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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 52666 | 5 April 2017 19:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Unanalyzed Condition for Both Edg Transfer Line Connections | Both Emergency Diesel Generators (EDGs) have truck connections connected to transfer lines that are potentially not compliant with general design criteria. A potentially unanalyzed condition exists due to threat of tornado generated missiles. While in Mode 1 at 100% power, the Control Room was notified that the outdoor portion of the line upstream of JEV0001, EMERG FUEL OIL STORAGE TK A TRUCK CONN ISO, and the outdoor portion of the line upstream of JEV0002, EMERG FUEL OIL STORAGE TK B TRUCK CONN ISO, potentially have not been reviewed to meet general design criteria. No major equipment was out of service. No systems were required to respond to this event. The unit remains in Mode 1 at 100% power. The NRC Senior Resident Inspector has been notified. Compensatory measures have been established IAW (in accordance with) EGM 15-002. The Unit entered Tech Spec 3.8.1 Condition B and D for approximately 45 minutes until compensatory measures were put into effect. The licensee identified this condition during a design review and is currently identifying long-term corrective actions.
The Unit entered Tech Spec 3.8.1 Condition B and F, not Tech Spec 3.8.1 Condition B and D mentioned earlier. Notified R4DO (Vasquez). | Emergency Diesel Generator | 05000482/LER-2017-003 05000482/LER-2017-002 |
| ENS 50628 | 19 November 2014 21:37:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | On 11/19/2014, Wolf Creek determined that the alternative shutdown procedure for responding to a fire in the control room does not address all potential operating modes of the credited emergency diesel generator. During the 2014 Triennial Fire Protection Inspection, it was postulated that a loss of offsite power following a fire in the control room would cause the Train B emergency diesel generator (EDG) to start and load to the bus. The fire is also postulated to cause damage to the Train B essential service water (ESW) pump control circuit and prevent the pump from automatically starting and cooling the EDG. Additionally, the fire is postulated to cause spurious operation of another large load that is not normally sequenced onto the bus following a loss of offsite power. This postulated scenario loads the EDG to approximately 53% of its rated load without cooling. This scenario was not considered when developing the control room fire response strategy. Preliminary calculations show that operators have 3.6 minutes to establish cooling prior to the EDG tripping on high jacket water temperature. Performance timing determined that operators would establish cooling in approximately 10 minutes. Based on this information, it was determined that this condition is unanalyzed and is potentially reportable per 10 CFR 50.72(b)(3)(ii)(B). An hourly fire watch compensatory measure is in place in the control room, consistent with procedural requirements for a Post Fire Safe Shutdown circuit analysis deficiency. The licensee has notified the NRC Resident Inspector. | Service water Emergency Diesel Generator | |
| ENS 50468 | 18 September 2014 16:55:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | During a review of INPO Event Report 14-33, Direct Current Circuits Challenge Appendix R Fire Analysis, it was determined that portions of the control circuits for the Turbine Generator DC Emergency Lube Oil Pump and the Emergency DC Seal Oil Pump are not properly fused to prevent overload and possible secondary fires. The review found that a fire at the motor starter cabinet in the turbine building could cause specific smart hot shorts that could cause overheating of the control cable and result in secondary fires outside the turbine building. Based on this information, it was determined that this condition is unanalyzed and is potentially reportable per 10 CFR 50.72(b)(3)(ii)(B). Hourly fire watch compensatory measures are in place in the affected areas of the Turbine Building. The presence of compensatory measures in addition to automatic fire detection and suppression in these fire areas ensures protection of the equipment. The licensee has notified the NRC Resident Inspector. | ||
| ENS 49423 | 9 October 2013 22:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | A review of industry operating experience with respect to fire induced damage to unfused Direct Current (DC) ammeter circuits in the control room has determined that the condition described below is applicable to Wolf Creek Nuclear Generating Station. This condition results in an unanalyzed condition with respect to 10CFR50 Appendix R analysis requirements. The original plant wiring design did not include overcurrent protection features to limit the fault current in these circuits. The wiring design for the ammeters contains a shunt in the current flow from each NK direct current (DC) battery or charger. Two leads run from the shunt to a current meter in the main control room (MCR). These leads are tied to the positive polarity of the NK battery system. The ammeter wiring attached to the shunt is not overcurrent protected. It is postulated that a fire could cause one of these ammeter wires to short to ground at the same time the fire causes another DC wire from the opposite polarity on the same battery to also short to ground. This would cause a ground loop through the unfused ammeter cable. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment. Reference Palo Verde plant event #49411. A Breech Authorization with compensatory Control Room hourly fire watch for this issue is in place and will remain in effect until this deficiency is resolved. This condition has been discussed with the Resident Inspector. Similar Events #49422 and #49419 | ||
