Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57177 | 18 June 2024 07:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | The following information was provided by the licensee by email: On 06/18/2024 at 0317 EDT with Unit 2 at 18 percent power, the reactor was manually tripped due to elevated secondary chemistry levels (sodium and chlorides). The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in Mode 3. Decay heat is being removed by auxiliary feedwater and atmospheric dump valves. St. Lucie Unit 1 was unaffected and remains at 100 percent power. This event is being reported pursuant 10CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. An investigation is underway to determine the root cause of the elevated chemistry levels.
The following retraction was provided by the licensee via email: The purpose of this notification is to retract a previous report made on 06/18/2024 at 0652 (EDT) (EN# 57177). Notification of the event to the NRC was initially made because of inserting a manual reactor trip due to elevated secondary chemistry levels (sodium and chlorides). After the initial report, Florida Power and Light has concluded that the event did not meet the reporting requirements on 10 CFR 50.72(b)(2)(iv)(B) since it was part of a normal plant shutdown. Therefore, this event is not considered an unplanned scram and is not reportable to the NRC as a Licensee Event Report per 10 CFR 50.73. The NRC Senior Resident Inspector has been notified. | Auxiliary Feedwater | |
| ENS 57161 | 4 June 2024 17:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | The following information was provided by the licensee via email: At 1352 EDT, on June 4, 2024 with Unit 2 at 92 percent power, the reactor was manually tripped due to a loss of condenser vacuum resulting from a circulating water pump trip. The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat is being removed by steam discharge to the main condenser using the turbine bypass valves. St. Lucie Unit 1 was not affected and remains at 100 percent power. This event is being reported pursuant to 10 CFR 50.72 (b)(2)(iv)(B). The NRC Resident Inspector has been notified. | Main Condenser | |
| 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 55641 | 10 December 2021 15:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | On 12/10/2021, at 1024 EST, with Unit 1 at 100 percent power, the reactor was manually tripped due to lowering level in the steam generators. All systems responded as expected following the trip. The reactor is currently stable in Mode 3 and operators restored steam generator level utilizing main feedwater. The cause of the reduction in feedwater flow is under investigation. St. Lucie Unit 2 was not affected and remains at 100 percent power. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip. The NRC Resident Inspector has been notified. All rods inserted into the core during the trip. The plant is in its normal shutdown electrical lineup. Decay heat is being maintained by steam discharge to the main condenser using the turbine bypass valves. | Steam Generator Feedwater Main Condenser | |
| 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 54262 | 7 September 2019 12:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On September 09, 2019 at 0824 EDT, with St. Lucie Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped on Low Reactor Coolant System Flow due to a trip of the 1A1 reactor coolant pump. The trip was uncomplicated with all systems responding normally post-trip. Operators responded and stabilized the plant in Mode 3. The cause of the loss of the 1A1 reactor coolant pump is currently under investigation. St. Lucie Unit 2 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. Decay heat removal is being accomplished by main feed water and the main condenser using the turbine steam bypass valves. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Decay Heat Removal Main Condenser | |
| ENS 54027 | 25 April 2019 13:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Turbine Generator Trip | At 0918 (EDT) on 4/25/19, with (Saint Lucie) Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a Turbine Trip. The reactor trip was uncomplicated with all systems responding normally. Operations is maintaining the plant stable in Mode 3. Decay heat removal is being accomplished by main feed water and the main condenser using the turbine steam bypass valves. Unit 2 is not affected and remains at 100% power. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. | Decay Heat Removal Main Condenser | |
