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| Entered date | Event description | |
|---|---|---|
| ENS 46338 | 17 October 2010 14:35:00 | At 0953 CDT, Wolf Creek experienced a reactor trip due to low steam generator level. At the time of the trip the plant was in Mode 1 approximately 16% power following a forced outage. A feedwater isolation signal (FWIS) was generated due to high S/G level (P-14) in 'B' S/G (steam generator). The FWIS resulted in a low S/G level. Although a manual reactor trip was ordered by the duty Shift Manager, the manual trip signal was not inserted before the reactor automatically tripped on low steam generator level. Auxiliary Feed Water systems actuated as expected due to low steam generator levels. The plant is presently in Mode 3 at normal operating temperature and pressure. The Senior Resident Inspector has been notified. All control rods fully inserted during the trip. All three Auxiliary Feedwater Pumps started to maintain S/G levels. The plant was stabilized with the motor driven startup feedwater pump maintaining S/G level. Decay heat is being removed using the atmospheric steam dumps. There is no primary to secondary leakage. The plant is in its normal shutdown electrical lineup. A press release will be issued. |
| ENS 45453 | 21 October 2009 21:29:00 | At approximately 1755 CDT on 10/21/09, control room personnel were notified of a potentially contaminated individual suffering from a personal medical emergency inside the containment building. The individual was treated by the onsite first responders who administered CPR and AED (Automatic External Defibrillator). At 1816 (CDT), Coffey County medical personnel arrived on site. The individual was transported via ambulance to the Coffey County Hospital where the individual was pronounced dead at approximately 1900 (CDT) from an apparent cardiac arrest. A WCNOC (Wolf Creek Nuclear Operating Corp.) health physics technician accompanied the individual to the hospital and determined that the individual was not contaminated. The individual was a supplemental employee supporting refueling activities. OSHA will be notified of the work place fatality. State and Local Government Agencies will be notified. No news release is planned. The licensee notified the NRC Resident Inspector. |
| ENS 45281 | 21 August 2009 11:22:00 | Turbine trip and Feed (Water) Isolation Signal generated as result of high level in 'A' Steam Generator. At approximately 0648 CDT a turbine trip and Feed Water Isolation Signal was generated when the level in the 'A' Steam Generator exceeded the initiation set point of 78%. Steam Generator level returned below the initiation setpoint at approximately 0656 CDT. The turbine was already in the tripped condition and Feed Water Isolation Valves closed due to the plant trip that occurred at 1549 CDT on 8/19/2009 (Reference EN# 45278). Steam Generators are being fed with the motor driven Auxiliary Feed Water Pumps. Decay heat removal is via the Steam Generator Atmospheric Relief Valves. The plant is in Mode 3 with Reactor Coolant System pressure at approximately 2235 psig and temperature at approximately 561 F. The licensee has notified the NRC Resident Inspector. |
| ENS 45278 | 19 August 2009 20:38:00 | Wolf Creek experienced a reactor trip at 1549 CDT. The first out annunciator was TURBINE TRIP and P9 Reactor TRIP. At approximately the same time the unit experienced a momentary loss of offsite power. The emergency diesel generators started (and loaded) as expected to supply power to the safety busses due to the loss of offsite power. Auxiliary Feedwater and Feedwater Isolation actuations occurred as expected. All control rods inserted into the core during the trip. All Reactor Coolant Pumps tripped due to the loss of offsite power. Decay heat was initially being removed by the Steam Generator Atmospheric Relief Valves. Presently the plant is stable in Mode 3. The 'D' Reactor Coolant Pump has been restarted. The licensee is continuing to investigate the cause of the trip. The atmospheric relief valves lifted for approximately 10 minutes, however there was no primary to secondary leakage. Both A & B EDG's loaded for about 2 minutes. At the present time the electrical lineup is normal and the EDG's are shutdown. Plant is at Normal Operating Pressure and just below Normal Operating Temperature. Decay heat and S/G levels are being maintained with the Auxiliary Feedwater pumps. The licensee has notified the NRC Resident Inspector. The Licensee may issue a press release on this event.
The atmospheric relief valves lifted for approximately 2 minutes, not 10 minutes as stated above. Also the A & B EDG's did not load for only 2 minutes. Actually "the B Safety Bus was paralleled to its normal off site source and the B Emergency Diesel was realigned for auto-start at 1740. The A Safety Bus was paralleled to its normal off site source and the A Emergency Diesel was realigned for auto-start at 1844. Notified the R4DO (Jones).
