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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55390 | 31 July 2021 21:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip of Unit 1 | At 1646 (CDT) on 7/31/21, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to control board indications of a Unit 1 'B' Main Feed Pump trip. After the reactor trip, one of the Condenser Steam Dump valves cycled to intermediate and remained stuck. The Condenser Steam Dump Valve was isolated locally using manual isolation valves. The 'B' Feed Regulating Bypass Valve did not control in automatic and was taken to manual to control the level in 'B' Steam Generator. The Auxiliary Feedwater System automatically actuated as designed when the valid actuation signal was received. Operations stabilized the plant in Mode 3. Decay heat is being removed by atmospheric dump valves due to condenser unavailability. Unit 2 is unaffected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. During the transient, all control rods inserted into the core. There is no known primary to secondary leakage. During the transient, no relief valves or safeties lifted. The plant is currently maintaining normal operating temperature and pressure with all safety equipment available. The plant is in its normal shutdown electrical lineup. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 53779 | 5 December 2018 06:00:00 | 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 Loss of Main Condenser Vacuum | At 1539 (CST) December 5, 2018, with Unit 1 at 100 percent power, the reactor was manually tripped due to degrading condenser vacuum. The trip was uncomplicated with all systems responding normally, post-trip. An actuation of the auxiliary feedwater system occurred during the manual trip. The auxiliary feedwater system automatically started as designed when the valid actuation signal was received. Operations stabilized the plant in mode 3 (hot standby). Decay heat is being removed by atmospheric dump valves. Unit 2 is not affected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The loss of condenser vacuum resulted because one of two circulating water pumps was running and its discharge valve shut. The cause for the valve shutting is under investigation. There is no primary to secondary leakage. The licensee notified the NRC Resident Inspector | Auxiliary Feedwater Main Condenser | |
| ENS 51570 | 29 November 2015 01:12:00 | 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 After Voltage Regulator Malfunction | Unit 1 automatic reactor trip actuated due to an automatic voltage regulator (AVR) malfunction which caused a generator lockout and turbine trip. The cause of the AVR malfunction is unknown at this time. All control rods fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feed water system. The auxiliary feed water system actuated based on low steam generator level, but since has been secured. Off-site power remains available. No release is occurring and emergency core cooling systems did not actuate. Emergency plan entry was not required. The plant is in its normal shutdown electrical lineup at normal operating temperature and pressure. Unit 2 was not affected by the Unit 1 transient. The licensee has notified the NRC Resident Inspector | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 50649 | 3 December 2014 02:50:00 | 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 Failure of Condensate Pump | Initiated a manual Unit 1 reactor trip due to imminent failure of 1P-25B Condensate Pump. Unit 1 had commenced a rapid down power due to the degradation of the pump. The event is reportable under 10CFR50.72(b)(2)(iv)(B) for a manual actuation of the reactor protection system when the reactor is critical and 10CFR50.72(b)(3)(iv)(A) for an actuation of specified system (6) PWR auxiliary feedwater system. Auxiliary feedwater system actuation was due to low steam generator water levels in both 'A' and 'B' Steam Generators, an expected system response during a reactor trip. Decay heat removal is by forced circulation and is being controlled by auxiliary feedwater system and condenser steam dumps. After the trip, both main steam generator feedwater pumps were secured due to feed pump suction pressure remaining low post trip. All other plant systems functioned as required. All control rods fully inserted in the core due to the manual trip. There was no ECCS actuation. Off-site power has been maintained throughout the event. No primary or secondary safety relief valves lifted during the reactor trip. Unit 1 is in a normal shutdown electrical lineup. There was no effect on Unit 2. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Decay Heat Removal Safety Relief Valve Control Rod | |
| ENS 48189 | 15 August 2012 01:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manually Tripped in Anticipation of an Automatic Turbine Trip | Unit 1 Manual Reactor Trip was initiated in anticipation of an auto turbine trip due to operators noticing the turbine governor valves closing in response to an Electro-Hydraulic Control System signal. All Control Rods are fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feedwater system. The auxiliary feedwater system actuated based on low steam generator level, but has since been secured. There were no unexpected (inconsistent with nature of trip) pressure or level transients. Offsite power remains available. No release occurred nor is ongoing. Emergency Core Cooling did not actuate. No unexpected isolations occurred. Emergency Plan entry was not required. The licensee informed the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 48053 | 28 June 2012 01:46:00 | 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 100% Load Reject | Unit 2 Manual Reactor Trip was actuated due to indications of a 100% Load Rejection. The cause of the Load Rejection is not known at this time. All Control Rods are fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feedwater system. The auxiliary feedwater system actuated based on low steam generator level, but has since been secured. There were no unexpected (inconsistent with nature of trip) pressure or level transients. Off site power remains available. No release occurred nor is ongoing. Emergency Core Cooling Systems did not actuate. No unexpected isolations occurred. Emergency Plan entry was not required. The plant is stable at normal temperature and pressure. The electrical system is in a normal offsite power alignment. The Unit 2 Reactor Trip had no effect on Unit 1 which continues to operate at 100% power. The licensee has notified the NRC Resident Inspector.
