|Entered date||Site||Region||Reactor type||Event description|
|ENS 49422||25 February 2020 08:26:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined that the condition described below to be applicable to Callaway Nuclear Plant resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E Train B batteries and chargers (including the B Swing charger) control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane; simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This would cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. Similar Events: EN #49411 and EN #49419|
|ENS 54069||17 May 2019 03:35:00||Callaway||NRC Region 4|
EN Revision Text: REACTOR TRIP DUE TO SOURCE RANGE HI FLUX SIGNAL This is an 8-hour, non-emergency notification for a valid reactor trip signal with the reactor not critical, and a valid auxiliary feedwater system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) - Valid System Actuation.
At 2303 (CDT) on May 16, 2019, the plant was administratively in mode 2 due to withdrawing control rods for startup following refuel. The reactor had not been declared critical. The P-6 permissive at 10E-10 Amps was met for one of two Intermediate Range detectors allowing for block of the Source Range high flux trip (1E5CPS). Prior to performing the block, the Source Range high flux trip setpoint was exceeded and a reactor trip received. All systems responded as expected. A feedwater isolation signal was received due to the reactor trip with feedwater temperature less than 564 degrees Fahrenheit. Auxiliary feedwater was started to maintain steam generator levels. The plant is being maintained stable in mode 3 with no complications. The NRC Resident Inspector was present during the startup and was notified of the reactor trip.
A correction is being made for the sixth sentence in the second paragraph above, which states, 'A Feedwater Isolation signal was received due to the reactor trip with feedwater temperature less than 564 degrees Fahrenheit.' Within this sentence, 'feedwater temperature' is to be replaced with 'reactor coolant system temperature.' The licensee has notified the NRC Senior Resident Inspector.
|ENS 54061||12 May 2019 04:51:00||Callaway||NRC Region 4||On 5/11/19, Callaway Energy Center entered Mode 4 at 1217 (CDT). At 2305, the door from the Auxiliary Building to the RAM Storage building was found blocked open. This door is an Auxiliary Building pressure boundary for the Emergency Exhaust system. The Emergency Exhaust system is required in Modes 1,2,3,4, and during movement of irradiated fuel assemblies in the Fuel Building. The door was being blocked open with a large ramp. This rendered the Emergency Exhaust system not capable of performing its design safety function. LCO (Limiting Conditions for Operation) 3.7.13.B was entered, and preparations to move the ramp commenced. LCO 3.7.13.B is for two Emergency Exhaust trains being inoperable due to an inoperable auxiliary building boundary. The allowed outage time is 24 hrs. to restore the boundary to Operable. The door was closed and LCO 3.7.13.B was exited at 0111 on 5/12/19. This event is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) control the release of radioactive material, or (D) mitigate the consequences of an accident. The NRC Senior Resident has been notified.|
|ENS 54005||17 April 2019 06:25:00||Callaway||NRC Region 4||At approximately 0137 CDT, with the Plant (Callaway) in No Mode (Defueled) the "B" Switchyard Bus cleared resulting in a loss of normal power to "A" Train Safety Related Transformer XNB01. This resulted in an under voltage condition on Safety Related Bus NB01. The "A" Emergency Diesel started per design and re-energized Bus NB01. This actuated the shutdown sequencer which first sheds loads including the "A" Spent Fuel Pool Cooling Pump and started "A" Essential Service Water Pump, "A" Component Cooling Water Pump, "A" Control Room A/C and other design loads. No complications were identified. The "A" Switchyard Bus remained energized at all times. The "A" Spent Fuel Pool Cooling Pump was restarted per off normal procedure response at 0149 CDT. Spent Fuel Pool water temperature started at 102 F and rose to 103 F prior to restart. There was no movement of irradiated fuel in progress in the Fuel Building during this time. The plant remains stable in No Mode (Defueled). At the time of the loss of "B" Switchyard Bus, the plant was closing Generator Output breaker MDV53 to establish a backfeed alignment. Further investigation is in progress. This event is reportable per 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident has been notified.|
|ENS 53864||7 February 2019 17:05:00||Callaway||NRC Region 4||A non-licensed and non-supervisory employee inadvertently possessed and consumed alcohol within the protected area. The employee's access to the plant has been suspended. The Senior Resident Inspector has been notified by the licensee.|
|ENS 53759||28 November 2018 17:25:00||Callaway||NRC Region 4||On November 28, 2018, while performing an engineering review of the bases for environmental qualification (EQ) requirements for the Atmospheric Steam Dumps (ASDs), it was determined that applicable EQ requirements had not been applied to a key component of each of the ASDs. The result of this issue is that it the availability of the ASDs for a controlled plant cooldown following a postulated steam line break outside containment cannot be assured. Callaway is developing a compensatory action temporary plant modification to install insulation that will protect the affected ASD components from the post Main Steam Line Break temperature. This condition is reportable 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (B) remove residual heat, or (D) mitigate the consequences of an accident. The issue places the plant in a 24-hour Technical Specification (TS) Limiting Condition for Operations (LCO), 3.7.4. The licensee has notified the NRC Resident Inspector.|
|ENS 53717||5 November 2018 15:38:00||Callaway||NRC Region 4||This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) which requires notification to the NRC of notifications made to other government agencies. On November 3, 2018, a Fyrquel hydraulic fluid leak was discovered at the plant's intake structure (on the Missouri River). The leakage was initially believed to have only flowed into the sump located at the lower level of the intake structure. On November 5, 2018, during subsequent investigation, plant staff recovered approximately 40 gallons from the sump. Approximately 200 gallons of hydraulic fluid could not be accounted for, so it was assumed to be released to the Missouri River. The leak was either caused by the hydraulic fluid being pumped to the stilling chamber, which drained to the river before operators were able to secure the intake pumps, or by leakage from hydraulic lines located near the free discharge valve, which would have then been carried by water leaking from piping associated with the free discharge valve into the stilling chamber (and then to the river). The Environmental Protection Agency (EPA), Unites States Coast Guard National Response Center (NRC), and Missouri Department of Natural Resources (DNR) were notified on November 5, 2018. The NRC Senior Resident Inspector has been notified by the licensee. The licensee confirmed that this leakage was above the reportable quantity. Notified EPA, USDA, and FEMA. Notified DOE and HHS via e-mail.|
|ENS 53485||3 July 2018 19:07:00||Callaway||NRC Region 4|
EN Revision Text: DISCOVERY OF AN UNANALYZED CONDITION THAT SIGNIFICANTLY DEGRADES PLANT SAFETY On July 3, 2018, while performing a review of Emergency Operating Procedures, a concern was identified regarding the potential for excessive loss of ultimate heat sink inventory (UHS) through the auxiliary feedwater (AFW) system mini-flow recirculation pathway. This condition would have the potential to prevent the ultimate heat sink from providing an adequate inventory of water for a 30-day mission time.
The normal water supply for the Callaway AFW system is the condensate storage tank (CST). The CST is a non-safety grade component. The safety-grade supply for AFW is the essential service water (ESW) system. The ESW system is supplied by the UHS. The UHS thermal performance analysis accounts for a loss of UHS inventory to the AFW system up until the point of the accident sequence that the AFW pumps would be secured. The analysis did not include an allowance for loss of UHS inventory through the AFW mini-flow recirculation pathway following the AFW pumps being secured. The EOP guidance that secures the AFW pumps does not isolate the mini-flow recirculation pathway.
Initial estimates indicate that loss of UHS inventory through the mini-flow recirculation pathway, if not isolated, would preclude the UHS from completing its 30-day mission time. This potential for depletion of the UHS placed the plant in an unanalyzed condition that significantly degraded safety.
Callaway has issued interim guidance to the on-shift personnel regarding this concern to ensure that the ultimate heat sink water level is maintained at a level that will be adequate to mitigate the potential loss of inventory.
This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspectors have been notified of this condition.
