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| Entered date | Site | Scram | Region | Reactor type | Event description | |
|---|---|---|---|---|---|---|
| ENS 57032 | 16 March 2024 18:36:00 | Waterford | Manual Scram | NRC Region 4 | CE | The following information was provided by the licensee via phone and email: At 1449 CDT, Waterford 3 Steam Electric Station was operating at 100 percent power when a manual reactor trip was initiated due to main feed isolation valve (FW-184B) and main steam isolation valve (MS-124B) going closed unexpectedly. Emergency feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system. 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 through the turbine bypass valves and the atmospheric dump valve on loop '2'. There is no primary to secondary system leakage. The cause of the isolations is still being investigated. |
| ENS 56991 | 24 February 2024 18:08:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | The following information was provided by the licensee via email: At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56856 | 16 November 2023 05:15:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report The NRC Senior 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. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). |
| ENS 56839 | 7 November 2023 18:42:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. |
| ENS 56459 | 9 April 2023 04:42:00 | Palo Verde | Automatic Scram | NRC Region 4 | CE | The following information was provided by the licensee via email: 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 2144 MST on April 8, 2023, the Unit 1 reactor automatically tripped due to the loss of reactor coolant pumps stemming from the loss of 13.8 kV power to the pumps. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip as expected so the control room operators opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses failed to complete a fast bus transfer, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. Following the reactor trip, all control element assemblies inserted fully into the core. No automatic specified system actuation was required or occurred. No emergency plan classification was required per the Emergency Plan. Safety related buses remained powered from offsite power during the event and the offsite power grid is stable. Unit 1 is stable and in Mode 3. Decay heat is being removed by the atmospheric dump valves and the class 1E powered motor driven auxiliary feedwater pump. The loss of hydraulic pressure, the main generator output breakers failing to automatically open and the fast bus transfer not actuating are being investigated. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B). The NRC Senior Resident Inspector has been informed. Unit 2 is in a refueling outage in Mode 5 and Unit 3 is in Mode 1 at 100 percent power.
This update is being made to report the manual actuation of the B-train auxiliary feedwater pump and manual main steam isolation signal (MSIS) actuation affecting multiple main steam isolation valves (MSIVs) following the reactor trip. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Warnick)
This update is intended to clarify the initial description of the event that occurred on 4/8/2023. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip. The control room operators manually opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses de-energized. A fast bus transfer did not occur per design, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Gaddy) |
| ENS 56087 | 3 September 2022 01:58:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 55963 | 25 June 2022 00:44:00 | Waterford | Automatic Scram | NRC Region 4 | CE | The following information was provided by the licensee via email: At 2012 CDT, Waterford 3 Steam Electric Station, Unit 3 was operating at 100 percent power when an automatic reactor trip occurred due to Main Steam Isolation Valve MS-124B going closed unexpectedly. Subsequently, both main feedwater isolation valves shut. Emergency Feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected and all other plant equipment functioned as expected. This was an uncomplicated scram. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system. The NRC Resident Inspector has been notified. |
| ENS 55684 | 3 January 2022 15:58:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | The following information was provided by the licensee via email: At 1223 (EST) on 01/03/2022, Calvert Cliffs Unit 2 automatically tripped from 100 percent power due to loss of electrical load. The cause is under investigation. The site Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods inserted and decay heat is being removed via the condenser. The plant is in a normal shutdown electrical lineup. There was no impact on Unit 1. |
| ENS 55641 | 10 December 2021 13:35:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 55597 | 21 November 2021 14:28:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | At 1046 EST on November 21, 2021, with Calvert Cliffs Nuclear Power Plant Unit 2 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering levels in both steam generators following a loss of the 21 and 22 steam generator feed pumps. An Auxiliary Feedwater System actuation occurred to restore steam generator water levels. The trip was not complicated, with all systems responding normally. Decay heat is being removed by the Auxiliary Feedwater System. Calvert Cliffs Nuclear Power Plant Unit 1 is unaffected and remains in Mode 1 at 100 percent power. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification. RPS actuation, per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight-hour, non-emergency notification, Specific System Actuation, per 10 CFR 50.72(b)(3)(vi)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55265 | 19 May 2021 08:35:00 | Palo Verde | Automatic Scram | NRC Region 4 | CE | At 0315 MST on May 19, 2021, Unit 2 reactor automatically tripped during testing of the Plant Protection System. The Reactor Protection System actuated to trip the reactor on High Pressurizer Pressure, although no plant protection setpoints were exceeded. Main Steam Isolation Signal (MSIS), Safety Injection Actuation Signal (SIAS), and Containment Isolation Actuation Signal (CIAS) were received. No injection of water into the Reactor Coolant System occurred. Auxiliary Feedwater Actuation Signals (AFAS) 1 and 2 actuated on low Steam Generator water level post trip as designed. This event is being reported as a reactor protection system and a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Following the reactor trip, all (Control Element Assemblies) CEAs inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. Unit 2 is stable and in Mode 3. Steam Generator heat removal is via the class 1 E powered motor driven auxiliary feedwater pump and Atmospheric Dump Valves. The NRC Senior Resident Inspector has been informed.
