|Entered date||Site||Scram||Region||Reactor type||Event description|
|ENS 51036||3 May 2015 12:56:00||Wolf Creek||Manual Scram|
|NRC Region 4||Westinghouse PWR 4-Loop||On 5/3/2015 during power ascension following Refueling Outage 20, Steam Generator 'C' water level increased rapidly, causing a Feedwater isolation on high Steam Generator water level and an associated Turbine trip. The reactor was subsequently manually tripped. At the start of the event, reactor power was approximately 22%. Plant staff was in the process of transferring from Main Feedwater Bypass Feed Regulating Valve control, used for low power control, to Main Feedwater Regulating Valve control as part of power ascension. When the Main Feedwater Regulating Valve for 'C' Steam Generator (AEFCV-530) was opened, it went to about 80% open, causing an overfeed of the 'C' Steam Generator. High Steam Generator water level in 'C' Steam Generator initiated an automatic Feedwater Isolation Signal, automatic Turbine Trip and automatic trip of the operating main feed pump. The operating crew initiated a manual reactor trip. The Auxiliary Feedwater System automatically initiated as part of the plant response to the feedwater system transient. The plant is presently stable in Mode 3. All equipment functioned normally, except the 'C' Main Feedwater Regulating Valve (AEFCV0530) which did not function to properly control Steam Generator level. This valve did function as designed to close on the Feedwater Isolation Signal. NRC Resident Inspector has been contacted.|
|ENS 50855||28 February 2015 05:24:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop|
Wolf Creek Generating Station performed a planned shutdown for the start of a refueling outage. As part of the procedure GEN 00-005, Minimum Load to Hot Standby, a planned manual reactor trip was initiated from 25 percent power level with the plant in Mode 1. As part of this planned shut down sequence, an anticipated automatic Auxiliary Feedwater (AFW) actuation signal was generated. This is a non-emergency event notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to preplanned manual actuation of the reactor protection system and auto-initiation of auxiliary feed water system. All systems and components operated as designed with the exception of the main generator output breakers. They did not open as designed. They were manually opened using the main control room hand switches as directed by EMG ES-02, Reactor Trip Response. The plant is stable in Mode 3 with AFW secured, with plans to cool down and enter Mode 5 for the planned refueling outage. The NRC Resident Inspector has been notified.
WCNOC (Wolf Creek Nuclear Operating Corporation) is retracting the 10 CFR 50.72(b)(3)(iv)(A) notification based on further review of the event. A valid AFAS (Auxiliary Feedwater Actuation signal) actuation occurred during the planned shutdown as a result of SG (Steam Generator) water level reaching the lo lo level setpoint. The AFAS actuation was an expected actuation that occurred due to preplanned activities covered by GEN 00-005. On-shift control room personnel were aware that an AFAS actuation would occur as a result of tripping the plant between 30% and 25% power. The AFAS actuated consistent with the planned shutdown with no anomalies. This is consistent with NUREG-1022, Rev. 3, Section 3.2.6 that states, in part: 'With regard to preplanned actuations, operation of a system as part of a planned test or operational evolution need not be reported. Preplanned actuations are those that are expected to actually occur due to preplanned activities covered by procedures. Such actuations are those for which a procedural step or other appropriate documentation indicates that the specific actuation is actually expected to occur. Control room personnel are aware of the specific signal generation before its occurrence or indication in the control room.' The NRC Resident Inspector was notified. Notified R4DO(Okeefe).
|ENS 49339||11 September 2013 20:37:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop|
Wolf Creek has commenced a plant shutdown in accordance with Technical Specifications. The A Train Class 1E Electrical Equipment Air Conditioning unit was declared non-functional due to a possible failed compressor cylinder, as indicated by increased vibration. This failure could prevent the unit from performing its required function over its required mission time, as required by Technical Specifications 3.8.4, 3.8.7, and 3.8.9. The following safety related electrical equipment was declared inoperable: 4.16KV Bus NB01; 480 Volt AC buses NG01 and NG03; 120 Volt Instrument AC inverters and buses NN11, NN13, NN01 and NN03; 125 VDC chargers and buses NK11, NK13, NN01 and NN03. Technical Specification 3.0.3 was entered at 1645 CDT on 9/11/2013 from Technical Specification 3.8.7 due to two out of four 120 VAC inverters (NN11 and NN13) being inoperable. Plant shutdown to Mode 5 commenced at 1731 CDT. The unit is currently at approximately 50% power. All electrical systems listed above remain available but are declared inoperable due to inadequate room cooling capability. No major equipment is out of service. The NRC Resident Inspector has been notified. No switchgear room temperature limits were challenged. See EN #49008 (May 6, 2013) and EN #49126 (June 17, 2013) for similar events.
