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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56190 | 31 October 2022 00:57:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: At 2057 Eastern Daylight Time (EDT), with Unit 1 in Mode 3 at 0 percent power, an actuation of the Auxiliary Feedwater (AFW) System occurred during an attempt to start the 'B' Main Feed Pump. The reason for the AFW system auto-start was due to the 'A' electrical bus being under clearance and the 'B' Main Feed Pump not starting, resulting in a valid actuation signal for loss of both Main Feedwater pumps. The 'A' and 'B' motor-driven AFW (MDAFW) pumps were running prior to the attempted start of the B Main Feedwater pump and continued to run. The MDAFW Flow Control Valves (FCVs) went full open automatically as designed when the MDAFW actuation signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Feedwater Auxiliary Feedwater | |
| ENS 56189 | 30 October 2022 10:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip and Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: At 0653 Eastern Daylight Time (EDT), with Unit 1 in Mode 1, at 16 percent power, an automatic reactor trip occurred due to an under-voltage condition on the 'A' reactor coolant pump (RCP) and the 'C' RCP. Power was lost from the 'A' auxiliary bus while performing an operating procedure to transfer power from the 'A' start-up transformer to the 'A' unit auxiliary transformer. Operations responded and stabilized the plant. Decay heat is being removed by the main steam system to the atmosphere using the steam generator power-operated relief valves. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Steam | |
| ENS 56075 | 28 August 2022 07:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: (On 8/28/2022) at 0329 Eastern Daylight Time (EDT), with Unit 1 in Mode 1 at 100% power, the reactor was manually tripped due to a 'B' train main feedwater pump trip. The trip was not complex with all systems responding normally post-trip. The auxiliary feedwater (AFW) system started automatically as expected. Operations responded and stabilized the plant. Steam generator levels are being maintained by AFW through the AFW flow control valves. Decay heat is being removed by using the steam generator power-operated relief valves. The reason for the 'B' train main feedwater pump trip is under investigation. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'B' train main feedwater pump trip is suspected to be the result of an electrical transient due to the alarms that the operators received. In addition, the 'A' train main feedwater pump also tripped subsequent to the reactor trip and that cause is still under investigation. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 55868 | 29 April 2022 08:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: At 0405 Eastern Daylight Time (EDT), with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to degrading condenser vacuum. The trip was not complex, with all systems responding normally post-trip. The Auxiliary Feedwater System started automatically as expected. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam System to the main condenser using the turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No Tech Spec limits were exceeded. Offsite power is available. The suspected cause for the loss of condenser vacuum is when performing the scheduled monthly swap of condenser vacuum pumps, a suction valve failed to shut. | Reactor Protection System Auxiliary Feedwater Main Condenser Main Steam | |
| ENS 55038 | 16 December 2020 13:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Generator Lockout | On December 16, 2020 at 0851 EST, with Harris Nuclear Plant Unit 1 in Mode 1 at 80 percent power, an automatic reactor trip occurred due to lockout of the main generator. The trip was not complex, with all systems responding normally post-trip. The initial assessment of this event indicates that there was a ground fault on the 'B' train of the non-safety electrical distribution system that caused the main generator lockout. Steam generator levels are being maintained by normal feedwater through the feedwater regulator bypass valves. Decay heat is being removed by using the condenser steam dump flow path. Due to the unplanned Reactor Protection System actuation while critical, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This condition does not affect the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All rods inserted into the core during the trip. The electrical grid is stable and all safe shutdown equipment is available for service. No reliefs lifted during the transient. | Steam Generator Feedwater Reactor Protection System | |
