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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55272 | 24 May 2021 13:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | on 5/24/21 at 0916 Edt, an Automatic Reactor Trip on Unit 1 Occurred. All Safety Systems Responded Normally and the Plant Is Currently Stable in Mode 3 (Hot Standby) at Normal Operating Temperature and Pressure. Preliminary Indications Are That the Unit Trip Was Caused by a High Neutron Flux Rate Detected by the Power Range Nuclear Instruments. Troubleshooting and Investigation Are Ongoing to Determine the Initiating Cause. | Unit 2 is not impacted and remains stable in Mode 1 at 100 percent power. 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) and 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. No relief valves opened. All Rods fully inserted. Decay heat is being removed by Auxiliary Feedwater via the steam dumps. The plant is in a normal post-trip electrical line-up. | Reactor Protection System Auxiliary Feedwater | |
| ENS 54446 | 16 December 2019 08:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Autostart Due to Loss of Power to 1B Shutdown Board | At 0358 EST, on 12/16/2019, with Unit 1 in Mode 1 at 100 (percent) power and Unit 2 in Mode 1 at 47 (percent) power, a valid actuation of the Emergency Diesel Generators (EDG) occurred. The reason for the emergency diesel generator auto start was that the normal feeder breaker from the 1C 6.9KV Unit Board to the 1B-B 6.9KV Shutdown Board (SDBD) tripped due to the breaker's 51G relay actuating causing an under-voltage signal on the 1B-B 6.9KV Shutdown Board. All 4 Emergency Diesel Generators automatically started as designed when the 6.9KV Shutdown Board under-voltage signal was received. The 1B-B 6.9KV Shutdown Board was automatically energized from the 1B-B 6.9KV Diesel Generator. All required 6.9KV loads were sequenced back on to the 1B-B 6.9KV Shutdown Board as designed after the board was energized from its emergency diesel generator. The remainder of the electrical system is in normal alignment. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Emergency Diesel Generators. There was no impact to the health and safety of the public or plant personnel. The NRC Senior Resident has been notified. | Emergency Diesel Generator | |
| ENS 54438 | 12 December 2019 09:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | EN Revision Imported Date : 3/4/2020 MANUAL REACTOR TRIP DUE TO A LOSS OF HEATER DRAIN TANK PUMP FLOW At 0432 EST, on 12/12/19, Sequoyah Unit 2 experienced a manual reactor trip. The trip was initiated due to a loss all number 3 Feedwater Heater Drain Tank pump flow; plant procedures directed a manual reactor trip if power is greater than 80 percent. The Auxiliary Feedwater System (AFW) automatically actuated as required when the expected post trip feedwater isolation actuation actuated. Reactor Coolant System (RCS) temperature is being maintained by the steam dump system with all 4 Reactor Coolant Pumps (RCPs) in service. All control and shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the transient. Unit 2 is currently stable at normal operating temperature and normal operating pressure in Mode 3. The electrical system is in a normal alignment. There was no impact on U1. There was no impact to the health and safety of the public or plant personnel. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10CFR50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification accordance with 10CFR50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. The NRC Resident Inspector was notified.
The following update to the EN submitted on 12/12/19 is being made to provide clarification on reporting criteria originally described in paragraph five of EN 54438: 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 was notified. Notifed R2DO (Davis). | Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater | ||
| ENS 54242 | 27 August 2019 05:09:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 0109 EDT, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a dropped rod causing a negative rate trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 2 is not affected. 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) and 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. | Reactor Protection System Auxiliary Feedwater | |
| ENS 53999 | 14 April 2019 07:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | En Revision Imported Date 8/7/2019 | EN Revision Text: AUTOMATIC REACTOR TRIP DUE TO MAIN FEEDWATER PUMP TRIP At 0320 EDT, April 14, 2019, Sequoyah Unit 1 experienced an automatic reactor trip. The event was initiated by the trip of the 1A main feedwater pump. During the automatic unit runback, an automatic reactor trip was initiated due to low-low level in Steam Generator number 3. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. During this operational cycle, one control Rod Position Indicator (RPI) for core position E-5 in shutdown bank 'A' has been inoperable, and the appropriate Condition and Required Actions of (Technical Specification Limiting Condition of Operation) 3.1.7 were complied with. Due to this inoperable RPI, the associated shutdown rod is conservatively assumed to be full out and untrippable. Consequently, boration was required to establish adequate shutdown margin. All other Control and Shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. There was no impact on Unit 2. 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) and 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 Senior Resident Inspector has been notified.
