterson). | Steam Generator Feedwater Service water Reactor Protection System Auxiliary Feedwater Main Turbine Emergency Feedwater System | |
| ENS 42533 | 28 April 2006 03:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Both Safety Injection Trains Declared Inoperable | At 22:30 on 04/27/2006 the Kewaunee Power Station declared both trains of Safety Injection (SI) inoperable. SI Pumps A & B were declared out of service per Surveillance Procedure SP 87-125 (Shift Instrument Channel Checks-Operating) due to SI Accumulator B level change of greater than or equal to 3%. SI Accumulator B level indicator (LI-935) indicated 30%. Previous indication recorded for SI Accumulator B was 33%. Per SP-87-125 'If the SI pumps are not vented prior to a level decrease of greater than or equal to 3%, the SI pumps become INOPERABLE.' TS 3.0.c (Standard Shutdown Sequence) was entered. At 0130 on 4/28/2006, venting of both SI Pumps was completed. SI Pump A was returned to service and SI Pump B remains out of service due to SW Train B inoperability. The licensee notified the NRC Resident Inspector. * * * UPDATE ON 6/21/06 AT 0922 EDT FROM JERRY RISTE TO GERRY WAIG * * * On 4/28/06, with Train B of SI already inoperable for other reasons, EN#42533 was made under 10CFR 50.72(b)(3)(ii)(B) for both trains of Safety Injection (SI) being inoperable due to excessive level drop in an SI Accumulator which had the potential to gas-bind both trains. Subsequent evaluation has determined that, under the existing plant conditions, an SI train should only be declared inoperable if it is found that gas is actually present in the SI pump. Since venting after the event revealed no gas present, neither SI train was rendered inoperable by the Accumulator level drop, and thus both trains were never inoperable. Therefore, EN#42533 is being retracted. The licensee will notify the NRC Resident Inspector. Notified R3DO (K. O'Brien) | ||
| ENS 42534 | 28 April 2006 00:29:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Two Trains of Shield Building Ventilation Inoperable | At 19:29 on 04/27/2006 the Kewaunee Power Station declared two trains of Shield Building Ventilation (SBV) inoperable. Train B SBV was declared inoperable on 4/25/2006 at 17:25 when SW Train B was declared out of service for a leak that developed on the branch header to Diesel Generator B and TS 3.3.e.2 was entered. On 4/27/2006 at 19:29 QA typing discrepancies in RR-119 were discovered. QA-2 components were used in a QA-1 system, which could potentially cause a failure of some safety related equipment powered from this relay rack. As a result, all safety related equipment powered from RR-119 were declared inoperable and the appropriate technical specifications entered. SBV Train A Damper Control is powered from RR-119 and was declared inoperable at 19:29 on 4/27/2006 resulting in two trains of SBV being inoperable and TS 3.6.c.1 entered. The QA-2 components that may have an adverse affect on QA-1 safety related components in RR-119 have been removed and both trains of SBV were declared operable at 00:29 on 4/28/2006. The licensee notified the NRC Resident Inspector.