| ENS 54834 | 13 August 2020 13:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Dropped Control Rod | On August 13, 2020, at 0938 EDT, with Harris Nuclear Plant Unit 1 in Mode 1 at 100 percent power, a control rod dropped during control rod testing. This is considered to be an unanalyzed condition and requires a manual reactor trip in accordance with plant procedure. All safety systems functioned as expected. Auxiliary Feedwater started as designed and was secured. Steam generator levels are being maintained by Main Feedwater through the feedwater regulator bypass valves. Decay heat is being removed by using the condenser steam dump flow path. Due to the RPS actuation, this event is being reported as a four hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the unanalyzed condition and unplanned Auxiliary Feedwater actuation, this event is also being reported as an eight hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 54599 | 23 March 2020 14:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On March 23, 2020, at 1013 EDT, with Harris Nuclear Plant Unit 1 in Mode 1, at 100 percent power, an unplanned actuation of the reactor protection system occurred. This resulted in an automatic reactor trip. The trip occurred during the restoration of the auto-stop turbine trip function during a planned maintenance evolution. All safety systems functioned as expected. Auxiliary Feedwater started as designed and was secured. Steam generator levels are being maintained by normal feedwater through the feedwater regulator bypass valves. Decay heat is being removed by using the condenser steam dump flow path. Due to the unplanned Reactor Protection System actuation while critical and the expected Auxiliary Feedwater actuation, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 53318 | 7 April 2018 08:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Auxiliary Feedwater System (Afw) | On April 7, 2018 at 0451 EDT, with Unit 1 in Mode 3 at 0 percent power, an auto actuation of 'A' and 'B' Motor Driven Auxiliary Feedwater (MDAFW) pumps occurred during the shutdown of Unit 1 for Harris Nuclear Plant's refueling outage. Plant Operators successfully took control of the AFW flow and noted the 'B' Main Feed pump was still running with proper suction and discharge pressures of 430 lbs. and 1000 lbs. The 'A' and 'B' Motor Driven Auxiliary Feedwater (MDAFW) pumps automatically started as designed when the 'Loss of Both Main Feedwater Pumps' signal was received. The cause of the actuation is still being evaluated. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Feedwater Auxiliary Feedwater | |
| ENS 52335 | 26 October 2016 15:42:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Containment Sump Suction Valve Opened During Containment Spray Pump Testing | On October 26, 2016, the Harris Nuclear Plant was in Mode 6 with core reload complete, the reactor head removed, and reactor cavity water level greater than 23 feet. The refueling water storage tank (RWST) was less than 23.4% level as expected for the refueling conditions. During surveillance testing to adjust the eductor flow throttle position, the containment spray pump was started in recirculation mode with the discharge valve shut. With RWST level less than 23.4%, logic was satisfied to actuate Engineered Safety Features Actuation System (ESFAS) Functional Unit 8, containment spray switchover to containment sump. The containment sump suction valve opened in accordance with the design, however the action was unexpected by the operators. Therefore, operators secured the containment spray pump and shut the containment sump suction valve. ESFAS Functional Unit 2, Containment Spray, was not actuated and water did not flow through the containment spray nozzles. This event is reported as a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to the opening of the containment sump suction valve. This event did not impact the health and safety of the public. The licensee has notified the NRC Resident Inspector. | Containment Spray | |
| ENS 52291 | 8 October 2016 17:28:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Unusual Event Declared Due to Loss of Offsite Power | Loss of all offsite power capability, Table S-5, to 6.9kV emergency buses 1A-SA and 1B-SB for greater than or equal to 15 minutes. At 1328 EDT, while shutdown in Mode 4 (Hot Shutdown), Harris declared an Unusual Event due to a loss of offsite power. Following the loss of offsite power (LOOP), the Emergency Diesel Generators started and loaded onto their respective emergency buses. The reactor remains stable and shutdown in Mode 4. The licensee is currently investigating the cause of the LOOP and the emergency buses will continue to be powered by the EDGs until the licensee has determined the cause for the LOOP. Offsite power is currently available into the switchyard. The licensee notified the state government, the local government, and the NRC Resident Inspector. Notified DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only).
The cause (of the LOOP) is not known. Duke Energy Control Center has evaluated the grid and is comfortable with Harris connecting emergency buses back to the grid. Harris Plant is evaluating restoration. Faults were validated on the 115kV Cape Fear North and South supply lines into the Harris switchyard. This notification also addresses various valid actuations of safety systems, including the Emergency Diesel Generators, as well as, potential loss of Emergency Assessment Capabilities due to the LOOP impacting Emergency Planning equipment. The licensee notified the NRC Resident Inspector. Notified R2DO (Bonser), IRD (Grant), and NRR EO (King).
