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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56087 | 3 September 2022 02:48:00 | 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 Feed Pump | The following information was provided by the licensee via email: On 09/02/2022 at 22:48 with Unit 1 at 40% power, the reactor was manually tripped due to a loss of the only operating main feed pump which caused lowering level in the steam generators. All systems responded as expected following the trip. Auxiliary feed actuation signal occurred due to lowering steam generator levels. The cause of the main feedwater pump trip is under investigation. St. Lucie Unit 2 was not affected and remains at 100% power. This event is being reported pursuant to 10 CFR 50.72 (b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72 (b)(3)(iv)(A) for the auxiliary feed actuation. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat is being removed by using the atmospheric dump valves. | Steam Generator Feedwater | |
| ENS 55078 | 20 January 2021 23:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Trip of Motor Control Center | On 1/20/2021 at 1822 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a loss of Motor Control Center 2B2. The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in Mode 3. Auxiliary feed-water automatically actuated on the 2A Steam Generator post trip. Current decay heat removal is the 2B main feedwater pump to both steam generators and the Steam Bypass Control System to the main condenser. Unit 1 is not affected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. | Steam Generator Feedwater Steam Bypass Control System Decay Heat Removal Main Condenser | |
| ENS 53036 | 26 October 2017 06:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Following a Loss of Load | On October 26, 2017 at 0212 EDT St. Lucie Unit 2 experienced a reactor trip due to a loss of load event resulting in an RPS (Reactor Protection System) actuation. The cause of the loss of load is currently under investigation. Following the reactor trip, an Auxiliary Feedwater Actuation Signal occurred due to low level in the 2A Steam Generator. One of the two Main Feed Isolation Valves to the 2A Steam Generator did not close on the Auxiliary Feedwater Actuation Signal. 2A Steam Generator level was restored by Auxiliary Feedwater. The 2B Steam Generator level is being maintained by Main Feedwater. All CEAs (Control Element Assemblies) fully inserted into the core. Decay heat removal is being accomplished through forced circulation with stable conditions from Auxiliary Feedwater/Main Feedwater and Steam Bypass Control System. Currently maintaining pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Steam Bypass Control System Decay Heat Removal | |
| ENS 52757 | 15 May 2017 22:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Emergency Diesel Generator Signal Generated Upon Loss of 4160V Power | On May 15, 2017 at 1800 hours EDT, the '2A3' 4.16 KV safety related bus unexpectedly de-energized. The '2A' emergency diesel generator (EDG) system received a valid start signal from the undervoltage condition on the '2A3' bus but did not start as the EDG had been removed from service for maintenance. Loss of the '2A3' 4.16 KV bus resulted in a valid actuation of the undervoltage protection relays. The direct cause of the de-energization was determined to be failed secondary side potential transformer fuses. The 'B' train safety related electrical busses were unaffected by the event. The '2A3' 4.16 KV bus was reenergized at 2340. This event was determined to be reportable pursuant to 10CFR50.72(b)(3)(iv)(A). During the electrical transient, the licensee briefly entered Technical Specification 3.0.3 but plant conditions were restored, all required LCOs were satisfied, and Technical Specification 3.0.3 was exited before the plant was required to downpower. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 52191 | 21 August 2016 23:37:00 | 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 | Unusual Event - Loss of Offsite Power | At 35 percent power, a main generator lockout caused the main generator to trip, resulting in a reactor trip of Unit 1. Because of the lockout, power did not transfer to the startup transformers. Both emergency diesel generators started and aligned to the emergency busses. During the trip all control rods fully inserted and no safety or relief valves lifted. The plant is in Mode 3 steaming through the atmospheric relief valves and feeding the steam generators using auxiliary feedwater. There is no reported primary to secondary leakage. Primary coolant is being moved using natural circulation cooling. The trip of Unit 1 had no effect on Unit 2. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC, and Nuclear SSA (via e-mail).
The Unusual Event was terminated at 2125 EDT after the plant restored normal offsite power. The licensee notified the NRC Resident Inspector. Notified the R2DO (Sandal), IRD (Gott), NRR EO (Miller), DHS SWO, FEMA, DHS NICC, and Nuclear SSA (via e-mail).