This event was reported on April 27, 2006 (Event Number 42534) for two trains of shield building ventilation (SBV) being declared inoperable. Train B SBV was declared inoperable due to loss of the service water spray system for the SBV charcoal filters and Train A SBV was declared inoperable due to relay rack RR-119 Quality Assurance typing discrepancies (relay rack provides power for SBV Damper Control). Subsequent review of analysis determined that the SBV spray system is not required for post-LOCA operation to control the release of radioactive material. Therefore, Train B of SBV was not required to be declared inoperable when Train B service water was declared out of service. With only one Train of SBV being inoperable, this event is not reportable and is being retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (P. Louden). | Service water Shield Building | |
| ENS 42557 | 5 May 2006 21:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Potential Loss of Rhr | At 1600 CST on May 5, 2006, the Kewaunee Power Station (KPS) declared both trains of Residual Heat Removal (RHR) inoperable due to a vulnerability to internal flooding caused by possible ruptures of non-seismically qualified piping during a seismic event. Since the KPS RHR system is not protected from non-seismically induced piping breaks in the auxiliary building basement KPS does not presently meet the design criteria in the KPS USAR for the RHR system. The specific design criteria is stated in USAR section B.5 'Protection of Class I Items' and states that Class I items are protected against damage from 'Rupture of a pipe or tank resulting in serious flooding or excessive steam release to the extent that the Class I function is impaired.' With KPS in Intermediate Shutdown and both trains of RHR inoperable, KPS meets the Technical Specification requirement for Decay Heat Removal Capability with two (2) Steam Generators operable to remove decay heat. This event is being reported under 10CFR50.72(b)(3)(v)(B) for 'Any event that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to remove residual heat.' The licensee notified the NRC Resident Inspector. | Steam Generator Residual Heat Removal Decay Heat Removal | 05000305/LER-2006-003 |
| ENS 42589 | 18 May 2006 18:23:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Postulated Flooding That Could Impact Both Trains of the Residual Heat Removal (Rhr) System | At 1323 the Shift Manger was notified of a flooding concern that could impact both trains of the Residual Heat Removal (RHR) System. At this time both trains of RHR were declared inoperable and actions were taken to isolate the lines that could flood the RHR Pump Pits, The lines in question were from the Spent Fuel Pool Cleanup system specifically associated with the Spent Fuel Pool Dernineralizers and the Pre-filters and Post-filters. These lines were Isolated at 1345. Spent Fuel Pool cooling remained in service following isolation of the Spent Fuel Pool Cleanup system. Both RHR trains were returned to service at 1345. Initial piping analysis determined that these lines would have remained intact during and following a seismic event. The final evaluation s expected to be completed within two weeks. The NRC Resident Inspector was notified of this event by the licensee. | Residual Heat Removal | |
| ENS 42762 | 10 August 2006 05:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Elevated Tritium Levels in Onsite Sample Locations | At 1400 on 8/10/06 KPS (made) a notification to the State and Local Government that samples taken at settling plugs in the basement of the Auxiliary Building and Turbine Building show elevated tritium levels. Detected tritium levels were between 6,000 and 103,000 pico curies per liter. The Radiological Effluent Program has not detected any elevated tritium levels outside the plant. RCS leakage is .19 gpm, stable and within Technical Specification limits and there is currently no identified leakage from the Spent Fuel Pool. Investigation is continuing to identify the reason for the increase in tritium levels. The licensee notified the following agencies: State of Wisconsin: Department of Emergency Management and Department of Natural Resources Regional Office County governments: Kewaunee County Emergency Director and Manitowoc County Emergency Director The licensee notified the NRC Resident Inspector.
The licensee is planning a press release concerning the elevated tritium levels. This press release will be issued the morning of 8/11/2006. Notified R3DO (Kozak), NRR EO (Nieh), IRD (Wilson), R3 OPA (Mitlyng), HQ OPA (Hayden)
In the initial notification, KPS personnel stated that there was currently no identified leakage from the spent fuel pool. For clarification, there is no indication of gross leakage from the spent fuel pool. However, KPS personnel will be continuing an investigation by collecting and analyzing water droplets from the Spent Fuel Pool leakage detection telltale drains to determine if the water droplets are from the spent fuel pool. The licensee notified the NRC Resident Inspector. Notified R3DO (Kozak) | ||
| ENS 42884 | 7 October 2006 12:53:00 | 10 CFR 26.73, Applicability | Fitness for Duty - Confirmed Positive for Non-Licensed Supervisor | A non-licensed supervisor had a confirmed positive for alcohol. The supervisor's access to the plant has been denied while a review of this matter is performed. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified. | ||
| ENS 42947 | 30 October 2006 14:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip | During power operation, at 92% rated power, an automatic reactor trip occurred. The reactor protection signal that caused the reactor trip was steam generator 'B' steam flow greater than feedwater flow coincident with low water level on steam generator 'B.' The cause of the plant transient that led to the reactor trip was a loss of Instrument Bus 1 (Red Channel). Instrument Bus 1 unexpectedly deenergized during the performance of maintenance on the inverter (BRA-111) that feeds Instrument Bus 1. Following the reactor trip, the auxiliary feedwater pumps automatically started, as designed, due to a low level in the steam generators. After the trip, non-safety related 4160 Volt AC Bus 4 de-energized and secondary plant feedwater heater 15B relief valve lifted. The cause of the loss of Bus 4 and 158 feedwater heater relief lifting Is under Investigation. The plant is currently stable and in the hot shutdown (HSD) mode. Power was restored to Instrument Bus 1 at 1018. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical because of the automatic reactor trip and under 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed below because of the actions of the RPS and the automatic start of the AFW pumps. All control rods fully inserted. Decay heat is being removed by feeding the steam generators with AFW and steaming to the Condenser Dump. The licensee notified the NRC Resident Inspector.