The cause of the LOOP has been determined to be a momentary electricity loss on the 115kV Cape Fear North and South supply lines into the Harris switchyard. This event notification also addresses the loss of safety function of the offsite power system which occurred as a result of grid perturbations. The licensee notified the NRC Resident Inspector. Notified R2DO (Bonser), IRD (Grant), and NRR EO (King).
Based on the grid being stable and the 115kV Cape Fear North and South lines being available, the licensee terminated the Unusual Event at 2049 EDT on 10/8/16. The licensee notified the NRC Resident Inspector. Notified R2DO (Bonser), IRD (Grant), and NRR EO (King). Notified DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only).
10 CFR 50.72(b)(2)(XI) - OFFSITE NOTIFICATION At approximately 1305 EDT on October 9, 2016, Duke Energy personnel notified the North Carolina Department of Environment and Natural Resources of a spill of untreated sanitary wastewater. During a significant rainfall event associated with Hurricane Matthew, wastewater was released from the overflow of a lift station that did not function as a result of a power outage. The untreated sanitary wastewater entered the plant's storm drain system. The release has been stopped and the lift station power is restored. An investigation is in progress to further determine the cause and additional corrective actions. There is no impact to public health and safety or the environment due to this incident. This event is reportable per 10 CFR 50.72(b)(2)(xi), an event related to protection of the environment for which a notification to other government agencies has been made. The NRC Resident Inspector has been notified. Notified R2DO (Bonser). | Emergency Diesel Generator | |
| ENS 52289 | 8 October 2016 05:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Unplanned Reactor Trip and Safety Injection Due to Turbine Control Valve Transient | On October 8, 2016, while reducing power for a planned refueling outage, the unit was taken offline by opening the main generator output breakers. With the reactor at approximately 7 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 0150 (EDT), an unexpected steam valve transient occurred while main turbine valve control was being transferred from throttle valve to governor valves during main turbine overspeed testing. This resulted in an automatic low steamline pressure Safety Injection and Reactor Trip. All safety systems functioned as expected. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system (RCS) temperature and pressure following the reactor trip, with decay heat being removed using steam generator power operated relief valves. Steam generator water levels are being maintained using auxiliary feedwater. All emergency core cooling system (ECCS) equipment is available. The cause of the steam valve transient is under investigation. This condition is being reported as an ECCS discharge to RCS, an unplanned reactor protection system actuation, and a specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B). and 10 CFR 50.72(b)(3)(iv)(A). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. The Safety Injection occurred for approximately 6 minutes and Pressurizer level increased to approximately 71%. The Main Steam Isolation Valves closed as a result of the Safety Injection and Decay Heat is being removed using the Steam Generator Atmospheric Relief Valves. There is no known primary to secondary leakage. | Steam Generator Reactor Coolant System Reactor Protection System Main Steam Isolation Valve Auxiliary Feedwater Main Turbine Emergency Core Cooling System | |
| ENS 49742 | 18 January 2014 15:16:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Alert Declared Due to Fire in 480V Bus | (At 1016 EST, an) Alert (was declared) based on EAL # HA 2.1 Fire or explosion resulting in either: visible damage to any table H-1 structure or system/component required for safe shutdown of the plant, or control room indication of degraded performance of any safe shutdown structure, system, or component within any table H-1 area. Fire in 480V bus 1D2. Reactor was manually tripped 480 VAC safety related transformer fire in switchgear room. Plant reduced power and tripped the reactor manually. Reactor trip was uncomplicated. Fire was extinguished when the 480 VAC bus was de-energized. The licensee has notified the NRC Resident Inspector, the State of North Carolina, and other local authorities. Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail.
