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The following information was provided by the licensee via email: While operating at 100 percent reactor power, the Control Room received indications of a feedwater transient, and indications of decreasing level on Steam Generator `B.' Reactor Trip occurred approximately 30 seconds after initial indications of transient at 1803 CDT on 7/18/22. All Safety Related Equipment responded as expected, including actuation of Auxiliary Feedwater. Control Room responded properly and progressed through Emergency Operating Procedures. The Unit is Stable in Mode 3. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The plant is in a normal post-trip electrical line-up. Wolf Creek intends to make a press release.

• * * UPDATE FROM JOSHUA TURNER TO DONALD NORWOOD AT 1538 EDT ON 7/19/2022 * * *

The original event notification inadvertently indicated that a media / press release would be provided. However, no media / press release is planned. Notified R4DO (Gaddy).

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.

• * * UPDATE ON 9/12/13 AT 0215 EDT FROM MARCY BLOW TO HUFFMAN * * *

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.

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.

• * * UPDATE FROM MARCY BLOW TO DONALD NORWOOD AT 1923 EDT ON 6/26/2011 * * *

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).

Following a scheduled plant shutdown for refueling the operators were forced to close the Main Steam Isolation Valves (MSIV's) to limit plant cooldown. While opening MSIV's to restore steam to the secondary, a Reactor Trip and Safety Injection (SI) occurred. The MSIV bypass valves were opened to equalize pressure across the MSIV's. Steam header pressure dipped when the MSIV for 'C' Steam Generator (S/G) was opened. The low steamline pressure bistables are rate sensitive and actuated to cause the SI when steam pressure dipped. Lowest steamline pressure was 1040 psig, the low steam line pressure SI actuates at 615 psig. During the SI the PZR (Pressurizer) PORV's cycled approximately 10 times to limit RCS pressure. When the PORV's opened the 'B' PZR Code Safety Main Control Board (MCB) and plant computer alarm actuated but the actual MCB indication did not change nor does plant response indicate that a PZR Code Safety opened. This appears to be an indication problem related to the PORV's cycling. All equipment functioned as required. The station electric buses are aligned to normal offsite power. Decay heat removal is being discharged to the atmospheric relief valves with no indication of primary to secondary leakage. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM MARCY BLOW TO DONALD NORWOOD AT 1221 EDT ON 3/19/2011 * * *

1. The expected system actuations that occurred when the plant experienced a Safety Injection (SI) 03/19/11 at 04:04 CDT, previously reported on EN 46685 for 10CFR50.72(b)(2)(iv)(A), is also reportable under 10CFR50.72(b)(3)(iv)(A) for Specified System Actuation.

2. During the recovery of the Safety Injection (SI) actuation that occurred 03/19/11 at 04:04 CDT and previously reported on EN 46685, the Safety Injection Signal was reset which blocked any further automatic actuation. This was directed per the appropriate procedure step. There is no Technical Specifications allowed condition for both trains of ECCS to be inoperable, therefore the unit entered Tech. Spec. 3.0.3 due to the Auto SI feature being blocked. LCO 3.5.2 action C.1. directs immediate entry into LCO 3.0.3. The entry into TS 3.0.3 was made at 0411 CDT and exited at 0639 CDT when the Reactor Trip Breakers were reclosed which re-enabled the automatic SI signal. This is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. NRC Resident was notified of the update. Notified R4DO(Cain).

While operating in Mode 1, 100% rated thermal power (RTP), Wolf Creek received a Containment Purge Isolation Signal (CPIS) caused by Containment Purge Exhaust Radiation Monitor GT RE-33 exceeding the high radiation trip setpoint. There was no containment purge in progress at the time of the CPIS so no containment dampers actuated or were required to actuate. Control Room Ventilation Isolation Signal (CRVIS) was also received, as expected, from the actuation of the CPIS. All CRVIS components actuated as required. Review of GT RE-33 identified that the radiation monitor spiked high causing the CPIS then returned to normal values. All other containment radiation monitors are indicating normal values. The NRC Resident has been notified of this event by the Licensee. GT RE-33 is currently removed from service.