Due to the loss of Bus 4, the running Circulating Water Pump was lost resulting in a loss of normal heat sink to the condenser. The standby Circulating Water Pump was started at 1002 CST and the condenser heat sink was restored. Until the condenser steam dump was restored, the plant was steaming through the steam generator PORVs (atmospheric steam dumps). There is no steam generator tube leakage. The licensee notified the NRC Resident Inspector. Notified R3DO (M. Phillips) | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 42983 | 10 November 2006 20:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip at Low Power | On 11/10/2006 with a shutdown in progress to repair a degraded bearing on the turbine generator, an automatic reactor trip occurred due to a power range nuclear instrumentation (NI) low range - high flux trip. Reactor power had just been lowered to below 10% power (P-10) where the power range (NI) low range trips become active. The bistable for power range NI N-42 had been tripped due to an unrelated failure on 11/09/2006. When P-10 automatically unblocked, a power range NI low range high flux reactor trip was generated. At the time of the trip, reactor power was well below the trip setpoint of 24.5% power. Following the trip, Main Feedwater Regulating Valve, FW-7A, did not automatically close as required on the reactor trip coincident with Low Tave (554F). The Reactor Operator reported FW-7A was mid-position and attempted to manually close FW-7A. It did not respond. As a result, levels in steam generator A rose to greater than 67%, which initiated feedwater isolation. The feedwater isolation signal tripped the running feedwater pump. With no feedwater pumps running, both Auxiliary Feedwater Pump A and Auxiliary Feedwater Pump B automatically started as required. The High-High steam generator level also resulted in a second reactor trip initiation signal. The Reactor Operator manually controlled Auxiliary Feedwater flow to steam generator A to restore normal level. Following the feedwater isolation, FW-7A fully closed. Following the trip, MS-201B1, the steam supply to main steam reheater B1 was locally isolated to limit the RCS cool down. This was a previously discussed contingency action. Main steam isolation valves remained open and normal condenser heat sink remained available. Further investigation as to the cause of the trip is in progress. Recovery actions per normal operating procedures are in progress. The plant was being shut down at a rate of a half percent power per minute at the time of the trip. All control rods fully inserted on the reactor trip and no safety or relief valves lifted. The plant was aligned for the normal shutdown electrical lineup prior to the trip. The temperature on the generator bearing reached a maximum of 190F with trip guidance set at 225F. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Main Steam Isolation Valve Auxiliary Feedwater Control Rod Main Steam | |
| ENS 43096 | 12 January 2007 16:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip/Reactor Trip | At 1039 Kewaunee Power Station experienced a loss of auto stop oil pressure on the main turbine while performing the monthly Turbine Trip Mechanism Test resulting in a reactor trip. No safeguards equipment was out of service at time of trip. Following the trip one of the moisture separators associated with the main turbine had its associated steam inlet valve fail open, which resulted in RCS temperature decreasing to 536 degrees F. This valve was manually isolated and RCS temperature returned to normal 547 degrees F. Normal heat sink via the condenser was available during the event. Investigation is continuing into the cause of the turbine trip. This event is being reported under 10CFR50.72(b)(2)(iv)(B) for actuation of the Reactor Protection System and 10CFR50.72(b)(3)(iv)(A) for actuation of the Auxiliary Feedwater System." All rods fully inserted and no primary or secondary PORV's lifted. There was no injection from any safety systems, other than AFW. Decay heat is being dumped to the condenser via steam dumps. The reactor coolant pressure and temperature are stable. The licensee notified the NRC Resident Inspector. | Reactor Protection System Auxiliary Feedwater Main Turbine | |
