Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 56968 | 15 February 2024 08:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip/Reactor Trip | The following information was provided by the licensee via email: At 0247 CST on 2/15/2024, Callaway Plant was in mode 1 at approximately 100 percent power when a turbine trip and reactor trip occurred. All safety systems responded as expected with the exception of an indication issue on the feedwater isolation valves, which were confirmed closed. A valid feedwater isolation signal and auxiliary feedwater actuation signal were also received as a result of the reactor trip. The plant is being maintained stable in mode 3. All control rods fully inserted from the reactor trip signal and decay heat is being removed via the auxiliary feedwater system and steam dumps. The NRC Resident Inspector was notified. | |
| ENS 56723 | 6 September 2023 13:30:00 | 10 CFR 26.719, FFD Reporting requirements | Controlled Substance Found in the Protected Area | The following information was provided by the licensee via email: On 09/06/2023, at approximately 0830 (CDT), a bottle of vanilla extract, intended for use in cooking, with an alcohol content of greater than 0.5 percent by volume was found in the protected area. An immediate extent-of-condition search of other kitchen areas within the protected area identified four additional bottles of vanilla extract or imitation vanilla extract, for a total of five bottles identified. The alcohol content by volume (ABV) of these extracts ranged from an unlisted percentage (with ethyl alcohol as a listed ingredient) up to 41 percent ABV. The volume capacities of the bottles ranged from 2 to 8 fluid ounces, with varying volumes of remaining contents. | |
| ENS 56561 | 6 June 2023 17:33:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report | The following information is summary provided by the licensee via email: A non-licensed supervisor was found to have falsified fitness for duty reports for a period of two months. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. | |
| ENS 56401 | 9 March 2023 17:53:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Injured Worker | The following information was provided by the licensee via email: At approximately 1153 CST on March 9, 2023, a contract worker was injured requiring an ambulance transport to a local hospital. Union Electric (Ameren Missouri) subsequently learned that the individual requires an overnight hospital stay. This event is reportable to the Occupational Safety and Health Administration per 29 CFR 1904.39(a)(2) by the contract worker's employer and is reportable to the Missouri Public Service Commission in accordance with Missouri regulation 20 CSR 4240-3.190(3)(A). This notification is being made to the NRC pursuant to 10 CFR 50.72(b)(2)(xi) due to other government notification. The individual was not working in a radiologically controlled area. The NRC Senior Resident Inspector has been notified of this event. | |
| ENS 55972 | 2 May 2022 04:05:00 | Other Unspec Reqmnt 10 CFR 50.73(a)(1), Submit an LER | Invalid Specified System Actuation | The following information was provided by the licensee via phone and email: This non-emergency notification is being made pursuant to the provisions of 10 CFR 50.73(a)(1) to report the occurrence of an invalid automatic actuation satisfying the reporting criterion of 10 CFR 50.73(a)(2)(iv)(A), specifically for the actuation of one train of the Essential Service Water (ESW) system that occurred on May 1, 2022. On May 1, 2022, with the plant shut down and the core offloaded, control room personnel were performing a fast power transfer from Engineered Safety Feature (ESF) transformer XNB02 to ESF transformer XNB01. In anticipation of this activity, the `B' load shedder and emergency load sequencer (LSELS) had been removed from service. Also, at the time, a portion of the `A' ESW train was isolated to support performance of a local leak rate test (LLRT) of a containment isolation valve in the affected portion of `A' ESW train piping. Service Water was supplying cooling water flow to `A' train loads (in lieu of ESW cooling water). When the power transfer was performed, an unexpected automatic start of the `A' ESW pump, along with some associated, automatic valve repositioning, occurred. The actuation occurred due to inadvertent satisfaction of automatic start logic for the ESW pump. The logic is intended to detect loss of ESW flow when the opposite train LSELS isolates Service Water during an undervoltage condition on a safety bus. The flow transmitter involved in the actuation is situated in a portion of the ESW piping that was isolated for the LLRT. The low-flow signal from the transmitter was consequently not reflective of low cooling water flow to plant loads in light of the fact that cooling water flow was being supplied to plant loads and the transmitter was locally isolated. In regard to the ESW train actuation, therefore, although the undervoltage signal was considered a valid signal due to the voltage drop caused by the fast transfer activity, the low-flow signal from the noted transmitter was considered to be invalid. For this invalid actuation, it was concluded that the actuation was not part of a pre-planned sequence, that the affected system was not properly removed from service during the occurrence, and that the safety function had not already been performed relative to the occurrence. (The) NRC Resident Inspector has been notified and an email of this report has been sent to hoo.hoc@nrc.gov. | |
