|Report date||Site||Event description|
|05000482/LER-2017-003||2 November 2017||Wolf Creek|
On September 7, 2017, Wolf Creek Generating Station (WCGS) was in Mode 1 at 100 percent power. During evaluation of protection for safety-related equipment from the damaging effects of tornados, WCGS personnel determined that the non safety-related exhaust lines from safety-related atmospheric relief valves (ARVs) and main steam safety valves (MSSVs) could be crimped by tornado generated missiles. If these are crimped completely, these components may be unable to perform their safety functions. The ARVs and MSSVs were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado- Generated Missile Protection Noncompliance," Revision 1 was applied. Immediate compensatory measures consistent with EGM 15-002 were implemented within the time allowed by the applicable Technical Specification Limiting Condition(s) for Operation. The ARVs and MSSVs were subsequently declared operable but nonconforming. These tornado missile vulnerabilities existed since the original plant construction. Actions will be taken to establish compliance for these components either by a plant modification or employing a methodology for addressing tornado missile non-conformances.
On April 5, 2017, WCGS personnel provided a 10 CFR 50.72 notification in Event Notification (EN) 52666 concerning tornado missile protection issues known at that time. As stated in EGM 15-002, the NRC will exercise enforcement discretion for subsequent tornado missile 10 CFR 50.72 notifications. Therefore, no 10 CFR 50.72 notification was made for this condition.
|05000482/LER-2017-002||31 May 2017||Wolf Creek|
On April 5, 2017, Wolf Creek Generating Station (WCGS) was in Mode 1 at 100 percent power. During evaluation of protection for safety-related equipment from the damaging effects of tornados, WCGS personnel determined that safety-related fuel oil transfer lines inside the Diesel Generator Building could be damaged if tornado generated missiles struck the non-safety related truck connections to these transfer lines. As these safety-related transfer lines are required to supply fuel to the Emergency Diesel Generators (EDGs) from the fuel oil storage tank, Operations declared both of the EDGs inoperable. On April 5, 2017, at 1632 Central Daylight Time, an 8- hour, non-emergency report to the Nuclear Regulatory Commission (NRC) (reference NRC Event Notification Number 52666) was made in accordance with 10 CFR 50.72. Compensatory measures were implemented consistent with Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance." The EDGs were then declared operable but non-conforming.
These tornado missile vulnerabilities existed since the original plant construction. Immediate compensatory measures included verification that the severe weather procedures were up-to-date, ensuring that Operations personnel were current on their training to these procedures, and implementing measures to heighten station awareness until the vulnerabilities had been corrected. These vulnerabilities have now been permanently eliminated.
|05000482/LER-2017-001||17 May 2017||Wolf Creek|
On March 22, 2017, with the plant in Mode 1, a work activity isolated the emergency makeup from the Essential Service Water (ESW) to the "A" Train Component Cooling Water (CCW). An individual questioned whether the "A" Train CCW was operable in this configuration. Technical Specification (TS) 3.7.7 requires two CCW trains to be operable in Modes 1 through 4.
It was determined that the "A" train CCW was inoperable when emergency makeup from ESW was isolated and that the allowed outage time was exceeded. The cause of the event was that adequate information was not incorporated into TS Bases 3.7.7, stating that emergency makeup from ESW is required to maintain a CCW train operable.
|05000482/LER-2016-002||16 January 2017||Wolf Creek|
On November 16, 2016, at approximately 2109 Central Standard Time, while in Mode 5, a fault occurred which isolated the East Switchyard Bus from the Train "A" emergency AC plant bus NB01. During Refueling Outage 21, a modification to Transformer #7 allowed the offsite power through Transformer #7 to NB01 to be fed from either the East or West Switchyard Busses through 2 different breakers. After the loss of the East Switchyard Bus, the second breaker unexpectedly tripped which resulted in a loss of offsite power to NB01. An undervoltage condition was detected on NB01, which caused the Train "A" emergency diesel generator to start and to power NB01 as designed. The apparent cause of this event was that wiring in the Transformer #7 primary differential protective relay was landed on the incorrect termination point.
The wiring error on the primary differential protective relay was corrected and its functionality was verified.
