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| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 55626 | 6 December 2021 19:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 3 Automatic Trip | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. At 1203 MST on December 6, 2021, the Unit 3 reactor automatically tripped on low departure from nucleate boiling ratio. A part-strength control element assembly was being moved at the time of the trip. Unit 3 is stable and in Mode 3. In response to the reactor trip, all control element assemblies inserted fully into the core. Safety-related electrical power remains energized from off-site power sources and reactor coolant pumps continue to provide forced circulation through the reactor. Decay heat is being removed by the steam bypass control system and main feedwater system. Required systems operated as expected. No emergency classification was required per the Emergency Plan. The NRC Senior Resident Inspector has been informed. Units 1 and 2 were unaffected by this transient. |
| ENS 55114 | 26 February 2021 17:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to Loss of Two Reactor Coolant Pumps | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. At 1033 MST on February 26, 2021 the Unit 2 reactor automatically tripped due to a loss of two Reactor Coolant Pumps stemming from a loss of a 13.8 kV non-class bus during maintenance. Following the reactor trip, all Control Element Assemblies inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Unit 2 is stable and in Mode 3. Decay heat is being removed by the Steam Bypass Control System and Main Feedwater System. The NRC Senior Resident Inspector has been informed. Unit 1 and Unit 3 were not affected by the Unit 2 reactor trip. |
| 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. |
| ENS 54621 | 19 March 2020 07:00:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 Notification - Transducer Defect | The following is a summary of information obtained from Palo Verde via email: During pre-installation testing, five Masoneilan Model transducers were identified as unable to be calibrated prior to installation. These transducers were received by the vendor as safety-related components and are used to provide remote control operation of the Atmospheric Dump Valves (ADVs). The ADVs provide a safety grade method for cooling the unit to Shutdown Cooling System entry conditions, should Palo Verde's preferred heat sink via the Steam Bypass Control System to the condenser and/or atmosphere not be available. The transducer receives a 4-20 mA signal and translates it to a 3-15 psi output to the ADV positioner. This is accomplished by varying the supply air from 23-30 psi down to the appropriate 3-15 psi signal. Palo Verde provided an Interim 10 CFR 21 Report for the five Masoneilan Model 8005N transducers, Part 21 log number 2019-36-00, ML 19323C971, which was submitted by Arizona Public Service (APS) Company on November 15, 2019. An evaluation of the transducers was completed on March 19, 2020. The evaluation concluded that the inability of the transducers to be calibrated represented a defect. The licensees affected are undetermined at this time. Palo Verde has been in communication with the vendor. The vendor has currently not provided an extent of condition. Point of Contact: Lorraine Weaver (602) 448-5915 |
| 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. |
| 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. |
| 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. |
| ENS 52321 | 27 October 2016 05:21:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of Charging and Letdown Systems from the Reactor Coolant System | At 0021 (CDT) on 10/27/16, Waterford 3 (WF3) experienced a loss of the charging and letdown systems from the Reactor Coolant System (RCS). Technical Specification (TS) 3.0.3 was entered due to the loss of all three charging pumps. Charging Pump AB was restored and aligned to replace Charging Pump A and WF3 exited TS 3.0.3 at 0055 on 10/27/16. The cause of the loss of charging pumps was due to Refueling Water Storage Pool to Charging Pumps Suction Isolation, CVC-507, not opening as expected following a loss of Static Uninterruptible Power Supply (SUPS) 014AB during electrical troubleshooting. The cause of CVC-507 not opening is being investigated. Power was restored to SUPS 014AB at 0022. WF3 is currently stable in Mode 3 with decay heat being removed by the Steam Bypass Control System. Pressurizer Level was maintained throughout the event. WF3 was previously shut down for reasons unrelated to this event. The NRC Resident Inspector has been notified. Valve CVC-183 closed when the power supply was lost. CVC-183 is the Volume Control Tank outlet isolation valve. Waterford 3 will remain in mode 3 until the issue has been corrected.
