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| Start date | Reporting criterion | Title | Event description | System | 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. | Feedwater Auxiliary Feedwater Control Rod | |
| 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. | Feedwater Auxiliary Feedwater Control Rod | |
| 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. | Feedwater Auxiliary Feedwater Control Rod | |
| 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. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| 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. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| 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. | Steam Generator Auxiliary Feedwater Control Rod | 05000483/LER-2015-004 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. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod | |
| 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). | Steam Generator Control Rod | 05000483/LER-2013-008 |
| ENS 44719 | 14 December 2008 23:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Electrical Fault on an Operating Condensate Pump | Callaway Plant was at 98% power with two condensate pumps in service. At 1714 on 12-14-08, the reactor was manually tripped due to a motor ground fault on the 'B' condensate pump. The 'C' condensate pump was not available for service (See EN #44714). This left only one condensate pump available so the reactor was manually tripped. The plant is in mode 3 and stable. Reactor trip procedures have been implemented and exited and normal operating procedures are in progress at this time. All safety systems actuated as designed. All control rods inserted during the trip. Offsite power is available and powering safety loads. There is no primary to secondary leakage. Decay heat is being removed via steam dumps to the condenser with makeup provided by Auxiliary Feedwater to the steam generators. The standby diesel generators and safety systems are available. The NRC Senior Resident Inspector has been notified. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 44714 | 12 December 2008 05:01:00 | 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 a Feedwater Transient | The 'C' condensate pump tripped due to an electrical problem which caused a feedwater transient. The unit (automatically) tripped on high steam generator level at approximately 2301 CST. The unit is in mode 3 and stable. The reactor trip procedures are in progress at this time. All control rods fully inserted into the core during the reactor trip. Offsite power is available and powering safety loads. The steam generator atmospheric steam dumps lifted momentarily during the transient and reseated. There is no known primary to secondary leakage. Decay heat is being removed via steam dumps to the condenser with makeup provided by auxiliary feedwater to the steam generator. The standby diesel generators and safety systems are available. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 44652 | 12 November 2008 00:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to a Feedpump Trip | At 1846 the reactor was manually tripped due to 'B' Main Feed Pump (MFP) trip on low lube oil pressure. The plant was operating at 97 percent (power) when annunciation was received that 'B' MFP was experiencing low lube oil pressure. Subsequently the 'B' MFP tripped on low lube oil pressure. The operating crew manually tripped the reactor per procedure OTO-AE-00001. The Auxiliary Feedwater System automatically actuated. All control rods fully inserted during the event and all safety systems responded as designed. The unit is removing decay heat using steam dumps to the Main Condenser. No primary relief valves or Main Steam relief valves lifted during the event. (The licensee is) currently investigating low lube oil pressure on 'B' MFP. Aux Feed is supplying make up to the steam generators. The plant is in normal electrical shutdown line-ups. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 43227 | 9 March 2007 16:43:00 | 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(b)(3)(iv)(A), System Actuation | Manual Reactor Trip at Reduced Power Due to Inadequate Feedwater Control | At approximately 0803 Central Standard Time (CST) on 3/9/2007, Callaway Plant received indications of a condenser tube leak. At approximately 0826 CST, a power reduction from 100 percent to less than 5 percent was initiated in accordance with the Secondary Water Chemistry Program. At 1043 CST, with the reactor at approximately 30 percent power, a manual reactor trip was initiated due to inadequate feedwater control to the C steam generator. There was indication of feedwater flow to and increasing level in the C steam generator with both the main feedwater reg valve and its bypass valve closed. Cause of the inadequate feedwater control is under investigation. All safety systems functioned as designed with the exception of the digital rod position indication for control rod K10 which failed during the trip due to an existing problem. Control rod K10 is believed to be fully inserted, but is being treated as indeterminate at this time. Main feedwater isolation and motor-driven auxiliary feedwater actuations occurred as expected. Both trains of the emergency diesel generators and the offsite power supplies are available. Callaway plant is currently stable in Mode 3 at normal reactor coolant system parameters. This notification is being made in accordance with 10CFR50.72(b)(2)(iv)(B) for manual reactor trip (RPS actuation), 10CFR50.72(b)(2)(xi) for offsite notification and 10CFR50.72(b)(3)(iv)(A) Specified System Actuation. The licensee informed the resident inspector of the event and will be making State notifications. A press release is expected.