On August 21, 2016 at 1926 EDT, St. Lucie Unit 1 experienced a reactor trip and a loss of offsite power due to a main generator inadvertent Energization Lockout Relay actuation. The cause of the lockout is currently under investigation. Coincident with the loss of offsite power, the four reactor coolant pumps deenergized. Both the 1A and 1B Emergency Diesel Generators started on demand and powered the safety related AC buses. All CEAs (Control Element Assemblies) fully inserted into the core. Offsite power to the switchyard remained available during the event, and at 2036, restoration of offsite power to St. Lucie Unit 1 was completed. Decay heat removal is being accomplished through natural circulation with stable conditions from Auxiliary Feedwater and Atmospheric Dump Valves. Currently maintaining pressurizer pressure at 1850 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 2 was unaffected and remains in Mode 1 at 100% power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation. The licensee notified the NRC Resident Inspector. Notified the R2DO (Sandal).
On August 21, 2016 at 2330 EDT, St. Lucie Unit 1 started two Reactor Coolant Pumps to establish Forced Circulation in order to enhance Decay Heat removal. Plant conditions remain stable with Auxiliary Feedwater and Atmospheric Dump Valves in service. This report is submitted in accordance with 10 CFR 50.72(c)(2)(ii) as a follow up notification of protective measures taken. The licensee notified the NRC Resident Inspector. Notified the R2DO (Sandal). | Steam Generator Reactor Coolant System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 51423 | 17 September 2015 16:22:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of Uv Relays Following Loss of Unit 2 Startup Transformer | On September 17, 2015, at 1222 hours, with Unit 2 in Mode 5 at the beginning of a refueling outage, an electrical fault on the 2A 6.9 kV bus resulted in the loss of the 2A startup transformer and its associated non-safety related 2A2 and safety related 2A3 busses. At the time of the event, the 2A Emergency Diesel Generator (EDG) had been properly removed from service for scheduled maintenance. The loss of the 2A start up transformer initiated the under voltage relays, which resulted in a valid actuation signal that would have started the 2A EDG. Additionally, the 2A train of shutdown cooling (SDC) was de-energized; the 2B (protected) train of SDC was not affected by the event and remained in service to remove decay heat. The 2A shutdown cooling train was restored and made available on September 19, 2015 at 0030. The 2B EDG and 2B startup transformer remained operable. St. Lucie did not report this event within 8 hours of occurrence, however, this event was subsequently determined to be reportable pursuant to 10CFR50.72(b )(3)(iv)(A). During this event, Unit 1 experienced a loss of the 1A startup transformer. There was no effect on Unit 1 operation, as its associated non-safety and safety-related busses remained powered by the auxiliary transformer. The 1A startup transformer returned to service on September 18, 2015 at 2103. The licensee informed the NRC Resident Inspector. | Emergency Diesel Generator Shutdown Cooling | |
| ENS 51302 | 10 August 2015 02:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Testing | On August 9, 2015, during the performance of Reactor Protection System Logic Matrix Testing, a reactor trip occurred. All CEA's (control rods) fully inserted into the core. Decay Heat removal is from Main Feedwater and Steam Bypass to the Main Condenser. All equipment operated as expected. Currently maintaining pressurizer pressure at 2250 psia, temperature maintaining at 532 degrees F. Unit 2 was unaffected and remains in Mode 1 at 100% power. This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the Reactor Trip and 10CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation. The plant is in its normal shutdown electrical lineup. No safety or relief valves lifted during this event. The cause of the trip is under investigation. The licensee notified the NRC Resident Inspector. | Feedwater Reactor Protection System Decay Heat Removal Main Condenser | |