This is a retraction of EN 52321 which was reported as an 8 hour Non-Emergency on October 27th at 0826 EST. At 0021 (EST) on 10/27/16, Waterford 3 (WF3) experienced a loss of the charging and letdown systems due to an electrical transient on a Static Uninterruptable Power Supply that was being worked under a maintenance work order. The cause of the loss of charging pumps was due to Refueling Water Storage Pool to Charging Pumps Suction Isolation, CVC-507, not opening as expected following a loss of Static Uninterruptible Power Supply (SUPS) 014AB. Both operating charging pumps automatically secured due to low suction pressure trips as designed. Post event investigation determined that a relay that had failed affected only the normal suction path isolation valves to the charging pumps and did not impact the safety related flow path that is required during a Safety Injection Actuation Signal (SIAS). Had an SIAS occurred during the period when no active suction path was aligned, the low pressure trip would have been blocked and the pumps selected to start on an SIAS would have auto started. The SIAS would have aligned the Boric Acid Make-up system for Emergency Boration. This would have resulted in the Charging Pumps being aligned to take suction from the Boric Acid Make-up pumps and/or Boric Acid Gravity Feed valves. A function of the charging systems is to inject concentrated boric acid into RCS upon an SIAS. As discussed in FSAR Section 6.3.3.3.1, the injection flow from the charging pumps is not credited in the small break LOCA analysis. Charging is however credited for natural circulation cooldown without letdown in order to meet the safe shutdown requirements of NRC Branch Technical Position (RSB) 5-4. This analysis assumes that the charging source is initially Boric Acid Makeup Tanks followed by Refueling Water Storage Pool. Both sources were available. The charging system was fully capable of performing its safety function following the relay failure. The charging pumps remained capable of starting on an SIAS and the flow path from the Boric Acid Management system remained operable. In addition the flow path from the RWSP was not affected since the outlet isolation valve could be manually opened. The NRC Resident Inspector has been notified. Notified the R4DO (Taylor). |
| 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. |
| 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. |
| 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. |
| 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. |
| 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. |
| 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. |
| ENS 45200 | 13 July 2009 11:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Identified | On 07/13/2009 at 00:38 EDT a planned manual reactor trip was initiated on Unit 2 based on increased, unidentified RCS leakage. The RCS leakage had increased over a four day period. The downpower was initiated at 21:32 EDT from 100% power and proceeded in a controlled manner to 25% power. The reactor was manually tripped at 00:38 EDT, as planned. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed and the unit was stabilized in Mode 3. Decay heat was removed by the Steam Bypass Control System (SBCS) to the condenser. The Main Feedwater Pump unexpectedly tripped on low suction pressure just prior to the reactor trip, but was recovered prior to levels reaching the Auxiliary Feedwater Actuation (AFAS) setpoint. There were no further major equipment failures identified post trip. On 7/13/2009 at 07:00 EDT a Containment walkdown confirmed the leak was caused by a J-joint weld defect on the 2B2 RCP Seal injection line. The leak is classified as reactor coolant pressure boundary leakage. The plant is in the process of cooling down to enter Mode 5 as required by Technical Specification 3.4.6.2.a. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(A). Additional investigation is being performed to determine the cause of the leak. The initial leak rate was less than 0.2 gpm. The repair will require taking the plant to a reduced inventory. The licensee notified the NRC Resident Inspector |
| 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. |
| ENS 44525 | 28 September 2008 04:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 3 Reactor Manually Tripped Following Automatic Turbine Trip on High Vibration | The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS under the reporting requirements of 10CFR50.73. On September 27, 2008, at approximately 21:51 Mountain Standard Time (MST) Palo Verde Unit 3 operators manually tripped the reactor from approximately 34% rated thermal power in response to an automatic trip of the main turbine. The unit was being shutdown from 100% power due to a chemistry excursion in the secondary plant. Following the manual reactor trip all CEAs fully inserted into the reactor core. This was an uncomplicated reactor trip. No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 3 is stable at normal operating temperature and pressure in Mode 3. Decay heat removal is via the Steam Generators and Main Feedwater to the Main Condenser using the Steam Bypass Control System. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The main turbine tripped on high vibration. The Senior Resident Inspector was informed of the Unit 3 reactor trip. |
| ENS 44496 | 16 September 2008 21:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip During Troubleshooting of Control Rod Drive Mechanism Motor Generators | The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On September 16, 2008 at approximately 1400 hours Mountain Standard Time (MST) Palo Verde Nuclear Station Unit 3 experienced a manual reactor trip. At the time of the trip trouble shooting efforts were in progress to investigate abnormalities with the control element drive mechanism (CEDM) motor generators (MG). The MGs (2 - 100% each) normally operate in parallel to supply the necessary power to grip, move and hold the control element assemblies (CEA). During the trouble shooting activities the 'B' MG was removed from service. The 'A' MG initially provided power to the CEDMs but did not maintain the power. As a result of the pre-job briefing contingency actions, a manual reactor trip was ordered. All CEAs fully inserted into the core. No other emergency actuation signals were initiated and none were required. Off site power provided power to the class buses during and after the event. Decay heat removal is being provided by the steam bypass control system to the main condenser. No major equipment was inoperable prior to the event that contributed to the event. The unit is in Mode 3, Hot Standby, at normal temperature and pressure. The cause of the MG abnormality is under investigation. Normal feedwater remained in service providing water to the steam generators. The licensee notified the NRC Resident Inspector. |