| ENS 53703 | 29 October 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Inadequate Feedwater Flow | On October 29, 2018 at 1317 EDT, with St. Lucie Unit 1 in Mode 1 at 100% power, the reactor was manually tripped due to inadequate feedwater flow to both 1A and 1B Steam Generators (S/Gs). The trip was uncomplicated with all systems responding normally post-trip. (All control rods fully inserted and there were no specified system actuations.) Operators responded and stabilized the plant in Mode 3. The cause of the inadequate feed flow to the 1A and 1B Steam Generators is currently under investigation. Decay Heat removal is being accomplished through forced circulation with stable conditions from Main Feedwater and the Steam Bypass Control System to the Main Condenser. Currently maintaining Pressurizer pressure at 2250 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 10CFR50.72(b)(2)(iv)(B) for the reactor trip. The NRC Senior Resident Inspector has been notified. | Steam Generator Reactor Coolant System Feedwater Steam Bypass Control System Decay Heat Removal Main Condenser Control Rod | |
| ENS 53665 | 12 October 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On October 12, 2018 at 1353 EDT, St. Lucie Unit 2 experienced an automatic RPS actuation and Reactor Trip due to a fault on the 2A1 6.9kv bus during a transfer of the bus power supply from the 2A Auxiliary Transformer to the 2A Startup Transformer. The bus fault caused a fire in the 2A1 6.9kv switchgear that has been extinguished. Offsite support was not required to extinguish the fire. The specific cause of the fault is currently under investigation. Following the reactor trip, both Steam Generators are being supplied by main feedwater. All (Control Element Assemblies) (CEAs) fully inserted into the core. Decay Heat removal is being accomplished through forced circulation. Main Feedwater and Steam Bypass Control Systems are maintaining stable conditions in Mode 3. 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. The fire was extinguished within 28 minutes. Plant loads are being supplied by the 2B Auxiliary Transformer. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Steam Bypass Control System Decay Heat Removal | |
| 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 49528 | 12 November 2013 05:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Unisolable Leak in Digital Electro-Hydraulic System | At 0002 EST, Unit 1 Manually tripped the Reactor from 90% power due to an unisolable leak in the Digital Electro-Hydraulic (DEH) System. All CEAs fully inserted into the Reactor Core. All systems responded as expected on the trip. Decay Heat removal currently using Main Feedwater and Steam Bypass Control System. After the trip, DEH pumps were secured to stop the transfer of fluid from the DEH system to the Turbine Building. Investigation ongoing to determine exact location of the leak. This condition is reportable pursuant to 10CFR50.72(b)(2)(iv)(B). The was no impact on Unit 2. The NRC Resident Inspector has been notified. | Feedwater Steam Bypass Control System Decay Heat Removal | |
| ENS 49082 | 31 May 2013 11:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to an Anticipated Loss of Condenser Cooling | On May 31, 2013 at 0712 (EDT), Unit 2 (reactor) was manually tripped due to high differential pressure on the debris filter for the 2A1 Condenser Waterbox which required a trip of the 2A1 Circulating Water Pump. The 2A2 Condenser Waterbox and the 2A2 Circulating Water Pump were already removed from service due to a suspected condenser tube leak. All CEAs (Control Element Assembly) fully inserted into the core. Decay heat removal is from main feedwater and steam bypass to the main condenser. The cause of the rising differential pressure on the 2A1 debris filter was potentially due to an influx of algae. This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the reactor trip. The reactor trip response is considered uncomplicated and the unit is stable in Mode 3 at normal temperature and pressure. Unit 2 is in a normal shutdown electrical lineup. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. | Feedwater Decay Heat Removal Main Condenser | |
| ENS 48818 | 12 March 2013 18:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip on Thermal Margin/Low Pressure | On 3/12/13 at 1451 EDT, during normal full power operations, Unit 1 automatically tripped due to the Thermal Margin/ Low Pressure trip setpoint being exceeded. The trip was uncomplicated and all CEAs (control element assembly) fully inserted when the reactor was tripped. Main Steam Safety valves lifted momentarily post trip and reseated. No automatic safety system actuations were required and none occurred. The cause and details of the automatic trip are under investigation. The plant is stable in Mode 3 at normal operating temperature and pressure. RCS Heat Removal is being maintained with Main Feedwater and Atmospheric Dump Valves in operation. The operation of the Steam Bypass Control System is under review by Engineering. The Offsite power grid is available and stable. The licensee has notified the NRC Resident Inspector and there was no impact on Unit 2. | Feedwater Steam Bypass Control System Main Steam Safety Valve | |