At 00:36 CDT 9/12/13, Wolf Creek had an Auxiliary Feedwater Actuation during a plant shutdown in accordance with Technical Specifications. The plant was in Mode 3, all control rods inserted, with reactor trip breakers closed when low steam generator levels prompted a manual reactor trip. A Valid Auxiliary Feed Actuation signal was received due to low steam generator levels. All Auxiliary Feedwater pumps started and operated as expected. The licensee has informed the NRC Resident Inspector. R4DO (Hay) notified.
|ENS 46990||26 June 2011 18:00:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop|
6/26/11 at 1609 CDT, the reactor was manually tripped due to the trip of the 'B' Main Feed Pump while operating in Mode 1 at approximately 82% reactor power. The unit was increasing power to 95% after the current refuel outage. The cause of the trip of the 'B' Main Feed Pump is not known at this time. All equipment functioned normally as expected. The investigation into the cause of the 'B' Main Feed Pump trip is ongoing at this time. Current plant status is Mode 3. The NRC Senior Resident has been contacted. All rods fully inserted upon reactor trip. The unit is stable with Auxiliary Feedwater supplying the Steam Generators. Decay heat is being removed to the Main Condenser via steam dumps. The electrical system is in a normal post-trip alignment. The licensee characterized the reactor trip as uncomplicated.
A valid Auxiliary Feed(water) actuation signal (occurred) due to trip of both of the Main Feed pumps from a turbine trip and low steam generator levels. This is reportable under 10CFR50.72 (b)(3)(iv)(A), 8-hour report. All auxiliary feed pumps started and operated as expected. The licensee will notify the NRC Resident Inspector. Notified R4DO (Deese).
|ENS 46338||17 October 2010 14:35:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop||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 45749||8 March 2010 05:46:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop||Wolf Creek experienced an 'A' Main Feedwater Pump (MFP) trip at 0332 CST during a plant startup. Reactor power was 42% at the time of the manual reactor trip. The loss of the 'A' MFP resulted in no feedwater flow to the Steam Generators and a manual reactor trip was ordered due to the impending reactor trip on low-low Steam Generator level. Post-trip decay heat was being removed by the Condenser and Auxiliary Feedwater System. Reactor Coolant System (RCS) temperature dropped below 550 degrees Fahrenheit, which required emergency boration to ensure shutdown margin was maintained. Emergency boration was secured at 0404 CST when RCS temperature was raised to greater than 550 degrees Fahrenheit. The lowest RCS temperature reached following the reactor trip was 544 degrees Fahrenheit. The cause of the 'A' Main Feedwater Pump trip is not known at this time. Primary to secondary leakage is less than 0.224 gallons per day. All systems functioned as designed. The plant is being maintained at normal Mode 3 pressure and temperature. The licensee has notified the Senior NRC Resident Inspector. All control rods fully inserted. The plant is in a normal post-trip electrical line-up.|
|ENS 45739||2 March 2010 18:48:00||Wolf Creek||Automatic Scram||NRC Region 4||Westinghouse PWR 4-Loop||Wolf Creek experienced a reactor trip at 1458 CST. The first out annunciator was Steam Generator Level Lo Lo Reactor Trip. The trip was caused by the loss of the 'A' Main Feed Pump. The cause of the loss of the feed pump was due to the loss of 120 VAC non-safety instrument inverter PN09. PN09 supplies the Main Feed Pump Speed Control Circuitry. The loss of the PN09 also resulted in the loss of the ability to dump steam to the main condenser. Initial post trip decay heat was being removed with the Steam Generator Atmospheric Relief Valves and Auxiliary Feed Water. The Atmospheric Relief Valves cycled from approximately 1458 CST until approximately 1504 CST. Primary to Secondary leakage is less than 2.68 gallons per day. PN09 was re-energized at 1554 CST. All systems functioned as designed with the exception of the instrumentation powered by PN09. At the time of the trip the 'A' Emergency Diesel Generator and the 'A' Class IE Air Conditioning Unit were out of service for maintenance. The plant is being maintained at normal Mode 3 pressure and temperature. The reactor trip was uncomplicated and decay heat is currently being removed by steam dumps to the main condensers. The licensee has notified the NRC Senior Resident Inspector.|
|ENS 45278||19 August 2009 20:38:00||Wolf Creek||Automatic Scram||NRC Region 4||Westinghouse PWR 4-Loop|