Event Notification (EN) 53485, made on July 3, 2018, is being retracted because re-evaluation performed subsequent to the notification has demonstrated, based on actual plant equipment and environmental conditions, that the unanalyzed inventory losses previously reported by EN 53485 would not have depleted the UHS inventory to an unacceptable level during its 30-day mission time. The re-evaluation has led to the conclusion that the previously unanalyzed losses of UHS inventory would not have prevented the UHS from performing its specified safety functions and meeting its 30-day mission time requirements. With the UHS capable of performing its specified safety functions and meeting its 30-day mission time requirements, the systems supported by the UHS would have remained capable of performing their specified safety functions. Based on these considerations, it has been determined that the condition reported in EN 53485 did not result in the plant being in an unanalyzed condition that significantly degraded safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R4DO (Gaddy).
|ENS 53203||12 February 2018 11:21:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On February 12, 2018, during performance of a TSC Diesel Generator Functional Test, the TSC Ventilation could not be placed in Filter Mode. Filter Mode operation is credited in the TSC habitability analysis of record. The TSC was declared non-functional due to the unavailability of the filtration system. The Emergency Operations Facility (EOF) is available for use as a backup TSC. Additionally, the TSC would be available for emergency response purposes for events that do not involve a release in progress. The NRC Resident Inspector has been notified.
At 1258 CST, TSC ventilation was declared operable. A start permissive lever was adjusted to remedy interference. The licensee notified the NRC Resident Inspector.
|ENS 53119||14 December 2017 11:29:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||On December 13, 2017, Callaway determined that a violation of one provision of the site's Fitness For Duty (FFD) policy had occurred. FFD testing confirmed the use of a controlled substance. The violation was committed by a non-licensed, supervisory employee. The individual's unescorted access to the plant has been removed. The NRC Resident Inspector has been notified by the licensee.|
|ENS 52998||3 October 2017 15:52:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||This report is made per 10 CFR Part 21.21(d)(3)(i) on the identification of a defect or a failure to comply. By the letter dated August 31, 2017, Westinghouse notified Callaway Plant that they were unable to complete a 10 CFR 21.21(a) evaluation for a product advisory which was issued by Cameron Measurement Systems for a concern with the Model 752B transmitter product line. The product advisory identified the potential for instability in the transmitter output signal under certain grounding conditions. Callaway was supplied transmitters that could be affected under Westinghouse part numbers 8765D64G03, 8765D64G04, and 8765D64G05. The Westinghouse notification letter served as the transfer of reporting responsibilities for this concern from Westinghouse to Callaway in accordance with 10 CFR 21.21(b). The 10 CFR Part 21 evaluation is based on the technical information provided by Westinghouse. The notification letter describes that shifts up to 10-20 percent of instrument scale (1.6-3.2mA) can be observed within the transmitter output under the grounding conditions such as those introduced by the original equipment manufacturer during testing. Westinghouse evaluations concluded that such instabilities would be self-revealing within plant applications for which the transmitter output signal was supplied to a Westinghouse 7300 system, assuming the transmitter stanchion was grounded. Not all transmitters within this product line were subject to this concern. On 10/02/2017 Callaway personnel completed the 10 CFR Part 21 evaluation. Of the transmitters identified above, only one is currently installed at Callaway in location BNLT0930, Refueling Water Storage Tank Protection A Level Transmitter. No instabilities (oscillations) have been observed in this transmitter but plans are being made to replace the transmitter. If one of the susceptible transmitters were installed for RCS flow application (low flow reactor trip function), the allowed sensor drift to accommodate changes in transmitter performance would be required to be limited to 1percent of instrument span per an 18-month operating period. Current margin available within the set point uncertainty analysis for this Reactor Trip function is 0.62 percent of instrument span. The observed shift in output of 10-20 percent of instrument span would thus exceed the available margin for this protective function. Therefore, the safety function for this device could not be assured for all transmitter configurations. This condition could create a substantial safety hazard due to the loss of safety function of a basic component which meets the criteria of major degradation of essential safety-related equipment. For the situation at Callaway, no other reporting criteria apply since there is no evidence that the installed transmitter BNLT0930 is susceptible to the concerns noted in the product advisory. The NRC Resident Inspector has been notified of this issue.|
|ENS 52607||13 March 2017 16:47:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On March 13, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified non-conforming conditions in the plant design such that specific Technical Specification equipment is considered to not be adequately protected from tornado missiles. The Emergency Fuel Oil Truck Connection Lines for both redundant Emergency Fuel Oil trains extend through the Plant South wall of the Diesel Generator Building structure where they may be exposed to design bases tornado missile impact. The direct impact by the design basis missile could result in damage to the fuel oil transfer lines, thereby preventing delivery of the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. This condition affects the fuel supply to both supported Emergency Diesel Generator trains. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified.
Event Notification EN # 52607, made on 03/13/2017, is being retracted because new information has been obtained that negates the originally reported condition. Specifically, subsequent to the Event Notification, an engineering analysis was performed which confirmed that a design basis missile strike on either of the unprotected truck connection lines would not result in damage to the extent that the affected fuel oil transfer lines would be prevented from delivering the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. (The analysis showed that although bending / deformation of the lines would occur in response to the postulated missile strike, integrity of the lines would remain.) Based on the above, the unanalyzed condition did not prevent the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition does not meet the criteria for 8-hour notification that are provided in 10 CFR 50.72(b)(3)(ii)(B) or 10 CFR 50.72(b)(3)(v)(A), (B), or (D). As the condition does not require enforcement discretion, the provisions of EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents) need not be invoked. However, the immediate and long-term compensatory actions that have been taken following discovery of the condition will remain in place until the condition is fully resolved. The NRC Resident Inspector has been informed of this Event Notification retraction. Notified R4DO (Azua).
|ENS 51874||20 April 2016 22:04:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Recent Operating Experience at Callaway has shown that the pressure transient experienced in the Essential Service Water (ESW) system during Engineered Safety Feature Actuation System (ESFAS) testing can result in gasket failure on the Control Room Air Conditioners rendering the units nonfunctional. Previously, this pressure transient was considered to be the result of the system alignment used to perform the surveillance test, which is not the same as the system lineup which would occur on an actual Loss of Offsite Power (LOOP) or Safety Injection Signal Design Basis Accident (DBA) event. However, on April 20th, 2016, Callaway received preliminary analysis results that predict the Control Room Air Conditioners would actually experience a greater pressure transient during a DBA than what is currently experienced during ESFAS testing. This condition could result in the Control Room Air Conditioners not being capable of performing their safety function following a DBA event, and challenge Control Room Habitability. Therefore, this condition meets the reporting criteria of 10 CFR 50.72(b)(3)(ii)(B). Based on current conditions (i.e., the plant is not in Power Operation), this condition does not present an immediate safety concern. The analysis of the pressure transient experienced by the ESW system during a postulated DBA event is preliminary and further evaluation of the analysis is ongoing. The NRC Resident Inspector has been notified.|
|ENS 51846||3 April 2016 07:13:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 2302 (CDT) on April 2, 2016, with the plant shutdown, (with) all control rods inserted in the reactor and while attempting to reset the reactor trip breakers to support outage activities (reset of the main turbine), the reactor trip breakers reopened. This was identified to be due to having both trains of Solid State Protection System (SSPS) out of service while in Mode 5. With both trains of SSPS out of service, a condition was met that would cause a reactor trip signal due to having a general warning condition on both trains. Per procedure, the control rods were incapable of withdrawal and fully inserted. Reactor Coolant System boron was 2280 ppm. There were no actuations as a result of the reactor trip breakers opening due to SSPS being removed from service. The licensee will be notifying the NRC Resident Inspector.