| ENS 43196 | 28 February 2007 05:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip During the Performance of a Surveillance | On 2/27/2007 at 2333 CST a Reactor Trip occurred during performance of a surveillance procedure calibrating a Nuclear Power Range instrument. The Reactor trip resulted in an automatic Turbine Trip and actuation of the Auxiliary Feedwater System. No safeguards equipment was out of service at the time of the trip. Following the trip, a steam inlet valve on a Moisture Separator associated with the main turbine failed to close which resulted in RCS temperature decreasing to 537 degF. This valve was manually isolated and RCS temperature returned to normal 547 degF. Normal heat sink to the Main Condenser was available during the event. Investigation is continuing into the exact cause of the Reactor trip. This event is being reported under 10CFR50.72(b)(2)(iv)(B) for actuation of the Reactor Protection System and 10CFR50.72(b)(3)(iv)(A) for actuation of the Auxiliary Feedwater System. All control rods fully inserted on the reactor trip. Decay heat is being removed by Auxiliary Feedwater feeding the steam generators, steaming to the main condenser. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Turbine Main Condenser Control Rod | |
| ENS 43324 | 29 April 2007 20:26:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Shutdown Due to Both Emergency Diesels Declared Inoperable | At 1453, (the licensee) declared D/G (Emergency Diesel Generator) - A inoperable due to outside air temperature exceeding 78.2 degrees Fahrenheit and entered Technical Specification 3.7.b.2 (which has a) 7-day LCO. At 1526, outside air temperature increased to greater than 81.4 degrees Fahrenheit and D/G - B was declared inoperable. With both D/Gs declared inoperable, Technical Specification 3.0.c , Standard Shutdown Sequence was entered. No other safety related components are inoperable at this time. Plant shutdown was initiated at 1624 on 4/29/07. Based on recent findings by the licensee, the EDGs may not be capable of generating the minimum electrical power output needed to support a safety bus when outside air temperature exceeds certain limits (Operability Determination 151 dated April 26, 2007) The licensee is in the process of making ventilation modifications that will correct this condition. The licensee expects that the temperatures will drop below the EDG inoperability limits before the plant actually shuts down. The licensee notified the NRC Resident Inspector.
EN 43324 provided notification of initiation of a unit shutdown due to both emergency diesel generators (EDGs) being declared inoperable due to outside air temperature exceeding allowable limits. Unit shutdown had been initiated per Technical Specification (TS) 3.0.c, Standard Shutdown Sequence, due to TS 3.7 not being met. A subsequent engineering evaluation allowed for increasing the allowable limit for outside ambient air temperature at which the EDGs are operable. The highest outside ambient air temperature recorded on 04/29/2007 (83.5 F) did not exceed the new allowable limit. Therefore, based on the new evaluation, both EDGs remained operable. TS 3.7 remained met and no TS shutdown was required. Consequently, this condition did not meet any reportability criteria in 10 CFR 50.72. As a result, the notification made on 4/29/07 (EN 43324) is hereby retracted." NRC R3DO (Patrick Louden) notified. The NRC Resident Inspector was notified of this retraction by the licensee. | Emergency Diesel Generator | |
| ENS 43633 | 11 September 2007 19:39:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Auxiliary Building Special Ventilation and Shield Building Ventilation Systems Inoperable | At 1439 the Control Room received a Condition Report that identified an unanalyzed condition associated with the Auxiliary Building Special Ventilation and Shield Building Ventilation Systems which resulted in declaring this equipment inoperable. The issue was recirculation flow from Auxiliary Building Special Ventilation back to the Auxiliary Building added additional area heat gain resulting in exceeding the design capacity for the area fan coil units. Technical Specification LCO was entered per TS 3.6.c requiring a reactor shutdown within 12 hours. At 1654 Engineering provided information that if the control switch for Auxiliary Building Special Ventilation Train A was maintained in the off position, the area heat load was within the capacity of area fan coil units. This was completed and both trains of Shield Building Ventilation and Auxiliary Building Special Ventilation Train B was returned to operable. Technical Specification requirements to shutdown were exited. The plant remains in a 7 day Action Statement per TS 3.6.c with Auxiliary Building Special Ventilation Train A inoperable. This event is reportable under 10CFR50.72(b)(3)(v)(C), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.' The licensee notified the NRC Resident Inspector.