The licensee provided an update to paragraph 2. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. All Control and Shutdown rods fully inserted, except E-5 which was previously identified and conservatively assumed to be in a full out position. Applicable TS actions were performed to maintain shutdown margin. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. Notified the R2DO (Gerald McCoy) | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 52769 | 24 May 2017 03:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | All Four Edgs Autostarted During Transfer of Shutdown Board to Normal Power Source for Unit 2 Testing | On May 23, 2017 at 2330, while transferring 2A-A 6.9 kV Shutdown Board from its alternate power source to its normal power source in support of outage testing, a failure occurred which resulted in the loss of the Shutdown Board, emergency start of all 4 Emergency Diesel Generators (EDGs), and required the manual emergency stop of 2A-A EDG. During transfer of the 2A-A 6.9kV Shutdown Board, the hand switch for the normal feeder breaker on the shutdown board was being maintained in the 'CLOSE' position while the alternate feeder breaker hand switch was placed in 'TRIP.' As expected, the alternate feeder breaker opened and the normal feeder breaker closed. However, the upstream supply breaker to the normal feeder breaker immediately tripped due to an overcurrent relay actuation on a single phase. As a result, the 2A-A 6.9 kV Shutdown Board deenergized, initiating a blackout signal which started all 4 of the station's EDGs. During board stripping (opening of all feeder and load breakers, to prepare the board for automatic reenergization from the EDG), the normal feeder breaker to the Shutdown Board failed to trip. This failure to trip prevented the emergency feeder breaker in the output of 2A-A EDG from closing, in accordance with interlock logic. As a result, 2A-A 6.9 kV Shutdown Board remained deenergized which prevented the cooling water supply valve for the EDG from opening due to loss of motive power. This lack of cooling caused operators to emergency stop the 2A-A EDG. Power was restored to the Shutdown Board on May 24, 2017 at 0037. Unit 1 is currently stable in Mode 1, at 100% power and Unit 2 is stable in Mode 5 with RCS at 164 F and 340 psig. The cause of the breaker trip on overcurrent and the failure of the normal feeder to trip on load shedding are under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | 05000327/LER-2017-002 |
| ENS 52469 | 30 December 2016 18:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Control Rod Not Withdrawing as Expected | On 12/30/16 at 1302 EST, Unit 1 began withdrawing control bank rods for an approach to criticality following a refueling outage. At 1305 EST, operators observed that control rod H-6 did not withdraw. Operators entered the applicable Abnormal Operating Procedure. Operators tripped the reactor as required by plant procedures and entered applicable Emergency Operating Procedures. The act of manually tripping the reactor generated a valid trip signal in the plant Reactor Protection System. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater was already supplying the Steam Generators; a Feedwater Isolation occurred due to plant conditions. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators, and decay heat removal via the steam dumps. Method of decay heat removal is Steam Generators via the steam dumps. Current reactor coolant system conditions: Temperature at 548 degrees F and stable, pressure 2235 psig and stable. All control and shutdown banks are inserted. Electrical alignment is normal, supplied by offsite power. No impact to Unit 2, Unit 2 is in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 51720 | 9 February 2016 19:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Eccs Discharge to Rcs Via Charging System | At 1415 EST on 02/09/2016, Sequoyah Unit 1 was at 0 percent power (mode 3, 526F, 2235 psig) when a low steam line pressure Safety Injection actuated from Loop 2 Steam Generator. Prior to this event, the Loop 2 Main Steam Isolation Valve bypass was opened at 1413 EST for main steam line warm up in preparation for unit startup. Loop 2 Main Steam Isolation Valve bypass closed automatically following low steam line pressure Safety Injection. Following the Safety Injection, all safety-related equipment operated as designed. Current Reactor Coolant System temperature and pressure - Unit 1 is currently being maintained in Mode 3 at approximately 517 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via steam generator atmospheric relief valves. There is no indication of any primary to secondary leakage. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. The cause of the Safety Injection actuation is under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. Due to RCS pressure, the only system that injected into the RCS was the charging system. The AFW system initiated to feed the steam generators and the Emergency Diesel Generators started but did not load. | Steam Generator Reactor Coolant System Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Main Steam Line | |
| ENS 51559 | 23 November 2015 13:44:00 | 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 Partial Msiv Closure | At 0820 EST on 11/23/2015, Sequoyah Unit 1 was at 100% power when operators identified the Loop #3 Main Steam Isolation Valve (MSIV) came off its full open seat. This was evidenced by no OPEN indication on the main control board, dual indication on the post accident monitoring panel, and a change in both Tavg and steam pressure. Operators were dispatched locally to the MSIV and to the battery board room to investigate if a cause could be identified for the MSIV movement. The field investigation identified no issues. The operating crew manually tripped the reactor at 0844 EST due to an increasing Tavg-Tref mismatch. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the feedwater isolation signal. Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure, approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. There is no indication of any primary to secondary leakage. All control rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2 as it continues through the refueling outage with the core off-loaded. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. All control rods fully inserted during the reactor trip. The atmospheric steam dumps did operate during the transient and then shut. After the trip, the MSIV re-opened. | Steam Generator Feedwater Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 51392 | 14 September 2015 08:26:00 | 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 Power to Vital Instrument Board | At 0426 EDT on 9/14/2015, Sequoyah Unit 1 was at 100% power when the Vital Instrument Power Board (VIPB) 1-II deenergized. A manual reactor trip was initiated in accordance with the Abnormal Operating Procedure for the loss of VIPB 1-II. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP (Normal Operating Temperature/Normal Operating Pressure), approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 547 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary to secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100 (percent). There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 51265 | 27 July 2015 14:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor/Turbine Trip | At 1043 EDT on 7/27/2015, Sequoyah Unit 1 was at 82% power and continuing to perform a startup when the reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3, at NOT/NOP (normal operating temperature and normal operating pressure), approximately 545 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. The immediate cause of the trip was an electrically-induced turbine trip. Due to fluctuating voltage the main generator voltage regulator was taken to manual; immediately after this the unit tripped. Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 545 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 51259 | 24 July 2015 17:51:00 | 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 on Main Generator Lockout | At 1351 EDT on 7/24/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. The immediate cause of the trip was an electrically-induced turbine trip. There was no associated work in progress related to this and all systems were normally aligned. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. The 2B-B Emergency Diesel Generator is currently in service for the performance of an unrelated surveillance. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. The cause of the main generator lockout is under investigation. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal | |
| ENS 50878 | 11 March 2015 10:21:00 | 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 Nuclear Instrumentation Signal | At (0621 EDT) on 3/11/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a Power Range Nuclear Instrumentation negative rate signal, caused by a malfunction in the rod control system. There was no associated work in progress related to this and all systems were normally aligned. Current Temperature and Pressure - temperature is 547 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 50856 | 2 March 2015 11:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 06:45 EST on 3/2/2015, Sequoyah Unit 2 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 2 is currently being maintained in Mode 3 at NOT/NOP (normal operating temperature and pressure), approximately 548 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a main generator C phase differential relay actuation. There was no associated work in progress related to this and all systems were normally aligned. It is currently not understood why the relay actuated. Current Temperature and Pressure - temperature is 548 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. The 2B Emergency Diesel Generator is currently unavailable for planned maintenance and will be returned to service prior to unit restart. There is no operational impact to Unit 1. Unit 1 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector" and the state. No primary or secondary safety valves lifted during this event. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal | |
| ENS 48778 | 24 February 2013 17:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Condenser Vacuum Indication | On February 24, 2013 at 1205 (EST) with reactor power at 25% and the turbine offline, a manual reactor trip for Sequoyah Unit 2 was initiated due to loss of condenser vacuum indication causing closure of condenser steam dumps, opening of the Steam Generator Atmospheric Relief Valves, and lowering hotwell level resulting in imminent loss of hotwell pumps. The cause of the event was determined to be a faulty test connection on B Condenser vacuum pressure switch. During the event, steam pressure rose to the setpoint for the first Steam Generator code safety valve (1064 psig). (The safety valve opened, then reseated). Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. The reactor trip was uncomplicated. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (Normal Pressure and Temperature), with auxiliary feedwater supplying the steam generators and maintaining level at approximately 33% narrow range. Method of decay heat removal is via atmospheric reliefs to the atmosphere. Current RCS conditions: temperature (is) 547 degrees F and stable. Pressure (is) 2235 psig and stable. (There is) no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal and supplied from offsite power. (There is) no impact to Unit 1. Unit 1 is operating at 100% power / Mode 1. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart is 02/25/2013. (The licensee plans a press release.) The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 48198 | 16 August 2012 23:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip from a Reactor Coolant Pump Trip | At 1926 EDT on 8/16/2012, Unit 2 reactor automatically tripped on single loop loss of flow, following #4 RCP trip. Cause of RCP trip is under investigation. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (normal operating pressure/normal operating temperature), with auxiliary feedwater supplying the steam generator. Method of decay heat removal is via steam dumps to condenser. Current RCS conditions: Temp = 547 degrees F and stable, Pressure = 2235 psig and stable. No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from offsite power. No impact to Unit 1: Unit 1 is operating at 100% power / Mode 1. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 47169 | 19 August 2011 02:50:00 | 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 Reactor Coolant Pump Bus Undervoltage | At 2250 EDT on 8/18/2011, Unit 1 Reactor/Turbine automatically tripped on RCP (Reactor Coolant Pump) Busses UV (Under-Voltage) trip. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. No primary PORVs and/or Safety Valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F (degrees) and 2233 psig, with Auxiliary Feedwater supplying the Steam Generators. At the time of the trip, a 50G (instantaneous overcurrent ground) relay flag was found dropped on the '1A' 6.9 KV unit board. Subsequently, the '1A' 6.9 KV start bus was found to have transferred to its alternate supply, 'B' CSST (Common Station Service Transformer). 1A condenser circulating water pump motor trip out was also received in the MCR (Main Control Room). The method of decay heat removal is via steam dumps to the condenser with MSIVs open. The current temperature and pressure is stable. There is no indications of any primary/secondary leakage. All control rods inserted. The electrical alignment is normal with the exception of the above mentioned items, supplied from off-site power. There is no impact to Unit 2. Unit 2 is operating at 100% power/ Mode 1. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 47081 | 21 July 2011 01:29:00 | 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 Negative Rate Trip Following Rapid Turbine Load Reduction | At 2129 EST on 7/20/2011, Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. Primary PORVs and/or safety valves lifted and reseated as indicated by tailpipe temperatures and PRT pressure. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with Auxiliary Feedwater supplying the steam generators. At the time of the trip, maintenance was in progress on Preferred Inverter #1. AOP-P.09 'Loss of 120VAC Preferred Power' was used to restore power to #1 Preferred board after the trip. Method of decay heat removal is via steam dumps to the condenser with MSIVs open. Current temperature and pressure: Temperature - 548 degrees Fahrenheit and stable, Pressure 2235 - psig and stable No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from off-site power. No impact on Unit 2. Unit 2 is operating at 100% power/Mode 1 The NRC Resident Inspector has been informed. The licensee notified the State of Tennessee. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 46991 | 26 June 2011 20:14:00 | 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 | At 1614 EDT on 6/26/2011, Unit 1 reactor automatically tripped following a turbine trip from greater than 50% rated thermal power (P-9 interlock). Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected on loss of the operating main feedwater pumps. Initially, the steam dump system functioned as expected (all valves opened). Subsequently, the steam dump system was manually turned off when 3 of the valves did not close when expected. Consequently, decay heat removal is via the Steam Generators' atmospheric relief valves. No primary or secondary safety valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators. There is no indications of any primary to secondary leakage. All control and shutdown rods are inserted. The electrical alignment is normal, supplied from off-site power. There was no impact to Unit 2. Unit 2 is operating at 100% power / Mode 1 The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 46492 | 20 December 2010 05:50:00 | 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 a Fire in the Main Generator Bus Duct | At 0050 EST on 12/20/2010, Unit 1 reactor was manually tripped due to a reported fire inside the Unit 1 main generator bus duct. A fire was reported at 0045 EST in the Unit 1 bus duct which is located inside the turbine building. The Unit 1 reactor was tripped to remove power from the generator bus. After the reactor trip, the fire was extinguished by the application of water to the bus duct by the fire brigade. The fire was reported extinguished at 0100 EST. Method of decay heat removal is via steam dumps. Current temperature and pressure: Temp. 548 degrees and stable; Pressure 2239 psig and stable. No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from off-site power. No impact to Unit 2, (and) Unit 2 is operating at 100% power (in) Mode 1. The NRC Resident Inspector has been notified. | Decay Heat Removal | |
| ENS 46424 | 17 November 2010 03:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Based on Decreasing Steam Generator Level | At 2210 EST on 11/16/2010, the Unit 1 reactor was manually tripped based on decreasing S/G level in Loop 4. Prior to the reactor trip power was at 26% and ascending following completion of a scheduled refueling outage. Moisture Separator Reheater 1C1 safety valve lifted and would not reseat. Turbine was tripped at 2206 EST to isolate steam leak. Following the Turbine trip, automatic S/G level control did not maintain S/G level. Manual control was taken however, S/G level could not be recovered. A manual reactor trip was initiated due to low narrow range S/G level. All other plant systems responded as expected. All rods fully inserted into the core during the trip. The reactor is at normal operating pressure and temperature and operators are removing decay heat via the steam dumps to condenser. As expected, the auxiliary feedwater system actuated during the transient. The grid is stable and the plant is in its normal shutdown electrical lineup. There is no known primary-to-secondary leakage. Plant response to the trip was considered normal and uncomplicated. Unit 2 was not affected by this event. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 45520 | 26 November 2009 07:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Trip Due to Loss of Vacuum on Turbine Driven Feedwater Pump | At 0242 EST, 11/26/2009 Unit 2 reactor was manually tripped based upon indications that the 2A Main Feedwater Pump Turbine was losing vacuum. Prior to the trip, the reactor was at 30% RTP and ascending following completion of a scheduled refueling outage. 2A Main Feedwater Pump was in service; preparations were in progress to start 2B Main Feedwater Pump to support continued power ascension. Subsequent to the trip, the Auxiliary Feedwater Pumps (motor-driven and turbine-driven) started as expected in response to the trip of both main feedwater pumps following receipt of a normal feedwater isolation signal. No primary or secondary plant safety valves operated during the transient. All plant system responses to the trip were as expected. An investigation will be conducted to identify the cause of the indicated loss of vacuum and the required corrective actions. Expected restart date to be determined. All control rods fully inserted. The licensee has notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 45097 | 27 May 2009 23:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip from Power Range High Neutron Flux Rate | At approximately 1904 on 05/27/09, Sequoyah Unit 2 received an Automatic Reactor trip on Power Range High