The secondary differential protective relay wiring was verified to be correct. The East Switchyard Bus, Transformer #7, and its differential relays were all restored to service.
|05000482/LER-2016-001||10 November 2016||Wolf Creek|
On October 6, 2014, at approximately 1326 Central Daylight Savings Time (CDT) during a scheduled 24-hour Run, the 'B' Emergency Diesel Generator (EDG) unexpectedly tripped and a fire was observed in the electrical cabinet (NE106). This resulted in an unplanned entry into a 72 hour shutdown Limiting Condition of Operation (LCO) and an ALERT emergency classification. The source of the fire was the Power Potential Transformer (PPT). On 1/28/16, a Hardware Failure Analysis concluded that the failure of the PPT was due to overloading which resulted from failure of a diode in the power rectifier of the EDG excitation system. Failure of the diode was induced by a governor actuator malfunction on June 9, 2014.
The PPT and associated cabling were replaced. All power diodes in each power rectifier were replaced. Further corrective actions are being tracked by Condition Report (CR) 88665.
|05000482/LER-2015-004, Incorrect Decision Results in Two Containment Isolation Valves being in a Condition Prohibited by Technical Specifications||14 September 2015||Wolf Creek|
On May 5, 2015, it was discovered that the motive force (air supply) was not removed for two containment shutdown purge valves as required by Technical Specification 3.6.3, "Containment Isolation Valves." The motive force was restored to allow the performance of procedure STS KJ-001A, "Integrated Diesel Generator and Safeguards Actuation Test - Train A," on April 26, 2015. After performance of procedure STS KJ-001A, the motive force was not removed for the two containment shutdown purge valves. The plant entered Mode 4 on April 28, 2015 at 0144 Central Daylight Time (CDT).
Upon discovery, the air supply valves for the two containment shutdown purge valves were locked closed, removing the motive force. The cause of the event was the decision to only track components listed in the manual Locked Component Log during plant start-up, allowed a mode change with components out of position.
The safety significance of the event was low as each penetration flow path had a redundant valve that was closed with the motive force removed.
|05000482/LER-2015-002, Two Control Room Air Conditioning Trains Inoperable Due to Failure to Meet Surveillance Requirement||26 August 2015||Wolf Creek|
From January 3, 2013, through August 13, 2013, the Conditions and Required Actions of limiting condition for operation (LCO) LCO 3.7.11, LCO 3.0.3 and LCO 3.0.4 were not met.
On April 16, 2015, an apparent cause evaluation on Condition Report (CR) 92274, "Application of TS SR 3.0.1," identified the potential that the acceptance criteria in procedures STS PE-010A/B, "Control Room NC System Flow Rate Verification (A/B) Train," may not have been met when the acceptance criteria was revised on January 3, 2013 from > 18,360 cfm and information determined that the prior performances of STS PE-010A/B did not meet the new acceptance criteria. Additionally, procedure STS PE-010B was not performed until July 8, 2013 and procedure STS PE- 010A was not performed until August 13, 2013.
The apparent cause of this event is the information in Operability Evaluation OE GK-12-017, which addressed a separate issue on the same equipment, enabled control room operators and engineering personnel to rationalize the assumption that the change to the acceptance criteria was bounded and did not impact the ability to meet SR 126.96.36.199.
|05000482/LER-2015-003, Manual Reactor Trip due to High Steam Generator Level Transient at Low Power||1 July 2015||Wolf Creek|
On May 3, 2015 at 1021 Central Daylight Time (CDT), Wolf Creek Generating Station (WCGS) was at approximately 25% power during startup from Refueling Outage 20. The 'C' steam generator received a Hi-Hi level signal while the Reactor Operator was manually transferring feedwater control from the Main Feedwater Regulating Valve (MFRV) bypass valves to the MFRVs. This resulted in a main turbine trip, feedwater isolation, and auxiliary feedwater actuation. A manual reactor trip was initiated at 1022 CDT. The root cause was the lack of a reliable and consistent standardization between the operating crews to control the transfer of the steam generator feedwater flow between the MFRV bypass valves and MFRVs and a sense of urgency by the control room operators.