Repairs were made to the Digital Rod Position Indication (DRPI) for control rod K10. The cause was determined to be a bent pin on a cable connector. When DRPI was restored, control rod K10 was verified to be fully inserted. Post maintenance testing was completed satisfactorily. The cause of the inadequate feedwater control for the C steam generator was a failure of the positioner for the C main feedwater reg valve. The positioner was replaced and tested satisfactory. The plant is currently in Mode 2 (Startup) The resident inspector was informed of the update. R4DO (Mike Shannon) notified. | Steam Generator Reactor Coolant System Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 42571 | 12 May 2006 05:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip on High Steam Generator Level | While lowering turbine load to 45% for planned maintenance to replace an RCS Loop flow transmitter, Callaway Plant experienced high vibration on two main turbine bearings. The main turbine was manually tripped at 0047 in accordance with off-normal procedures. At 0052, received a Steam Generator High-High Level (P-14) on the "A" S/G resulting in a Feed Water Isolation Signal (FWIS) and Auxiliary Feed Water Actuation (AFAS). The reactor was manually tripped at 0053. Emergency Operating Procedures were completed and exited at 0115. After receiving the main turbine high vibration alarm, the plant reduced power to below the reactor trip/turbine trip setpoint and manually tripped the main turbine. Steam generator level rose to the P-14 feedwater isolation setpoint at which time, the reactor was manually tripped. All rods fully inserted on the trip. Decay heat is being removed by condenser steam dumps. Steam generator level is being maintained with the startup feed pump. The electrical grid is stable. No relief valves or safety valves lifted. The cause of the high vibration is being investigated. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine | |
| ENS 41347 | 19 January 2005 18:51:00 | 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(b)(3)(iv)(A), System Actuation | Callaway Automatic Reactor Trip on Low Steam Generator Level | Unit experienced an automatic reactor trip following a momentary loss of power to a process control system power supply that resulted in a low Steam Generator water level. This event is being reported in accordance with 10CFR50.72(b)(2)(iv)(B) due to an automatic Reactor Trip and subsequent Aux Feedwater actuation and Feedwater isolation. On 1/19/2005 at 12:51 CST with Callaway plant at 100% reactor power, a Reactor Trip occurred on low Steam Generator 'A' level due to an apparent momentary loss of power in Control Cabinet (Relay Panel) RP043. A failed Primary Power Supply to RP043 replacement was in progress at the time of the trip. A momentary power spike on the in-service backup power supply caused the Main Feedwater Regulating valve (MFRV) 'A' to shut and both Main Feedwater Pumps (MFPs) went to their low speed stops. Attempts by the Reactor Operator to take manual control of 'A' MFRV and the MFPs were unsuccessful resulting in the Reactor Trip. Following the reactor trip all systems responded as designed with no abnormal responses. The plant is currently stable in Mode 3 with RCS pressure at 2237 psig and RCS temperature at 557.8 F. The Auxiliary Feedwater is currently supplying the Steam Generators with steam flow to the Main Condenser for decay heat removal. All control rods fully inserted. Both NRC Resident Inspectors were contacted and responded to the Control Room. Corporate Communications will be providing public information on the plant trip to the media. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | 05000483/LER-2005-001 |
| ENS 40522 | 15 February 2004 21:19:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip with Auxiliary Feedwater Actuation | The following information was received from the licensee via facsimile: At 1519 (CDT) on 2-15-04, the plant was at 20 power, increasing load. During the transfer from the bypass feedwater regulating valves to the main feedwater regulating valves, the main turbine tripped due to a P-14 S/G Hi-Hi level signal. The P-14 signal also caused a Feedwater Isolation Signal (FWIS), which resulted in a Motor Driven Aux Feedwater Actuation (MDAFAS), and Steam Generator Blowdown Isolation Signal. Steam Generator levels subsequently shrunk to low levels causing a Steam Generator low level reactor trip at 1524 (hrs). Sometime after the turbine trip, the turbine driven aux feed pump ((TDAFP)) was manually started to help restore S/G levels. The pump was later manually secured, but then received a subsequent auto-start signal and tripped on overspeed. The TDAFP was declared inoperable at 1535 (hrs). With the exception of the TDAFP, all safety systems operated properly. The plant is currently stable in Mode 3, with the primary plant at 557 deg F and 2235 psig, and pressurizer level at 25%. The licensee has notified the NRC Resident Inspector.