| ENS 50607 | 12 November 2014 20:48:00 | 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 Water Level | On November 12, 2014 at 1548 (EST), Unit 2 was manually tripped due to a lowering 2B steam generator level caused by the spurious (slow) closure of 2B Main Feedwater Isolation Valve, HCV-09-2B. All CEAs (control element assemblies) fully inserted into the core. All safety systems responded as expected with the 2B Auxiliary Feedwater Actuation System Channel 2 (AFAS 2) actuating on low 2B steam generator level. Decay heat removal is from main feedwater to the 2A steam generator and manual control of auxiliary feedwater to the 2B steam generator, with steam bypass to the main condenser. This event is reportable pursuant to 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the AFAS 2 actuation. During the transient, no relief or safety valves lifted. The grid is stable and the plant is in its normal shutdown electrical lineup at normal operating pressure and temperature. The cause of the feedwater isolation valve malfunction is under investigation. There was no effect on Unit 1. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 49536 | 14 November 2013 17:18:00 | 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 Low Level in the 2B Steam Generator | On November 14, 2013 at 1218 EST, Unit 2 was manually tripped due to a lowering 2B Steam Generator level caused by the spurious closure of 2B Main Feedwater Isolation Valve HCV-09-2A. All CEAs (Control Element Assemblies) fully inserted into the core. All safety systems responded as expected with the 2B Train Auxiliary Feedwater Actuation System Channel 2 (AFAS 2) actuating on low 2B Steam Generator level. Decay Heat Removal is from Main Feedwater to the 2A Steam Generator and Auxiliary Feedwater to the 2B Steam Generator with Steam Bypass to the Main Condenser. This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the Reactor Trip and 10CFR 50.72(b)(3)(iv)(A) for the AFAS 2 actuation. The plant is in its normal shutdown electrical lineup. No safeties or relief valves lifted during this event. The NRC Resident Inspector has been notified by the licensee. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 48388 | 8 October 2012 01:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Essential Bus Deenergized While Defueled | On October 7, 2012, with Unit 2 in a defueled condition, a differential current lockout occurred on the 2B3 4.16kV essential bus, causing a deenergization of the 2B3 4.16kV essential bus. At the time of the event, the 2B Emergency Diesel Generator (EDG) was loaded to the essential bus. Due to the differential current lockout, all bus loads were lost and the 2B EDG output breaker feeding the essential bus opened and the 2B EDG transferred to emergency mode. The 2A EDG is operable and in standby. All equipment responded as expected. The plant is currently being maintained in a defueled condition. Decay heat removal is being supplied by the 2A Fuel Pool Cooling train. The cause of the differential current lockout of the 2B3 4.16kv bus is under investigation. This event is reportable pursuant to 10CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Decay Heat Removal | |
| ENS 48369 | 3 October 2012 12:43:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Failure of Startup Transformer Caused Undervoltage Condition on Essential Bus | On October 3, 2012, with Unit 2 in a defueled condition, a failure occurred on the 2B Startup Transformer, causing an undervoltage condition on an essential bus and resulted in the automatic start and loading of the 2B Emergency Diesel Generator (EDG). Prior to the event, the 2B EDG was available and not required by Technical Specifications; however, the 2B EDG was inoperable. Additionally, the 2A EDG is available. All equipment responded as expected. Currently maintaining the plant in a defueled condition. Decay heat removal is being supplied by the 2A Fuel Pool Cooling train and was never interrupted. There was no impact on the Shutdown Safety Assessment. This event is reportable pursuant to 10CFR50.72(b)(3)(iv)(A). Due to common high side feed, the loss of the 2B Startup Transformer resulted in the loss of the 1B Startup Transformer. Prior to the event, the 1A EDG was out of service for maintenance. As a result, Unit 1 entered Technical Specification 3.8.1.1 Action C. due to the loss of one offsite AC circuit and one diesel generator inoperable. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator Decay Heat Removal | |