On June 27, 2012 at 2046 CDT, a Unit 2 Manual Reactor Trip was initiated in anticipation of an automatic turbine trip due to operators noticing the turbine governor valves closing and turbine load reduction. It was originally reported that the cause was not known. This notification is updated to provide information to the cause of the load reduction. Troubleshooting has shown that the reduction of turbine load was due to a turbine speed channel card failing in the turbine control system. The card failure resulted in sending the auxiliary governor in the turbine control system an incorrect indicated overspeed condition and throttling the turbine governor valves to reduce the turbine speed. There was no actual turbine overspeed condition. The auxiliary governor is not part of the reactor protection system. No reactor protection setpoints were exceeded. All other plant systems functioned as required, including the Reactor Protection System. All control rods fully inserted into the core due to the manual reactor trip. There was no Emergency Core Cooling System actuation. No Emergency Diesel Generators were started and power continued to be supplied from off site. The reactor coolant system had forced circulation and the condenser steam dumps were used for decay heat removal from the steam generators. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley). | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Decay Heat Removal Control Rod | |
| ENS 46957 | 14 June 2011 00:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip During Startup Physics Testing | On June 13, 2011 at 1924 CDT, an automatic reactor trip was actuated during Shutdown Bank Insertion for Beginning of Life Physics Testing. Neutron flux lowered from the intermediate range to the point of automatic energization of the Source Range Nuclear Instruments. Energization of the Source Range Nuclear Instruments during this step is anticipated and permissible per the test procedure. When it energized, the N-31 Source Range Nuclear Instrument (Red Channel) failed high which initiated an automatic reactor trip. All plant systems functioned as required, including the Reactor Protection System. All control rods fully inserted into the core. There was no Emergency Core Cooling System or Auxiliary Feedwater System actuation. No Emergency Diesel Generators were started and power continues to be supplied from off site. The reactor coolant system has forced circulation and the atmospheric steam dumps are currently being used for decay heat removal from the steam generators. There was no radiological release and emergency plan implementation was not required. An investigation of the failure of the N-31 Source Range Nuclear Instrument is in progress There is no known primary to secondary leakage. The licensee notified the NRC Resident Inspector.
At the time of the trip, the reactor was subcritical. This notification is updated to reflect reporting in accordance with 10CFR50.72(b)(3)(iv)(A). An investigation of the failure of the N-31 source range nuclear instrument continues. This event was initially reported by the licensee as a four-hour report in accordance with 10CFR50.72(b)(2)(iv)(B). The licensee notified the NRC Resident Inspector. Notified R3DO (Kozak).
'On June 14, 2011 at 2055 EDT, EN 46957 was amended to state the reactor was subcritical at the time of the trip. Subsequent review of this situation by the licensee resulted in a determination that at the time the test sequence was initiated, the reactor was critical and in MODE 2. Accordingly, this non-emergency report is amended to reflect EN 46957 as originally submitted on 06/13/11 at 2229 CDT. LER 301/2011-004-01 (submitted on July 25, 2011) will be revised accordingly.' The licensee has notified the NRC Resident Inspector. R3DO (Peterson) notified. | Steam Generator Reactor Coolant System Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Decay Heat Removal Control Rod | 05000301/LER-2011-004 |
| ENS 46129 | 27 July 2010 01:01:00 | 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 Condenser Vacuum | On 7/26/2010 at 2001 (hrs. CDT), Control Room personnel initiated a manual reactor trip from approximately 19% reactor power. Unit 1 was in the process of coming off-line to support Main Generator repair. The generator breaker had just been opened and load transferred to condenser steam dumps when a loss of condenser vacuum occurred. The reactor was manually tripped due to a loss of main condenser vacuum with reactor power above P-10 permissive. All systems functioned as expected. All control rods fully inserted. Main Steam Isolation valves were manually shut. All reactor coolant system parameters are as expected, with reactor coolant temperature being maintained by atmospheric steam dumps. Currently, the plant is at normal operating temperature and pressure with the steam generators being fed by the main feed pumps. Feedwater is being supplied via the condenser and condensate storage tank. There were no lifts of safeties or reliefs during the transient. The plant is in its normal shutdown electrical line-up with no effect on Unit 2. There is no known primary-to-secondary leakage. The cause of the loss of vacuum is under investigation. The licensee has notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Main Steam Isolation Valve Main Condenser Control Rod | |
| ENS 46080 | 9 July 2010 11:47:00 | 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 Feedwater Regulating Valve Failure | On 07/09/10 at 0647 (CDT) hours, control room personnel initiated a manual reactor trip of Unit 2 from approximately 64% power as a result of a failure of the "A" feedwater regulating valve (FRV). All (other) systems functioned as expected. All rods fully inserted into the core. The unit is stable in MODE 3 at normal RCS (Reactor Coolant System) pressure and temperature. The cause of the FRV failure is being investigated. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit electrical power is lined up to offsite power in a normal configuration. Decay heat is being removed from the steam generator through the steam dumps to the main condenser. The FRV failed open and the valve controller was unable to place the FRV into the correct position. The steam generator HI-Hi level provided a feedwater isolation signal and a high level lockout of the FRV. Currently, the feedwater bypass valve is controlling steam generator level. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Main Condenser | 05000301/LER-2010-002 |