| ENS 48134 | 25 July 2012 21:02:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Identification of a Degraded or Unanalyzed Condition | At 1602 (CDT), Engineering personnel notified the control room that during review of a pipe stress calculation it was identified that non-conservative or incorrect methodologies were used in the calculation. This calculation was for a modification to install four; 3 (inch) drain lines between the Essential Service Water (ESW) (safety) and the Service Water (SW) (non-safety) in 1991. A preliminary ME101 stress analysis performed, which corrects the above-identified discrepancies, indicates that the pipe stresses at the drain line weldolet connection exceed the ASME code of record allowable stresses by approximately 50%, when the revised Stress Intensification Factor (SIF) is applied. This modification affected both trains (A & B) ESW trains. The normal system alignment uses the SW water to supply the ESW, then during accident conditions the SW and ESW systems isolate from each other so that two redundant separate train isolation valves isolate the ESW system. These 3 (inch) drain lines are located in the section of piping that is isolated from the ESW and SW systems. At the time of notification 'A' ESW was isolated from SW and 'B' ESW was in normal system alignment. 'B' ESW was declared inoperable and action was taken to separate the SW and ESW and isolate the 3 (inch) drain valves. With this action complete the non-conforming components have been removed from service and OPERABILITY of the ESW has been restored. This condition is been reported per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
Further engineering evaluation determined that the four drain lines (3-inch) between the Essential Service Water (ESW) (safety) and the Service Water (SW) (non-safety) were found to be within the allowable limits for operability and are acceptable. As a result, the condition has been determined to not be reportable per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. Notified R4DO (Miller). | Service water | |
| ENS 47412 | 3 November 2011 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition for a Postulated Control Room Fire | During the 2011 Triennial fire protection inspection, it was determined that the off normal procedure for control room evacuation due to fire has two defects. It does not adequately protect the steam generators from overfilling and possibly damaging the turbine driven auxiliary feedwater pump. In addition it does not protect the reactor coolant system pressurizer from filling to above 100% indicated water level, possibly causing the primary system to go solid. Both of these issues are results of inadequate assumptions used in the Post Fire Safe Shutdown Analysis of a fire in the Control Room. Compensatory measure of hourly fire watch for the control room is in place. The procedure for control room evacuation due to fire is being revised to include compensatory actions that will address the above events. The NRC Resident Inspector has been notified. | Steam Generator Reactor Coolant System Auxiliary Feedwater | |
| ENS 47076 | 20 July 2011 18:15:00 | 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 | Fire Protection Program Concern | On July 20, 2011 at 1315 hours it was determined that a design deficiency at the Wolf Creek Nuclear Operating Corporation (WCNOC) constituted a fire protection program concern which could adversely affect the ability to achieve and maintain safe shutdown in the event of a control room fire. In the event of a postulated control room fire, a normally closed valve (EFHV0060) on the Essential Service Water (ESW) system return from the Component Cooling Water (CCW) heat exchanger could spuriously open. If this occurs, the flow balance in the ESW system would be affected and cooling flow to other essential components could be reduced to below the minimum required flow. Present system operability is not affected as there has been no occurrence of a fire in the Control Room and compensatory actions are in place to detect and mitigate the effects of a fire in the Control Room. An hourly fire watch is currently in place in the control room and will remain in place until this issue is resolved. In addition to the hourly fire watch, alternate shutdown procedure OFN RP-017 is being revised to include interim compensatory actions to deenergize, and verify closed, valve EFHV0060. This event is being reported in accordance with 10 CFR 50.72 (b)(3)(ii)(B) and 10 CFR 50.72 (b)(3)(v)(A) as an unanalyzed condition that significantly degrades plant safety. A follow-up licensee event report will be made in 60 days. The NRC Resident Inspector has been notified. | Service water | |
| ENS 46909 | 1 June 2011 18:15:00 | 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)(ii)(B), Unanalyzed Condition | Both Trains of Component Cooling Water Declared Inoperable Due to Voiding | At 1315 (CDT), while in Mode 3 at normal operating pressure and 552?F, both trains of Component Cooling Water (CCW) were inoperable due to indications of voiding. The 'A' Train CCW had been declared inoperable at 1000 (CDT) when review of pump test data indicated a potential void and Technical Specification LCO 3.7.7, 'Component Cooling Water (CCW) System,' Condition A was entered. At 1315 (CDT), indications of voiding were identified in the common service loop piping, which was aligned to the 'B' Train CCW. The 'B' Train CCW was declared inoperable and the plant entered LCO 3.0.3 due to both CCW Trains being inoperable. At 1410 (CDT) plant cool down to Mode 4 was commenced. The 'A' Train CCW has subsequently been vented and void volume is currently within allowable limits for operability. However, further evaluation of this voiding is underway prior to declaring the 'A' Train CCW operable. The licensee has notified the NRC Resident Inspector.