The Unit 2 reactor tripped because of actual High Pressurizer Pressure that occurred as a result of a Main Steam Isolation Signal actuation. At 0337 MST, both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) were made inoperable when the injection valves were overridden and closed in accordance with station procedures. At 0346 MST, in accordance with station procedures, both trains of Containment Spray, LPSI, and HPSI pumps were overridden and stopped, rendering Containment Spray inoperable as well. This represents a condition that would have prevented the fulfillment of a safety function required to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). Additionally, at the time of the Safety Injection Actuation Signal (0315 MST), both trains of Emergency Diesel Generators actuated as required and both 4160 VAC busses remained energized from off-site power. The NRC Senior Resident Inspector has been informed. Notified R4DO (Young)
The inoperability of both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) and both trains of Containment Spray (CS) following the Unit 2 reactor trip has been determined to be an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Additionally, inoperability of both trains of HPSI resulted in a reportable condition that could prevent fulfillment of its credited safety function to maintain the reactor in a safe shutdown condition per 10 CFR 50. 72(b)(3)(v)(A). The additional reporting criteria were discovered during review of the event and corresponding safety analyses. The NRC Senior Resident Inspector has been informed. Notified R4DO (Werner) |
| ENS 55147 | 21 March 2021 23:57:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | At 2216 EDT on 3/21/2021, Calvert Cliffs Unit 2 was manually tripped from 37 percent power due to lowering level in the 21 Steam Generator. All systems responded per design. Main Feedwater was secured and Auxiliary Feedwater was manually initiated. The Site Senior Resident has been notified. The cause of the lowering level in the 21 Steam Generator is under investigation. |
| ENS 55078 | 20 January 2021 21:48:00 | Saint Lucie | Automatic Scram | NRC Region 2 | CE | 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 55028 | 10 December 2020 20:43:00 | Arkansas Nuclear | Automatic Scram | NRC Region 4 | CE | On December 10, 2020 at 1608 CST, Arkansas Nuclear One, Unit 2 (ANO-2) experienced an automatic reactor scram from 100 percent power due to Low Steam Generator Water Level in 2E-24A Steam Generator. Emergency Feedwater actuated automatically due to low water level in the A Steam Generator. Due to inadequate control of the B Main Feedwater Control System, water level in the B Steam generator rose to a level requiring manual trip of the B Main Feedwater pump. Emergency Feedwater responded as designed to feed both steam generators automatically. All other systems responded as designed. All electrical power is being supplied from offsite power and maintaining unit electrical loads as designed. Unit 2 is currently stable in Mode 3 (Hot Standby) maintaining pressure and temperature via Emergency Feedwater and secondary system steaming. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of both 10 CFR 50.72(b)(2)(iv)(6) for the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) for the actuation of the Emergency Feedwater System. The Arkansas Nuclear One NRC Senior Resident Inspector has been notified.