| ENS 55848 | 19 April 2022 15:30:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report - Failed FITNESS-FOR-DUTY Test | The following information was provided by the licensee via email: A contract employee supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | |
| ENS 55698 | 7 January 2022 18:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip / Reactor Trip | The following information was provided by the licensee via email: At 1223 CST on January 7, 2022, Callaway Plant was in Mode 1 at approximately 100 percent power when a turbine trip / reactor trip occurred. All safety systems responded as expected with the exception of an indication issue with the 'B' Feedwater Isolation Valve, which was confirmed closed. A valid Feedwater Isolation Signal and Auxiliary Feedwater Actuation Signal were also received as a result of the reactor trip. The plant is being maintained stable in Mode 3. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. The NRC Senior Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The plant is in a normal shutdown electrical lineup. | |
| ENS 55538 | 21 October 2021 18:03:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report | A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. | |
| ENS 55259 | 14 May 2021 18:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Individual Fall from Scaffold Ladder - Other Government Agency Notification | At approximately 1300 CDT on 05/14/2021, a contract worker, who was using a scaffold ladder to access their work area on the iso-phase bus duct system for the main transformers at the Callaway plant, fell approximately 27 feet to the ground. An ambulance was dispatched to transport the individual to a local hospital. Union Electric (Ameren Missouri) subsequently learned that the event caused the individual to have a serious injury that required an overnight hospital stay. This event is reportable to OSHA per 29 CFR 1904.39(a)(2) by the contract worker's employer and is reportable to the Missouri Public Service Commission in accordance with Missouri regulation 20 CSR 4240-3.190(3)(A). This notification is being made to the NRC pursuant to 10 CFR 50.72(b)(2)(xi) due to other government notifications that will occur as a result of a situation related to the health and safety of onsite personnel. The NRC Senior Resident Inspector has been notified of this event. The individual was not working in a contamination area. | |
| ENS 55247 | 10 May 2021 17:50:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Discovered Vulnerability in Fitness for Duty Program | On May 10, 2021, Callaway determined that a violation of 10 CFR 26.4(c) occurred. A licensee employee was assigned to perform Emergency Response Organization (ERO) duties that required that employee to be subject to the Fitness for Duty (FFD) program. However, the individual had been removed from the FFD program. The individual's unescorted access to the plant had been temporarily removed, but the individual was still required to report to the Emergency Operations Facility in accordance with the emergency plan procedures. The individual's ERO qualification has been deactivated. A review determined that this condition did not apply to any other ERO responders. This discovery is reported pursuant to 10 CFR 26.719(b)(4). The NRC Senior Resident Inspector has been notified of the event. | |
| ENS 55049 | 24 December 2020 18:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Turbine and Reactor Trip | At 1235 CST on December 24, 2020, Callaway Plant was in Mode 1 at approximately 90 percent power when a turbine trip/reactor trip, from a vital main generator trip signal, occurred. All safety systems responded as expected with exception of an indication issue with the 'B' Feedwater Isolation Valve, which was confirmed closed, and one intermediate range nuclear instrumentation channel which failed off-scale low following the trip. A valid Feedwater Isolation Signal and Auxiliary Feedwater Actuation Signal were also received as a result of the plant trip. The plant is being maintained stable in Mode 3. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. The NRC Senior Resident Inspector was notified. | |
| ENS 54916 | 27 September 2020 07:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | At 0203 (CDT) on September 27, 2020 the plant was in Mode 1 at 98 percent power when a turbine trip/reactor trip (from a generator trip signal) occurred. All systems responded as expected. A Feedwater Isolation signal was received due to the reactor trip with RCS average temperature less than 564 degrees Fahrenheit. The Auxiliary Feedwater system started on a valid actuation signal to restore and maintain steam generator levels. The plant is being maintained stable in Mode 3 with no complications. The NRC Resident Inspector has been notified of the reactor trip. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. | |
| ENS 54806 | 30 July 2020 13:15:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | EN Revision Imported Date : 8/18/2020 EMERGENCY PROCEDURE ERROR POTENTIALLY PREVENTING TIMELY COMPLETION OF EMERGENCY CORE COOLING SYSTEM RECIRCULATION ALIGNMENT At 0815 CDT on 7/30/2020, it was determined that a procedural error in emergency procedure ES1.3, Transfer to Cold Leg Recirculation, could delay realignment from emergency core cooling system (ECCS) injection phase to recirculation phase under lower plant operational modes. It is noted this scenario is postulated to occur only when the boron dilution mitigation system is operable in lower modes of operation as per Technical Specification 3.3.9 (required operable in Mode 2 (below P-6), 3, 4 and 5). Current plant conditions require this feature nonfunctional so this issue does not impact current plant conditions. This condition is not bounded by existing design and licensing documents; however, it poses no current impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
