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| Entered date | Event description | |
|---|---|---|
| ENS 52218 | 2 September 2016 11:29:00 | While operating in MODE 1 at 100 percent rated thermal power and placing Excess Letdown in service for Reactor Coolant System (RCS) leak detection, RCS operational leakage exceeded 1 gpm (gallon per minute) unidentified leakage as identified by performing RCS Water Inventory Balance using the Nuclear Plant Information System Computer. This required the entry into Technical Specification (TS) Limiting Condition of Operation (LCO) 3.4.13 Condition B at 0808 (CDT) on 9/2/16. The associated action is to place the unit into Mode 3 in 6 hours. Trending of containment sump level indicates the leakage is inside containment with the exact location within containment unknown. Containment inspection is being performed to try and identify the source of Reactor Coolant System leakage. NRC Resident Inspector has been notified. Re-alignment of the Letdown System back to its normal arrangement has subsequently reduced RCS leak rate to 0.521 gpm at 0652 CDT on 9/2/16. Unusual or Not Understood - Leak Location is not known at this time. Maximum leak rate recorded was 1.358 gpm. The leak was first discovered at 08/31/16 at 1519 CDT. Safety Related Equipment not operational - Reactor Vessel Level Indicating System (TS 3.3.3).
Wolf Creek Nuclear Operating Corporation is retracting the 10 CFR 50.72(b)(2)(i) notification based on subsequent review of the event. The calculation of unidentified leak rate which triggered entry into the Mode 3 Required Action Statement was performed immediately after placing RCS Excess Letdown in service. An evaluation of the leak rate calculation determined that the leak rate was invalid due to performance of the RCS water inventory balance during non-steady state operating conditions. This was contrary to the requirements of TS Surveillance Requirement 3.4.13.1, as this test was performed while charging and letdown flows were being stabilized following the alignment of excess letdown. A walk down of the Excess Letdown system while in-service determined no leakage. Subsequent RCS water inventory balances performed with Excess Letdown in service under steady state operating conditions while in Mode 3 at normal operating pressure and temperature determined the maximum calculated unidentified leak rate was 0.675 gpm. After the plant entered Mode 3 a non-RCS pressure boundary leak was identified during equipment walk downs on a seal weld from the reactor vessel head core exit thermocouple nozzle assembly 77. The leakage did not impact the ability to shut down the unit. No TS limits were exceeded during this event. Therefore, the plant shutdown to investigate and correct leakage past the seal weld of a threaded connection does not meet the reporting requirements of 10 CFR 50.72. The NRC Resident Inspector has been notified. Notified the R4DO (Kramer). |
| ENS 52184 | 16 August 2016 15:43:00 | On 8/16/2016, while operating at 100% in Mode 1, routine testing of the off-site sirens for Wolf Creek was scheduled. The county dispatcher was unable to actuate any sirens. The dispatch supervisor was contacted to ensure there was not a personnel qualification issue. Both the dispatcher and the supervisor were unable to actuate any sirens, either manually or using the normal computer controls. Coffey county personnel, assisted by Wolf Creek personnel, determined that a battery had failed causing a fuse to blow and de-energizing the equipment needed to actuate the sirens. The battery has been replaced, the fuse has been replaced, and the system has been tested satisfactorily. The emergency sirens were restored to service at 12:20 CDT 8/16/2016. NRC resident has been notified. No plant systems were affected by this failure of notification equipment. The plant remained at 100% power Mode 1 throughout this event. |
| ENS 51036 | 3 May 2015 12:56:00 | 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 50468 | 18 September 2014 14:05:00 | 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. |