| ENS 47915 | 11 May 2012 07:55:00 | 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 Feedwater Control Valve Failure | On May 11, 2012, a failure of the High Power Feed Regulating Valve FCV-9011 resulted in '2A' S/G water level lowering. Manual operator control of the Main Feed Regulating system was unsuccessful in stabilizing S/G water level. '2A' S/G level lowered to the procedurally required manual reactor trip criteria. The crew inserted a manual trip. All CEAs fully inserted into the core. Following the trip, Auxiliary Feedwater actuated as designed and decay heat removal was via Auxiliary Feedwater and Steam Bypass to the Main Condenser. All equipment operated as expected. Currently, Unit 2 is maintaining pressurizer pressure at 2250 psia, temperature at 532 degrees F on Main Feedwater (using Low Power Feed Regulating Valves LCV-9005/9006) and Steam Bypass Control. 'Unit 1 was unaffected and remains in Mode 1 at 29% power. This event is reportable pursuant to 10CFR50.72(b)(2)(iv)(B) for the Reactor Trip, as well as 10CFR50.72(b)(3)(iv)(A) for specified system actuation (Auxiliary Feedwater). The licensee has notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 47657 | 10 February 2012 06:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Maintenance Activities Cause an Inadvertent Emergency Diesel Generator Start | On February 10, 2012, with Unit 1 in Mode 5, while performing scheduled maintenance, a technician inadvertently made contact with a component that caused an undervoltage condition on an essential bus, resulting in the automatic start and loading of the 1B Emergency Diesel Generator (EDG). Prior to the event the 1B EDG was inoperable and not required by Technical Specifications; however, the 1B EDG was available. All equipment responded as expected. Currently maintaining the plant in Mode 5. Decay heat removal is being supplied by the 1A Shutdown Cooling train and was never interrupted. There was no impact on the Shutdown Safety Assessment. Unit 2 was unaffected and remains in Mode 1 at 100% power. This event is reportable pursuant to 10CFR 50.72(b)(3)(iv(A). The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Shutdown Cooling Decay Heat Removal | |
| ENS 46928 | 6 June 2011 07:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Error During Reactor Protection System Surveillance Leads to Reactor Trip | On June 6, 2011, during the performance of Reactor Protection System Logic Matrix Testing, a Reactor Trip occurred. All CEA's (Control Element Assemblies) fully inserted into the core. Decay Heat removal was initially from Auxiliary Feedwater and Steam Bypass to the Main Condenser. All equipment operated as expected. Currently maintaining pressurizer pressure at 2250 psia, temperature maintaining at 532 degrees F. Auxiliary Feedwater actuated as designed. As of 0435, decay heat removal is via Main Feedwater and Steam Bypass control to Main Condenser. During the transfer of Auxiliary Feedwater to Main Feedwater a second AFAS actuation occurred. Unit 1 was unaffected and remains in Mode 1 at 100% power. This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the Reactor Trip, as well as 10CFR50.72(b)(3)(iv)(A) for specified system actuation (Auxiliary Feedwater). The licensee notified the NRC Resident Inspector. | Feedwater Reactor Protection System Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 45843 | 15 April 2010 19:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Manual Reactor Trip | At 1539 (EDT), Unit 2 was manually tripped due to lifting of the 2B moisture separator reheater relief valve. The Unit commenced a rapid downpower and then a manual reactor trip was initiated at approximately 95% power. All CEA's (control element assemblies) fully inserted on the trip. Auxiliary feedwater automatically initiated on low steam generator level due the 2A steam generator 15% feedwater bypass not opening. No pressurizer power operated relief valves (PORVs) opened. RCS heat removal is now being maintained with auxiliary feedwater and the steam bypass control system. Main feedwater is available. All other systems functioned normally, and the plant is stabilized at normal operating temperature and pressure in Mode 3. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to manual RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) due to auxiliary feedwater system actuation. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Steam Bypass Control System | |
| ENS 45374 | 21 September 2009 17:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Manual Reactor Trip During Reactor Startup | At 1227 EDT, a reactor startup was commenced on Unit 2. Mode 2 was entered at 1325 EDT. At 1333, a Reactor Control Operator noted that Primary Safety Valve V1202 had indications that it was leaking past its seat. Plant procedures required reducing RCS (Reactor Coolant System) pressure in 100 psi increments until the safety reseated. This event required the plant pressure to be reduced to 200 psi below Normal Operating Pressure. Prior to commencing the depressurization, a manual reactor trip was ordered by the Unit Supervisor as discussed in the pre-evolution brief. The unit was in Mode 2 approaching criticality at the time of the trip. The unit is currently stable in Mode 3, Hot Standby. The reactor trip was uncomplicated. All equipment operated as expected. Main feedwater remained available during the entire event. Auxiliary Feedwater and Atmospheric Dump Valves remained in service during the entire event. Unit 1 was unaffected by the event and remained at full power. The grid remained stable throughout the event. All control rods fully inserted. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 44952 | 1 April 2009 22:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Manual Reactor Trip | At 1805, due to lowering Condenser Vacuum caused by ingress of algae and seaweed, Unit 2 was manually tripped. Power had been reduced to 94% for the securing of one Circulating Water Pump (2A1). It was then identified that 2A2 Circulating Water Debris filter differential pressure was above administrative limits of 200 inches water. While the station was making preparations to reduce Circulating water flow on the 2A2 Circulating Water Pump, the unit began losing condenser vacuum. Plant was manually tripped at 92% power. All CEA's fully inserted on the trip. Auxiliary Feedwater automatically initiated on Low Steam Generator Level. No PZR PORVS opened. RCS Heat removal is now being maintained with Main Feedwater and Steam Bypass control system. All systems functioned normally, and plant is stabilized at normal operating Temperature and Pressure. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to manual RPS actuation and 10 CFR 50.72(b)(3)(iv)(B) due to PWR auxiliary feedwater system actuation. There was no impact on Unit 1. The licensee informed the Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Steam Bypass Control System | |
| ENS 44276 | 7 June 2008 12:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following the Trip of a Condensate Pump | On 6/7/08 at 0818 hours, an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 100% power due to a trip of the 2B Condensate Pump, which led to a trip of the 2B Main Feedwater Pump (MFP) and decreasing Steam Generator (S/G) levels. The reactor was manually tripped due to decreasing S/G levels. Following the reactor trip, EOP-1, Standard Post Trip Actions and EOP-2, Reactor Trip Recovery procedures were completed and Unit 2 was stabilized in Mode 3. All control rods fully inserted. The Main Steam Safety Valves lifted as expected. Feedwater to the S/Gs was initially supplied by the 2A (MFP) until Auxiliary Feedwater Actuation System (AFAS) actuated as expected on low S/G level. Subsequently, the Auxiliary Feedwater Pumps restored S/G levels. Unit 2 electrical requirements were provided from offsite power. Other than the trip of the 2B Condensate Pump (initiating event) there were no major equipment failures. Unit 1 was not affected by this event. The grid is stable. Decay heat is being removed by the Auxiliary Feedwater Pumps feeding the S/Gs steaming to the bypass valves in the Condenser. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Steam Safety Valve Control Rod | |
| ENS 43874 | 29 December 2007 06:31:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip When Five Control Rods Unexpectedly Dropped 20 Inches | At 2320, on 12/28/07, a Reactor Startup was commenced. At 0025, on 12/29/07 Subgroup #15, of Regulating Group #3, was placed on the hold bus. Placing the Subgroup on the hold bus was a pre-planned action that was briefed prior to the reactor startup, in accordance with an approved interim engineering disposition. The interim engineering disposition was written and approved on 12/28/07 for concerns over CEA #1, of Subgroup #15, dropping into the core unexpectedly. Subgroup #15, of Regulating Group #3, contains five CEA's (CEA # 60, 62, 64, 66 and 1). At 0047, all Regulating Group CEA's, with the exception of Regulating Group #5, were placed at the Upper Electrical Limit (136 inches withdrawn). Regulating Group #5 was at 120 inches withdrawn in preparation for diluting to criticality. At 0107, the dilution to criticality was commenced. At 0131, all 5 CEA's in Subgroup #15 slipped into the core approximately 20 inches. A manual reactor trip was then ordered by the unit supervisor. 2-EOP-1, 'Standard Post Trip Actions' was then performed. The unit was borated to shutdown boron concentration. All Safety Functions were met satisfactorily and 2-EOP-1 was exited. The unit was in Mode 3 approaching Mode 2 at the time of the trip. The unit is currently stable in Mode 3, Hot Standby. Reactor coolant pump heat is being removed using the atmospheric steam dumps. The licensee notified the NRC Resident Inspector. | Control Rod | |