Neutron Rate Reactor Trip. The automatic trip occurred during severe weather and lightning onsite. ESF functions initiated as designed including Aux Feed Water auto-start and automatic Feedwater isolation. All control rods fully inserted. The plant is currently being maintained in Mode 3 at approximately 547 degrees F 2235 PSIG. Decay heat is being removed by the Auxiliary Feedwater system and Steam Dump valves. The cause of the Power Range Neutron Flux High Rate Reactor Trip is not known at this time and investigation is ongoing. Unit 1 remains at 100% power". No relief or safety valves lifted. The electrical lineup is normal and the 2B Diesel Generator is tagged out for maintenance. The licensee has contacted the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 45045 | 7 May 2009 02:56:00 | 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 Feed Regulating Valve Failure | On 5/6/09 at 2256 Sequoyah Unit-1 was manually tripped from 100% reactor power. The manual trip was in response to the failure of Loop 1 Feed Water Regulating Valve (FRV). Manual control was attempted to control level in Loop 1 Steam Generator however Loop 1 FRV failed to respond. A manual reactor trip was initiated as a result of this failure. In addition, Auxiliary Feedwater (AFW) initiated as required due to a Feedwater Isolation signal. The Loop 1 FRV did not isolate from the Feedwater Isolation signal, however the Loop 1 Feedwater Isolation Valve closed as designed. The Plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 (degrees) F and 2235 psig, with Auxiliary Feedwater supplying the steam generators and Steam Dumps to Main Condenser removing decay heat. Maintenance activities have been initiated to repair the Loop 1 FRV. All rods inserted on the trip. No safety or relief valves lifted as a result of the transient. The plant is in its normal shutdown electrical lineup. No grid instabilities exist and there was no effect on Unit 2. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 45029 | 29 April 2009 01:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Initiated as a Result of a Loss of Condensate Flow | At 2159 Sequoyah Unit 1 was manually tripped from 18% reactor power. The manual trip was in response to automatic isolation of all intermediate pressure feedwater heater strings resulting in a loss of all condensate flow. The feedwater heater strings isolated following a manual trip of the main turbine. The main turbine was manually tripped in response to a failed open moisture separator reheater relief valve. At the time of the event Sequoyah Unit 1 was raising power following a refueling outage. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary feedwater automatically actuated as expected on loss of the operating main feedwater pump. The plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps to main condenser removing decay heat. Maintenance activities have been initiated to repair the MSR relief valve. Electrical lineup is normal with all safety related equipment operable. Licensee is investigation the cause for the relief valve failure. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine Main Condenser | |
| ENS 44935 | 26 March 2009 08:52:00 | 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 Aux Feed Water Start Due to Loss of an Off Site Power Source | At approximately 0452 on 03/26/09, Sequoyah Unit 2 received an Automatic Reactor trip on Reactor Coolant Pump Busses Undervoltage. A loss of Common Service Station Transformer C caused a loss of power to the 2B and 2D Unit Boards. The 2B and 2D unit boards are the 6.9Kv electrical feeds to the 2-2 and 2-4 RCPs, respectively. RCPs 2-1 and 2-3 are running. ESF functions initiated as designed including Aux Feed Water auto-start, automatic Feedwater isolation, and auto-start of all four EDGs. The 2A Shutdown Board is being powered from the 2A EDG. The plant is currently being maintained in Mode 3 at approximately 547 degrees F /2235 PSIG. Decay heat is being removed by the auxiliary feedwater system and Steam Generator Atmospheric Relief valves. The cause of the loss of Common Service Station Transformer C is not known at this time and investigation is ongoing. All rods inserted as expected. No safety related equipment is out of service. Unit 2 has no known Steam Generator Tube leaks. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 44934 | 26 March 2009 08:52:00 | 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 Aux Feed Water Start Due to Loss of an Off Site Power Source | At approximately 0452 on 03/26/09, Sequoyah Unit 1 received an Automatic Reactor trip on Reactor Coolant Pump Busses Undervoltage. A loss of Common Service Station Transformer C caused a loss of power to the 1B and 1D Unit Boards. The 1B and 1D unit boards are the 6.9Kv electrical feeds to the 1-2 and 1-4 RCPs, respectively. RCPs 1-1 and 1-3 are running. ESF functions initiated as designed including Aux Feed Water auto-start, automatic Feedwater isolation, and auto-start of all four EDGs. The 1A Shutdown Board is being powered from the 1A EDG. The plant is currently being maintained in Mode 3 at approximately 547 degrees F /2235 PSIG. Decay heat is being removed by the auxiliary feedwater system and Steam Generator Atmospheric Relief valves. The cause of the loss of Common Service Station Transformer C is not known at this time and investigation is ongoing. All rods inserted as expected. No safety related equipment is out of service. Unit 1 has no known Steam Generator Tube leaks. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 44649 | 9 November 2008 23:21:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Breakers Opened Manually Due to Indicated Dropped Control Rod | On 11/19/08 at 1821 with unit 2 in Mode 3, the unit 2 reactor trip breakers were opened from the Main Control Room (MCR) due to indications of a Shutdown Bank "A" Rod E-11 dropping into the Reactor Core. At the time the reactor trip breakers were opened, the MCR Operators were in the process of withdrawing Shutdown Banks in preparations for entry into Mode 2. All other Shutdown Banks and Control Banks were inserted at the time the reactor trip breakers were opened. In addition, a Feedwater Isolation Signal was generated as designed. All safety related equipment operated as designed. The Plant is being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with AFW supplying the S/G's and Steam Dumps to Main Condenser removing decay heat. No primary system or steam generator safety valves opened due to this trip. An investigation has been initiated to determine the cause of the indications of Shutdown Bank 'A' Rod E-11 dropping into the Reactor Core. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Main Condenser Control Rod | |