The event is bounded by analyses as reported in the Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR) Section 15.2.7, "Loss of Normal Feedwater Flow.
|05000482/LER-2015-001, Personnel Error Causes Two Inoperable Residual Heat Removal Trains||25 March 2015||Wolf Creek|
On January 28, 2015, the nightshift operations crew implemented a clearance order to support planned maintenance on two Residual Heat Removal (RHR) System valves that included closing valves EJHV8716A and EJHV8809A. At 0534 hours on January 28, 2015, Condition C of Limiting Condition for Operation (LCO) 3.5.2 was entered upon determining that less than 100% of the Emergency Core Cooling System (ECCS) flow equivalent to a single OPERABLE ECCS train was available with valve EJHV8716A closed. Entry into LCO 3.0.3 in accordance with Required Action C.1 was made on two separate occasions. Action was taken to restore valves EJHV8716A and EJHV8809A to the open position and exit LCO 3.0.3.
The licensed operators involved with the preparation and implementation of the clearance orders did not recognize that current plant conditions could not support the proposed activity. The individuals involved with this event had their qualifications removed until remediation occurred. Red switch boxes have been placed on the control boards in the control room on the valves in procedure AP 26C-004, "Operability Determination and Functionality Assessment," Section A.16 that can cause an entry into LCO 3.0.3.
|05000482/LER-2013-010||17 December 2013||Wolf Creek|
On October 18, 2013 at 1141 Central Daylight Time (CDT), the Class 1E electrical equipment air conditioning unit, SGKO5A, was declared nonfunctional due to low lube oil pressure on the SGKO5A compressor. As a result, Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 was entered and a plant shutdown was commenced. Mode 3 was entered on October 18, 2013 at 1735 CDT.
The cause of the SGKO5A failure was a loss of lube oil pressure sensing to the pressure switch of the SGKO5A compressor. Contaminates in the system caused the loss of lube oil pressure sensing to the pressure switch. An inadequate flush and restoration of the system in May 2013 allowed contaminates to remain in the system.
SGKO5A was returned to a functional status on October 21, 2013 at 1915 CDT. Wolf Creek Generating Station returned to Mode 1 on October 27, 2013 at 2007 CDT.
|05000482/LER-2013-009||5 December 2013||Wolf Creek|
On October 9, 2013 during a review of external operating experience, Wolf Creek Nuclear Operating Corporation (WCNOC) determined that a fire in some areas containing direct current (DC) ammeter circuits could result in secondary fires outside the primary fire area. This condition could impact the ability to achieve safe shutdown. The cause of this deficiency is that the original plant design did not specify fuse protection of these DC shunt ammeter circuits. The postulated scenario is a fire that causes a short to ground on cables associated with the DC ammeters, concurrent with a short to ground on a safety related 125 VDC circuit on the negative side of the same battery source. Since the ammeter circuits are not overcurrent protected, they could overheat and ignite anywhere along their route. The negative side of the 125 VDC circuit is fuse protected, but the fuse may not clear prior to ammeter cable ignition if a high resistance short to ground exists.
The safety significance of this event is considered low. An hourly fire watch was established in the affected areas. A modification will be implemented to correct this deficiency.
|05000482/LER-2013-007||14 November 2013||Wolf Creek|
On June 17, 2013 at 1111 Central Daylight Time (CDT), the Class 1 E electrical equipment air conditioning unit, SGKO5A, was declared nonfunctional due to an analysis of an oil sample that showed elevated levels of aluminum. As a result, Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 was entered and a plant shutdown was commenced.
On June 17, 2013, enforcement discretion was sought to permit noncompliance with TS 3.8.4, TS 3.8.7, TS 3.8.9, as well as LCO 3.0.3, to permit additional time to complete repairs and restoration of SGKO5A before a plant shutdown would be required. An additional 168 hours was requested to restore SGKO5A to a functional status such that the Completion Time of LCO 3.0.3 would expire at 1111 CDT on June 24, 2013. The Nuclear Regulatory Commission granted approval of the requested enforcement discretion on June 17, 2013. The replacement of the SGKO5A compressor and testing to restore functionality was completed at 2220 CDT on June 21, 2013.
|05000482/LER-2013-008||12 November 2013||Wolf Creek|
On September 11, 2013 at 1645 Central Daylight Time (CDT), the Class 1E electrical equipment air conditioning unit, SGKO5A, was declared nonfunctional due to low oil level on the SGKO5A compressor, elevated vibration and an increase in motor current. As a result, Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 was entered and a plant shutdown was commenced. Mode 3 was entered on September 11, 2013 at 2312 CDT.
Following the plant shutdown, while in Mode 3, the 'A' steam generator (SG) level approached the Auxiliary Feedwater Actuation Signal setpoint of 23.5 % level. The Control Room operators initiated a manual reactor trip. As a result of the trip, a feedwater isolation signal and a motor- driven auxiliary feedwater actuation signal was generated.