At 0713 EDT on April 3, 2016, EN #51846 provided notification of a Reactor Protection System actuation as revealed by the reactor trip breakers opening. Upon further investigation, it has been determined that the system actuated during maintenance activities due to a reactor trip signal caused by both trains of the Solid State Protection System (SSPS) being in test. This signal was not in response to actual plant conditions or parameters satisfying the requirements for initiation of the system and was therefore invalid. As such, the notification made by EN #51846 for a valid actuation of a specified system is hereby retracted. In addition, an editorial change to the first sentence of the original notification description is hereby made. The first sentence is revised to read as follows: At 2303 EDT on April 2, 2016, with the plant shut down and all control rods inserted into the reactor, while attempting to reset the reactor trip breakers to support outage activities (reset of the main turbine), the reactor trip breakers reopened. The NRC Resident Inspector will be notified. Notified the R4DO (Kellar).
|ENS 51649||12 January 2016 18:37:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On January 12, 2016, during performance of an EOF (Emergency Off-Site Facility) Diesel Operability Test, the EOF Air Conditioning unit and EOF Air Return fan did not run as expected. Per plant procedure, the operators placed the system in filtration mode and then back in normal mode. Again, both units did not run as expected. The EOF was declared non-functional due to the failure of the air conditioning unit and fan. The EOF is available for emergency response purposes unless the temperature can not be maintained or a release is in progress. The backup EOF is available for use if needed. The licensee notified the NRC Resident Inspector.
The Emergency Off-Site Facility (EOF) was restored to functional status at 1223 CST on 01/13/2016. The licensee notified the NRC Resident Inspector. Notified R4DO (Kellar)
|ENS 51646||12 January 2016 11:14:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Event Report per 10 CFR 26.719(b)(2)(ii) On January 11, 2016, Callaway determined a violation of two provisions of the site Fitness For Duty policy were committed offsite by a non-licensed supervisory employee. Unescorted access for the employee has been denied. The licensee notified the NRC Resident Inspector.|
|ENS 51308||11 August 2015 05:19:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Reactor trip caused by turbine trip. Turbine tripped immediately following the trip of one of four 345KV offsite lines. The reason for protective relaying not preventing the grid disturbance from tripping the turbine generator is not known at this time. All normal offsite and onsite power sources are available. Auxiliary Feedwater actuated as expected on low steam generator level following the trip from 100% power. All systems functioned as expected in response to the trip. The NRC Senior Resident Inspector has been notified. An electrical fault on a 345 kV line 2 miles from the site caused the bus to strip and reclose, which cleared the fault. All control rods fully inserted and the plant is in its normal shutdown electrical lineup.|
|ENS 51257||23 July 2015 19:39:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||During plant cooldown in response to conditions reported to the NRC in Event Notification 51253, Callaway was in Mode 3 (Hot Standby) and on the way to Mode 5 (Cold Shutdown). In accordance with cooldown procedures, Callaway was operating with one Main Feedwater Pump when the pump speed unexpectedly lowered to 0 RPM. The pump was manually tripped in response to the condition. This led to a decrease in water level in the steam generators. In response, operators manually activated the Auxiliary Feedwater System. All systems and components functioned normally in response to the event, and plant operators are currently continuing the controlled shutdown from Mode 3 to Mode 5. Safety related buses are receiving normal off-site power and the grid is currently stable. The NRC Resident Inspector was notified.|
|ENS 51253||23 July 2015 04:21:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On July 23, 2015 at 0115 (CDT), Callaway Plant initiated a shutdown required by Technical Specifications (TS). At 2139 (CDT) on July 22, 2015, TS 3.4.13 Condition A was entered due to unidentified RCS leakage being in excess of the 1 gpm TS limit. The leak was indicated by an increase in containment radiation readings, increasing sump levels, and decreasing levels in the Volume Control tank (VCT). A containment entry identified a steam plume; due to personnel safety the exact location of the leak inside the containment building could not be determined. At this time radiation levels inside (the) containment are stable and slightly above normal. There have been no releases from the plant above normal levels. The (NRC) Senior Resident Inspector was notified.
Callaway entered TS 3.4.13 Condition B at 0053 (CDT on July 23, 2015) for the subject leakage since reactor coolant pressure boundary leakage could not be ruled out by visual inspection. The estimated leak rate when the decision was made to shut down the plant was approximately 1.8 gpm. The plant entered Mode 3 at 0600 CDT. Additionally, at approximately 1315, it was determined that the duration of the required outage would be greater than three days, thus requiring notification to the Missouri Public Service Commission. This offsite notification is reportable to the NRC (per 10CFR50.72(b)(2)(xi)), and the above table has been updated to reflect this reporting requirement. The licensee notified the NRC Resident Inspector. Notified R4DO (Gepford).
Clarification to the initial event notification: the term 'RCS' used above means 'Reactor Coolant System.' Therefore the second sentence from the initial notification is clarified to read, 'At 2139 (CDT) on July 22, 2015, TS 3.4.13 Condition A was entered due to unidentified Reactor Coolant System (RCS) leakage being in excess of the 1 gallon per minute (gpm) TS limit.' The licensee notified the NRC Resident Inspector. Notified the R4DO (Gepford).
|ENS 50625||19 November 2014 03:39:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
Following shift turnover from days to nights on 11/18/2014, it was discovered that all (4) of the Safety Injection (SI) Accumulator Outlet Isolation Valve breakers were unlocked and closed. At the time of discovery, 3 of the safety injection accumulator valves were open and 1 was closed for testing. At that time the plant was in MODE 3 at normal operating pressure and temperature. The plant had been performing RCS pressure isolation valve testing prior to shift turnover. The condition was discovered during testing of valves associated with the 'C' safety injection accumulator. After discovery of the condition, Operations directed that the 'A', 'B', and 'D' SI Accumulator Outlet Isolation Valve breakers be opened and locked. This action was completed by approximately 1930 (CST) on 11/18/2014.
The NRC Resident Inspector was notified. The plant entered T.S. 3.0.3 for approximately 30 minutes while restoring the 'A', 'B' and 'D' accumulators to operable (breakers opened and locked with their associated outlet valves open).
|ENS 50465||17 September 2014 13:59:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||At approximately 0913 CDT, the Callaway Plant Control Room was notified that an oil tanker truck overturned on plant property. The location of the incident was inside the owner controlled area but outside of the protected area. The truck was making a delivery to the plant. The Callaway County Emergency Operation Center was contacted at approximately 0930 to request an ambulance for the driver. The driver of the truck was transported offsite for medical treatment. The drivers injuries are not life threatening. The incident has resulted in a slow leak of diesel fuel that is being contained onsite. Missouri Department of Natural Resources was notified of the spill. The Missouri Highway Patrol is onsite assisting with the incident and restoration efforts. Both NRC Resident Inspectors were notified.|
|ENS 50333||31 July 2014 19:02:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Voluntary notification per the NEI Groundwater Protection Initiative. On July 31, 2014, Callaway Plant received results of a sample from a new ground water monitoring well. The sample was taken on July 25, 2014. The sample results indicated a tritium concentration of approximately 1.6 E6 picocuries/liter and a Co-60 concentration of approximately 12 picocuries/liter. The new monitoring well is located within the plant's property and is adjacent to a manhole where the plant's discharge piping joins with the cooling tower blowdown piping. Both the plant discharge piping and the cooling tower discharge piping are buried. Releases from the plant discharge line have been suspended. A backup sample taken on July 25, 2014, will be sent to a lab for analysis. Another sample will be taken on August 1, 2014. There is no effect on drinking water, and therefore, no dose to the general public or plant staff. The licensee will notify the Missouri State Department of Natural Resources and Callaway County officials. The licensee will notify the NRC Resident Inspector.|
|ENS 50056||24 April 2014 14:08:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Contrary to the requirements in 10 CFR 26.137(b), a Department of Health and Human Services (DHHS) certified laboratory returned results for a blind specimen that was inconsistent with what was expected. On 04/22/2014, blind specimens from the same lot number were sent to the two contracted DHHS laboratories. On 04/23/2014, one of the labs reported unexpected results while the other laboratory reported the expected results. At approximately 1530 (CDT) on 04/23/2014, the lab report was reviewed by Fitness For Duty Management at Callaway Plant and the inaccurate result was identified. On 04/24/2014, the Medical Review Officer (MRO) contacted Clinical Reference Laboratory (CRL) to discuss the testing discrepancy and directed the lab to retest the specimen. The MRO requested that CRL initiate an investigation to determine the reason for the inaccurate result and provide a report of the results of that investigation within 20 days. 10 CFR 26.719(c)(3), 'Reporting Requirements,' requires that 'if a false negative error occurs on a quality assurance check of validity screening tests, as required in � 26.137(b), the licensee or other entity shall notify the NRC within 24 hours after discovery of the error.' The licensee has notified the NRC Resident Inspector.|
|ENS 49857||26 February 2014 18:18:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Contrary to the requirements in 10 CFR 26.137(b), a DHHS (Department of Health and Human Services) certified laboratory returned results for a blind specimen that were inconsistent with what was expected. On 02/25/2014, dilute blind specimens from the same lot # were sent to the three contracted DHHS laboratories. Upon review by the Callaway MRO (Medical Review Officer) at approximately 07:30 (CST) on 2/26/2014, it was discovered that one of the laboratories (Toxicology) reported results of negative. That result was inconsistent with the certification received from the blind provider (ProTox) certifying the specimen as negative and dilute. Later in the day on 2/26/14, the remaining two labs (Quest and CRL) also returned results of negative instead of negative and dilute. 10 CFR 26.719(c)(3), reporting requirements requires that 'If a false negative error occurs on a quality assurance check of validity screening tests, as required in � 26.137(b), the licensee or other entity shall notify the NRC within 24 hours after discovery of the error.' While it initially appears that the blind specimen certification provided by ProTox may be in error, since all three DHHS labs obtained the same testing result, additional investigation is necessary. This report is being made conservatively until the cause can be determined. The licensee informed the NRC Resident Inspector.|
|ENS 49398||1 October 2013 01:05:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 2257 CDT on September 30, 2013 the Callaway Plant Emergency Off-Site Facility (EOF) was declared nonfunctional due to air in-leakage outside acceptance criteria while ventilation is in filtration mode. Efforts are underway to restore the air in-leakage within acceptance criteria at the EOF. If EOF activation is necessary during the period of EOF non-functionality, the Recovery Manager will evaluate the suitability of the facility for the specific conditions of the event. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the unavailability of an emergency response facility. The NRC Senior Resident Inspector has been notified.