The licensee is retracting this event based on the following: Further engineering evaluation concluded the amount of gas ingested by the 'A' CCW pump would not cause any degradation to the pump. The remainder of the gas in the system was less than the acceptance criteria for the CCW system. The 'A' CCW train was capable of performing its specified safety function and therefore would have been considered operable. The condition would not have prevented the CCW System from fulfilling its safety function and would not be reportable under 10 CFR 50.72. The licensee has notified the NRC Resident Inspector. R4DO (Gaddy) notified. | ||
| ENS 46488 | 17 December 2010 21:35:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Potential Equipment Actuations Due to Electrical Short Circuits | A post-fire safe shutdown (PFSSD) circuit analysis identified the potential for a fire in the Control Room to cause multiple proper polarity hot shorts causing the Pressurizer Power Operated Relief Valves (PORV) to open. This postulated failure would require at least three separate 'smart' hot shorts including 1) a positive external hot short, 2) a negative external hot short, and 3) a short across the control room hand switch to cause each PORV to open. This could result in a loss of coolant accident inside the containment building. This circuit analysis also identified that a fire in the control room could cause the recirculation dampers associated with the 'B' Essential Service Water Pump Room and 'B' Emergency Diesel Generator Room to fail open, closed, or anywhere in between. This could cause the room temperatures to drop below the minimum or rise above maximum design temperatures depending on ambient conditions. Compensatory measures are established for early detection and extinguishment of a fire associated with these circuits in the Control Room. Procedure changes are also being developed for Control Room Evacuation. The licensee notified the NRC Resident Inspector. | Service water Emergency Diesel Generator | |
| ENS 46428 | 18 November 2010 22:26:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Post Fire Safe Shutdown Unanalyzed Condition | A Post Fire Safe Shutdown (PFSSD) circuit analysis identified that certain fuses installed within Train 'B' Exciter/Voltage Regulator cabinet NE106 are susceptible to failure in the event of postulated fire-induced hot shorts within the control room. Loss of power to this circuit will prevent operation of these functions. The emergency pre-positions and manual voltage adjustment circuits are not credited for PFSSD following a control room fire. However, field flashing is credited following a fire in the control room. Failure of field flashing after a postulated fire will prevent voltage generation on the 'B' diesel generator. This could result in the inability of the 'B' train EDG to supply its associated safety bus during the postulated fire. Compensatory measures are established for early detection and extinguishment of a fire associated with this circuit in the Control Room. Additional compensatory measures are being developed. The licensee is not in any tech. spec LCO's as a result of this condition and has established fire watches as a compensatory measure. The licensee notified the NRC Resident Inspector. | 05000482/LER-2010-013 | |
| ENS 46196 | 21 August 2010 22:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | One Train Eccs Declared Inoperable Due to Void in "a" Rhr Heat Exchanger | On 8/21/10 at 1700 (CDT) with the Unit in Mode 1, 100% power, Engineering personnel provided information to the Control Room that a known void in the 'A' Residual Heat Removal (RHR) heat exchanger could move from the heat exchanger to other locations in the Emergency Core Cooling System (ECCS). The analysis provided by a vendor indicates that during a specific RCS leak scenario with a failure of the A' RHR Pump, our (Unit 1) piping configuration could potentially allow the void to be swept from the 'A' RHR heat exchanger and be transported to the 'A' Centrifugal Charging Pump (CCP) and 'A' Safety Injection Pump (SIP) while in the cold leg recirculation mode of operation. This condition results in one train of ECCS being inoperable. The ECCS safety function is maintained due to redundant and interconnecting piping, in both injection and recirculation phases of operation. The NRC Resident has been informed. The licensee declared one train of ECCS inoperable placing Unit 1 in the 72-hour Technical Specification (TS) 3.5.2, Condition A, Action A.1. Corrective actions are on going to remove the void and exit the TS Action Statement.