| ENS 41911 | 11 August 2005 14:48:00 | 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 a Condensate Pump Bus Lockout | On 8/11/2005, a manual reactor trip was initiated due to lowering steam generator level caused by a partial loss of main Feedwater. The partial loss of feedwater was caused by the differential lock out of the non-vital 2A2 4160 V bus which resulted in loss of the 2A Condensate Pump that tripped the 2A Main Feedwater Pump. All rods inserted and no Steam Generator Safety Valves lifted. The differential lock out of the non-vital 2A2 4160 V (bus) deenergized the 2A3 vital 4160V bus, starting the 2A Emergency Diesel Generator and the 2A3 loads were sequenced on the Emergency Diesel Generator per design. Subsequently, the Auxiliary Feedwater System was automatically initiated due to lowering steam generator levels. All safe shutdown equipment operated as expected. The plant is stable in Mode 3, Hot Standby conditions, with decay heat removal being accomplished by steaming to the Main Condenser and Feedwater to the steam generators supplied by the Main Feedwater system. The Offsite power grid is available and stable. The '2C' Auxiliary Feedwater Pump was out of service for routine surveillance and it had no effect on the cause of the trip nor had any effect on the trip recovery. St. Lucie is investigating the cause of the lockout on the 2A2 4160V bus. Unit 1 was not affected by this event. At the time of this report, the 2A Emergency Diesel Generator was still loaded while investigations were underway. Steam generator level is being maintained using main feed. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal Main Condenser | 05000389/LER-2005-003 |
| ENS 41715 | 19 May 2005 00:59:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent De-Energization of 4160 V Safety Related A.C. Bus with Edg Auto Start | On 5/18/05 at 20:59 the 1A3 4160 V safety related A.C. bus inadvertently de-energized and the 1A Emergency Diesel Generator (EDG) automatically started and loaded onto the bus. The inadvertent de-energization of the 1A3, 4160 V bus appears to have resulted from testing of the 4160 V under voltage relays. Currently, normal power has been restored to the 1A3, 4160 V bus and the 1A Emergency Diesel Generator has been secured. This notification is being made pursuant to 10 CFR 50.72(b)(3)(iv)(A) to be completed within 8 hours as a safety systems actuation of the 1A3, 4160 V under voltage relaying and inadvertent start and load of the 1A Emergency Diesel Generator. The licensee notified the NRC Resident Inspector.
On 5/18/05 at 20:59 the 1A3 4,160 KV safety related AC bus inadvertently de-energized and the 1A Emergency Diesel Generator (EDG) automatically started and loaded onto the bus. This event was initiated during the performance of a plant surveillance 1-OSP-100.07, to test the 1A3 4,160 KV Bus Under Voltage Relay. The 1A EDG loaded and carried the 1A3 bus. The 1B3 4,160 KV bus was unaffected and the "B" side power remained energized. This update is to provide the following additional information identified during the follow up investigation. This update is to identify that HVS-1B, Containment Fan Cooler, did not start as expected after the 1A EDG automatically loaded on the 1A3 4,160 KV Bus. The HVS-1A and HVS-1B were both load shed from the bus prior to closure of the 1A EDG output breaker. The HVS-1A did start as expected following closure of the EDG output breaker. The HVS-1B is on the three (3) second load block for the 1A EDG to restart, but did not start. A Root Cause Team has been formed to identify the cause of the initiating event and the auto-start failure of HVS-1B. A Condition Report was generated and a troubleshooting plan has been developed to determine the cause of the initiating event and failure of the HVS-1B to automatically restart. The 1B3 4,160 KV safety related AC bus and associated EDG were not affected by this event and remained operable during and following the event. Troubleshooting for the subject failed equipment is ongoing. The licensee notified the NRC Resident Inspector. The R2DO (Ogle) was notified. | Emergency Diesel Generator | |
| ENS 41293 | 25 December 2004 11:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Manually Tripped to Remove Failing Condensate Pump from Service | Manual reactor trip due to condensate pump failure. All systems functioned as required post trip. The plant is currently stable in Mode 3. On 12/25/04 Unit 2 at St. Lucie experienced a high amperage reading on a condensate pump. Visual observation of the pump indicated blistering paint at the electrical connections. The condensate pump was taken out of service and Unit 2 was manually shut down. Decay heat is being removed via normal means to the condenser. The AFW system started as expected. All rods inserted correctly and all systems functioned as designed. The licensee notified the NRC Resident Inspector.