The cause of the SGKO5A failure was an inadequate flush and restoration of the system following actions taken to restore SGKO5A in May 2013. The cause of the manual reactor trip and auxiliary feedwater actuation was the lack of crew proficiency to maintain SG levels in Mode 3, immediately following a rapid shutdown.
|05000482/LER-2013-006||14 August 2013||Wolf Creek|
On May 6, 2013 at 1733 Central Daylight Time (CDT), the Class 1 E electrical equipment air conditioning unit, SGKO5A, was declared nonfunctional due to an increasing temperature trend in the 'A' train safety related electrical equipment room. As a result, Technical Specification Limiting Condition for Operation 3.0.3 was entered and a plant shutdown was commenced.
Wolf Creek Generating Station (WCGS) entered Mode 3 on May 7, 2013 at 0009 CDT.
Damage to the liquid line filter drier assembly was caused by over tightening of the wing screw that holds the assembly together. This caused the partial blockage of the thermostatic expansion valves feeding the SGKO5A evaporator coils resulting in the temperature increase in the 'A' train safety related electrical equipment room.
The thermostatic expansion valves and filter drier assembly were replaced and WCGS returned to Mode 1 on 5/13/2013 at 0832 CDT.
|05000482/LER-2013-005||13 May 2013||Wolf Creek|
At 0134 Central Daylight Time (CDT) on 03/13/2013, Control Room annunciators 23B, "DG NE02 UV or UF," and 23D, "DG NE02 Trouble," were received for the 'B' diesel generator (DG). At 0149 CDT on 3/13/2013 the Shift Manager declared a Notification of Unusual Event (NUE) for Loss of Electrical Power/Assessment Capability, as both DGs were not available. At the time of the event, the reactor vessel was defueled with all fuel located in the spent fuel pool. The 'A' DG was out of service for maintenance. Power to the safety related busses were being supplied from the offsite power sources.
The cause of the event was failure of the 'B' DG Jacket Water Pressure Switch (KJPS0162) due to water intrusion in the electrical portion of the switch. Excessive pressure oscillations in the jacket water pressure sensing line led to high cycle fatigue failure of the KJPS0162 diaphragm. The pressure switch was replaced.
The 'B' DG was returned to service at 0221 CDT on 03/14/2013. The NUE was terminated at 0239 CDT on 03/14/2013.
|05000482/LER-2013-004||3 May 2013||Wolf Creek|
On 03/08/2013, Wolf Creek Generating Station was in Refueling Outage 19 with the core defueled and no movement of irradiated fuel assemblies. During replacement of the SGKO5A Class lE electrical equipment air conditioning compressor, it was discovered that the compressor terminal box mounting screws were over torqued to 50 inch pounds. A work history review determined that SGKO5B and the SGKO4B control room air conditioning compressor also had over torqued mounting screws. SGKO4A was unaffected as no maintenance had been performed on the compressor terminal box that specified torque requirements for the mounting screws. As a result, it was determined that SGKO5B had been nonfunctional and SGKO4B had been inoperable during the previous cycle. This resulted in a condition prohibited by Technical Specification and a condition that could have prevented the fulfillment of a safety function.
The mounting screws were replaced with safety related screws and torqued to the proper value.
|05000482/LER-2013-003||15 April 2013||Wolf Creek|
At 1500 hours Central Standard Time on February 16, 2013, Limiting Condition for Operation (LCO) 3.9.3 was declared not met when it was determined that the source range monitors SENI0060A and SENI0061A were not both coupled to the core and source range neutron flux monitor SENI0031 was inoperable. On February 15, offloaded. On February 15, 2013, at 2134 hours, SENI0031 was declared inoperable for performance of maintenance activities. On February 16, 2013, fuel offload occurred from 0426 hours to 0933 hours and 1137 hours to 1500 hours.
Condition A of Technical Specification 3.9.3 was entered and core alterations were suspended and operations that would cause positive reactivity additions were suspended per Required Action A.1 and Required Action A.2. On February 16, 2013, at 1620 hours, cabinet SE054A and monitor SENI0031 were returned to service.