On September 11, 2007 Kewaunee Power Station reported an unanalyzed condition associated with the Auxiliary Building Special Ventilation and Shield Building Ventilation Systems, which resulted in declaring the equipment inoperable. The issue involved recirculation flow from the Auxiliary Building Special Ventilation System back into the Auxiliary Building causing additional area heat gain resulting in exceeding the design capacity for the area fan coil units. A subsequent engineering review concluded that required equipment in the Auxiliary Building Fan Floor area would have remained functional during a design basis accident provided service water temperature does not exceed 78.14 degrees Fahrenheit. A review of service water temperatures over the past three years did not identify any occurrences exceeding this value. Therefore, the safety function of the systems would have been met. Based on the analysis performed, Event Notification EN# 43633 is hereby retracted. The Auxiliary Building Special Ventilation and Shield Building Ventilation Systems remain in a non-conforming condition because the maximum service water design temperature is 80 degrees Fahrenheit. Service water temperature is being administratively limited to 71 degrees Fahrenheit pending final corrective action. The licensee has notified the Point Beach NRC Senior Resident Inspector. Notified R3DO (Madera). | Service water Shield Building | |
| ENS 43651 | 18 September 2007 17:00:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Shield Building Ventilation System Inoperable | At 1200 a condition associated with the Shield Building Ventilation Train A damper controller was identified that rendered the system inoperable. This condition was identified while Shield Building Ventilation Train B was inoperable for routine maintenance. Consequently, both trains of Shield Building Ventilation were simultaneously inoperable. Technical Specification TS 3.6.c requires a reactor shutdown within 12 hours of this condition. The routine maintenance on Shield Building Ventilation Train B was completed and retest performed at 1331. Shield Building Ventilation Train B was returned to operable and Technical Specification requirements to shutdown were exited. The plant remains in a 7 day Action Statement per TS 3.6.c with Shield Building Ventilation Train A inoperable . This event is reportable under 10CFR50.72(b)(3)(v)(C), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.' I&C technician was walking down a different job in the area and noticed a light illuminated on the A control board which should not have been illuminated. Additional investigation revealed an electrical condition would have prevented the modulation of the dampers thereby rendering the A train system inoperable. The licensee notified the NRC Resident Inspector. | Shield Building | |
| ENS 43689 | 3 October 2007 20:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Potential Unavailability of Two Way Radio System to Support Safe Shutdown | In response to Self-Assessment Report SA014668 and preparation for the 2008 Triennial Fire Protection Inspection, a Fire Protection Improvement Plan implementation is in progress. Preliminary analysis of the major equipment, power supplies and cables associated with the two-way radio communication system (DCR-3341) indicates the potential for the plant two-way radio system to be adversely impacted and potentially unavailable to support post-fire safe shutdown operator actions and/or fire brigade fire fighting activities for a fire location in: - Fire Zone AX-35 (Control Room and AC Equipment Room) - Fire Zone TU-22 (Turbine Room) at the mezzanine elevation just outside Battery Room 1B - Fire Zone TU-98 (Battery Room 1B) This review has identified discrepancies regarding the credited means of communication required for use by Operators in response to an Appendix R fire. The safe shutdown procedures E-O-06 and E-O-07, and the Manual Action Feasibility Study (Fire Protection Engineering Evaluation FPEE-003) only credit the plant two-way radio system. However, upper tier program documents (e.g., Fire Protection Program Plan, Appendix R Design Description) do not consistently contain the same requirements. For example, the Fire Plan, Rev 7, Section 12.9 requires both the 5-channel Gai-Tronics system between key shutdown locations AND the multi-channel portable radio communications equipped with repeaters and provided for use by the plant fire brigade shall be operable at all times. The Appendix R Design Description. Rev. 5 is silent on safe shutdown communications. It is not clear at this time whether the 5-channel Gai-Tronics should also be credited in the safe shutdown procedures. If so, then the cables supporting operation of the Gai-Tronics would need to be identified and located by fire zone to determine their availability in lieu of two-way radio communications for a fire in any of the three fire zones identified above. Until this is verified, the Appendix R timeline for achieving sate shutdown may not be able to be met. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(V)(A), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) shutdown the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, (C) control the release of radioactive material, OR (D) mitigate the consequences of an accident. The licensee will notify the NRC Resident Inspector.