| ENS 44627 | 4 November 2008 04:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Feedwater Reg Valve Failure | On 11/3/08 at 2322, Unit 2 was manually tripped due to the failure of Loop 4 Feed Water Reg. Valve (FRV Controller). Manual control was attempted to control level in Loop 4 Steam Generator (S/G); however, Loop 4 FRV failed to respond. A manual Reactor Trip was initiated as a result of this failure. In addition, Auxiliary Feedwater (AFW) initiated as required due to a Feedwater isolation signal The Loop 4 FRV did isolate from the Feedwater isolation signal. The Plant is being maintained in Mode 3 at NOT/NOP, 547 F and 2235 psig, with AFW supplying the S/G's and Steam Dumps removing decay heat. Additionally, Unit 2 has an indication of a primary leak inside lower containment. The leak rate is calculated to be approximately 2.0 gallons per minute. Based on current indications, the leak is suspected to be from a Pressurizer level transmitter No primary system or steam generator safety valves opened due to this trip. An investigation has been initiated to determine the cause of the Loop 4 Feed Water Reg. Valve (FRV Controller) failure and the source of the approximately 2.0 gallons per minute primary leak. A recovery plan will be developed. Pressurizer (PRZ) level is stable. All safety related systems are available and OPERABLE for safe plant shutdown. There is no impact on Unit 1. The loop 2 S/G blowdown sample line valve did not go closed as expected on the Feedwater isolation signal. The licensee is taking action to isolate it. The licensee notified the NRC Resident Inspector.
The Sequoyah Nuclear Plant Unit - 2 reactor coolant system unidentified leakage was terminated on Nov. 4, 2008 @ 2205 hours when a manual valve was closed isolating the reactor coolant system from the leak location." The licensee closed the root valve to the PZR pressure channel 2-VLV-68-446A. The licensee will notify the NRC Resident Inspector. Notified R2DO(Desai) | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater | |
| ENS 43909 | 17 January 2008 00:25:00 | 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 Lowering Steam Generator Level | During a surveillance test on the Loop 3 Steam Generator pressure channel, the loop 3 Steam Generator main feedwater regulating valve went closed. The unit was manually tripped on lowering steam generator level. All control rods fully inserted on the reactor trip. Both motor driven and the turbine driven AFW pumps started successfully. The turbine driven AFW pump was secured and the motor driven AFW pumps are providing feedwater to the steam generators. Decay heat is being removed to the main condenser via the turbine steam dump valves. No PORV, safety relief valves, or atmospheric dump valves opened on the reactor trip. Unit 1 is in a normal shutdown electrical lineup. There was no effect on Unit 2 during this event. The licensee is investigating the unexpected closure of the Loop 3 Steam Generator main feedwater regulating valve. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Safety Relief Valve Main Condenser Control Rod | |
| ENS 43233 | 13 March 2007 19:27:00 | 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" Main Feedwater Pump | At 1527 EST on 3/13/07 a manual reactor trip was actuated on Unit 2 due to 'A' MFPT (main feedwater pump turbine) malfunction. All systems responded as expected following the manual trip. The plant is currently stable in Mode 3 (Hot Standby) at 547 (degrees) F. No primary system or steam generator safety valves opened due to this trip. AFW (auxiliary feedwater pumps) system started and operated as designed. All emergency core cooling systems, emergency diesel generators are fully operable if needed and the electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System | |
| ENS 43115 | 23 January 2007 17:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Feedwater Regulating Valve Failure | An automatic reactor scram occurred resulting from loop #2 steam generator low level. The main feedwater regulating valve 2-FCV-3-48 failed closed when the air supply line to the valve severed. A significant leak had been identified and efforts to repair had been initiated. All systems responded as designed. The plant is stable on Auxiliary Feed Water (AFW) and steam generator cooling (to the condenser via steam dump valve) while repairs to the failed air line are planned. All control rods fully inserted into the core. No steam generator or pressurizer relief valves lifted during the transient. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 42444 | 22 March 2006 21:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip / Reactor Trip Caused by Main Generator Trip | This notification fulfills the 4-hr. non-emergency reporting requirement for Reactor Protection System actuation and the 8-hr. non-emergency reporting requirement for an ESF actuation (specifically Auxiliary Feedwater System start). At 1624 EST, Unit 2 reactor tripped from full power due to main generator trip caused by actuation of the main generator neutral overvoltage relay (the 100% stator ground fault relay also actuated). Safety Injection did not actuate and was not required to actuate. All automatic actions occurred as designed (e.g., steam dump operation, feedwater system, etc.). The operating crew responded to the trip using emergency procedures. The unit is being maintained in Mode 3 (Hot Standby) using General Operating Instructions. No primary steam or steam generator safety valves opened due to this trip. An investigation has been initiated to determine the cause of this unit trip. A recovery plan will be developed. All control rods fully inserted on the reactor trip. Decay heat is being removed by auxiliary feedwater feeding the steam generators and steaming through the steam dump valves to the main condenser. Offsite power is stable and all emergency busses are being powered via offsite power. All Emergency Diesel Generators are available if required. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 41583 | 9 April 2005 15:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Turbine Trip Due to Low Auto Stop Oil Pressure on Pressure Switches | At 11:11 Unit 1 Turbine Tripped initiating Reactor Trip. Initial Indications are Turbine Trip due to Lo Auto Stop Oil Pressure on 2 pressure switches. Investigating Switch Actuations.