Repairs were made to the EOF ventilation system and all required post-maintenance testing has been completed satisfactorily. The EOF has been restored to a functional status. The licensee will notify the NRC Resident Inspector. Notified R4DO (Gepford).
|ENS 49219||27 July 2013 01:14:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On July 26, 2013, at 2349 CDT, the Callaway nuclear power plant declared an Unusual Event due to a fire not extinguished within 15 minutes of control room notification, EAL HU 2.1. The fire was located in the turbine building near the main generator. Concurrent with the fire, the reactor tripped due to a turbine trip. All control rods fully inserted and all reactor coolant pumps (RCPs) tripped. The fire has been extinguished and the licensee is in progress of restoring RCPs. The licensee notified the NRC Resident Inspector, State Emergency Management Agency and Local Authorities. Notified DHS SWO, FEMA, and DHS NICC.
The licensee terminated the Unusual Event at 0101 CDT. Decay heat is being removed via aux feed water from the steam generators to the condenser. Visual inspection determined the location of the fire to be in the phase B generator bus duct. Notified R4DO (Allen), NRR EO (Monninger), IRD (Marshall), DHS SWO, FEMA, and DHS NICC.
The licensee made notifications under 10CFR50.72(b)(2)(iv)(B) (RPS Actuation), 10CFR50.72(b)(2)(xi) (Offsite Notification) and 10CFR50.72(b)(3)(iv)(A) (ESF Actuation - AFW). The licensee will be making a press release and notifying the NRC Resident Inspector. Notified R4DO (Allen).
Upon further review, the licensee believes that the initially reported EAL for the UE notification, HU 2.1, was not applicable. Although indications of a fire were present for greater than 15 minutes, the criteria at Callaway apply to a fire within 50 feet of safety related equipment. There was no safety related equipment within 50 feet of where the fire occurred. The proper EAL classification should have been HU 3.1 due to release of potentially toxic gas or asphyxiant or flammable gas that could impact plant operation. This EAL is applicable due to the heavy smoke release from burning electrical insulation and melted bus and ductwork which prevented access to the turbine building area where the fire took place. The licensee will notify the NRC Resident Inspector of this update. Notified R4DO (Allen).
|ENS 49141||22 June 2013 13:04:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report an event involving an invalid actuation signal affecting the Auxiliary Feedwater (AFW) and Essential Service Water (ESW) systems. Initial conditions on 04/24/2013: refueling outage was in progress, there was no fuel in the reactor vessel (No MODE), a B safety-related train outage was in progress, and the A ESW train was in operation to support cooling of the A train safety-related equipment. Some separation group 2 bistables were in a tripped condition because instrument power bus NN02 was de-energized. At approximately 0400 (CDT) on 04/24/2013, Separation Group 4 DC bus NK04 experienced a ground condition. Plant personnel were using a plant procedure to search for the ground. When breaker NK5409 was opened, some unexpected Engineered Safety Features Actuation System (ESFAS) signals occurred. Opening the breaker removed power to the B ESFAS cabinet. With power removed to the B ESFAS cabinet, the circuit cards that generate cross-train trips failed to a tripped condition (thus generating cross-train trip signals) which resulted in some A train ESFAS actuations, in particular, auxiliary feedwater actuations for the A motor-driven and the turbine-driven AFW pumps. Additionally, an AFW Low Suction Pressure (LSP) circuit card tripped, and when combined with the bi-stable that was in a tripped state because bus NN02 was de-energized, the 2-out-of-3 logic was made up, resulting in an auxiliary feedwater LSP actuation. The LSP actuation resulted in the A Train ESW pump receiving a start signal, and the A motor-driven and the turbine-driven AFW pump suction supply valves receiving an actuation signal to transfer the suction supply from the normal source to the ESW system. Neither the motor-driven nor the turbine-driven auxiliary feedwater pumps started because they had been properly removed from service earlier in the outage. The A ESW pump was already running. No water was transferred from the ESW system to the AFW system since system tagging had been previously placed to isolate the two systems. The actuations were considered invalid because they were caused by opening breaker NK5409 which resulted in loss of power to the B ESFAS cabinet. The Senior Resident Inspector was notified.|
|ENS 49103||10 June 2013 17:06:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation signal affecting the emergency feedwater system. While the plant was in Mode 5 on 4/11/2013, during performance of a maintenance procedure for AMSAC system logic verification, an invalid MDAFAS occurred. (Note: AMSAC is ATWAS Mitigation System Actuation Circuitry and MDAFAS is Motor Driven Auxiliary Feedwater Actuation Signal). Both trains of the Motor Driven Auxiliary Feedwater Pumps (MDAFPs) started. While generation of the actuation signal is an expected result of the procedure, the actuation occurred several steps earlier in the procedure than expected. Additionally, the Control Room Operators were not expecting the MDAFPs to start. The pumps were manually stopped. The actuation was caused by procedural guidance not containing a sufficiently prescribed sequence of activities that should occur when simulating plant conditions leading to the intended actuation of the AMSAC system. The plant was not in a condition where feedwater was required. The Senior NRC Resident Inspector was notified.|
|ENS 49014||9 May 2013 07:18:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||During Callaway refueling outage 19 on 5/8/13 at approximately 1900 hour CDT, water was observed dripping from piping insulation in the overhead by RCS loop 4. Further investigation determined it was near Safety Injection (EP) vent valve EPV0109. A scaffold was built and insulation was removed to perform an inspection. At approximately 0509 hours CDT on 5/9/13, engineering inspected the piping and determined there was a crack in the socket weld where 3/4 inch vent valve EPV0109 is connected to the 'B' train injection piping to RCS loop 4 Cold Leg. The estimated leakage rate through the crack is 6 (six) drops per minute. The configuration of this vent valve is a 3/8 inch flow restrictor socket welded to the six inch piping and a 3/4 inch vent valve socket welded to the flow restrictor. The crack is in the socket weld between the ASME code class 1 flow restrictor socket and the ASME code class 2 vent piping. Callaway plant was in mode 6 with refueling pool level greater than 23 feet above the reactor vessel flange at the time of the discovery. The 'A' RHR train which discharges to RCS loops 1 and 3 Cold Legs is the currently operable RHR train. 'B' RHR train was declared inoperable when the weld crack was identified. Only one RHR train is required to be operable at the present plant Mode of applicability. Repair plans are being developed. Basis for Reportability: This condition constitutes abnormal degradation of a principle safety barrier due to unacceptable welding defects within the primary coolant system. There is a check valve between this leak and the reactor coolant system. Therefore, this is considered unisolable and pressure boundary leakage. The licensee notified the NRC Resident Inspector.|
|ENS 48879||2 April 2013 20:30:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 1707 CDT on 4/2/13 an arc flash occurred at the 'B' safeguards transformer (XMDV24) in the plant switchyard at Callaway. At the time of the flash, ground straps were being placed on the 'B' safeguards transformer which had been removed from service for maintenance. The event resulted in a loss of power to areas/buildings outside the power block. There was no impact to equipment and systems in the plant. Four workers were injured or affected by the flash. The extent of the electrical-related injuries has not been determined. However, based on reports from the scene, all of the workers were conscious and walked away from the scene. One person was transported by helicopter and two by ambulance to a local hospital. The fourth person experienced only a minor injury. The hazard has been isolated and investigation of the cause is in progress. Notifications of this event are planned to be made to OSHA and the Missouri Public Service Commission. The licensee notified the NRC Resident Inspector.