Additional information has been obtained by Engineering. The largest void identified in the 'A' RHR heat exchanger would not prevent the RHR system from performing its safety function. Hydraulic analysis models of this void moving through the system have shown that no pressure pulses will occur that could challenge the structural integrity of the system. The model results also show that ingestion of the void by the higher head ECCS pumps during the ECCS recirculation mode has no adverse affect on the ECCS high head pumps and they would remain capable of performing their safety function. As a result, this event is not reportable per 50.72(b)(3)(ii)(B). This event is being retracted. The licensee notified the NRC Resident Inspector. R4DO (Lantz) has been notified. | Residual Heat Removal Emergency Core Cooling System | |
| ENS 46136 | 28 July 2010 19:22:00 | 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)(ii)(B), Unanalyzed Condition | Three Auxiliary Feedwater (Afw) Trains Declared Inoperable | At 1422 CDT, while operating in MODE 1 at approximately 100% rated thermal power (RTP), Wolf Creek Generating Station (WCGS) entered Technical Specification 3.7.5, 'Auxiliary Feedwater (AFW) System,' Condition D for three AFW trains inoperable. The AFW pumps are equipped with recirculation lines to prevent pump operation against a closed system. The recirculation line for each pump combines into a common header that flows to the condensate storage tank (CST). The common recirculation line transitions from safety-related to non-safety related at the AFW to CST pipe tunnel. During a review of the system design information, it was identified that a seismic event or tornado could result in damage to this portion of the recirculation line and thereby restrict or block the recirculation flow causing pump damage. TS 3.7.5, Required Action D.1, requires the immediate initiation of action to restore one AFW train to OPERABLE status. The Note to Required Action D.1 specifies that LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. The plant remains in MODE 1 at approximately 100% RTP with activities that could lead to the need to maneuver the plant being suspended. WCGS personnel are developing an evaluation to support the OPERABILITY of the AFW trains with the reliance on compensatory measures as an interim action until final corrective action to resolve the condition are completed. The NRC Resident Inspector has been notified.
Further evaluation by the Engineering Department determined that the capability of the AFW system was not reduced. The postulated condition is considered a design deficiency (nonconforming condition) to be corrected before considering the equipment fully qualified. The safety functions and the capability of the AFW system to mitigate a design basis event are not impaired. As a result, the condition has been determined to not be reportable per 10 CFR 50.72(b)(3)(ii)(B), 10 CFR 50.72(b)(3)(v)(A) or 10 CFR 50.72(b)(3)(v)(B). The licensee will notify the NRC Resident Inspector. Notified the R4DO (Farnholtz). | Auxiliary Feedwater | |
| ENS 45859 | 20 April 2010 18:07:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Five Potential Issues Discovered During Post-Fire Safe Shutdown Review | While performing a post-fire safe shutdown (PFSSD) review per Enforcement Guidance Memorandum (EGM) 09-002, 'Enforcement Discretion for Fire Induced Circuit Faults', and Regulatory Guide (RG) 1.189, Revision 2, five issues were identified that could be potentially reportable. EGM 09-002 provides enforcement discretion for non-compliances related to fire-induced multiple spurious operations (MSOs). The enforcement discretion period allows 6 months from the date of issuance of RG 1.189, Rev. 2 to perform the following: - Identify noncompliances related to multiple fire induced circuit failures - Implement compensatory measures for the noncompliances, and - Place the noncompliances in the licensees' corrective action program The five fire-induced issues that were identified are: - Three issues which could potentially prevent operation of the Residual Heat Removal Pump(s), - An issue where all four Reactor Coolant Pumps may not stop from the Control Room, and - An issue where a Centrifugal Charging Pump could spuriously start and fill the pressurizer solid. RG 1.189 was issued on November 2, 2009. Therefore, the 6-month identification period ends on May 2, 2010. For issues identified during the 6-month identification period, enforcement discretion continues for an additional 30 months to resolve identified issues. The 10 CFR 50, Appendix R Fire is not a Design Basis Accident. While Appendix R fires are postulated to cause a Technical Specification (TS) inoperability of a (System, Structure or Component) SSC, the failure to previously identify these issues does not constitute an actual TS inoperability. Rather, the issues identified during this MSO are noncompliances with the Current Licensing Basis (CLB). Therefore, the affected systems are considered operable but degraded or non-conforming as defined in Regulatory Issue Summary (RIS) 2005-20. The licensee has notified the NRC Resident Inspector. | Residual Heat Removal | |
| ENS 45692 | 10 February 2010 00:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Post Fire Safe Shutdown Unanalyzed Condition | A potential Post-Fire Safe Shutdown issue associated with the 'B' Emergency Diesel Generator Voltage Control Circuitry has been discovered, should a Control Room fire cause a hot short in the 'B' EDG Voltage Control Circuitry. A fire in the control circuit that renders the 'B' EDG inoperable and is large enough to require the Control Room to be evacuated during a loss of offsite power is an unanalyzed condition. This condition was discovered during a planned maintenance outage for the 'B' EDG. Compensatory measures are being established for early detection and extinguishment of a fire associated with this circuit in the Control Room. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | 05000482/LER-2010-003 |