Current Conditions: U1 (Unit 1) Mode 3, Stm Dumps In Service Normal, RCS temp at 547.4F, Pressure 2238 psig. 2 AFW (Auxiliary Feedwater) trains In Service at Normal Operation. All Rods Fully Inserted. Yes. The licensee indicated there was not excessive cooldown or depressurization in the steam generators whose atmospheric relief valves had to be manually closed. The NRC Resident Inspector has been notified.
AFW Actuation - due to FW (Feedwater) Isol(ation) tripping both MFPs (Main Feed Pumps) (trips start AFW) AFW includes 2 motor-driven and 1 turbine-driven pump. R2DO (Rogers) notified. | Steam Generator | |
| ENS 41437 | 24 February 2005 02:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Low-Low Steam Generator Level | The following information, in addition to the phone report, was obtained from the licensee via facsimile: While performing maintenance, (Maintenance) inadvertently opened a 125 VDC battery board (four) breaker which resulted in a reactor trip from steam generator low-low level. The plant is being maintained in Mode 3 at NOP/NOT (Normal Operating Pressure/Normal Operating Temperature), 547 (degrees) and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps removing the decay heat. All rods inserted on the trip. No relief valves lifted during the transient. The electric plant is stable with 125 VDC restored to a normal configuration. Steam generator level has been restored to normal levels. Unit 1 was not affected by the transient. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 40589 | 15 March 2004 20:18:00 | 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 Generator Electrical Fault Signal | At 1518 on 3/15/04, Unit 1 Reactor tripped due to a Main Turbine trip. The Main Turbine tripped due to a Main Generator electrical fault, Auxiliary Feedwater (AFW) started when both Main Feedwater Pumps tripped on Low Tave Feedwater isolation. The AFW start was expected on the Reactor Trip. All safety systems performed as required. Investigation is in progress to determine and correct the cause of this trip. All control rods fully inserted; the electrical grid is stable; ECCS systems remain operable; decay heat is being removed via AFW and steam dumps. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Main Turbine Control Rod | |
| ENS 40475 | 25 January 2004 08:16:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Auto Start of Turbine Driven Auxiliary Feedwater Pump | At 0316, EST, on 25 January 2004, the Unit 2 turbine driven auxiliary feedwater pump (TDAFWP) was automatically started upon receipt of a valid start signal. The start of the pump was unplanned. At the time, the Unit 2 reactor was critical at (approximately) 0.001 % power as indicated on the Intermediate Range Nuclear Instruments (Unit 2 is emerging from a forced outage following main generator repairs). The steam generators were being supplied water by the motor-driven auxiliary feedwater pumps. Shift operators were in the process of aligning the Unit 2 condensate system for cleanup and closed both (out of service) main feedwater pumps' suction isolation valves, as required by the procedure. Closure of a main feedwater pump's suction valve generates a main feedwater pump trip signal. Trip of both main feedwater pumps generates an automatic auxiliary feedwater system start signal. The start of the TDAFWP had no adverse effect on the unit. Operators reset one of the main feedwater pumps and secured the TDAFWP. The Licensee will notify the NRC Resident Inspector.
Background: On January 25, 2004, Unit 2 was being returned to operation following a maintenance outage. Unit 2 was critical (Mode 2) with auxiliary feedwater pumps 2A and 2B in service and both MFW pumps trip buses energized. Operations continued to align the condensate system, on long cycle for clean-up. This alignment requires the closure of the main feedwater pumps suction valves. The main feedwater pumps had already been aligned for service with the trip buses energized. The plant design results in an automatic start of the auxiliary feedwater pumps when both MFW pumps trip buses trip. On January 25, 2004 at approximately 0316 hours, the Unit-2 TDAFW pump automatically started when an operator inappropriately closed the suction valves to both MFW pumps. The closure of the pumps suction valves initiated an invalid AFW pump start signal, resulting in the start of the TDAFW pump and the steam generator level control valves went to automatic. The two motor driven AFW pumps received a start signal though they were already in service for plant startup. This actuation was not the result of actual plant conditions or parameters (there was no feedwater flow through the MFW pumps); rather it was the result of an inappropriate action in closure of the MFW pumps suction valves with both MFW pumps trip buses energized. This event was originally reported on January 25, 2004 as a valid auxiliary feedwater system start signal. Further review has determined the signal to be invalid as described above. REPORT: This event was originally reported on January 25, 2004 as a valid auxiliary feedwater system start signal. Further review has determined the signal to be invalid as described above. The following information is provided in accordance with 10 CFR 50.73(a)(2)(iv)(A): 1. Both the Unit 2 Train A and B motor-driven auxiliary feedwater pumps were running, the turbine driven auxiliary feedwater pump started and the steam generator level control valves went to automatic. 2. Complete train actuation occurred. 3. The turbine-driven auxiliary feedwater pump start and the steam generator level control valves going to automatic were successful. Notified the R2DO (OGLE). The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 40305 | 10 November 2003 04:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Sequoyah Unit 2 Auxiliary Feedwater Actuation During Outage | On 11/9/03 at 2315 (EST), with Unit 2 in Mode 3 in preparation for a Refueling Outage, an unanticipated ESF actuation occurred. With one Main Feedwater Pump (MFP) reset and one MFP tripped per the General Operating Procedures, Main Condenser Vacuum was being broken. Low Condenser vacuum caused the trip of the reset MFP. The signal for trip of two MFPs caused an Aux Feedwater (AFW) start. AFW was in service in manual control per procedure when the start signal was received; however, the start signal caused the Level Control Valves (LCVs) to reposition to their accident position. The licensee notified the NRC Resident Inspector.