| ENS 46032 | 19 June 2010 11:36:00 | 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 Generator Lockout | At 0636 hours, control room personnel initiated a manual reactor trip of Unit 2 from MODE 2 at 0% power. Just prior to the initiation of the manual reactor trip, an automatic turbine trip had occurred as a result of the receipt of a generator lockout signal. Prior to the automatic turbine trip, an orderly power reduction to 44% power had been completed and the unit was being maintained in a stable condition in MODE 1. At the time of the turbine trip, two sets of condenser steam dump valves were isolated, in preparation for scheduled condenser waterbox tube cleaning. Following the manual reactor trip, all safety systems and equipment operated as expected. The cause of the turbine trip, including receipt of the generator lockout signal, is under investigation. The unit is stable in MODE 3 at normal RCS temperature and pressure. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) t and 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified. Just prior to the trip, the unit was critical in Mode 2. All rods are fully inserted into the core. | ||
| ENS 43407 | 5 June 2007 20:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After Feedwater Valve Failure | On 06/05/07 at 1512 hours CDT, operators observed that the Unit 1 main feedwater regulating valve (1 FD-476B) was going full open to full shut and entered abnormal operating procedure (AOP) 2B for a feedwater system malfunction. An immediate inspection of the valve determined that the valve positioner arm was disconnected, with the positioner arm locknut found on the floor adjacent to the valve. Operators manually tripped the Unit 1 reactor at 1517 hours CDT in response to the loss of 'B' train main feedwater control. During the trip, the auxiliary feedwater system actuated due to low level in the 'B' steam generator and an actuation of the ATWS mitigating system (AMSAC). Plant systems and equipment functioned properly following the manual trip with the following exceptions: A switchyard bus section 2 lockout occurred, resulting in loss of 345 Kv Line 121; 1FD-2603 bleeder trip valve (1 HX-22A moisture reheater drain) stuck open; and 1 P-129A turbine bearing oil lift pump did not automatically start, but was successfully started manually. Troubleshooting and additional investigations are continuing. The affected equipment has been quarantined. Unit 1 is in MODE 3 and is stable. Unit 2 was unaffected by the Unit 1 manual trip. All control rods fully inserted on the trip. No safety or relief valves lifted from the trip. Reactor pressure and temperature are being maintained with main feedwater and steaming to the main condenser. Emergency Diesel Generator GO-1 (two EDGs per train) and one Service Water Pump are out of service for replacement. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Service water Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42199 | 13 December 2005 09:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Loss of Condenser Vacuum | Point Beach Unit 1 was manually tripped at 0339 CST on 12/13/05 due to a loss of condenser vacuum caused by a mechanical failure of the running circulating water pump. All plant systems responded normally, including an auxiliary feedwater actuation. The trip was uncomplicated. All rods fully inserted. MSIVs were isolated due to the loss of condenser vacuum so decay heat is being removed by the atmospheric dump valves. The licensee indicated that there are no known steam generator tube leak issues. All systems functioned as required. The licensee was not in any significant LCO at the time of the trip. The trip had no impact on the electrical lineup or on Unit 2 operations. The cause of the circ water pump failure is still under investigation but there is evidence of sheared bolts on the pump coupling. The licensee noted the turbine condenser rupture disks blew due to high pressure. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | 05000266/LER-2005-008 |
| ENS 40754 | 15 May 2004 16:54:00 | 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 Diver Entangled in Intake Crib | A Unit 2 manual reactor trip was initiated when the control room was notified that a diver was entangled in the intake crib. Divers were being used to inspect the intake crib, install buoys, and the fish deterrent system. The diver's umbilical cord became snagged and attempts to free it were unsuccessful. The Unit 2 circulating water system was secured to aid in removing the diver from the water. The diver still had breathing air available during the transient. The diver was subsequently removed from the water unhurt. Plant systems functioned as required, including the Reactor Protection and Auxiliary Feedwater Systems. There was no Emergency Core Cooling System actuation. Note: The condenser was unavailable because circulating water was secured. This caused a loss of condenser vacuum and its use as a heat sink. The atmospheric steam dumps are currently being used for heat removal from the steam generators. The circulating water system was subsequently restored to service. This event is reportable pursuant to 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 50.72(b)(3)(iv)(A), PWR auxiliary feedwater system. All control rods inserted into the core. The electrical busses are in a normal shutdown line up. The licensee notified the NRC Resident inspector. | Steam Generator Reactor Protection System Auxiliary Feedwater Circulating Water System Emergency Core Cooling System Control Rod | 05000301/LER-2004-002 |