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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57155 | 30 May 2024 23:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email and phone: On May 30, 2024, at 1949 EDT, Unit 1 automatically tripped from 100 percent power due to an electrical fault on the B unit auxiliary transformer. The unit has been stabilized in mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). Decay heat is being removed by the condenser steam dump system and Unit 1 is in a normal shutdown electrical lineup. There was no impact on the health and safety of the public or personnel. The NRC Resident Inspector has been notified. | Reactor Protection System Control Rod | |
| 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 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 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 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 Emergency Core Cooling System 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 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 Emergency Core Cooling System 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 Emergency Core Cooling System | 05000400/LER-2003-005 |