At 1633 hours, fuel offload was recommenced. The cause of the event was non-conservative decision making by an Instrument and Control Supervisor and the Work Controls Senior Reactor Operator to perform work on nuclear instrumentation cabinet SE054A and source range monitor SENI031 when fuel offload had not been completed.
|05000482/LER-2013-001||11 March 2013||Wolf Creek|
On January 8, 2013, at 0500 hours Central Standard Time (CST), Required Action B.4.1 of Technical Specification (TS) 3.8.1 was entered when the 'B' diesel generator (DG) was declared inoperable for performing voluntary preplanned maintenance activities. During the planned replacement of the o-ring between the cylinder head and valve housing on the number 7 cylinder, maintenance personnel contacted one of the eight cylinder head studs and identified that the stud was loose. The broken stud was removed by hand.
On January 10, 2013, enforcement discretion was sought to permit noncompliance with TS 3.8.1, i.e., to permit additional time to complete repairs and restoration of the 'B' DG before a plant shutdown was required. An additional 96 hours was requested to restore the 'B' DG to operable status such that the Completion Time of Required Action B.4.1 would expire at 0500 hours CST on January 15, 2013. The NRC granted approval of the requested enforcement discretion on January 10, 2013. The replacement of the broken stud and two adjacent studs and testing to restore operability was completed at 2254 hours CST on January 12, 2013.
|05000482/LER-2010-013||17 January 2011||Wolf Creek|
During performance of a post-fire safe shutdown (PFSSD) circuit analysis, a commitment from Wolf Creek Generating Station LER 2010-003, the following issues were identified for a potential fire in the control room: (1) Certain fuses associated with the Train B emergency diesel generator (EDG) exciter/voltage regulator could fail, that would result in no power being available to supply credited PFSSD equipment. (2) The pressurizer power operated relief valves could fail open and not be closed within 3 minutes as required by the thermal hydraulic analysis. (3) Dampers in the Train B essential service water and Train B EDG rooms could fail and cause the room temperature to exceed the maximum design temperature for the rooms or drop below the minimum design temperature.
The cause was determined to be an inadequate review of control room circuitry for impact on PFSSD. A number of opportunities should have identified the issues including a root cause evaluation conducted in 2007. An hourly fire watch is in place in the control room and procedure OFN RP-017, "Control Room Evacuation," has been revised to address these issues. These issues have low safety significance.
|05000482/LER-2010-001||22 March 2010||Wolf Creek|
Industry operating experience, 0E30255 dated December 16, 2009, identified that a design feature of each main feedwater (MFW) pump can provide a status indication that the MFW pump is in service when the MFW pump may not actually be supplying water to the steam generator. In this condition, if the MFW pump in service tripped, the motor driven auxiliary feedwater (AFW) pumps would not receive an auto-start signal from the trip of both MFW pumps as required by technical specifications. The operating experience provided reference to Watts Bar Nuclear Plant licensee event reports for a similar event. Further research identified on August 7, 2008, the Nuclear Regulatory Commission (NRC) issued Watts Bar Nuclear Plant - NRC Integrated Inspection Report 05000390. The inspection report identified that a non-operating MFW pump in the reset condition impacts operability of the AFW auto-start channel due to the false (i.e., invalid) indication of the MFW pump status.
On January 21, 2010, the operating experience was placed into the corrective action program. On February 4, 2010, a review of the operating experience determined that the design and normal operation of the MFW pumps at Wolf Creek Generating Station (WCGS) could result in a condition that does not conform to the WCGS Technical Specification (TS) Table 3.3.2-1, Function 6.g., Trip of All Main Feedwater Pumps. During plant startups, the non-operating MFW pump in the reset condition results in two inoperable AFW auto-start channels. There is no TS Condition for two MFW pump trip channels inoperable. Limiting Condition for Operation (LCO) 3.0.3 specifies that when an associated Action is not provided, action shall be initiated within 1 hour to place the plant in Mode 3 in 7 hours. Action had not previously been taken as required by the TSs.
|05000482/LER-2005-005||28 November 2005||Wolf Creek|
On September 29, 2005, conditions were discovered where a postulated design basis fire could cause the loss of a safe shut down success path.
During reviews associated with post fire safe shutdown reanalysis work, Wolf Creek personnel discovered that the ability to perform diverse means (operator manual actions) required to mitigate spurious actuations came into question for Fire Area A-8. This could cause the loss of the centrifugal charging pump's (CCP's) capability to successfully inject borated water into the reactor, due to a potential for gas intrusion into the suction of the pumps. This does not meet Wolf Creek's commitments to 10 CFR 50 Appendix R.III.G as reflected in the approved Fire Protection Plan.