EN# 43689 provided notification that the Appendix R timeline for achieving safe shutdown may not be achievable due to potential loss of the credited two-way radio communication system due to fire and consequent need for face-to-face communications between the Operators. This position was adopted pending verification of the availability of the Gai-Tronics paging system. A subsequent engineering analysis of the major equipment, power supplies and cables associated with the Fire Protection/Appendix R two-way radio communication system, the Gai-Tronics plant paging system and the dedicated Emergency Gai-Tronics System, determined that adequate communications would have been available during a fire in the subject fire zones. Operators are familiar with and skilled in the use of the Gai-Tronics system as part of their job function. Interviews with on-shift Operators confirmed that operations would use the two-way radios; and if they failed, would then use the nearest Gai-Tronics handset station and then Emergency Gai-Tronics at specific locations for an Appendix R Dedicated Shutdown scenario. Where Operator manual actions are in close proximity to the Dedicated Shutdown panel, face-to-face communications would be achievable and timely. Consequently, the assumption of only face-to-face communications described in the Event Report would not have been necessary for all safe shutdown actions. Use of redundant communications systems (Gai-Tronics and Emergency Gai-Tronics) would have been available for fires in the subject fire zones such that the Appendix R safe shutdown time requirements would not have been significantly impacted. The loss of the two-way radio system for the identified fire zones would not have prevented the fulfillment of the safety function of systems that are needed to shut down the reactor and maintain it in a safe shutdown condition. As such, this condition is not reportable under 10CFR50.72(b)(3)(v)(A) as previously stated. Consequently, the notification made on 10/03/2007 (EN 43689) is hereby retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (Lipa). | ||
| ENS 44027 | 4 March 2008 07:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Maintenance on Mlshlcot Substation by Wps Results in Greater than 50% Siren Coverage Loss | On March 4, 2008, at 0100 CST, ten emergency notification system sirens in the Kewaunee Power Station Emergency Planning Zone (one in Kewaunee County, nine in Manitowoc County) were de-energized due to a planned power outage. The loss of power to these emergency notification sirens resulted in a lost population coverage of 51.2%. As a result, this event is being reported under 10 CFR 50.72(b)(3)(xiii) and guidance in NUREG-1022 as a major loss of off-site communications capability. The expected duration of the planned power outage is 2 hours. Kewaunee and Manitowoc County Emergency Management agencies have already established appropriate contingency measures for the planned loss of the emergency notification system. Contingency measures include route alerting and an available plane with loud speaker for the areas affected by the power outage. The State of Wisconsin and the Federal Emergency Management Agency were previously informed of this planned outage. The NRC Resident Inspector has been notified. | ||
| ENS 44144 | 16 April 2008 19:26:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Two Gallons of Radioactive Water Spilled in Protected Area | A small quantity of radioactive water spilled on the concrete in the Protected Area and is in the process of being cleaned up. Approximately two gallons of water leaked from a shipping container being moved from Containment. The water was contained to the local concrete and tractor trailer receiving the load. The tractor trailer remains in the Protected Area and will be processed for free release. Initial analysis shows the presence of Co-58, Mn-54, Co-60, and Cs-137. Specific activity levels are not available at this time. No individual contaminations occurred during this event. The licensee has reported this event to the Wisconsin Department of Natural Resources (WDNR) pursuant to WDNR Regulation NR706.03 at 1426 CDT. Subsequent review has determined that this event was not reportable to the WDNR and an update was provided to the WDNR at 1800 CDT. This event is reportable under 10 CFR 50.72 (b)(2)(xi), 'Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The Licensee notified the NRC Resident Inspector. | ||