Ameren Missouri issued a press release about the event described above at approximately 1507 CDT on April 4, 2013.
The NRC Resident Inspector was notified. Notified the R4DO (Kellar).
|ENS 48268||2 September 2012 14:45:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||In response to identification of a non-radiological leak from the Neutralization Tank at Callaway Plant today (9/2/2012), notification was made to the EPA National Spill Response Center at 1031 CDT and to the Missouri Department of Natural Resources at 1043 CDT. The leak was initially identified at 0926 CDT. From testing of a sample taken from the tank, the pH of the tank fluid was reported to be 1.9. Initially, the leak was to the ground and into a ditch that is part of a flow path that ultimately leads off site via a storm sewer. However, there is no indication of any of the leakage flowing beyond the site boundary via that pathway since action was promptly taken to divert the leakage to the sump area of the equalization tank (on site). The leakage will thus be collected there until it terminates. At 0926 CDT, the leakage rate was estimated to be approximately 20 gpm; at 1025 CDT the leakage was estimated to be approximately 50 gpm. The leak is at the bottom of the Neutralization Tank, and thus will terminate when the tank is emptied. At 1218 CDT, the fluid level in the Neutralization Tank was at 17%. The initial quantity of fluid in the tank (at the onset of the leak) was approximately 110000 gallons. This spill was reported to offsite organizations, as noted. This event is reportable to the NRC pursuant to 10CFR50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified of the event and this ENS notification. 17% tank fluid level corresponds to approximately 25000 gallons.|
|ENS 48236||27 August 2012 18:36:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||Notification was made to Missouri Department of Natural Resources and the EPA National Spill Response Center of a Neutralization Tank discharge piping leak on 8/27/12 at 1526 hrs. (CDT). Subsequent testing of the leaking fluid at 1430 hrs. revealed the pH of the leaking fluid was 13, which is a characteristic hazard waste. Reportable quantity is 100 lbs. The estimated total volume released was less than 100 gallons, but greater than 100 lbs. This spill was reported to offsite organizations. Therefore, this event is reportable to the NRC per 10CFR50.72(b)(2)(xi). Mitigating strategy is to neutralize the contents of the tank, and the leaking fluid is being absorbed by absorbent materials. The licensee has notified the NRC Resident Inspector.|
|ENS 47930||17 May 2012 12:11:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 0438 (CDT) on Thursday, May 17, the Callaway Plant Emergency Operations Facility (EOF) was declared non-functional when the building's return fan was found not running. Loss of the EOF return fan, results in an inability to maintain a positive pressure on the facility. Efforts are underway to return this fan to service. If an emergency is declared requiring EOF activation while the EOF is non-functional, EOF emergency response personnel will report to their backup locations in accordance with Callaway Plant emergency planning procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the unavailability of an emergency response facility. The NRC Resident Inspector has been notified.
At 1330 (CDT) on 05/17/12, the Callaway Plant EOF was been returned to service. The licensee will notify the NRC Resident Inspector. Notified R4DO (Powers).
|ENS 47885||1 May 2012 22:01:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 1300 on May 1, 2012, as a result of fire water flushing operations, it was observed that the floor drains in the 'A' and 'B' ESF (Engineered Safety Features) 4160 VAC switchgear rooms were draining extremely slow. Engineering was consulted and it was identified that the floor drains in these rooms are credited with preventing any water accumulation in these rooms as a result of internal flooding due to a pipe break. It is expected that the floor drains in the 'A' ESF switchgear room can drain approximately 134 gallons per minute (gpm) and the floor drains in the 'B' ESF switchgear room can drain approximately 208 gpm. With the floor drains partially blocked, a break in the 'A' Essential Service Water pipe in the 'B' ESF Switchgear Room would result in flood levels in the 'B' ESF Switchgear Room to exceed the maximum levels calculated in the current flooding analysis. The higher flood level may result in the inoperability of 'B' train Electrical Switchgear. The 'A' train Essential Service Water supplied equipment would be adversely affected due to the reduced flow. Consequently the pipe break would result in both ESF trains being adversely affected. Compensatory measures have been taken to restore system operability. The NRC Resident Inspector has been notified.