A 1-hour fire watch was implemented in the fire area.
The safety significance for this event is low.
|05000482/LER-2004-003||5 May 2004||Wolf Creek|
On March 6, 2004, Wolf Creek Generating Station (WCGS) was operating at 100 percent steady state power. At 9:47 P.M. Central Standard Time, startup transformer neutral ground relays 251N-2T1 and 251N-4T1 tripped which initiated protective lockout and transformer trip relays for the startup transformer, tripping the transformer and resulting in a loss of power to the Engineered Safeguards Feature (ESF) bus NB02. The loss of power to the transformer that normally powers this bus resulted in the automatic start, and subsequent loading of the "B" Emergency Diesel Generator (EDG.) The cause of the transformer trip was due to a cable fault in the cable connected to the stamp transformer's "X" winding "C" phase (X-C) bushing, caused by water that had collected in the cable through connecting lug inspection holes.
The safety significance of this event is low. This event is bounded by the current licensing basis analysis as reported in WCGS Updated Safety Analysis Report (USAR) section 15.2.6 "Loss of Non Emergency AC Power to the Station Auxiliaries." All safety-related equipment operated as expected. There were no adverse effects to the health and safety of the public.
|05000482/LER-2004-002||9 April 2004||Wolf Creek|
On February 13, 2004, at 8:04 AM CST, Wolf Creek Generating Station (WCGS) experienced an automatic actuation of the Reactor Protection System (RPS) including an automatic reactor trip due to Lo-Lo water level in the "D" steam generator (SG). This actuation occurred following failure of the "D" SG main feedwater regulating valve (MFRV) resulting in valve closure. When the MFRV failed closed, "D" SG level decreased below the reactor trip setpoint, initiating a reactor trip. The unit received an expected feedwater isolation and auxiliary feedwater actuation (both motor and turbine driven pumps) because of the Lo-Lo SG level. All control rods fully inserted, and the RPS and Engineered Safety Features (ESF) systems performed as expected.
The cause of the MFRV closure was separation of the valve plug from the valve stem, which caused the plug to fall into its seated (closed) position. The stem/plug assemblies were subsequently replaced in all four (4) MFRVs.
The safety significance of this event is low. This event is bounded by the current licensing basis analyses as reported in WCGS Updated Safety Analysis Report (USAR) section 15.2.7 "Loss of Normal Feedwater Flow." All safety related equipment performed as expected. There were no adverse effects on the health and safety of the public.
|05000482/LER-2004-001||12 March 2004||Wolf Creek|
On January 15, 2004, at 8:30 PM, central standard time, containment spray system (EIIS: BE) isolation valve (EIIS: ISV) EN-V025 was found locked in the fully open position. The containment spray system was declared inoperable in accordance with Technical Specification (TS) 3.6.6. Valve EN-V025 was closed and the containment spray system was declared operable at 8:51 PM on January 15, 2004.
This valve was unlocked and opened on December 30, 2003, during the performance of the containment spray pump inservice pump test for train B, which is controlled by procedure STS EN-100B. This procedure requires that after this test has been completed, valve EN-V025 is to be restored to closed and locked, and subsequently verified to be closed and locked.
This condition existed prior to discovery for a period greater than 72 hours, the allowed Completion Time associated with Required Action A.1 of TS 3.6.6. As such, this situation represents a condition prohibited by Technical Specifications, and is reportable per 10 CFR 50.73(a)(2)(i)(B).
The root cause for valve EN-V025 being in the incorrect position is lack of information validation or verification. The two nuclear station operators did not adequately validate information, and did not adequately verify the information (i.e., valve position).
|05000482/LER-2003-004||15 January 2004||Wolf Creek|
On November 17.2003, during Wolf Creek Generating Station (WCGS) refueling outage number 13, a non-licensed Nuclear Station Operator observed water leaking approximately 25 to 30 drops per minute from the ASME Code Class 2, 3/4-inch line upstream of Safety Injection (EP) system valve EPV0109. This Is the 3/4-Inch vent line on the combined SI/RHR outlet piping to EP Accumulator Tank D. The Reactor Coolant System (RCS) leakage constitutes degradation of a principal safety barrier and Is considered reportable to the requirements of 10 CFR 50.73(a)(2XiiXA), `Degraded or Unanalyzed Condition.' Conditions that represent welding or material defects in the primary coolant system which cannot be found acceptable under ASME Section XI standards are reportable to this criterion.