| ENS 45277 | 19 August 2009 20:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Time to Establish Charging Flow Exceeded Thermal Hydraulic Analysis | Valve BNHV8812A RWST to RHR Pump A Isolation is manually closed in OFN RP-017, Control Room Evacuation, in response to a control room fire. Simulated timing assumed approximately 1 minute to close the valve. Discussions with the MOV engineer revealed that the valve takes 600 turns to close. Operations Standing Order 1 limits the number of hand wheel turns to 60 per minute. At a closure rate of 60 turns per minute, it will take a minimum of 10 minutes to close the valve, rather than the assumed 1-minute. The additional 9 minutes necessary to close the valve will delay the completion of subsequent steps. Most significantly, the time to establish charging flow increases from 20 minutes to a minimum of 29 minutes. Thermal hydraulic analysis documented in SA-08-006 assumed the charging flow will be established within 28 minutes. Therefore the time to establish charging flow exceeds that assumed in the thermal hydraulic analysis by at least 1 minute. Compensatory actions have been put in place in order to accomplish this task within the time allowed by the thermal hydraulic analysis. The licensee has notified the NRC Resident Inspector. | ||
| ENS 44586 | 21 October 2008 05:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition During Design Basis Fire in Area A-27 of Auxiliary Building | During a design-basis fire in area A-27 (2026' level Auxiliary Building) actions designated in OFN KC-016, FIRE RESPONSE, remove safety-related 125 VDC Power from the B Train Pressurizer PORV, BB PCV-456A, by opening fused disconnects NK0404 and NK0405. The intent of this action is to remove all power from the affected cable tray. This action removes control power to several safety-related components, which are assumed to remain available to the Control Room since the analyzed fire in area A-27 relies on the B Train to maintain safe-shutdown conditions. Opening NK0404 and NK0405 and removing control power fm the assumed safe-shutdown train was not fully analyzed as to the impact on Control Room Actions and subsequent affects on availability of the assumed safe-shutdown train. Actions taken or planned: 1) Fire detection and suppression in area A-27 are functional. 2) Established an hourly firewatch in area A-27 per the AP 10-104, Breach Authorizations. 3) Changed OFN KC-016, FIRE RESPONSE, to isolate the specific fused disconnect for BB PCV-456A, NK4421, and added a note to the operator to be aware that the PORV could reopen due to a cable-to-cable hot short. The licensee notified the NRC Resident Inspector. | ||
| ENS 44172 | 26 April 2008 18:18:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Discovered Due to Improperly Installed Fuse Wiring in Class 1E Air Conditioning Unit | While trouble shooting, it was discovered that fuses for a remote switch to start the 'B' Class 1E air conditioning unit were installed in series and therefore not installed per design. Drawings show these fuses wired in parallel, rather than in series. In the series configuration, if an electrical short occurs on the Control Room side of the circuit and blew the fuse, the 'isolated' portion of the circuit, fed from the next fuse, would also loose power. Therefore, the circuit could not be re-energized and the unit could not be started in the event of a Control Room Fire. The Class 1E electrical equipment air conditioning units are included in the post fire safe shutdown design to ensure adequate room cooling for the operating train of safety related electrical equipment. The air conditioning units provide a support function for Class 1E electrical equipment required for safe shutdown. The loss of the Class 1E air conditioning units does not directly result in loss of capability to achieve and maintain safe shutdown in the event of a fire. Rather, room heating beyond design limits could reduce the life of electrical components within the switchgear. Qualification data exists to show that some components within the switchgear will survive the expected room temperatures and be functional following a loss of room cooling. Data does not exist for each and every component within the switchgear, so a conclusive argument cannot be made regarding equipment functionality following a loss of room cooling. Therefore, this condition resulted in an unanalyzed condition that could potentially affect post-fire safe shutdown equipment availability. There are no compensatory measures required since the unit is not operating. The configuration will be corrected prior to changing to mode 4. The licensee notified the NRC Resident Inspector. | ||
| ENS 44131 | 10 April 2008 15:15:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | "Analysis Indicates Containment Coolers May Not Have Automatically Started in Slow Speed Following Postulated Main Steam Line Break | Information provided by another plant identified a condition where Containment Fan Coolers may trip if running in fast speed and not start automatically in slow speed, following a Loss of Coolant (LOCA) or Main Steam Line Break (MSLB). Wolf Creek commenced evaluation to determine if our Containment Coolers were susceptible to this condition. Investigation and analysis as of this date indicate that in the case of a Main Steam Line Break, the coolers could trip while running in fast speed and not be able to be automatically started by the sequencer in slow speed due to present electrical design configuration. The peak containment pressure in this case would exceed the analysis of record by approximately 5 psig. For the spectrum of the smaller break LOCA's, where actuation is delayed until pressurizer low pressure Safety Injection or the Containment Hi 1 pressure signal, analysis shows that the Containment Pressures would not be exceeded if the Containment Fan Coolers did not automatically start in slow speed. For a Large Break LOCA, analysis shows that the Containment Cooler Fans would be shed prior to tripping, and would be automatically started in slow speed by the sequencer. Wolf Creek is currently in a defueled condition, for Refueling Outage 16. Continued evaluation and analysis of this issue is ongoing. This issue will be resolved prior to the plant entering Mode 4, where Technical Specification 3.6.6 requires Containment Fan Coolers be operable. The information was originally received from Callaway. The licensee notified the NRC Resident Inspector.