Event Notification (EN) 54806, made on 7/30/2020, is being retracted because re-evaluation performed subsequent to the notification has demonstrated that the error in Emergency Operating Procedure ES1.3 would not have resulted in a condition outside of the current licensing basis analyses of record for the Callaway Plant. This re-evaluation addressed core effects, containment pressure-temperature and radiological consequences analyses, documented in the plant's corrective action program. The re-evaluation has led to the conclusion that the procedural error in ES1.3 would not have prevented any system required to be OPERABLE by the Technical Specifications from performing its specified safety functions. With all systems capable of performing their specified safety functions, the current licensing basis analyses of record for Callaway Plant remain valid and bounding. Based on these considerations, it has been determined that the condition reported in EN 54806 did not result in the plant being in an unanalyzed condition that significantly degraded plant 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 this Event Notification retraction. Notified R4DO (Taylor) | ||
| ENS 54639 | 4 April 2020 06:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | At 0114 (CDT) on April 04, 2020 the plant was in Mode 1 at 100 percent power when a Digital Feedwater Trouble alarm was received unexpectedly. Operators identified a full feedwater demand signal and lowering level in the 'C' Steam Generator. At 0115 a Reactor Trip signal for Low Steam Generator Level was received. All systems responded as expected. A Feedwater Isolation signal was received due to the reactor trip with RCS average temperature less than 564 degrees Fahrenheit. Auxiliary Feedwater started on a valid actuation signal to restore and maintain Steam Generator levels. The Plant is being maintained stable in Mode 3 with no complications. The NRC Resident Inspector has been notified of the Reactor Trip. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. | |
| ENS 54069 | 17 May 2019 04:03:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | En Revision Imported Date 6/6/2019 | 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 04:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | 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 06:37:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Undervoltage to a Safety Related Bus Resulting in Valid System Actuation | 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 16:50:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty - Alcohol Consumed in Protected Area | 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 06:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | 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 05:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Chemical Spill Offsite | 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 05:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | En Revision Imported Date 8/1/2018 | 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 53426 | 23 May 2018 12:30:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | On May 23, 2018 Callaway determined that a violation of one provision of the site's Fitness for Duty (FFD) policy occurred. FFD pre-access testing confirmed a test failure for alcohol. The violation was committed by a non-licensed supervisory employee. The individual did not hold unescorted access to the plant but did perform behavioral observation program (BOP) duties. The BOP qualification has been removed. The NRC Resident Inspector has been notified by the licensee. | |
| ENS 53388 | 7 May 2018 18:35:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition 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 (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident." The licensee notified the NRC Resident Inspector of this condition. | |
| ENS 53223 | 20 February 2018 18:25:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | All Three Auxiliary Feedwater Pumps Inoperable Due to Helb Door Being Open | At 1225 CST, all three Auxiliary Feedwater (AFW) pumps were declared inoperable at the Callaway Plant upon discovery that a door (DSK13311) credited for protection of equipment from the effects of a high-energy line break (HELB) hazard had come partially open due to vibration harmonics from the running turbine-driven Auxiliary Feedwater pump (TDAFP). Immediate investigation identified that play in the mechanism that holds the door closed had rendered it susceptible to movement from the vibration harmonics. The affected HELB door specifically protects safety-related instruments that provide a swap-over signal upon detection of a low suction pressure condition for the AFW pumps and thereby automatically effect a suction transfer for the AFW pumps from the condensate storage tank (normal/standby source) to the Essential Service Water (ESW) system (credited safety-related source). All of the AFW pump suction transfer instrument