The cracked vent line and associated socket weld were removed and repaired on November 18, 2003. The completed weld repair was inspected, tested and found acceptable.
Initial evaluation concluded that this was a fatigue crack initiating from a filing groove that induced a stress concentration that, after several years of cycling, propagated through the vent line.
The safety significance of this event Is low. The RCS leakage that resulted from the cracked vent line is within the capability of reactor makeup systems and a complete failure Is bounded by the plant Loss of Coolant Accident (LOCA) analysis.
NRC FORM SSO (7-2001)
|05000482/LER-2003-002||23 May 2003||Wolf Creek|
At 1000 on March 25, 2003, the Wolf Creek Generating Station (WCGS) Shift Manager was notified that both "A" and "B" wide range trains of Reactor Vessel Water Level Indicating System (RVLIS) were not functioning as designed. B Both "A" and "B" wide range trains of RVLIS in the Main Control Room were off-scale high (>120%), consistent with the plant computer point indication of 122.8%. Based upon discussion with vendor representatives and industry peers, WCGS concluded that this reading was approximately 12% higher than should be expected. The Shift Manager declared both "A" and "B" trains of RVLIS inoperable and entered Condition C of Technical Specification 3.3.3, Post Accident Monitoring (PAM) Instrumentation.
Although WCGS entered the applicable Technical Specification Required Action immediately following recognition of this condition, further evaluation determined that this condition existed for a time longer than permitted by Technical Specifications. As such, this issue represents a condition prohibited by Technical Specifications, and is reportable per 10 CFR 50.73 (a)(2)(i)(B).
Because various other means are available to plant operators to determine reactor vessel level, there is minimal safety significance associated with this issue.
|05000482/LER-2002-005||24 October 2002||Wolf Creek|
On September 9, 2002, at 3:58 p.m. Central Daylight Time (CDT), Wolf Creek Generating Station (WCGS) experienced a Degraded Voltage signal on Emergency Power Bus NB01 causing an automatic actuation of the "A" Emergency Diesel Generator ("A" EDG) and the associated Load Shedder and Emergency Load Sequencer (LSELS). The cause of the degraded voltage signal was determined to be due to a hardware failure in an electronic relay driver card in the LSELS degraded voltage circuitry. An actual degraded emergency bus voltage situation did not exist.
The safety significance of this event is low. All safety related equipment responded to the actuation signal as expected. There were no adverse effects on the health and safety of the public.
|05000482/LER-2002-004||17 October 2002||Wolf Creek|
At 0815 on June 7, 2002, the Wolf Creek Generating Station (WCGS) Shift Manager was notified of a postulated fire event that could cause a cable-to-cable hot short. If cable-to-cable hot shorts are assumed to occur, this event has the potential to cause water in the refueling water storage tank (RWST) to drain to the containment recirculation sump. It was discovered that the control cables for two redundant motor operated valves are routed in the same electrical raceway. The two valves are in the same electrical separation group, but are redundant in their function of conserving water inventory in the RWST. Further investigation determined that the control cables for motor operated valves in the opposite electrical separation group have the same configuration. The cause of this condition is that cable-to-cable interactions were not considered in the initial design of the plant.
At 0945 on July 19, 2002, the Shift Manager was notified of conditions where a postulated fire event could lead to the loss of motor control centers that power post-fire safe shutdown equipment in both trains due to inadequate horizontal separation in conjunction with improper breaker coordination.
At 1030 on August 20, 2002, conditions were discovered where a postulated fire could cause the loss of both centrifugal charging pump's (CCP) capability to successfully inject borated water into the reactor.
The safety significance of these events is low.
|05000482/LER-2002-003||28 June 2002||Wolf Creek|
On May 8, 2002, at 5:07 p.m., Wolf Creek experienced an automatic actuation of the Reactor Protection System (RPS) including an automatic reactor trip due to low water level in the "D" steam generator (SG). This actuation occurred following the closure of the "D" SG main feedwater regulating valve (FRV) during surveillance testing.