The licensee called to provide some additional information concerning this event: 1) The manual trip of Unit 2 was due to the failure of the 2B condensate pump. 2) The condensate pump failure was a result of a failed termination of the "A" phase motor lead to the field cable. 3) The failure of the motor lead is considered a random event and does not have an generic implications. The licensee also noted that the reactor power had actually been reduced to 95% immediately prior to the manual trip. The NRC Resident Inspector and R2DO (Julian) have been notified.
The following information was obtained from the licensee via facsimile: This notification is an amended notification to the original notification of unit trip and RPS actuation due to failure of 2B condensate pump. The amended notification includes an 8-hour notification non-emergency 10 CFR 50.72 (b)(3)(iv)(A) to identify Aux Feedwater Actuation automatically actuated post manual reactor trip on 12/25/04. Specified System Actuation per 50.72 (b)(3)(iv)(B)(6) AFAS (Auxiliary Feedwater Actuation System). The licensee notified the NRC Resident inspector. R2DO (Julian) notified of this update. | Feedwater | |
| ENS 41072 | 26 September 2004 03:56:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to a Loss of Offsite Power | At 2356 (EDT), September 25, 2004, off-site power was lost for both units. At the time of the event both units were in mode 4 and cooling down to meet shutdown entry conditions. The shutdown of both units was due to hurricane conditions from hurricane Jeanne. All four emergency diesels (2 per unit) started and properly loaded. Both units are stable in natural circulation cooling. Efforts continue to place both units on shutdown cooling. All systems operated as expected. The plant was already in a Notification of Unusual Event due to the Hurricane. Shutdown Cooling was established on Unit 1 at approximately 0020 EDT on 9/26/04. The licensee informed both state and local agencies and the NRC Resident Inspector. A conference call was held at approximately 0005 EDT with R2 Response Manager (Len Wert), NRR EO (Stu Richards) and IRD (Susan Frant) to discuss the loss of offsite power. The participants concluded that the NRC Monitoring Mode entered at 1515 EDT on 9/25 for the Hurricane was appropriate.
At 2356, September 25, 2004, St. Lucie Station, Units 1 and 2, experienced a loss of off -site power (LOOP). At the time of the LOOP all four emergency diesel generators started and loaded the safety related buses. In it's initial notification of the LOOP (0035 09/26/04, EN# 41067), FPL reported that all systems had performed as expected. However, during it's post LOOP walkdowns of the control room boards the control room operators determined that the 1 B Intake Cooling Water (ICW) pump had not automatically started as expected. The pump was subsequently started from the control room using the manual control switch. The cause for the failure to automatically start will be investigated and corrected prior to returning the unit to service. Notified R2 Response Manager (Len Wert)
At 1050 EDT both units recovered offsite power exiting the criteria for the UE based on LOOP. The licensee informed state and local agencies and the NRC Resident Inspector. HOO Note: See related ENs # 41067, 41071 and 41073. | Emergency Diesel Generator Shutdown Cooling | |
| ENS 40403 | 20 December 2003 14:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation Due to Loss of Turbine Generator Excitation | On December 20, 2003, at 0949 hours, an automatic reactor trip occurred due to a loss of excitation of the turbine generator. All plant safety functions were maintained throughout the event. The plant was stabilized in Mode 3. All plant safety systems responded normally with the exception of the 2C Auxiliary Feedwater Pump (steam driven) which tripped on mechanical overspeed. The 2A and 2B Auxiliary Feedwater Pumps (electric driven) functioned normally to restore the 2A and 2B Steam Generator levels. Post trip system anomalies include RCS Letdown isolated, Steam Generator Blowdown isolation valves closed, Control Room ventilation system swapped to recirculation mode, and the Fuel Handling Building ventilation system swapped to the Shield Building. An Emergency Response Team has been formed to review these conditions prior to plant restart. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to the automatic RPS Reactor Trip. All controls inserted properly. Decay heat is being removed using the turbine bypass valves. The Licensee notified the NRC Resident Inspector.
This update is provided to include the 10CFR50.72(b)(3)(iv)(A) notification criterion for the auxiliary feedwater actuation. The NRC Resident Inspector was notified., Reg 2 RDO(Fredrickson) was informed. | Steam Generator Auxiliary Feedwater Shield Building |