| 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 47987 | 2 June 2012 23:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to a Turbine Control System Failure | At 1935 during normal full power operations, Unit 1 automatically tripped due to a loss of load caused by an instantaneous failure of the turbine control system. The trip was uncomplicated and all CEAs (control rods) fully inserted when the reactor was tripped. No automatic safety system actuations were required and none occurred. The cause and details of the turbine control system failure are under investigation. The plant is stable in Mode 3 at normal operating temperature and pressure. RCS Heat Removal is being maintained with Main Feedwater and Steam Bypass Control Systems with condenser vacuum. The offsite power grid is available and stable. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to RPS actuation with the reactor critical. The licensee notified the NRC Resident Inspector. | Feedwater Steam Bypass Control System | |
| 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 47793 | 31 March 2012 04:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Uncontrolled Cooldown | At 0022 (EDT) on 03/31/12 while maintaining power stable at 10% for Steam Bypass Control System testing, Unit 1 was manually tripped due to an uncontrolled cooldown caused by PCV-8802 (Steam Bypass Control Valve) unexpectedly opening. Following the trip, PCV-8802 closed and the secondary was isolated by closing the Main Steam Isolation Valves per Standard Post Trip Actions. Following isolation of the steam demand, the trip was uncomplicated with all CEAs fully inserted. No automatic safety system actuations were required and none occurred. The cause of the unexpected opening of the Steam Bypass Control System valve is under investigation. The plant is stable in Mode 3 at normal operating temperature and pressure. RCS Heat Removal is being maintained with Auxiliary Feedwater and Atmospheric Dump Valves. The Offsite power grid is available and stable. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to manual RPS actuation with the reactor at power. The RCS cooled down from 532 degree to 515 degrees over a period of approximately 2 minutes and 40 seconds. The reactor was manually tripped when RCS temperature reached 515 degrees and the lowest RCS temperature observed after the trip was 505 degrees. The licensee has notified the NRC Resident Inspector. | Main Steam Isolation Valve Auxiliary Feedwater Steam Bypass Control System | |
| ENS 47752 | 19 March 2012 03:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Control Element Assembly Anomalous Behavior | At 2336 EDT, during performance at Low Power Physics Testing, Unit 1 was manually tripped while the reactor was critical at less than 1% power due to Control Element Assembly (CEA) Regulating Group #3 exhibiting anomalous behavior (continued to insert with no operator action). The trip was uncomplicated and all CEAs fully inserted when the reactor was tripped. No automatic safety system actuations were required and none occurred. The cause for the abnormal CEA performance is under investigation. The plant is stable in Mode 3 at normal operating temperature and pressure. RCS Heat Removal is being maintained with Auxiliary Feedwater and Atmospheric Dump Valves. The offsite power grid is available and stable. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) due to manual RPS actuation with the reactor critical The NRC Resident Inspector has been informed. | Auxiliary Feedwater | |
| 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 47353 | 19 October 2011 09:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip on Loss of Condenser Vacuum | On October 19, 2011, at 0528, Unit 1 was manually tripped due to rising condenser backpressure. All (Control Element Assembly) CEAs fully inserted into the core. Decay Heat Removal is from Main Feedwater and Steam Bypass to the Main Condenser. The cause of the rising backpressure was an unplanned trip the Circulating Water Pump 1A1, which degraded the Circulating Water System performance. At the time of the trip, an additional Circulating Water Pump 1A2 was secured for planned maintenance. The cause of the Circulating Water Pump 1A1 trip is under investigation. This event is reportable pursuant to 10CFR 50.72(b)(2)(iv)(B) for the Reactor Trip. The plant is stable at normal operating temperature and pressure. The licensee notified the NRC Resident Inspector. | Feedwater Circulating Water System Decay Heat Removal Main Condenser | |