Harris Nuclear Plant secured from the Alert at 1551 EST, on 1/18/14. The plant is stable, the fire is out, the TSC and EOF have been secured and plant recovery has been transferred to the outage control center. There were no personnel injuries or radiological releases. Radiation monitor RM-*1TS-3653C (Technical Support Center Radiation Monitor) is out of service. The licensee has notified the NRC Resident Inspector. Notified the R2DO (King), R2RA (McCree), NRR (Leeds), IRD MOC (Grant), OPA (Brenner), NRR EO (Lee) Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail. | ||
| ENS 46396 | 5 November 2010 05:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of an Emergency Diesel Generator Following Inadvertent De-Energization of Safety Bus | This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A), an event or condition that resulted in a valid actuation of emergency AC electrical power systems. At 0104 EDT on November 5, 2010, the plant was in Mode 5 Cold Shutdown in a refueling outage. Maintenance personnel were performing post maintenance testing on the main generator relay panel when the 'B-SB' Safety Bus was inadvertently de-energized, resulting in automatic starting of the 'B-SB' Emergency Diesel Generator. Safety systems responded as expected during the event, and the 'A-SA' Emergency Bus remained operable with power available from both offsite power and the 'A-SA' Emergency Diesel Generator throughout the event. The 'B-SB' Safety Bus was powering shutdown cooling, which was restored using 'B-SB' Residual Heat Removal system at 0107 EDT with the 'B-SB' Emergency Diesel Generator supplying the bus. The testing procedure was not intended to result in the safety bus being de-energized nor automatic start of the Emergency Diesel Generator. Although the specific cause of the safety bus being de-energized is not known at the present time, it appears that it was directly related to the testing being performed. The licensee notified the NRC Resident Inspector. The 'B-SB' Safety Bus has been returned to its normal offsite power source and the 'B-SB' Emergency Diesel Generator secured and returned to standby status. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal | |
| ENS 45499 | 16 November 2009 03:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Main Generator Oil Leak | At 2242 (EST) on 11/15/09, the reactor was manually scrammed from 100% power due to a large oil leak on the main generator seal oil system. Condenser vacuum was broken immediately following the reactor trip, and the main turbine stopped rotating at 2324 (EST). Following the reactor trip, the 'B' steam generator Main Steam Isolation Valve (MSIV) failed to fully close on demand, but was closed due to field actions at 2303 (EST). The reactor remained stable at NOP/NOT following the reactor trip. Offsite power remained available throughout the event. This condition is being reported as actuation of the reactor protection system in accordance with 10CFR 50.72(b)(2)(iv)(B). All control rods fully inserted and decay heat is being removed through the S/G relief valves to the atmospheric dumps. No known primary to secondary leakage exists. The plant remains stable in Mode 3. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Protection System Main Steam Isolation Valve Main Turbine Control Rod | |
| ENS 44423 | 19 August 2008 13:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rod Control System Malfunctioned | On August 19, 2008, with the Unit shut down in Mode 3, post maintenance testing was being performed for the Digital Rod Position Indication System. While performing this test, a 'Rod Control Urgent Failure' alarm was received upon initial withdrawal of Control Bank-C. All other control and shutdown banks remained fully inserted in the core. Local inspection revealed a phase failure on movable gripper coils in a power cabinet. In accordance with plant procedures, at 0905 hours, a manual reactor trip was initiated by operators opening the reactor trip breakers. All Safety Systems functioned as designed and Rod Control System repairs are in progress. This event posed no significant safety implications because the reactor was subcritical when the reactor trip breakers were opened. Compliance with all Technical Specification requirements was maintained. The health and safety of the public were not affected by this event. The NRC Resident Inspector was notified. | ||
| ENS 44404 | 11 August 2008 04:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Degrading Condenser Vacuum | At 0049 EDT on 8/11/2008, the Harris Nuclear Plant was manually scrammed from 21% power due to indications of degrading condenser vacuum. At the time, a reactor shutdown was in progress with indications of a degraded condenser boot seal. The unit was stabilized in Mode 3 with no additional equipment failures or other complications. The reactor is currently at normal operating pressure and temperature. The highest vacuum observed was 8". All rods inserted into the core after the manual trip. Auxiliary feed water did not start as a result of the trip. Steam generator level is being maintained via normal feed water flow path. Decay heat is being removed via the steam dumps to atmosphere. There is no known primary to secondary leakage. No power-operated or manual reliefs lifted during the transient. The grid is stable and loads are being supplied via the station start-up transformer. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 43676 | 29 September 2007 02:32:00 | 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 Loss of a Startup Transformer | On September 28, 2007, while reducing power for a planned refueling outage with the reactor at approximately 30 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 2232 a fault pressure trip signal was received on the A Startup Transformer (SUT), causing a loss of power to Aux Buses D, A & C electrical buses as well as the A-SA safety bus. The loss of A & C buses initiated the RCP underfrequency trip which tripped the Reactor and all three RCPs as designed. The A Diesel Generator automatically started and reenergized bus A-SA as designed. The auxiliary feedwater system actuated as expected due to undervoltage on the A-SA safety bus and loss of the main feedwater pumps. All control rods inserted on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using auxiliary feedwater. All emergency core cooling system equipment is available. The plant electrical system is being restored at this time. The A SUT remains out of service. The cause of the loss of power from A SUT is under investigation. This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector was notified of this event by the licensee. The plant was in natural circulation for approximately 1 hour. The Main Steam Isolation Valves (MSIVs) were manually isolated per procedure due to loss of EHC indication. Presently the B RCP has been restored to service, MSIVs are still closed, and the motor driven auxiliary feedwater pumps are feeding the Main Steam Generators. There are not any leaking steam generator tubes. The A EDG will be secured after backfeeding of the deenergized buses have been established. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Main Steam Isolation Valve Auxiliary Feedwater Emergency Core Cooling System Control Rod | 05000400/LER-2007-003 |
| ENS 42848 | 19 September 2006 14:00:00 | 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 Generator Lockout Signal | At approximately 1000 EDT on September 19, 2006, with the reactor at 100 percent power in Mode 1, the reactor was automatically tripped from a turbine trip due to a generator lockout signal. The cause of this signal is under investigation. The auxiliary feedwater system actuated as expected to stabilize steam generator levels. All systems functioned as required and no other safety systems were actuated. All control rods inserted (fully) on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using normal feedwater. All emergency core cooling system equipment is available. The plant electrical system is available and in a normal configuration. This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A)." The MSIVs are open with the steam generators discharging steam to the main condenser using the steam dump valves. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser Control Rod | 05000400/LER-2006-003 |
| ENS 41654 | 1 May 2005 04:21:00 | 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 Loss of Condensate Pump | The plant was in Mode 1 at 100% power. At 0021 (EDT) the reactor was manually tripped following a loss of 1B Condensate pump per AOP-010, Feedwater Malfunctions. The cause of the 1B Condensate pump trip is not known at the present time. The plant is stable in Mode 3 at normal temperature and pressure. All safety systems functioned as expected; AFW automatically actuated due to low level in the steam generators to provide continued decay heat removal. All control rods fully inserted on the manual reactor trip. Secondary PORVs opened on the trip and reclosed. Steam generators are discharging steam to the main condenser using the turbine steam dump valves. AFW has been secured and main feedwater is operating to maintain SG levels. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Decay Heat Removal Main Condenser Control Rod | 05000400/LER-2005-002 |
| ENS 41179 | 7 November 2004 21:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | "a" Auxiliary Feedwater (Afw) Pump Manually Started Due to Decreasing Steam Generator Water Level. | The 'A' motor driven AFW pump was manually started in response to lowering Steam Generator (SG) level. The plant was in Mode 1 with reactor power at approximately 7%, preparing for startup of the turbine generator. Steam Generator Feedwater flow to the Steam Generators was being supplied by the main feedwater system through the main feedwater regulating bypass valves operating in automatic control. The operating crew observed a lowering trend in 'C' Steam Generator level. Upon reaching the lower end of the procedurally established normal control band (time=1633), the operating crew took the following actions in accordance with plant procedures: (1) the associated feedwater regulating bypass valve was taken to manual control to attempt restoration of normal SG level, (2) the 'A' motor driven AFW pump was manually started (time= 1635) to supply AFW to the SG and (3) reactor power lowered to reduce steam demand. These actions resulted in the restoration of 'C' SG level to normal. Normal operating SG level is 57%. Lo Lo Steam Generator Level Trip occurs at 25%. The lowest SG level observed during the evolution was approximately 43%. The cause of the irregular Feedwater control is currently being investigated. The plant is currently at approximately 2.5% power with all Steam Generator levels at normal operating levels. No automatic ESF Actuations occurred. The NRC Senior Resident was informed. | Steam Generator Feedwater Auxiliary Feedwater | 05000400/LER-2004-006 |