On November 9, 2003, after the planned Unit 2 reactor shutdown for refueling outage number 12, an auxiliary feedwater system start signal was initiated. During the shutdown, the auxiliary feedwater pumps (AFW) were started and both main feedwater pumps (MFP) had been tripped. Following transition from emergency operating procedure to the normal plant shutdown procedure, the feedwater isolation signal and one MFP trip were reset to allow shutdown of the turbine-driven pump, in accordance with plant procedures. As a result of the reactor coolant temperature decreasing below 540 degrees Fahrenheit, emergency boration was performed, and subsequently, control room personnel closed the main steam isolation valves (MSIVs) to limit cooldown of the RCS. After closure of the MSIVs, Operations personnel broke main condenser vacuum in accordance with plant procedures. The operators knew by their procedure that this would result in a low-condenser vacuum trip signal to the reset MFP trip bus and, with both MFPs trip bus tripped, the logic would be completed for initiation of an AFW start signal. The procedure states 'Breaking main condenser vacuum will cause loss of main feed pump turbine condenser vacuum to both MFPs resulting in an ESF actuation.' On November 10, 2003, at 0039 Eastern standard time, NRC was notified in accordance with 10 CFR 50.72 (b)(3)(iv)(A) of the ESF actuation (event notification No. 40305). 10 CFR 50.72 and 50.73 requires reporting of any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. NUREG 1022 states that the intent is to require reporting of actuations of systems that mitigate the consequences of significant events. The breaking of the condenser vacuum was driven by plant conditions. However, the start signal to the AFW system was not to mitigate an event for the condition the plant was in (i.e., both MFPs had been secured and decay heat was being controlled by AFW in accordance with normal operating procedures) and is therefore considered a preplanned event. Based on the above information, the initiation of the AFW start signal is not reportable and the notification is being withdrawn. The licensee has notified the NRC Resident Inspector. Notified R2DO(Ayres) | Feedwater Main Steam Isolation Valve Auxiliary Feedwater Main Condenser | |
| ENS 40113 | 28 August 2003 20:04:00 | 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 | Alert Declared Due to Reactor Protection System Failure | At 1618 EDT on 08/28/03, an Alert was declared at Sequoyah Unit 1 due to a failure of the Reactor Protection System (RPS) to auto-trip the reactor following a turbine trip. Turbine oil trip testing was in progress when at 1604, the main turbine tripped. The reactor trip breakers were opened by manual operator action approximately 20 seconds after the valid reactor trip signal. Following the trip, all systems operated as required. The main steam isolation valves were manually closed based on indication of steam flow and a throttled steam dump valve. No primary system or steam generator safety valves opened. The unit is currently stable with temperature controlled using atmospheric relief valves. The Auxiliary Feedwater System received an auto-start signal and successfully operated. All control rods were fully inserted into the core. There was no release of radioactive material. The licensee has initiated an investigation into the cause of the incident. The NRC Resident Inspector and the State of Tennessee were notified by the licensee. The NRC notified FEMA (Steiner), DOE (Morroni), USDA (Beers-Block), HHS (Williams), DHS (Glick), and DOT/EPA/NRC (Threatt). The NRC entered monitoring phase of normal mode with Region 2 in the lead at 1735 EDT.
At 2030 EDT on 08/28/03, the licensee terminated the Alert at Sequoyah Unit 1. The reactor is stable and in Mode 3, Hot Standby. This termination is based upon containment integrity, Emergency Core Cooling System and Engineered Safety Features operability, availability of heat sink, onsite and offsite electrical system and Emergency Diesel Generator operability, radiation monitor operability, availability of technical support personnel, and activation of a forced outage recovery team. Reactor coolant samples were taken, which showed no increase in activity. The NRC notified FEMA (Austin), DOE (Morroni), USDA (Beers-Block), HHS (Hogan), DHS (Van Buskirk), and DOT/EPA/NRC (White). The NRC exited monitoring phase of normal mode at 2045 EDT. | Steam Generator Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Main Turbine Control Rod |