| ENS 47178 | 22 August 2011 19:13:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Rising Condenser Backpressure | On August 22, 2011 at 1513 (hrs. EDT), Unit 1 was manually tripped due to rising condenser backpressure. All CEAs fully inserted into the core. Decay heat removal was initially from main feedwater and steam bypass to the main condenser. The cause of the rising back pressure was an influx of jellyfish into the intake structure, degrading the circulating water system performance. Subsequent to the manual trip, the 1B Main Feedwater Pump was manually secured due to a leak on the pump casing. The 1A Main Feedwater Pump subsequently tripped due to low suction pressure after manually securing the 1B Condensate Pump, per procedure. Decay heat removal was transitioned to atmospheric dump valves and auxiliary feedwater. Unit 2 is in Mode 1, currently at 70 % power. Unit 2 power is being reduced from 100% in response to the influx of jellyfish. This event is reportable pursuant to 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip. During the transient, no primary or secondary relief valves lifted. Offsite power is stable and the plant is in its normal shutdown electrical line-up with power being supplied from offsite. There is no known primary-to-secondary leakage. The cause of the 1A Main Feedwater Pump trip is under investigation. Unit 2 remained at 70% reactor power before and after the event. The licensee has notified the NRC Resident Inspector.
On August 22, 2011 an abnormal fish kill of at least 1000 lbs was observed in the combined unit's intake canal. The cause of the fish kill was related to an unusually large sustained influx of jellyfish into the intake canal. Per the plant's environmental permit, the Florida Fish and Wildlife Conservation Commission (FWCC) was notified at 1627 EDT. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) due to the notification of the FWCC. The licensee has notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Circulating Water System Decay Heat Removal Main Condenser | |
| 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 46018 | 16 June 2010 21:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip After Two Control Rods Dropped | At 1710 EDT, Unit 1 was manually tripped due to two dropped control rods. All CEAs (Control Element Assemblies) fully inserted on the trip. Steam generator level control responded as expected and no pressurizer or power operated relief valves opened. RCS heat removal is being maintained by main feedwater and steam bypass control systems. All other systems functioned normally and the plant has stabilized at normal operating temperature and pressure in Mode 3. This non-emergency notification is being made pursuant 10 CFR 50.72(b)(2)(iv)(B) due to manual actuation of RPS. The licensee characterized the manual trip as uncomplicated. The second rod dropped within a very short time of the first rod. The cause of the rod drops is still under investigation. The licensee noted that no activities involving the rod control system were in progress when the event occurred. The licensee was at 45% as part of its post outage power ascension unrelated to the rod drop. The manual reactor trip action was taken per procedure when the second rod dropped. The reactor trip had no impact on Unit 2 operation. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Steam Bypass Control System Control Rod | |
| 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 44268 | 4 June 2008 21:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Following Loss of Main Feedwater Pump | On 6/4/2008 at 17:30 hrs an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 100% power due to a trip of the 2B Heater Drain pump leading to a trip of the 2A Main Feedwater Pump and decreasing S/G levels. The Reactor was manually tripped due to decreasing Steam Generator 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 S/G Safety Valves lifted and the last valve reseated at 950 psig. Feedwater to the S/G was supplied by the 2B Main FW Pumps and then the Auxiliary Feedwater Pumps. Unit 2 electrical requirements are provided from offsite power. All safe shutdown equipment operated as expected. There were no major equipment failures. Unit 1 was not affected by this event. The Grid is stable. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 43941 | 29 January 2008 10:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Manual Reactor Trip to Repair Rcp "2B1" Seal Package | St Lucie Plant, Unit 2, manually tripped the reactor at 0531 hrs. EST, as part of a controlled reactor plant shutdown, in response to an RCS leak rate of 0.28 gpm, indicated to be associated with the 2B1 Reactor Coolant Pump seal package. All systems functioned as expected, satisfactorily. All CEAs fully inserted. Decay Heat removal is with Steam Generators and Main Feedwater. This report is made pursuant to 10 CFR 50.72(b)(2)(iv)(B). Unit 2 will cooldown to Mode 5 for repairs. The licensee informed the NRC Resident Inspector.