The licensee is retracting this event based on the following: Wolf Creek evaluated this concern and concluded that the condition did not exist at Wolf Creek. Further analysis of the Main Steam Line Break, if this concern had existed, showed that the calculated post-accident pressure and temperature peak values would not exceed the peak accident values in the Final Safety Analysis Report. Therefore, an unanalyzed condition did not exist and Wolf Creek is retracting the 50.72(b)(3)(ii)(B) notification. The licensee informed the NRC Resident Inspector. Notified R4DO (Farnholtz). | Main Steam Line | |
| ENS 42603 | 24 May 2006 15:38:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Discovered During Post-Fire Safe Shutdown Area Review | While performing a post-fire safe shutdown area review for fire area C-35, it was discovered that a design basis fire in this area could prevent operation of both Class 1E Electrical Equipment air-conditioning units (SGK05A and SGK05B). The Class 1E Electrical Equipment air conditioning units provide temperature control for equipment required for safe shutdown in the Class 1E switchgear rooms (ESF switchgear rooms, DC switchgear rooms, and NK battery rooms). A fire in fire area C-35 could impact fire protection panel KC230 and cables associated with the automatic shutdown feature for the air conditioning units. The automatic shutdown feature normally operates whenever a fire occurs in any of the Class 1E switchgear rooms prior to Halon fire suppression operation. A fire in fire area C-35 could result in the inability to operate the air conditioning units and provide the necessary temperature control for equipment required for safe shutdown. Compensatory measures: a continuous fire watch was established for fire area C-35, a temporary procedure change will be completed to allow use of a jumper to restore Class 1E air conditioning unit (SGK05A) to service, if required, at which time the continuous fire watch will be changed to an hourly watch. The licensee notified the NRC Resident Inspector. | ||
| ENS 42327 | 10 February 2006 17:00:00 | 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 | Potential for Loss of Safe Shutdown Capability | As a result of Wolf Creek's 2005 Triennial Fire Protection Inspection, a re-evaluation of concerns described in NRC Information Notice (IN 92-18), 'POTENTIAL FOR LOSS OF REMOTE SHUTDOWN CAPABILITY DURING A CONTROL ROOM FIRE' was performed. During that re-evaluation it was identified that in the event of a fire in the control room, 39 motor operated valves (MOVs) credited for post-fire safe shutdown could potentially fail in an unanalyzed condition. Of those 39 MOVs, failure of 4 of them could potentially prevent achieving and maintaining safe shutdown (SSD) conditions. In the 39 MOV circuits identified above, an intra-cable hot short between one conductor on the hot side of the indication circuit and another conductor on the load side of the control room hand switch could bypass the torque switch and energize either the open or close coil. If this occurs, the open or close contactor will close and the motor will operate in either the open or close direction until the motor stalls, possibly resulting in damage to the valve such that it cannot be manually operated. Present system operability is not affected as there has been no occurrence of a fire in the Control Room and compensatory actions are in place to detect and mitigate the effects of a fire in the Control Room. Actions taken or planned: An hourly fire watch was established in the Control Room due to a previous condition identified on 11/16/2005 in accordance with AP 10-104, Breech procedure. The condition identified today will be referenced on the active Breach Request that was implemented on 11/16/2005. The licensee notified the NRC Resident Inspector. | Remote shutdown | |
| ENS 42182 | 2 December 2005 00:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Emergency Diesel Capability | This report addresses a concern with the capability to field flash the diesel generator after the diesel engine has started in response to a fire in the Control Room. Step C9 in procedure OFN RP-017 opens NB0212 to drop power to NB02 and force an automatic start of the 'B' Train emergency diesel engine. Step C10 states to verify that the diesel engine is running and provides an RNO action to start the diesel if it is not running. Step C10.d in OFN RP-017 has the operator check NB02 voltage on breaker NB0201 NORMAL. There is no RNO action if the voltage is not normal. A fire in control room panel RL015 has the potential to cause a loss of diesel generator field flashing. For example, a short to ground on conductor 51 in cable 14KJK03AH with a simultaneous short to ground on conductor N1 in cable 14NEB11AA could blow the fuse(s) associated with the speed control relays (LSR and HSR). Loss of power to these relays will prevent field flashing, which will prevent voltage generation from the diesel generator, even though the diesel engine may be running. Step C10.b in OFN RP-017 has the operator place the master transfer switch (KJHS0109) in LOC/MAN. Based on drawing E-13KJ03A, this action will isolate the control room start circuit by opening the contact between G7 and H7, and effectively isolate the short circuit potential. However, the fuse may have already blown before this action is taken. The circuit is not provided with redundant fusing, so it may be necessary to replace one or both fuses (FU5 and FU6) in panel KJ122 to re-establish power to the speed relays. Wolf Creek has implemented the following compensatory measures: Staged 4 spare fuses in the emergency locker in the B Train Diesel Generator Room. Revised OFN RP-017, Control Room Evacuation. Continued hourly fire watch in the control room (established due to another concern). The licensee notified the NRC Resident Inspector. | ||