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 45027||28 April 2009 17:41:00||Wolf Creek||Automatic Scram||NRC Region 4||Westinghouse PWR 4-Loop||Wolf Creek experienced a reactor trip at 1527 CDT. The RX trip resulted from low level in 'B' S/G. (The) 'B' Main Feedwater Regulating Valve (FRV) AE FCV-520 closed causing a loss of feedwater to the 'B' S/G. Initial investigation indicates that the FRV controllers lost power due to blown primary and backup fuses associated with its card rack. One intermediate range instrument (SE NI-36) stuck at 10E-10 amps requiring manual energization of source range instruments. Aux Feedwater and Feedwater Isolation actuations occurred as expected. All control rods inserted into the core during the trip. There were no relief or safety valves actuated during the transient. The electrical plant is in the normal shutdown alignment with offsite power supply emergency busses. Decay heat is being removed via the steam dumps to condenser. The licensee is investigating the cause of the blown fuses. The licensee has notified the NRC Resident Inspector.|
|ENS 44072||17 March 2008 16:30:00||Wolf Creek||Manual Scram||NRC Region 4||Westinghouse PWR 4-Loop||While operating at 100% rated thermal power in Mode 1, a Manual Reactor Trip was initiated due to the lowering of Steam Generator (S/G) level due to the loss of 'B' Main Feed Water Pump. Initial investigation indicates that the loss of XPB03 transformer caused the loss of Non-Class 1E 4160VAC PB03 and PB04. At the time of the event, bus PB04 was cross-tied with bus PB03 for scheduled maintenance of transformer XPB04. This caused loss of all condensate and heater drain pumps. A manual reactor trip was actuated in anticipation of an automatic reactor trip. Aux Feed auto actuation did occur as required. The Non-Safety Related Charging pump was lost due to the loss of PB03 and charging flow was re-established to the Reactor Coolant System by starting 'A' Charging pump. All other plant equipment functioned as required. The plant is currently stable in Mode 3 at 560 degrees F and 2235 psig. Continuing to investigate. At 15:03 CDT, plant experienced an auto Feed Water Isolation signal due to Low-Low S/G level. Feed Water Isolation had already occurred with the initial event but the signal had been previously been reset. Manual feed water flow control has been established from Aux Feed Water. All control rods inserted into the core during the trip. Decay heat is being removed via the steam dumps to condenser using AFW to feed the steam generators. No primary or secondary relief valves lifted during the transient. The electrical grid is stable and supplying plant safety loads via the normal path. NRC Resident has been notified.|
|ENS 41100||7 October 2004 15:03:00||Wolf Creek||Automatic Scram||NRC Region 4||Westinghouse PWR 4-Loop||At 1148 CDT on 10-7-2004, Wolf Creek Generating Station experienced on automatic reactor trip as a result of a turbine trip. Initial indications are that the trip was associated with a lightning strike in the switchyard. All control rods fully inserted. The 'A' train of the Residual Heat Removal system and the 'A' train Class IE Safety Related Switchgear Air Conditioning Unit were out of service for scheduled maintenance at the time of this event. All other safety related equipment operated as expected. The Auxiliary Feedwater system actuated as designed. The plant is currently stable in Mode 3 at Normal Operating Temperature and Pressure while plant personnel investigate the cause of the reactor trip and formulate the repair and restart plan. The NRC Senior Resident inspector has been informed. A news statement relative to this event is planned. Secondary atmospheric relief valves lifted and reseated as a result of the trip, however there was no release to the environment. Decay heat is being removed via the steam dumps. The electrical grid is stable.|
|ENS 40086||18 August 2003 23:36:00||Wolf Creek||Automatic Scram||NRC Region 4||Westinghouse PWR 4-Loop||At 1554 CDT on 08/18/2003 all four Steam Generators alarmed with Steam Flow/Feed Flow Mismatch followed by indication that there was no feedwater flow to the "B" Steam Generator. The reactor tripped approximately 20 seconds later on Steam Generator Low-Low Level, 23.5% Narrow Range, as required. A review of the plant computer information shows that event-initiating cause may be the unexplained closure of the "B" Feedwater Isolation Valve. All plant safety related systems operated as required. The Steam Dumps, which function to remove excess heat as the secondary systems shutdown, exhibited control problems when they were shifted from temperature control mode to steam pressure mode, as required by the emergency operating procedures. The Steam Dumps are currently operating and controlling in manual mode. When the reactor tripped all four Steam Generator Atmospheric Relief Valves opened, a normal response for a trip from full power. The "A" Steam Generator Atmospheric Relief Valve was slow to close in automatic. The valve was closed in manual and returned to automatic and is controlling properly. The plant is currently stable in Mode 3 at NOP and NOT while plant personnel investigate the causes of the trip and formulate the repair/restart plan. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. The electrical grid is stable. The NRC Resident Inspector is in the Control Room.|