• * * UPDATE AT 1630 EDT ON 06/21/10 FROM RICK HUBBARD TO S. SANDIN * * *

The licensee is retracting this event based on the following: The Containment Purge Exhaust Radiation Monitor GT RE-33 failed due to a corrupted database initiated by a Radiation Monitor System (RMS) communication loop problem. The signal was not initiated by actual plant conditions or parameters, so this was an invalid actuation. This actuation did not involve a critical scram. So it is not reportable per 10 CFR 50.72. The licensee will inform the NRC Resident Inspector. Notified R4DO (Pick).

A Turbine trip and Feedwater Isolation Signal was generated as a result of high level in 'A' Steam Generator. At approximately 0007 CST, a turbine trip and Feedwater 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 0008 CST. Steam Generator 'A' high level resulted from swell of the Steam Generator after opening the 'A' Main Steam Isolation Valve. Level was approximately 50% prior to the event and returned to approximately 50% within one minute. The turbine was already in the tripped condition due to the plant trip that occurred at 1458 CST on 03/02/2010 (Reference EN# 45739). Steam Generators are being fed with the motor driven startup Feedwater pump. Decay heat removal is via the Steam Generator Atmospheric Relief Valves and steam dumps. The plant is in Mode 4 with Reactor Coolant System pressure at approximately 500 psig and temperature at approximately 332 degrees F. Two Main Feedwater isolation valves did not fully close but the actuators were being supplied with auxiliary medium (air) at the time of the actuation. Feedwater isolation valves are not required to be operable in Mode 4. The licensee will notify the NRC Resident Inspector.

* * * RETRACTION FROM RICK HUBBARD TO PETE SNYDER AT 1456 ON 4/29/2010 * * * 

This event (Event Number 45745) is being retracted. A valid actuation of any systems listed in paragraph 50.72 (b)(3)(iv)(B) did not occur. The plant was in Mode 4 and neither a reactor trip nor an auxiliary feedwater system actuation signal occurred. Additional evaluation determined that the turbine trip and feedwater isolation signal are a function of the ESFAS (Engineered Safety Features Actuation System) instrumentation and does not result in an actuation of the RPS (Reactor Protection System) or other systems listed in paragraph 50.72 (b)(3)(iv)(B). The licensee notified the NRC Resident Inspector. Notified R4DO(Clark).

At 1010 CDT on 8-22-2004, during the performance of STS IC-211B, 'ACTUATION LOGIC TEST TRAIN B SOLID STATE PROTECTION SYSTEM,' Control Room operators received a reactor trip concurrent with power range flux lo (low) setpoint trip and intermediate range hi (high) flux reactor trip alarms. All safety-related equipment operated as required. The Steam Dump valves, which function to remove excess heat as the secondary systems shutdown, did not initially operate as expected, resulting in the operation of the Steam Generator Atmospheric Relief Valves (ARVs) to control Reactor Coolant pressure for approximately three minutes following the reactor trip, at which time the Steam Dump valves responded as expected and the ARVs closed. The Steam Dumps are currently operating correctly in automatic. 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 causes of the reactor trip and formulate the repair/ restart plan. A press release is planned. During the reactor trip, all control rods fully inserted. The electric plant is in a stable shutdown plant lineup. Plant personnel also intend to investigate the unusual response of the steam dump valves. The licensee has notified the NRC Resident Inspector.

• * * Update on 08/23/04 at 1635 EDT by Steven Gifford taken by MacKinnon * * *

Upon data review related to the operation of the Steam Dump valves and the Atmospheric Relief Valves (ARVs), plant staff has determined that the ARVs and Steam Dump valves operated as expected. The Steam Dump valves opened immediately following the reactor trip, as designed, due to the difference between their setpoint and the Reactor Coolant System (RCS) temperature at the time of the reactor trip. The Steam Dump valves remained open for approximately 20 to 30 seconds, at which time the RCS temperature had decreased to a value within the control range for the Steam Dump valves, causing them to close. The opening of the ARVs was in response, as designed, to the increase in Steam Generator pressure that resulted from the turbine trip shortly after the reactor trip."

NRC R4DO (Tom Farnholtz) notified.

The NRC Resident Inspector was notified of this update by the licensee