| ENS 42146 | 16 November 2005 21:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Failure to Meet Required Response Times for Shutdown Outside Control Room | After a timed walk down of an Off Normal Operations procedure for shut down outside the Control Room, it appears that required response times were not met. This is due to a new assumption that all actions would not function due to plant configuration. Some valve circuits are not protected with a second fuse that is isolated from the control room portion of the circuitry. The only fuse in the circuit may blow during the early part of the event, prior to manipulating the isolation switch on the breaker. Since the fuse could blow, there would be no control power to electrically manipulate the valve and that would cause the operator to go to the 'Response not Obtained' column adding time to the actions taken. This issue was previously identified at Wolf Creek in 1985 in IE INFORMATION NOTICE NO. 85-09. However, additional research needs to be accomplished as to why some circuits have only one fuse. Compensatory actions include: 1) A change to the procedure to direct additional actions. 2) An hourly fire watch has been added to the control room. The licensee notified the NRC Resident Inspector. | ||
| ENS 42029 | 30 September 2005 14:45:00 | 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 | Potential Vulnerabilities Discovered During Design Basis Fire Review | During a design basis fire in area A-8 (2000 level auxiliary building), (the licensee determined that) a number of Train A components, including the Train A motor driven auxiliary feedwater pump (control and power cables), are affected. Also, the turbine driven auxiliary feedwater pump (control cables) may not be available. In addition, the following spurious actuations could occur: 1.Pressurizer PORV BBPCV0455A opens and block valve BBHV8000A fails to close (Train A). 2. Steam Generator A ARV ABPV0001 spuriously opens and cannot be controlled from the control room (Train A). 3. Steam Generator C ARV ABPV0003 spuriously opens and cannot be closed from the control room (Train A). 4. Both VCT outlet valves BGLCV0112B and BGLCVO112C fail to close and normal letdown isolates, causing a reducing inventory in the VCT and possible hydrogen intrusion into the charging pump suction (Trains A & B). 5. Normal charging pump power cables pass through this fire area and may be damaged, causing the NCP to trip. 6. RHR suction valve from the RWST, BNHV8812A, loses power and containment sump valve EJHV8811A opens, causing the RWST to drain to the containment sump (Train A). 7. BIT inlet valve EMHV8803B fails to open from the control room hand switch (Train B). Actions taken or planned: 1) Detection / Suppression systems available in area A-8 are functional. 2) Hourly fire watch established IAW AP 10-104, Breech procedure. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 41786 | 10 May 2005 21:55:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Emergency Safety System Actuation Dampers Inoperable | During integrated ESFAS System Testing coming out of RF 14, it was noted (on 5/10/05) that 4 dampers receiving an SI signal did not reposition full closed as required. The dampers serve to isolate the Condenser Air Removal Ductwork in the Turbine Building from the Filtration train in the Auxiliary Building. Failure of these dampers to close was determined on 05/27/2005 to constitute an unanalyzed condition in that this path is assumed isolated in the accident analysis. The dampers receive a signal to close but are powered from an AC source requiring a Diesel Generator to supply power from a safety related motor control center. The problem occurs when a signal for loss of offsite power is inserted coincident with an SI signal. The dampers start to close, but there was no seal in circuit so that motion to the closed condition would continue once the affected Diesel Generator reenergized the safety related bus and power supply. The plant was in Mode 5 at time of discovery, but has operated at power with this configuration since the initial startup of the unit. The dampers could have been closed as evidenced by troubleshooting at the time relying on operator action once the Diesel Generator had reenergized the bus. There is no requirement for Sl and Auxiliary Building isolation in Mode 5. A modification was completed to the circuitry prior to entering Mode 4 where automatic isolation is required. The Reportability Evaluation was completed on 5/27/2005 but was not reported within the required 8 hours. The reportability of 8 hours was discovered during review on 06/21/2005 of the completed evaluation. Licensing personnel were immediately contacted. The licensee notified the NRC Resident Inspector. | ||