| ENS 43580 | 19 August 2007 21:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Plant Shutdown Due to Unisolable Reactor Coolant System Leakage | On 08/18/2007 at 16:03 EDT a Planned Manual Reactor Trip was initiated on Unit 2 based on increased unidentified RCS leakage. The RCS leakage had increased over a two day period and surpassed the hard limit of 0.28 gpm established by a plant Operational Decision Making (ODM) plan. The downpower was initiated on 08/18/07 at 10:35 EDT from 100% power and proceeded in a controlled manner to 25% power. The Feedwater Control System was maintained in the automatic mode. Station Power was transferred by procedure from the Auxiliary to the Startup Transformers. The reactor was manually tripped at 16:03 EDT. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed without contingencies and the unit was stabilized in Mode 3. Decay heat was removed by the Steam Bypass Control System (SBCS) to the condenser. The Steam Generators were fed by the MFW System and levels remained above Auxiliary Feedwater Actuation (AFAS) setpoint. There were no significant equipment failures identified post trip. The plant was cooled down and entered Mode 5 on 8/19/07 at 15:09 EDT. On 8/19/07 at 17:00 EDT while in Mode 5, a Containment walkdown confirmed the 2B1 RCP Seal injection line to be the only active RCS leakage source. The leak is unisolable from the RCS and is located at or adjacent to a socket weld on the 3/4" Class 1 Schedule 160 seal injection line. Removal of obstructions and insulation must occur before the specific details of the leak can be identified. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(A). Additional investigation is being perform to determine the detailed configuration of the leak and the root cause. The licensee notified the NRC Resident Inspector. |
| ENS 42925 | 21 October 2006 22:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Control Element Assembly Position Transmitter Output Deviations | The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On October 21, 2006 at approximately 1549 MST Palo Verde Unit 1 experienced a reactor trip (RPS actuation) from 100% rated thermal power due to low departure from nucleate boiling (DNBR) trips on all four channels of the core protection calculators (CPCs). The unit was at normal temperature and pressure prior to the trip. Several minutes prior to the Reactor Trip, RSPT#1 (Reed Switch Position Transmitter) CEA deviation alarms were received for CEA#29 (Shutdown Group B CEA). Operators observed that the magnitude of the deviation was fluctuating erratically. While investigating the alarms CEAC #1 Sensor Fail alarms were received and at 1549 MST, the Reactor automatically tripped on a CPC generated Lo DNBR trips on all 4 channels of CPCs. The apparent cause is presently suspected to be a failure of RSPT#1. An investigation has commenced to determine the root cause of the reactor trip. All of the control rods fully inserted into the core. The Steam Bypass Control System operated as designed, directing steam flow to the condenser. No main steam or primary relief valves lifted and none were required. There was no loss of heat removal capability or loss of safety functions associated with the event. Electrical buses transferred to offsite power as designed. The Shift Manager determined this event was an uncomplicated reactor trip. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event nor contributed to the event. Unit 1 is stable at normal operating temperature and pressure in Mode 3. No ESF actuations occurred and none were required. The event did not result in any challenges to the fission product barrier or result in any releases of radioactive materials. There were no adverse safety consequences or implications as a result of this event. The event did not adversely affect the safe operation of the plant or health and safety of the public. The NRC Resident was notified at 17:38 MST. |
| 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. |
| 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). |
| ENS 40870 | 14 July 2004 08:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Following a Main Generator Trip | The following information was received from the licensee via facsimile: The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On July 14, 2004, at approximately 01:35 Mountain Standard Time (MST) Palo Verde Unit 2 experienced a Main Generator Trip immediately followed by an automatic Reactor Trip. The reactor was at approximately 100% power and normal operating temperature and pressure prior to the event. The cause of the Main Generator Trip was most likely the result of electrical storm conditions present at the site at the time of the trip. Unit 2 was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. Heat removal was maintained to the condenser via the steam bypass control system. This was an uncomplicated reactor trip, No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 2 Is stable at normal operating temperature and pressure in Mode 3. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The (NRC) Senior Resident Inspector was informed of the Unit 2 reactor trip. No primary or secondary power-operated or manual relief valves lifted as a result of the plant transient. Units 1 and 3 were unaffected by the trip on Unit 2. Offsite power was maintained to Unit 2 safety busses throughout the event. |
| ENS 40795 | 7 June 2004 21:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to Main Turbine Electro-Hydraulic Control System Failure | On June 7, 2004, at approximately 14:58 Mountain Standard Time (MST) while at 99% RTP (rated thermal power), Palo Verde Unit 3 experienced an apparent electro-hydraulic control (EHC) system fault resulting in Combined Intercept Valve (CIV) closure. This plant upset was followed by a Reactor Power Cutback System (RPCS) initiation. Several seconds later the Reactor automatically tripped on Lo DNBR from approximately 65% RTP. Unit 3 was at normal temperature and pressure prior to the trip. All CEAs (control rod assemblies) inserted fully into the reactor core. This was an uncomplicated reactor trip. No emergency classification was required per the Emergency Plan. No automatic ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The Emergency Diesel Generators did not start and were not required. The offsite power grid is stable. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event that contributed to the event. Unit 3 is stabilized at normal temperature and pressure at approximately 565 degrees F and 2250 psia in Mode 3. The reactor coolant system remains in normal forced circulation with heat removal via the steam bypass control system to the condenser and feedwater from the non-essential auxiliary feedwater system. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The Senior Resident Inspector was informed of the Unit 3 reactor trip and this notification. The Senior Resident Inspector was on-site at the time of the reactor trip. |
| 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. |