When the FRV closed, "D" SG level decreased below the reactor trip setpoint of 23.5 percent, initiating a reactor trip. The unit received a feedwater isolation and auxiliary feedwater actuation (both motor and turbine driven) because of the low-low SG levels. All control rods fully inserted, and the RPS and the Engineered Safety Features (ESF) Systems performed as expected.
The cause of the FRV closure was the failure of a Westinghouse 7300-series manual controller card for the "D" SG FRV. The failed controller card was replaced.
The safety significance of this event is low. This event is bounded by the current licensing basis analyses as reported in Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR) section 15.2.7 "Loss of Normal Feedwater Flow." All safety related equipment performed as expected. There were no adverse effects on the health and safety of the public.
|05000482/LER-2002-002||17 June 2002||Wolf Creek|
On April 24, 2002, while in Mode 4, Wolf Creek Nuclear Operating Corporation (WCNOC) personnel completed surveillance procedure STS BB-004, "RCS Water Inventory Balance." A calculational (rounding) error resulted in the conclusion that the Reactor Coolant System (RCS) (EIIS system code AB) unidentified leakage was within the Technical Specification (TS) limit of 1.0 gpm. The unidentified leakage was actually 1.091 gpm, or .091 gpm above the TS limit.
On April 25, 2002, Wolf Creek Generating Station (WCGS) entered Mode 3 from Mode 4 contrary to TS Limiting Condition of Operation (LCO) 3.0.4. This specifies that the plant cannot change modes when an LCO is not met. The calculational error made in STS BB-004 was not recognized until after the mode change. When the error was recognized, TS 3.4.13, "RCS Operational LEAKAGE," Condition A, was entered. The source of the leakage was identified and corrected. On April 26, 2002, the leak rate was 0.31 gpm, and Condition B of TS 3.4.13 was exited.
This event is reportable per 10 CFR 50.73(a)(2)(i)(B) as a violation of the plant's Technical Specifications.
The root cause of this event was inadequate procedural guidance for performing RCS leakage calculations.
No impact on personal, nuclear, or radiological safety resulted from this event.
|05000482/LER-2002-001||25 March 2002||Wolf Creek|
This information is reported voluntarily appropriate to the guidance provided in NUREG 1022, "Event Reporting Guidelines 10 CFR 50.72 and 50.73," revision 2, section 2.7 "Voluntary Reporting.
Heat exchanger tube degradation was identified on both of Wolf Creeks Generating Station's (WCGS) Emergency Diesel Generators (EDGs). WCGS initiated an investigation team to examine this issue. Although the team concluded the EDGs were operable at the time of discovery, heat exchangers on both EDGs were found to be in a degraded condition. Suspect tubes were plugged. The team determined that the Preventive Maintenance (PM) program in place during the mid1990s did not implement eddy current testing (ECT) to monitor structural integrity of EDG heat exchanger tubing. This condition monitoring is in addition to the thermal performance monitoring performed in response to NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment.
Specific information is included with respect to ECT programs, including the identification and disposition of test data containing absolute drift indication (ADI), suspected dealloying (SDA), and tube pitting indications in admiralty brass heat exchanger tubes.
Wolf Creek Nuclear Operating Corporation (WCNOC) believes this information may be of generic interest to the nuclear industry.
|05000482/LER-2001-001||14 January 2002||Wolf Creek|
On November 16, 2001, Wolf Creek Nuclear Operating Corporation (WCNOC) personnel identified an error in an internal flood calculation. The calculation under estimated the input flow rate from a pipe break in two Auxiliary Building rooms and over estimated the drain flow rate from these rooms. As a result of correcting these values in the calculation, the calculation results now predict a flood water level that would submerge safety-related equipment required to switch the suction of the Auxiliary Feedwater (AFW) pumps from the Condensate Storage Tank (CST) to the Essential Service Water (ESW) system. Evaluations of event scenarios indicate that flooding of safety-related AFW equipment could occur and/or the decay heat removal function of AFW would not be accomplished.
The isolation valve for this line has been throttled, thus reducing the input flow rate and the resultant flood water level to ensure components required to fulfill the AFW system safety function are not submerged.
The cause for including these incorrect assumptions in this calculation is indeterminate due to the historical nature of this condition (the calculation was performed by an Architect/Engineer's staff in 1986). Based on a sampling review of other calculations, this issue is believed to be an isolated case.
This condition is considered to be of minimal safety significance.