| ENS 41327 | 13 January 2005 01:45:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 24 Hour Condition of License Report Regarding Halon System- Actuator Connection Error | The following information was received via facsimile from the licensee: Based upon information provided by Callaway on 1/12/05, it was determined that the manual pneumatic actuators on the Halon suppression systems are piped incorrectly which may result in the inability to actuate the Halon suppression systems manually or automatically. Each manual pneumatic actuator has clearly marked "A" and "B" ports. Per the M-658 vendor manual series & drawing M-658-00025, the "A" port shall be connected to the actuation pilot manifold or top of cylinder valve and the "B" port shall be connected to the solenoid valve. A field walk down was performed and in all but one case, the manual pneumatic actuator has been piped in the opposite configuration. This effects the Halon suppression system protecting the ESF switchgear rooms, the Rod Drive MG Set room, the North Electrical Penetration room, the South Electrical Penetration Room, the Switchgear & Switchboard rooms, and the Control Room cable trenches & chases. Based upon preliminary evaluation, it appears that the Halon suppression systems are inoperable. Fire watches were implemented for the affected areas. The licensee notified the NRC resident inspector.
This event report is retracted based on the following information provided from Wolf Creek by facsimile: Investigation - Informational tests conducted by the Vendor (Chemetron) and witnessed by Wolf Creek, Callaway, and NRC personnel on January 26, 2005 determined that the Halon systems would have properly actuated in the as-found incorrect configuration (port 'A' and 'B' connections reversed). The only identified difference in the actuation sequence between the tests conducted in the incorrect configuration versus the correct configuration is a delay of less than 2 seconds from the time the solenoid received the discharge signal until the first cylinder actuated. There is no regulatory or National Fire Protection Association standard or guideline that places a time requirement on this interval. This very slight time delay would have had no effect on the designed function of the Halon suppression system to extinguish a fire. Additional details are provided in the Chemetron report, "Report on Actuation Arrangements for Halon Extinguishing System Units," (Wolf Creek correspondence 05-00072) that includes the test procedure and results. Regulatory Evaluation - Guidance for reporting to the criterion of 10 CFR 50.73(a)(2)(ii) is provided in section 3.2.4 of NUREG 1022, "Event Reporting Guidelines 10 CFR50.72 and 50.73." This guidance states that an LER is required for a seriously degraded principal safety barrier or an unanalyzed condition that significantly degrades plant safety. Operating License Condition 2.C(5)(a) states the following: The Operating Corporation shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPS Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, and as approved in the SER through Supplement 5, subject to provisions b & c below. Conclusion: - Based upon the information provided, the Halon suppression system would have operated to extinguish a fire. This condition is not considered reportable to the requirements of 10 CFR 50.72(b)(3)(ii)(B), 10 CFR 50.73(a)(2)(ii), nor is it a violation of the Operating License Condition 2.C(5)(a). Consistent with this conclusion, ENS notification number 41327 for this event is to be retracted. Notified R4DO (Whitten). NRC Resident Inspector will be notified. | ||
| ENS 41317 | 7 January 2005 23:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Loss of a Fire Safe Shutdown Success Path | On January 07, 2005, conditions were discovered where a postulated fire could cause the loss of a fire safe shut down success path. Wolf Creek is in Mode 1, at 100 % power. During reviews associated with post fire safe shutdown reanalysis work, Wolf Creek personnel discovered that power and control cables for Boron Injection tank (BIT) inlet valve EMHV8803A could be damaged by a fire in Fire area A-1. This valve is needed because the A train is the protected train for this fire area. This does not meet our commitments to 10 CFR50 Appendix R.III.G as reflected in our approved Fire Protection Plan. Following a fire in the plant that requires a plant shutdown the function of the CCP and Bit Inlet valve is to inject borated water into the reactor to maintain reactor water inventory. A fire in this fire area (A-1) has the potential to cause the above-mentioned valve to not function properly and cause a loss of the capability to maintain inventory. Based on the guidance provided in NUREG 1022, Revision 2, this situation meets the criterion of 10 CFR 50.72(b)(ii)(B) for an 8 hour ENS notification, as it relates to being in an unanalyzed condition. The licensee has implemented a 1-hour fire watch in this fire area. The licensee has notified the NRC Resident Inspector. |