On May 1, 2012, Callaway Plant made an ENS notification in accordance with 10 CFR 50.72(b)(3)(ii)(B) to report the discovery of partially blocked floor drains in the safety-related 4160 V switchgear rooms. At the time of the initial notification, preliminary information indicated that the partially blocked floor drains could have caused a postulated flooding event to adversely affect independent trains of safety-related equipment inside these rooms. Upon further analysis, Callaway Plant staff determined that the pipe break assumed in the flooding calculation of these rooms was overly conservative. Specifically, based on seismic qualifications, the guillotine break of Essential Service Water piping that was originally assumed is not required to be postulated. Instead, a much smaller, through-wall crack of fire protection system piping is the most severe break that must be postulated in the safety-related 4160 V switchgear rooms. An analysis of a postulated flood hazard in these rooms was performed based on the correct water source. Even if considering a complete blockage of the floor drains in these rooms, this analysis demonstrates that a postulated fire protection system piping crack would not have adversely affected safety-related equipment. Based on the results of this analysis, the partially-blocked floor drain condition described in EN 47885 did not meet the criteria for reportability as an unanalyzed condition that significantly degrades plant safety. Event Notification 47885 is hereby retracted. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Greg Pick)
|ENS 47783||28 March 2012 18:26:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||At 1500 on March 28, 2012, Callaway Plant personnel discovered that the installation of a modification on the two 'B' train containment cooler units had inadvertently introduced a potential failure mechanism to the 'B' train containment coolers. Specifically, with the containment cooling fans initially in fast-speed operation, combined with certain initial plant conditions, thermal overload tripping of the coolers could occur. In such an event, during some postulated accidents, slow-speed restart of the containment coolers by the Load Shedding and Emergency Load Sequencing system could be prevented. As a result, the 'B' train containment coolers could be rendered unavailable for a portion of a postulated accident. Thus, the safety function of the 'B' train containment cooling fans cannot be assured when this degraded equipment condition is present and the containment cooling fans are run in fast-speed operation. This condition existed for the 'B' train containment cooling units since they were restored to service from maintenance at 0400 on March 15, 2012. Upon identification of this condition, the 'B' train containment cooling fans were switched from fast-speed to slow-speed operation and restored to operable status at 1515. This action precludes this degraded equipment condition from adversely affecting containment cooling fan function during an accident. Concurrent with this condition, the opposite train of containment coolers was removed from service for scheduled maintenance at 0505 on March 27, 2012. As a result, from 0505 on March 27, 2012 until 1515 on March 28, 2012, the safety function of the containment cooling system could not be assured for certain postulated accident conditions. The NRC Resident Inspector has been notified.|
|ENS 47749||16 March 2012 14:29:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||On March 11, 2012, a maintenance test of the Technical Support Center (TSC) diesel generator was performed. This diesel generator supplies backup A/C power to the TSC in the event of a loss of normal power. During this test, with the diesel generator supplying power to the TSC, two unexpected momentary losses of normal lighting to the TSC were observed. Following this event, the TSC diesel generator was declared non-functional. High-priority actions to troubleshoot and repair the TSC diesel generator were taken following this event. However, as of 1200 CDT on March 16, 2012, troubleshooting activities are still ongoing, and the cause of the event has not yet been identified. Normal power to the TSC has been available throughout this event, and the TSC is considered functional. However, due to the duration of the TSC diesel generator non-functionality, Callaway Plant is reporting this condition in accordance with 10 CFR 50.72(b)(3)(xiii) for a loss of emergency assessment capability. Please note that, in the event of a complete loss of power to the TSC concurrent with a plant emergency, the designated backup facilities will be used in accordance with the Callaway Plant emergency preparedness program procedures. The NRC Resident Inspector has been notified.|
|ENS 47426||9 November 2011 21:43:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||On November 9, 2011 at 1715, Callaway Plant staff determined that a postulated design basis fire event in Fire Area C-1, (Control Building, elevation 1974, ESW Pipe Space, Room 3101) could result in failure of the High Density Polyethylene (HDPE) piping in the Essential Service Water (ESW) system. In 2008-2009 timeframe, Callaway Plant implemented a modification which replaced underground large bore carbon steel ESW piping with HDPE piping. Four short sections of this HDPE piping enter the Control Building and interface with steel piping in Room 3101. During the design of the modification, it was not recognized that a fire barrier should be installed to protect the HDPE piping from the consequences of a fire. As a result of the missing fire barrier, a postulated fire could cause a failure of one train of the large bore HDPE piping located within the fire area. The resultant pipe failure could lead to flooding in the fire area that could adversely affect both trains of ESW equipment required to achieve and maintain safe shutdown. An hourly fire watch has been imposed as a compensatory measure for this condition in accordance with the approved fire protection program. This condition is reported in accordance with 10 CFR 50.72 (b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified.|
|ENS 47315||3 October 2011 10:24:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 0850 (CDT) on Monday, October 03, the Callaway Plant Technical Support Center (TSC) is undergoing planned maintenance to repair MTSUB7001, Manual Transfer Switch between normal and emergency power. This maintenance is currently scheduled to last for 1 day, at which time the TSC will be restored to service. During this period, the TSC will be without either normal or emergency power, thus rendering it non-functional. If an emergency is declared requiring TSC activation while the TSC is non-functional, TSC emergency response personnel will report to their backup locations in accordance with Callaway Plant emergency planning procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the planned unavailability of an emergency response facility.
Callaway Plant Technical Support Center (TSC) normal power was restored at 1751 CDT. The TSC is functional and available as an emergency response facility. The NRC Resident Inspector has been notified. Notified R4DO (Whitten).
|ENS 47291||25 September 2011 19:48:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 1804 on Sunday, September 25, the Callaway Plant Technical Support Center (TSC) will undergo planned maintenance to replace the building's heating, ventilation, and air conditioning (HVAC) system. This maintenance is currently scheduled to last for approximately five days, at which time the TSC will be restored to service. During this period, the TSC's HVAC system will not be able to provide positive pressure to the TSC, thus rendering it non-functional. If an emergency is declared requiring TSC activation while the TSC is non-functional, TSC emergency response personnel will report to their backup locations in accordance with Callaway Plant emergency planning procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the planned unavailability of an emergency response facility. The NRC Resident Inspector has been notified.
At 1804 CDT on September 25, 2011, the Callaway Plant Technical Support Center (TSC) underwent planned maintenance to replace the building's heating, ventilation, and air conditioning (HVAC) system. While this maintenance was being performed, the TSC was unable to maintain positive pressure within the building, thus rendering it non-functional. This ENS update is to document that, at 1800 on September 30, 2011, the Callaway Plant TSC was returned to service following successful completion of planned maintenance on the building HVAC system. The NRC Senior Resident Inspector has been notified. Notified R4DO (Werner).
|ENS 47275||18 September 2011 12:41:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
An Alert was declared at Callaway Nuclear Plant at 1056 (CDT) due to EAL HA3.1. Access to an Auxiliary Building area which is prohibited due to release of toxic gas which jeopardizes operation of systems required to maintain safe operations or safely shutdown the reactor. EAL HU3.1 (Unusual Event) is also applicable at the same time. The cause of the toxic gas release was a Freon gas leak from the 'A' Control Room air conditioner unit. The licensee has notified the NRC Resident Inspector and state and local government. Also notified USDA (Pitt) and HHS (Emerson).
At 1737 CDT, Callaway Nuclear Plant exited from the Alert for EAL HA3.1, and exited from the Unusual Event for EAL HU3.1. The plant continues to operate at 100% power in Mode 1. There was no radiological release due to this event. Additionally, a press release will be performed after the event closeout. The licensee has notified the NRC Resident Inspector and state and local government. Notifications were also given to R4DO (Pick), NRR EO (Giitter), IRD-MOC (Morris), HQ PAO (Hayden), DHS (Gates), FEMA (Via), DOE (Foote), USDA (Sanders) and HHS (Hoskins).