channels were declared inoperable. Per Technical Specifications (TS) 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation,' the applicable Condition(s) and Required Action(s) for inoperable AFW pump suction transfer instrumentation only addresses a single channel being inoperable. Thus, the condition of having all three instrument channels inoperable required entry into TS Limiting Condition for Operation (LCO) 3.0.3. At the same time, however, with the automatic suction transfer capability rendered inoperable, all three AFW pumps, i.e., the TDAFP and the 'A' and 'B' motor-driven AFW pumps, were declared inoperable. Although LCO 3.0.3 was applicable, entry into the Required Actions of LCO 3.0.3 was suspended per the Note attached to Required Action E.1 of TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' which states, 'LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.' At 1336 CST, Operations took actions to prevent the TDAFP from running, so the remaining (AFW Pumps) could be returned to Operable status. Operations then declared the affected instrumentation and the 'A' and 'B' motor-driven AFW pumps Operable. This allowed LCO 3.0.3 and Conditions A, B, D, and E under TS 3.7.5 to be exited. With only the TDAFP inoperable, TS 3.7.5 Condition C and its Required Actions remain in effect. Due to the degraded HELB door rendering all three AFW pumps inoperable, the unidentified condition is being reported as an unanalyzed condition that significantly degraded plant safety (per 10 CFR 50.72(b)(3)(ii)(B)) as well as a condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident (per 10 CFR 50.72(b)(3)(v)(A), (B), and (D), respectively). The NRC Senior Resident Inspector has been notified.
Event Notification (EN) # 53223, made on 2/20/2018, is being retracted because new information has been obtained that negates the original basis for reporting the unanalyzed condition. Specifically, an evaluation of the HELB that is postulated to occur in the TDAFP room has determined that without crediting door DSK13311 for protection, the affected safety-related instruments would not be exposed to environmental conditions beyond their analyzed capability. This resulted in a conclusion that the unanalyzed condition of door DSK13311 being open did not prevent the affected safety-related instruments or their supported AFW pumps from performing their required safety functions to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and/or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A), (B), or (D). The licensee notified the NRC Resident Inspector. Notified R4DO (Drake). | |
| ENS 53203 | 12 February 2018 10:12:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Ventilation Identified During Technical Support Center Test | 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 | 13 December 2017 19:28:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report | 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 | 2 October 2017 16:15:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Initial Part 21 Notification - Cameron Model 752B Differential Pressure Transmitters | 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 52905 | 15 August 2017 16:46:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | On August 15, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified a non-conforming condition in the plant design such that specific Technical Specification equipment is considered not to be adequately protected from tornado missiles. The recirculation lines for all three independent trains of Auxiliary Feedwater (AFW) connect to the Condensate Storage Tank (CST) inside the CST Valve House, which is not a tornado missile-resistant structure. The direct impact by a design basis missile could result in crimping of the recirculation lines, thereby creating the potential to cause damage to the Train A and B Motor-Driven Auxiliary Feedwater Pumps (MDAFPs) and the Turbine-Driven Auxiliary Feedwater Pump (TDAFP) by restricting recirculation flow to less than the design requirements. 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 NRC's Enforcement Guidance Memorandum EGM 15-002 and Interim Staff Guidance DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified. | 05000483/LER-2017-002 |
| ENS 52607 | 13 March 2017 18:58:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | 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 19:05:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Essential Service Water Pressure Transient | 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 04:02:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Actuation While Reactor Shutdown | 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 18:53:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Ventilation System to Emergency Off-Site Facility | 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 | 11 January 2016 18:00:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | 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 06:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip After an Offsite Electrical Fault | 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. | 05000483/LER-2015-004 Auxiliary Feedwater Control Valve Inoperable Due To Faulty Electronic Positioner Card |