| ENS 41129 | 18 October 2004 11:41:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Emergency Diesel Start | At 0741 (EDT), an unplanned actuation of an ESF system occurred due to the unplanned start and loading of the 'A' Emergency Diesel Generator (EDG). The plant was in a refueling outage in Mode 5 with the Reactor Coolant System (RCS) depressurized. The 'B' Emergency Safety Bus was operable and protected. The 'A' EDG started when the feeder breaker to the 'A' train Emergency Safety Bus opened unexpectedly, de-energizing the bus. The 'A' Emergency Diesel Generator (EDG) started and re-energized the bus as designed, and the 'A' Safeguard Sequencer initiated loading of the bus. Two anomalies occurred during re-energization of the 'A' emergency bus loads. Bus 1A3-SA, which provides power to various safety related load centers and supplies support system loads, did not re-energize. Additionally, the 'A' Emergency Service Water (ESW) pump did not start. The cause of the initial bus feeder breaker opening, and the cause of the subsequent equipment failures is under investigation. Bus 1A3-SA was manually re-energized at 1029. The 'A' ESW pump was manually restarted at 1054." The 'A' EDG cooling water was provided by the service water system which was not affected by the trip. The 'A' RHR pump was in service to provide Shutdown cooling and the pump tripped when the 'A' Train Emergency Safety Bus was deenergized. Reactor plant temperature rose from 116 to 122 degrees F in the four minute period prior to the restart of the 'A' RHR pump. 'A' RHR pump was restarted at 0745, and 'B' RHR pump was started at 0843 as a backup. Shutdown cooling was transferred to the 'B' Shutdown cooling loop at 1343 and the 'A' RHR pump was secured. The licensee notified the NRC Resident inspector. | Reactor Coolant System Service water Emergency Diesel Generator Shutdown Cooling | 05000400/LER-2004-005 |
| ENS 40730 | 6 May 2004 16:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip and Actuation of Auxiliary Feedwater System | The following information was received from the licensee via facsimile: On May 6, 2004, with the reactor at 100 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 1252 (EDT) the reactor was automatically tripped from a power range negative flux rate trip signal. The auxiliary feedwater system actuated as expected to stabilize steam generator levels. All systems functioned as required and no other safety systems were actuated. All control rods inserted on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using normal main feedwater. All emergency core cooling system equipment is available. The plant electrical system is available and in a normal configuration. The cause of the plant trip is under investigation. This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and10 CFR 50.72(b)(3)(iv)(A) . During the transient, a steam generator power-operated relief valve lifted momentarily and then re-seated. No reportable radiological release occurred during the event. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | 05000400/LER-2004-003 |
| ENS 40084 | 17 August 2003 19:51:00 | 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 Loss of "a" Condensate Pump. | At 3:51 PM EDT, on August 17, 2003, with the reactor at 100% in Mode 1, the reactor was manually tripped in response to a trip of the A condensate pump and subsequent trip of the A main feed pump. Both motor-driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump started automatically due to Lo-Lo steam generator level. The operations crew responded to the event in accordance with the applicable plant procedures. The plant was stabilized at a normal operating no-load Reactor Coolant System (RCS) temperature and pressure following the reactor trip. The condensate pump electrical supply breaker tripped due to instantaneous overcurrent, possibly related to a severe electrical storm in the area at the time. The feed pump tripped due to the loss of the condensate pump. Offsite power remained available throughout the transient. A local fire department responded to a downed power line in the vicinity of the Harris plant but the response was not related to any onsite activities. This condition is being reported as actuation of the reactor protection system and AFW in accordance with 10CFR50.72(b)(2)(iv)(B), and 10CFR50.72(b)(3)(iv)(A). 10CFR50.72 requires an 8-hour report for "Any condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation." In this case, the AFW pumps start signal was due to Lo-Lo Steam Generator Level. A root cause team is being formed to identify the cause and corrective actions Due to low decay heat in the core, the Main Steam Isolation Valves were closed and the Steam Generator PORVs are being used to maintain the plant in a Hot Standby condition. No known leaking steam generator tubes are known. Both motor-driven and the turbine driven auxiliary feedwater pumps were secured after main feedwater was returned to service. All Emergency Core Cooling Systems and the Emergency Diesel Generators are fully operable if needed. The electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater | 05000400/LER-2003-005 |