|ENS 47084||21 July 2011 16:45:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||On July 21, 2011 at 1100 (CDT) Callaway Plant staff determined that a design deficiency could adversely affect the 'B' Train of Essential Service Water (ESW) in the event of a Control Room fire. 'B' Train is the credited train for completion of a post-fire safe shutdown as a result of a Control Room evacuation. As a result of this deficiency, normally closed valve EFHV0060 could spuriously open during a postulated control room fire. EFHV0060 is located on the ESW return line from the 'B' Component Cooling Water (CCW) heat exchanger. If EFHV0060 spuriously opened as a result of this postulated fire, the flow balance in the 'B' Train of the ESW system would be affected. In this scenario, cooling water flow to other essential components could be reduced to below the minimum requirements. A fire watch has been imposed as a compensatory measure for this condition. Additionally, EFHV0060 has been closed and de-energized to preclude spurious opening in the event of a postulated control room fire. This condition is reported in accordance with 10 CFR 50.72 (b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified.|
|ENS 46814||3 May 2011 17:55:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||The following is a non-emergency notification in accordance with 10CFR50.72(b)(2)(xi), Offsite Notification. At 1036, May 3, 2011, Callaway Plant notified the Missouri Department of Natural Resources (DNR) of an issue with the potable water supply system at the facility. This notification to the Missouri DNR was made in accordance with 10 CSR 60-4.055, 'Disinfection Requirements,' due to circumstances that adversely affect the quality of potable water. Specifically, loss of function of the potable water chlorine pumps resulted in residual chlorine in the potable water system falling below the levels specified by the regulation. DNR has not restricted drinking of Potable Water at Callaway Plant. Required compensatory actions have been initiated, including planned repairs of the potable water chlorination system. The NRC Resident Inspectors have been notified.|
|ENS 46715||31 March 2011 20:50:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
In response to a condition identified in late 2010 concerning the control and removal of hazard barriers in the plant, a review of the basis and analysis for high energy line breaks (HELBs) and the barriers for protecting against such events has been underway at Callaway in accordance with the plant's corrective action program. While following up on a question from the NRC Resident Inspector, and as a result of an additional question from the Nuclear Oversight organization at Callaway, it was identified that non-safety piping located in the valve room associated with the Refueling Water Storage Tank (RWST) could potentially (make) all four RWST low water level pressure transmitters inoperable in the event of a malfunction of the non-safety piping concurrent with a design-basis loss-of-coolant accident (LOCA) and/or following a seismic event. The RWST water level transmitters (which are located in the RWST valve room) perform a safety-related function for the emergency core cooling system (ECCS) by automatically swapping suction sources for the ECCS during a LOCA from the RWST to the containment sumps when a low water level condition is reached in the RWST. These instrument channels are required to be OPERABLE in Modes 1, 2, 3 and 4 per Callaway Technical Specification 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation.' The subject non-safety piping delivers steam supplied by the Auxiliary Steam system to (and from) heaters surrounding the RWST for maintaining RWST contents above the minimum required temperature during winter conditions. The piping passes through the RWST valve room containing the noted RWST water level transmitters which were designed only for a mild environment. It has been identified, however, that the non-safety Auxiliary Steam piping constitutes a high energy line and that its failure could create harsh (hot and wet) conditions in the valve room to which the RWST water level instrumentation was not designed. Per the Callaway FSAR, where non-safety piping interfaces with safety-related piping or systems, the design must be such that failure of the non-safety piping does not adversely affect the safety function(s) of the interfacing safety-related piping or system (since non-safety piping may be assumed to malfunction in conjunction with a design-basis accident). In this case, and based on a conservative interpretation of the FSAR, if the non-safety piping in the RWST valve room is assumed to malfunction (i.e., break), a failure of the RWST instrumentation could occur, thereby preventing the ECCS suction swap over from occurring as required or assumed for LOCA mitigation. This condition required declaring all four RWST water level channels inoperable. In light of recognizing that the RWST water level instruments could be subject to a harsh environment when they were only designed for a mild environment, and could thus fail as a result, this condition represents an unanalyzed condition that significantly degrades plant safety. With regard to the impact on the required ECCS suction swap over function that requires the RWST water level channels to be operable, the inoperability of all four instrument channels is a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe condition, remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident. Upon declaring the RWST water level instrument channels inoperable, TS Limiting Condition for Operation (LCO) 3.0.3 was entered at time 1432 CDT on 3/31/2011. At 1634 CDT, the Auxiliary Steam system was isolated and depressurized. This removed the energy that could be released from a break in the non-safety piping, thereby restoring Operability for the RWST water level instruments. The NRC Senior Resident Inspector was notified.
On March 31, 2011, event notification EN 46715 documented that a harsh environment from a postulated High Energy Line Break (HELB) in the Refueling Water Storage Tank (RWST) valve room could affect RWST level transmitters. These level transmitters provide RWST water level indication in the main control room, which is identified as a safe shutdown function in the Callaway FSAR. They also provide low RWST water level signals for effecting automatic swap over of suction sources for the Emergency Core Cooling System in the event of a loss-of-coolant accident (LOCA). This break may be postulated to occur on non-safety related auxiliary steam lines that run through the RWST valve room and on to the RWST heaters. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46715 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that the sections of auxiliary steam piping in the RWST valve room are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside the RWST valve room, and a postulated auxiliary steam line break outside of the room would not adversely affect the RWST level transmitters. Since none of the postulated break locations are located inside the RWST valve room, there exists reasonable assurance that the RWST level transmitters would have remained capable of performing their safe shutdown function following a postulated break of the subject auxiliary steam lines. Further, there is no adverse effect on the assumed response to a postulated design basis LOCA since a hazard (such as a break in an auxiliary steam line) is not assumed to occur concurrently with the LOCA. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46715 is hereby retracted. The NRC Senior Resident Inspector has been notified. Notified R4DO (Haire).
|ENS 46693||24 March 2011 06:02:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
While performing an extent of condition review of high energy line break (HELB) analyses, a detailed review of the auxiliary steam system was being performed. During this review, sections of pipe that run through rooms 1206/1207 in the Auxiliary Building were identified that have design ratings indicating that they could possibly be classified as high energy lines. The pipes were verified to have not been considered in the current HELB analyses. This condition affects pressure transmitters ALPT0037, 38, & 39 which are not qualified for operation in a harsh environment. These pressure transmitters provide the Auxiliary Feedwater Pump (AFW) Suction Transfer signal on low suction pressure from the non safety Condensate Storage Tank to the Safety Related supply (Essential Service Water). Technical Specification (TS) 3.3.2-6.h bases state: "since these detectors are in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties." Based upon the above bases, with the identified aux steam lines in service, the pressure transmitter's operability could not be assured. This represented an unanalyzed condition and had the potential to affect equipment used for accident mitigation. TS 3.0.3 was entered at time 2354 (CST) on 3/23/2011. At 0009 (CST) on 3/24/2011, Aux Steam valves FBV0158, FBV0I48, FAV0002, and FAV0003 were isolated, removing the HELB concern (TS 3.0.3 was exited at this time). These are the active feed (isolation valves) to the lines passing through the Aux Building Rooms 1206/1207. The licensee notified the NRC Resident Inspector.
On March 23, 2011, event notification EN 46693 documented that a harsh environment from a postulated High Energy Line Break (HELB) could affect pressure transmitter ALPT0037, 38 and 39. These pressure transmitters provide the Auxiliary Feedwater Pump suction transfer signal on low suction pressure from the Condensate Storage Tank to the safety-related water supply (Essential Service Water). This break was postulated to occur on auxiliary steam lines in Auxiliary Building rooms 1206 And 1207. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46693 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that these sections of auxiliary steam piping are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside rooms 1206 and 1207. Analysis has been performed on these auxiliary steam lines for the remaining break locations that are required to be postulated. This analysis demonstrates reasonable assurance that safety related equipment, including pressure transmitters ALPT0037, 38 and 39, would have performed their safety functions following a postulated break of these auxiliary steam lines. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46693 is hereby retracted. The NRC Resident Inspectors have been notified. Notified the R4DO (Shannon).
|ENS 46597||7 February 2011 17:18:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On December 15, 2009, Callaway Plant reported a condition in which valve FBV0146, an isolation valve on an auxiliary steam line in the Auxiliary Building, was found to be kept normally open. With FBV0146 open, the auxiliary steam line downstream of FBV0146 must be considered a high energy line. This configuration was not consistent with the analysis of record for High Energy Line Break (HELB) events. Valve FBV0146 was closed upon discovery of the condition. This condition was reported under EN # 45571 as an unanalyzed condition that significantly degrades plant safety. EN #45571 was then retracted on January 21, 2010 when analysis showed that no safety-related components would be rendered inoperable in a postulated HELB event due to the condition. Based on this analysis, FBV0146 was reopened. Subsequent review of this condition now shows that, with FCV0146 open, a harsh environment from a postulated HELB downstream of FBV0146 could be transmitted to other areas of the Auxiliary Building. This would occur via a flow path through door gaps and an Auxiliary Building elevator shaft. This flow path had not been considered by the previous analysis. The areas that could be affected by a postulated line break contain safe shutdown equipment (such as equipment for the Component Cooling Water system) that is not assumed to experience harsh conditions. Because of the potential impact on this equipment, this condition is considered to have met the criteria for reporting under 10 CFR 50.72(b)(3)(ii)(B). FBV0146 is now closed. This review was performed as part of the ongoing evaluation of HELB Program deficiencies described in Callaway Plant License Event Report (LER) 2010-009-00. The NRC Resident Inspector has been notified. The auxiliary steam line in the Auxiliary Building feeds non-safety related components.