St. Lucie Plant, Unit 2, reported their manual reactor trip of 0531 hrs. EST, at 0630 hrs. EST, for an unidentified leak associated with the 2B1 RCP seal package (EN 43941). Subsequent investigation, reported to the STA (Shift Technical Advisor) at 1015 hrs. EST, determined that this leak constitutes an RCS Pressure Boundary Leak. This leak is at a pipe-to-flange weld on the outboard side of the first flanged coupling of the 2B1 RCP Upper Cavity Seal pressure sensing line. Investigation is in progress to determine the detailed configuration at the leak and root cause. There were NO ESFAS signals or actuations. Decay Heat Removal continues on Steam Generators, with MFW (Main Feedwater) and SBCS (Steam Bypass Control System); Off-Site power continues (to be) available and stable; Unit 1 at 100% power with operations and conditions normal. This report is made pursuant to 10 CFR 50.72(b)(3)(ii)(A). The licensee informed the NRC Resident Inspector of this update. Notified R2DO (Bonser). | Steam Generator Feedwater Decay Heat Removal | |
| 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 42647 | 16 June 2006 02:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Digital Electro-Hydraulic Leak | On 6/15/06 at 2223 hrs an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 45% power due to severe DEH Leak on the #1 Throttle Valve. DEH Leak ceased upon Turbine Trip. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed without contingencies and Unit 2 was stabilized in Mode 3. All control rods fully inserted and no S/G Safety Valves Lifted. Feedwater to the S/G was supplied by the main FW Pumps. All safe shutdown equipment operated as expected. There were no major equipment failures. Decay heat is being removed with main feedwater and dumping steam to the condenser. The grid is stable. The fire brigade was activated following the trip and set a fire watch (no fire). The NRC Resident Inspector was notified of this event by the licensee. | Feedwater Control Rod | |
| ENS 42277 | 20 January 2006 13:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Rps Actuation | On 1/20/2006 at 0730 hrs, a Unit 2 control room annunciator was received indicating high conductivity/sodium in the main condenser. The Chemistry Dept. confirmed the sodium level was high indicating saltwater intrusion from a tube leak in the 2B2 Condenser Waterbox. Per the Secondary Chemistry Off-Normal Procedure, a rapid downpower to less than 5% power was initiated to remove the affected waterbox from service. It was decided to go ahead and remove Unit 2 from service rather than remain critical at a low power level. The downpower was planned to decrease power to approximately 25%, perform a manual transfer of plant electrics to the auxiliary transformers, and then manually trip the reactor in accordance with plant procedures. All systems worked as planned during the downpower and the reactor was manually tripped at approximately 25% power at 08:56 hrs. Standard Post Trip Actions and the Reactor Trip Recovery Procedure were carried out without incident. All control rods fully inserted and no Steam Generator (S/G) Safety Valves lifted. Feedwater to the S/G was supplied by the Main Feedwater pumps during the shutdown and later transferred to the Auxiliary Feedwater pumps. 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 through the Atmospheric Dump Valves. The Main Feedwater pumps remain available if needed. Unit 1 was not affected by this event. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser 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 41297 | 27 December 2004 22:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Following Problem with Steam Generator Water Level Control | The licensee reported a "manual reactor trip due to low steam generator level caused by feedwater control system malfunction." The licensee stated that it manually tripped the reactor with steam generator water level at approximately 40% and decreasing. Steam generator water level control was restored following the trip using auxiliary feedwater. All rods fully inserted on the trip. No safety or relief valves lifted. Auxiliary feedwater was manually actuated and decay heat is currently being discharged via the atmospheric dump valves. Unit 1 is at full power and unaffected and the grid is stable. The plant was in no major LCOs at the time. All systems functioned as required. The licensee is still investigating the feedwater control system malfunction. The NRC Resident Inspector has been notified.