| ENS 44182 | 30 April 2008 21:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Two Emergency Diesel Generators Inoperable | Two emergency diesel generators found inoperable while two trains of residual heat removal are required to be Operable for core cooling. On 4/30 at 1600 it was discovered that the level in Emergency Diesel Generator Fuel Oil Storage tanks were not trending together. The two underground fuel oil tanks have a capacity of 35,000 gallons each (useable volume of 34,674 gallons). The tanks are connected with a siphon line between the tanks to ensure a passive means of transfer so one emergency diesel can use both tanks. The siphon line does not appear to be transferring fuel oil. KPS is in a refueling outage and in REFUELING SHUTDOWN mode with the vessel drained to 6" below the flange for reactor vessel reassembly. Both trains of RHR are required to be OPERABLE per TS 3.1.a.2.B. Technical Specification 3.7.c states that when its normal or emergency power source is inoperable, a system, train or component may be considered OPERABLE for the purpose of satisfying the requirements of its applicable LCO provided: 1. Its corresponding normal and emergency power source is OPERABLE; and 2. Its redundant system, train, or component is OPERABLE. Kewaunee Power Station (KPS) Technical Specifications Section 3.7.a.7 states that the reactor shall not be made critical unless both diesel generators are operable. The two underground storage tanks combine to supply at least 35000 gallons of fuel oil for either diesel generator and the day tanks for each diesel generator contain at least 1,000 gallons of fuel oil. The current volume is 29,700 gallons in Tank A and 21,500 gallons in Tank B. Based on two trains found inoperable, this is reportable under: 10 CFR 50.72(b)(3)(ii) (8-Hr) 10 CFR 50.72(b)(3)(v)(B) (8-hr) Per TS 3.1.a.2.B the condition must be fixed immediately. Troubleshooting/investigation is in progress to determine repairs. Cause is being investigated. Condition was found during engineering review of logs taken during operator rounds. The licensee notified the NRC Resident Inspector.
EN 44182 provided notification that the emergency diesel generator (EDG) fuel oil tanks were not capable of supplying the combined volume of fuel oil (35,000 gallons) from two storage tanks to either EDG as required by Technical Specification (TS) 3.7.a.7 for EDG operability. That notification was based on a conservative determination that a siphon line, which connects the two fuel oil storage tanks, was nonfunctioning. TS 3.7.a.7 requires both EDGs to be operable when the reactor is critical. Although the reactor was in refueling shutdown mode at the time of this event (i.e., not critical, and therefore not in the mode of applicability for TS 3.7.a.7), the EDGs provide a support function (emergency power) for the Residual Heat Removal (RHR) system; two trains of which were required to be operable for core cooling per TS 3.1.a.2.B. A subsequent review confirmed that the capability to transfer fuel from either storage tank to either EDG existed independent of the siphon line. The siphon line is not needed for transferring fuel oil from either tank to either EDG. There are alternate means available to transfer fuel oil to the EDGs; sufficient time is available to implement these alternate means of fuel oil transfer within the time available. Additional fuel oil supply margin existed due to the mode of operation at the time (refueling shutdown). The 35,000 gallon TS requirement is based on the expected fuel consumption of one EDG operating for seven days at continuous rated load, thus ensuring adequate time to restore off-site power or to replenish fuel. The amount of fuel in the storage tanks was sufficient for the EDGs to be capable of performing their support function for RHR in the refueling shutdown mode. As such, both RHR trains remained operable and TS 3.1 a.2.B was satisfied. Consequently, this condition did not meet the reportability criteria in 10 CFR 50.72. As a result, the notification made on 4/30/2008 (EN 44182) is hereby retracted. The licensee notified the NRC Resident Inspector. Notified the R3DO (Peterson). | Emergency Diesel Generator Residual Heat Removal | |
| ENS 44409 | 13 August 2008 14:31:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | A licensed employee had a confirmed positive for alcohol during a follow-up fitness-for-duty test. The employee's access to the plant has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| ENS 44482 | 11 September 2008 18:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Related to Certain Fire Conditions | During an NRC (Triennial Fire) Inspection, a condition was identified in which, under certain fire conditions, possible damage scenarios, and required operator actions, the ability to meet the performance criteria of the approved fire protection program may be challenged. Specifically it was postulated under certain fire conditions affecting the relay room, the ability to maintain pressurizer level within indicated range (high) may be challenged due a spurious opening of a pressurizer power operated relief valve. Compensatory actions to address the fire in the area of concern are in place. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector. | 05000305/LER-2008-001 | |