The licensee is retracting this report based on the following: On 02/07/2011, EN #46597 documented that a harsh environment from a postulated High Energy Line Break (HELB) could be transmitted to areas of the Auxiliary Building not qualified for harsh environments. This break was postulated to occur on an Auxiliary Steam line downstream of valve FBV0146 when FBV0146 is open. This condition was initially reported as an unanalyzed condition that significantly degraded plant safety. When EN #46597 was reported, the analysis of the Auxiliary Steam line included postulated break locations at any intermediate fitting, welded attachment, or valve. Subsequent analysis shows that this section of Auxiliary Steam piping is able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, line breaks are only assumed to occur at terminal ends and at the locations specified for ASME Class 2 and 3 piping. Breaks at intermediate fittings, welded attachments, and valves are not required to be assumed. Postulated breaks of this Auxiliary Steam line at the locations described above have been analyzed. This analysis demonstrates reasonable assurance that safety-related equipment would have performed their safety functions following a postulated break of this Auxiliary Steam line. Therefore, this condition is not an unanalyzed condition that significantly degrades plant safety and does not meet the reporting requirements of 10 CFR 50.72(b)(3)(ii)(B). Event notification #46597, made on 02/07/2011, is hereby retracted. The NRC Resident Inspectors have been notified. Notified R4DO (O'Keefe).
|ENS 46581||1 February 2011 17:51:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
At 1430 on 2-1-2011, due to worsening road conditions from heavy snowfall, a determination was made that there is a loss of plant access for some individuals and some impairment of evacuation routes. This meets the criteria for an immediate notification (8 Hour) per 10 CFR 50.72 (b)(3)(xiii). There is no impact on plant operation, all T/S (Technical Specification) required minimum staffing requirements are satisfied. The weather is predicted to be hazardous until the evening of 2/2/2011. The Missouri Department of Transportation reported they have a plow truck running from Callaway Plant to Jefferson City, MO. They also reported visibility is extremely poor due to blowing snow. The Missouri State Emergency Management Agency duty officer has been notified of the degraded Callaway ERO (Emergency Response Organization) response time and evacuation route issues. The NRC Resident Inspector was notified of this event by the licensee. Two shifts of personnel will be on site throughout this event. Diesel fuel, food, and water are above required minimums.
As of 1600 (CST) on 2/2/2011, road conditions have significantly improved. Primary evacuation routes have been reported to be passable. There is reasonable assurance of normal ERO response time. NRC Resident Inspector has been informed of the intent of this change and will be notified of this communication. Notified R4DO (Howell).
|ENS 46156||5 August 2010 15:39:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
The Technical Support Center (TSC) will be without power during performance of planned maintenance activities starting at approximately 1500 CDT on August 5, 2010. The maintenance activities, including electrical isolation and restoration, are expected to last approximately 12 hours. Contingency plans for emergency (response) situations have been established and the Emergency Response Organization members have been notified of their contingency actions. This event is reportable per 10CFR50.72(b)(3)(xiii) since this constitutes a loss of an emergency response facility for the duration of the maintenance activities. Region IV was notified of this planned outage. The licensee notified (State and local agencies and) the NRC Resident Inspector.
The Technical Support Center was restored to functional status at 0407 EDT. All systems verified operational. The licensee will notify State and local agencies and the NRC Resident Inspector. R4DO (Hagar) notified.
|ENS 45989||8 June 2010 19:57:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop||At approximately 0430 CDT on June 8, 2010, during restoration following an addition of 10 gallons of hydrazine and a flush of 2 gallons of demineralized water to the Condensate Storage Tank (CST), a water-hydrazine mixture began to leak from check valve KHV0179. KHV0179 is a nitrogen supply check valve that can also be used for hydrazine addition. Chemistry technicians estimated that a water-hydrazine mixture on the order of 2 gallons leaked through KHV0179 before the line could be isolated. Samples taken from the atmosphere above the spill contained 0.25 ppm hydrazine. The fluid on the ground was measured to contain 15% hydrazine. The Department of Natural Resources (DNR) was notified of this event at approximately 1000 CDT on June 8, 2010. The Nuclear Regulatory Commission (NRC) Resident Inspectors will be notified.|
|ENS 45837||13 April 2010 13:17:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
On April 13, 2010 at 1200 CDT, Callaway Unit 1 declared an Unusual Event due to identified leakage of greater than 25 gallons per minute (EAL SU6.1). During the flushing of the 'A' Chemical and Volume Control System mixed bed demineralizer, a leak occurred from the vent valve exceeding 25 gallons per minute for approximately 5 minutes. The leak was isolated and the Unusual Event was terminated at 1223 CDT. No personnel contamination or outside release occurred. The licensee has notified the State and local authorities and the NRC Resident Inspector.
Callaway Plant declared an Unusual Event at 1200 CDT on 4/13/2010. The cause of the event was Emergency Action Level (EAL) SU6.1, identified leakage greater than 25 gallons per minute (gpm). When attempting to place the 'A' Chemical and Volume Control System (CVCS) mixed bed demineralizers in service, a drop in the Volume Control Tank (VCT) level was noted by the Reactor Operator (RO). Technical Specification (TS) 3.4.13 Condition A was entered at 1033 CDT upon identification of identified leakage of the Reactor Coolant System (RCS) greater than 10 gpm. TS 3.4.13 Condition A requires reduction of leakage to limits within 4 hours. The leakage was verified to be stopped at 1038 CDT, at which time TS 3.4.13.A was exited. The system was restored to a normal alignment. Upon review, it was determined that the VCT level dropped approximately 125 gallons in 5 minutes. EAL SU6.1 was declared at 1200 CDT and closed at 1223 CDT. The state and local counties (Callaway, Gasconade, Montgomery, and Osage) were notified of the event at 1210 CDT and of the event closeout at 1227 CDT. The NRC Resident Inspector was notified of the event. A news release will be made by Ameren Corporate Communications. Notified R4DO (Spitzberg), NRR EO (Nelson), and IRD (Marshall).
The declaration of Unusual Event SU6.1, RCS leakage, on 4/13/10 is being retracted, because the declaration was inaccurate. The leak was an isolable intersystem leak of the CVCS, not RCS leakage, and the leak was stopped within the 15 minute time period allowed to determine EAL applicability. Therefore, the leakage did not meet the Initiating Condition for the EAL. The NRC Resident Inspectors have been notified. Notified R4DO (J. Whitten).
|ENS 45747||5 March 2010 16:32:00||Callaway||NRC Region 4||Westinghouse PWR 4-Loop|
Valve FBV0147, Boric Acid Batch Tank Auxiliary Steam Supply Isolation Valve, is credited with being closed in the Callaway FSAR. This eliminated the need to analyze lines FB-081-HBD and FB-082-HBD for high energy line breaks (HELB). However, FBV0147 was found to be kept normally open to allow steam service for the boric acid batching tank. This is contrary to the normal position assumed in the FSAR and HELB analyses. With valve FBV0147 open, the lines must considered high energy lines. The lines are in the auxiliary building and they traverse rooms containing several components including the flow transmitters for Residual Heat Removal (RHR) to train `A' accumulator injection supply header, RHR train `A' and 'B' SIS hot leg recirculation supply header, and several safety related auxiliary feedwater components. These instruments are used to provide indications during post-accident conditions. This configuration is therefore outside of current HELB analysis and potentially represents a condition that significantly degrades plant safety. The condition was identified to Operations at 0740. Valve FBV0147 was closed at 0810. At 1325 CST, the issue was determined to be reportable. The NRC Resident Inspector has been notified.
On 03/05/2010, EN #45747 provided notification that valve FBV0147 was found to be kept normally open to allow steam service for the boric acid hatching tank. This configuration was not consistent with the normal position assumed in the FSAR and HELB analyses. An engineering evaluation was subsequently performed for the auxiliary steam inlet and outlet piping for the boric acid batching tank. This valuation identified four postulated break locations which have all been analyzed. The analyses determined that all safety-related components in the affected rooms would be able to perform their safety functions in the event of a line break. Pipe whip, jet impingement, compartment temperature and over pressurization, flooding, and internal missiles resulting from a postulated line break would not prevent any safety-related equipment from performing its design function. As supported by this evaluation, this condition does not meet the criteria for an unanalyzed condition that significantly degrades plant safety as stated in 10 CFR 50.72(b)(3)(ii)(B). Therefore, the notification made on 03/05/2010 is hereby retracted. The NRC Resident Inspector will be notified. R4DO (Farnholtz) notified.