The original notification stated that decay heat was being discharged via the atmospheric dump valves post-trip when the decay heat removal mechanism being used was steam dump to the condenser via the steam bypass control system. Additionally, although the auxiliary feedwater system was used to deliver water to the steam generators post-trip, the main feedwater system was available for this function. The investigation into the feedwater malfunction is still in progress. The NRC Resident Inspector has been informed. R2DO (Moorman) notified. | Steam Generator Feedwater Auxiliary Feedwater Steam Bypass Control System Decay Heat Removal | |
| 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 41018 | 4 September 2004 05:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Steam Generator Level Oscillations | While reducing power in response to expected severe weather conditions from hurricane Frances, the reactor was manually tripped from 21% reactor power. The reactor was tripped because of significant swings in the 'B' Steam Generator level caused by erratic operation of the 'B' Feedwater Regulating Valve (FRV). All Control Element Assemblies fully inserted. Decay heat is currently being removed by Main Feedwater using the low power FRV bypass valves and Steam Bypass Control System (SBCS), maintaining RCS temperature at 532 degrees Fahrenheit. The unit will be maintained in a shutdown condition until repairs to the FRV are completed and the severe weather from Hurricane Frances abates. All primary and secondary systems performed as expected with the exception of the SBCS. The SBCS consists of five pressure control valves, PCV-8801 through PCV-8805 with PCV-8801 a larger capacity valve and designed to open first. The remaining valves are designed to open in series with overlap through their operating ranges. PCV-8801 failed to open and PCV-8802, 8803, and 8804 did not appear to properly control RCS temperature. PCV-8805 was operated in manual to control (steam generator) pressure. The Licensee notified the NRC Resident Inspector.
Florida Power and Light (FPL) is retracting this notification because the feedwater issues during the shutdown were not causal to the manual reactor trip. The St. Lucie Emergency Plan, requires that the units be taken offline prior to the onset of hurricane force winds onsite. In accordance with these requirements, St. Lucie Unit 2 was being taken offline prior to the arrival of hurricane Frances. Although automatic feedwater issues occurred during the downpower, the operators successfully took manual control of the main feedwater system. A reactor trip was not necessary to mitigate the condition, and continued operation and on-line troubleshooting would have been practical had the plant not been required to be shutdown for the approaching hurricane. The main feedwater control issues were not relevant factors during the planned plant shutdown/manual reactor trip. Therefore FPL is retracting this notification. The licensee notified the NRC Resident Inspector. Notified R2DO(Boland). | Steam Generator Feedwater Steam Bypass Control System | |
| 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 | |
| ENS 40375 | 4 December 2003 21:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | St. Lucie Unit 2 Manual Reactor Trip Due to Loss of Condensate Pump | On December 4, 2003, at 1605 hours, a down power was initiated due to a failing bearing on the 2A Condensate Pump. The pump bearing was hot and smoking. The plant fire team was deployed as a precautionary action. Due to continued degradation of the Pump bearing, a Manual Reactor Trip was initiated at approximately 60% power. Feed to the 2A and 2B Steam Generators was maintained via the 2B Main Feedwater Pump. All plant safety systems responded normally and plant safety functions were maintained throughout the event. The Plant was stabilized In Mode 3. Plant post trip anomalies include Steam Generator Blowdown isolation valves closed, Control Room ventilation system swapped to recirculation mode, the Fuel Handling Building ventilation system swapped to the Shield Building, and it was necessary to take Steam Bypass Control System to manual. An Emergency Response Team was 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 manual initiation of the RPS Reactor Trip. All control rods fully inserted into the reactor on the trip. The emergency diesel generators are available and the offsite electrical grid is in a normal configuration. No safety relief valves or power operated relief valves were known to have actuated during this event. St. Lucie Unit 1 was not affected and continues to operate in mode 1 at 100% rated thermal power. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Shield Building Steam Bypass Control System Safety Relief Valve Control Rod |