The initial report stated that there was no primary to secondary leakage. Actual primary to secondary leakage as measured on 8/14/2009 was a value of less than 0.722 gallons per day. The licensee notified the NRC Resident Inspector. Notified the R4DO (Miller) |
| ENS 45277 | 19 August 2009 20:33:00 | 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 42327 | 10 February 2006 16:06:00 | 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. |
| ENS 42328 | 10 February 2006 16:13:00 | 10 CFR26.24 describes the chemical and drug testing programs for persons subject to this part. 10 CFR26.24(2) states that 'Random testing must be conducted at an annual rate equal to at least 50 percent of the workforce.' On 2/10/06 at 0830, Wolf Creek determined that the random testing rate was 49.41 percent for calendar year 2005, which is below 50 percent. 10 CFR26.73 states that notification to the NRC Operations Center must be made by telephone within 24 hours. The licensee notified the NRC Resident Inspector.
This event is being retracted. A manual verification of the 2005 data was conducted. It confirmed that the random test rate for 2005 exceeded 50%, therefore this event is not reportable per 10 CFR 26.73. The licensee notified the NRC Resident Inspector. Notified the R4DO (J. Whitten) |
| ENS 42004 | 21 September 2005 02:37:00 | At 2240 09/20/2005 an engineering evaluation concluded that the Residual Heat Removal Containment Sump Isolation Valves were inoperable based on their inability to open against a higher differential pressure than had been previously evaluated. Historically Wolf Creek has operated with the Component Cooling Water isolated to the Residual Heat Removal Heat Exchangers. If the pumps started during accident conditions there is a potential that the pump heat generated could cause system pressure to increase to the suction relief valve setting of 450 psig. Testing of the valves has shown that they are capable of opening with 207 psid across the valve. Technical Specification 3.0.3 was entered at 2240 9/20/2005 due to the inoperability of both Residual Heat Removal trains. Power reduction was commenced at 2340. Component Cooling Water was aligned to the Residual Heat Removal Heat Exchangers at 0004 9/21/2005 restoring operability of the Residual Heat Removal trains. At 0017 9/21/2005 actions were commenced to restore power to 100%. This event is reportable under 10 CFR50.72(b)(2)(i) Technical Specification required shut down (4 hour) and 10 CFR50.72(b)(3)(v)(B) Event or Condition that could have prevented fulfillment of a Safety Function (8 hour). The NRC Resident has been contacted.
Evaluation of the Residual Heat Removal Containment Sump Isolation Valves concluded that the valves were operable and the Residual Heat Removal trains were capable of fulfilling their safety function. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(B). The licensee notified the NRC Resident Inspector. Notified the R4DO (Whitten). |
| ENS 41825 | 7 July 2005 15:27:00 | On July 07, 2005 at approximately 1105 (hrs. CDT) a report was received in the Wolf Creek Generating Station Control Room of a non-work related medical emergency involving an employee. Onsite medical personnel responded and commenced first aid including CPR. An ambulance was dispatched and the individual was transported offsite to a local hospital. Hospital personnel notified Wolf Creek personnel at 1152 that the individual had passed away. At this time no press release is planned. OSHA has been informed of the fatality. The licensee notified the NRC Senior Resident Inspector. |
| ENS 40596 | 17 March 2004 21:25:00 | At 1457 on 3-17-2004, Wolf Creek Generating Station Fire Protection Engineering submitted information to the Control Room that there was a flooding analysis concern in the Engineered Safety Features Switchgear (NB) Rooms regarding the 2 � inch fire protection header that is located in the rooms. The capability of the floor drain system in the room is indeterminate in regards to draining all the water that could be expected to accumulate in the rooms if the fire main were to rupture during a seismic event. Compensatory actions have been put in place in the form of additional fire suppression equipment and administrative isolation of the fire protection header that is located in the NB switchgear rooms. Investigation continues into the assumptions used in the flooding and seismic analysis. The Senior Resident Inspector has been notified.
On March 17, 2004, at 2125 EST, Wolf Creek Nuclear Operating Corporation (WCNOC) made an event notification regarding a flooding analysis concern related to a 2 1/2 Inch fire protection header located in the Engineered Safety Features Switchgear (NB) Rooms (EN40596). This notification identified that the analysis with regards to draining all the water from the room in the event that the fire protection header pipe were to rupture during a seismic event was indeterminate. Subsequent evaluation has determined that the fire protection header piping located in the subject NB switchgear rooms is built to seismic Il/I requirements, and is therefore not subject to a pipe rupture in the event of a seismic event. Therefore, WCNOC is retracting the March 17, 2004 event notification. The NRC Senior Resident Inspector has been notified. Notified R4DO (L. Smith). |