| ENS 44572 | 16 October 2008 18:17:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Related to Certain Fire Conditions | While performing an extent of condition review resulting from the unanalyzed condition related to certain fire conditions identified in Event Notification #44482, another situation was identified in which, under certain fire conditions, the ability to meet the performance criteria of the approved fire protection program may be challenged. Specifically, it was postulated under certain fire conditions affecting the relay room, the ability to maintain pressurizer level within the indicated range may be challenged due to spurious opening of pressurizer vent and reactor head vent valves. Compensatory actions to address the fire in the area of concern are in place. (1-hour Fire Watch). This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). NRC Resident has been notified. | 05000305/LER-2008-001 | |
| ENS 44616 | 30 October 2008 14:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(ii)(B), Unanalyzed Condition | Non-Functional Steam Exclusion Barrier | On 10/30/08 at 0930 an RP Technician transiting though a steam exclusion door found a kickplate degraded. The kickplate is held on by two screws and one screw was missing. When the door was opened, the kickplate rotated and became lodged in the staircase grating. This prevented the door from closing until the technician physically lifted the plate out of the way to close the door. The door was open for less than a minute. This kickplate is not part of the door seal itself so when the door is closed it is functional. The kick plate was taped up as a temporary fix and access was restricted through the door until permanent repairs were complete. Permanent repairs were completed on the door at 1116 on 10/30/08. While the door was open and could not close automatically, the barrier was non-functional. In accordance with TRM 3.0.9 Section A.1 all equipment supported by that steam exclusion barrier was immediately declared inoperable. This zone includes both trains of ECCS and support equipment (i.e., SI, RHR, ICS, CCW, etc ). TS 3.0.c was entered and exited during the time the door was open with both trains of ECCS inoperable. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function', and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety'. The licensee notified the NRC Resident Inspector. | 05000305/LER-2008-002 | |
| ENS 44718 | 13 December 2008 20:20:00 | 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 | Steam Exclusion Boundary Door Latch Failure | At 1420 on 12/13/08 a plant electrician identified that a steam exclusion door would not close when he was transiting through the door. This door would not have closed and maintained the Steam Exclusion Boundary and could have led to steam through out the Auxiliary Building which could have resulted in both Trains of ESF Equipment failing to perform their required functions, i.e., SI, RHR, CC etc. A Unit Supervisor who was in the Auxiliary Building at the time immediately went up to the door and found the latch was broke and the broken piece was removed which then enabled the door to close and latch. At 1425 on 12/13/08 the Steam Exclusion Boundary was restored to functional and both Trains of ESF Equipment restored to operable. The licensee informed the NRC Resident Inspector.
01/22/2009 - Retraction of EN 44718, both trains of Engineered Safeguards Features (ESF) equipment inoperable due to a degraded steam exclusion boundary door. EN 44718 provided notification that both trains of ESF equipment (e.g. SI, RHR, CCW, etc.) were inoperable due to degradation of a steam exclusion boundary door in the auxiliary building on December 13, 2008. Subsequent engineering evaluation determined that the degraded door remained capable of fulfilling its steam exclusion function during the brief (five minute) period when it was degraded. A degraded latch mechanism prevented the door from latching closed. However, the door was closed against its door jam, which resulted in a sufficiently small gap such that total allowed leak path criteria were not exceeded. Additionally, the door swing was in the direction of postulated steam flow, such that the door would have been held in the closed position by any steam overpressure postulated under accident conditions. Therefore, the door remained functional and the supported ESF equipment in the auxiliary building remained operable. Consequently, this condition did not meet the reportabllity Criteria in 10 CFR 50.72. As a result, the notification made on 12/13/2008 (EN 44718) is hereby retracted. The licensee will notify the NRC Resident Inspector. Notified the R3DO (O'Brien). |