The following information was received from the licensee via facsimile as a "rewrite of above paragraph 2. Sometime after the turbine trip, the Turbine Driven Aux Feedwater Pump (TDAFP) was manually started to help restore S/G levels. Due to decreasing reactor coolant system (RCS) temperature, it was decided to stop the TDAFP since it is a steam load. The steam supply valves were remote manually closed from the control room. At the same time the steam supply valves were closing, the steam generator levels shrunk to low levels causing a Steam Generator low level reactor trip and Turbine Driven Aux Feedwater Actuation (TDAFAS) at 1524. The steam supply valves immediately reopened due to the TDAFAS. It is postulated that this transient in the steam supply caused the pump to overspeed trip. The TDAFP was declared inoperable at 1535. With the exception of the TDAFP, all safety systems operated properly. The licensee notified the NRC Resident Inspector. Notified R4DO (Cain). | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Main Turbine | |
| ENS 40500 | 3 February 2004 10:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Main Genrator Output Breakers Opening. | AT 0439, 02/03/04, Callaway Plant experienced a reactor trip while operating transmission breakers located in the plant switchyard. Auxiliary feedwater actuated as expected to stabilize steam generator water levels. All other plant systems responded as expected. At present, the cause of the reactor trip is undetermined. Initial indications are that there were no voltage or frequency perturbations on the electrical grid. Plant staff is investigating potential areas in an effort to identify the cause of the trip. The main generator output breakers opened when one of the transmission breakers in the switchyard was opened per the scheduled maintenance test. When the incoming Transmission line breaker was opened it should not have caused the main generator output breakers to open. All rods fully inserted into the core. Both Motor Driven Auxiliary Feedwater pumps automatically started on a loss of both main feedwater pumps and the Turbine Driven Auxiliary Feedwater pump started on low low steam generator water level. The "C" Steam Generator atmospheric steam dump valve opened for about 17 seconds before it closed (no leaking steam generator tubes in the "C" Steam Generator). All Emergency Core Cooling systems are fully operable except for the "A" Safety Injection System. It was declared inoperable when it failed a valve test last night at 0205 CST. All Emergency Diesel Generators are fully operable. Turbine Driven Auxiliary Feedwater Pump discharge valves were closed after it had operator for about two hours. Approximately one hour after the valves were closed the Turbine Driven Auxiliary Feedwater pump tripped. This incident is being investigated. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 40484 | 27 January 2004 23:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | At 1730, 1/27/04, Callaway Plant experienced a reactor trip. Auxiliary feedwater actuated as expected to stabilize steam generator levels. All other plant systems responded as expected. At present the cause of the reactor trip is undetermined. Initial indications are that there were no voltage or frequency perturbations on the electrical grid and that the trip signal originated from Callaway's switchyard breaker circuitry. Plant staff is currently engaged in a review of all indications that occurred at the time of the trip and systematically investigating potential problem areas in an effort to identify the cause of the trip. The licensee also reported that all control rods inserted on the reactor trip, no primary or secondary system relief valves operated, and that reactor temperature is being maintained using steam dump to the condenser. Steam generator water levels are being maintained using auxiliary feedwater. The station electrical system is available and in a normal configuration. The licensee notified the NRC Resident Inspector.
The licensee provided the following information as an update to their initial report: The cause of the reactor trip has been identified as a failed electrical relay in the main generator protection circuitry. The relay is designed to sense a fault in the main electrical generator and trip the generator output breakers. The failed relay was designated as a 321/G relay and provides phase fault backup protection to prevent exceeding thermal limits of the stator windings. After determining the cause of the reactor trip, preparations for a reactor startup are in progress. The licensee will inform the NRC Resident Inspector. Notified R4DO(Johnson). | Steam Generator Auxiliary Feedwater Control Rod |