| ENS 51257 | 23 July 2015 18:57:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater Manual Start Due to Loss of Main Feedwater Pump During Plant Cooldown | 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 06:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Initiation of Plant Shutdown Due to Rcs Leakage | 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). | 05000483/LER-2015-004 05000483/LER-2015-001 05000483/LER-2015-002 Completion of a Shutdown Required by the Technical Specifications - TS 3.4.13 Manual Auxiliary Feedwater System Actuation Auxiliary Feedwater Control Valve Inoperable Due To Faulty Electronic Positioner Card |
| ENS 50650 | 3 December 2014 06:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | An unexpected main turbine trip causing a reactor trip occurred on 12/03/2014 (at 0022 CST) with the plant operating in Mode 1 at 100 percent power. As part of the plant design, an expected, valid actuation of the Auxiliary Feedwater System occurred in response to the reactor trip. As part of the Auxiliary Feedwater actuation, the 'B' Motor Driven Auxiliary Feedwater Pump to 'D' Steam Generator throttle valve did not throttle as expected and was manually isolated. All other systems functioned normally in response to the plant conditions. The plant is currently stable in Mode 3 at 0 percent power. Safety related buses are receiving normal off-site power and the grid is currently stable. The NRC Resident Inspector was notified. All control rods fully inserted on the reactor trip. Steam generator levels are being maintained by the AFW system and decay heat is being removed by the main condenser. No primary or secondary safety relief valves lifted during the transient. | |
| ENS 50625 | 19 November 2014 01:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to All Four Safety Injection Accumulators Inoperable | 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 50474 | 19 September 2014 16:34:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to a Postulated Hot-Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | From review of Event Notification 50468 made by Wolf Creek Nuclear Operating Company on 9/18/2014, which in turn was based on review of INPO Event Report 14-33, 'Direct Current Circuits Challenge Appendix R Fire Analysis,' it was determined that portions of the control circuits for the main turbine-generator direct-current (DC) Emergency Lube Oil Pump and the Emergency DC Seal Oil Pump at Callaway Plant are not properly fused to prevent overload and possible secondary fires. The review found that a fire at the motor starter cabinet in the turbine building could cause specific 'smart' hot shorts that could cause overheating of the control cable and result in secondary fires outside the turbine building, including the Control Building, thereby potentially affecting safe shutdown capability for the plant. Based on this information, it has been determined that this condition is unanalyzed, and on a conservative basis, is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. As compensatory measures, hourly fire watches are in place in the affected areas of the Turbine Building and Control Building. These compensatory measures, in addition to automatic fire detection and suppression capability in these fire areas, ensure protection of the potentially affected equipment. The NRC Resident Inspector has been notified. The licensee continues to evaluate other control circuits to identify if this condition exists elsewhere. | |
| ENS 50465 | 17 September 2014 14:13:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Oil Tanker Truck Spill | 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 19:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Tritium and Cobalt Detected in Ground Water Monitoring Well | 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 | 23 April 2014 20:30:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report - False Negative Error Occurred on Quality Assurance Test | 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 13:30:00 | 10 CFR 26.719, FFD Reporting requirements | 24 Hour Fitness-For-Duty Report Involving False Negative Errors During Quality Assurance Testing | 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 49422 | 9 October 2013 20:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Fire Event Could Result in a Hot Short That Could Adversely Impact Safe Shutdown Equipment | 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 | 05000483/LER-2013-009 |
| ENS 49398 | 1 October 2013 03:57:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Off-Site Emergency Operations Facility Declared Inoperable Due to Air In-Leakage | 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 04:49:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to a Fire in the Turbine Building Lasting Greater than 15 Minutes | 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). | 05000483/LER-2013-008 |
| ENS 49014 | 9 May 2013 10:09:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant Pressure Boundary Leakage | 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. | 05000483/LER-2013-006 |
| ENS 49141 | 24 April 2013 09:04:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Telephone Notification for an Invalid Specified System Actuation | 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 | 11 April 2013 06:28:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Telephone Notification for an Invalid Specified System Actuation | 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 48879 | 2 April 2013 22:07:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Electrical Fault in Switchyard Resulting in Personnel Injuries | 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). |