The purpose of this (report) is to provide an update to NRC Event Number 55028. The cause of the inadequate control of the B Main Feedwater Control System to control B Steam Generator Level was verified to be associated with the failure that led to the A Steam Generator low level condition. After taking action to trip the B Main Feedwater Pump, Emergency Feedwater was manually actuated for the B Steam Generator and the Emergency Feedwater System was verified to maintain proper automatic control of both Steam Generator levels. At the time of the initial event notification, plant temperature and pressure control had been transferred from Emergency Feedwater to Auxiliary Feedwater along with secondary system steaming. The licensee notified the NRC Resident Inspector. Notified R4DO (Kellar). |
| ENS 54978 | 2 November 2020 08:10:00 | Waterford | Automatic Scram | NRC Region 4 | CE | On November 2, 2020, at 0419 CST, Waterford 3 experienced an automatic reactor trip due to a Control Element Drive Mechanism Control System timer failure while attempting to synchronize a second motor generator set. All control rods fully inserted. The plant is currently in Mode 3 and stable with normal feedwater feeding and maintaining both Steam Generators. The NRC Senior Resident Inspector has been notified. The cause of the failure is still under investigation. |
| ENS 53424 | 23 May 2018 17:37:00 | Palo Verde | Automatic Scram | NRC Region 4 | CE | 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 May 23, 2018, at approximately 1128 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 control room received reactor protection system alarms for low departure from nucleate boiling ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 2 was operating normally at 100 percent power. Plant operators entered the reactor trip procedures and diagnosed an uncomplicated reactor trip. All CEAs (control element assemblies) fully inserted into the core. No emergency classification was required per the PVNGS Emergency Plan. The Unit 2 safety-related electrical buses remained energized from normal offsite power during the event. There was no impact to the required circuits between the offsite transmission network and the onsite Class 1E Electrical Power Distribution System; the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 2 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. The cause of the reactor trip is under investigation. 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 Resident Inspector has been informed of the Unit 2 reactor trip. Decay is being removed via steam dumps to condenser. Units 1 and 3 at Palo Verde were unaffected by the transient and continue to operate at 100 percent power. |
| ENS 53215 | 16 February 2018 02:50:00 | Palo Verde | Automatic Scram | NRC Region 4 | CE | 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 February 15, 2018, at approximately 2153 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 1 Control Room received Reactor Protection System alarms for Low Departure from Nucleate Boiling Ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 1 was operating normally at 100 percent power. Plant operators entered the emergency operations procedures and diagnosed an uncomplicated reactor trip but noted that Reactor Coolant Pumps 1B and 2B were not running due to a loss of power. All CEAs (Control Element Assemblies) fully inserted into the core. Following the reactor trip, all nuclear instruments responded normally. No emergency classification was required per the PVGS Emergency Plan. The PVGS Unit 1 safety related electrical busses remained energized from normal offsite power during the event. The Unit 1 'B' Diesel Generator is currently removed from service for maintenance. Due to ongoing planned maintenance on NAN-X02, Startup Transformer 2, fast bus transfer for NAN-S02 (from NAN-S04) was blocked. This resulted in a loss of offsite power to NAN-S02 and NBN-S02. The offsite power grid is stable. Unit 1 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. 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 Resident Inspector has been informed of the Unit 1 reactor trip.
Unit 1 is stable in Mode 3 following an uncomplicated trip. Offsite power has been restored to non-safety related electrical busses. Troubleshooting continues to determine the cause of the event. During performance of the alarm response procedure, it was identified that the seismic monitoring (SM) system had been in alarm since the reactor trip and was incapable of performing its emergency plan function. Pursuant to 10 CFR 50.72(b)(3)(xiii), this condition constitutes a major loss of emergency assessment capability. Compensatory measures have been implemented in accordance with PVNGS procedures to provide alternative methods for HU2.1 event classification with the SM system out of service. Maintenance is currently in progress to restore SM system functionality. The licensee notified the NRC Resident Inspector. Notified R4DO (Werner).
Station staff completed an evaluation of event EN #53215 reported on February 15, 2018, and determined that the seismic monitoring system remained capable of assessing a seismic event following the reactor trip. Therefore, a major loss of emergency assessment capability pursuant to 10 CFR 50.72(b)(3)(xiii) did not occur as reported in the update on February 16, 2018. The NRC Resident Inspectors have been notified. Notified the R4DO (Gaddy). |
| ENS 53036 | 26 October 2017 05:54:00 | Saint Lucie | Automatic Scram | NRC Region 2 | CE | 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 52863 | 17 July 2017 17:37:00 | Waterford | Automatic Scram | NRC Region 4 | CE | During a rain and lightning storm, plant operators observed arcing from the main transformer bus duct and notified the control room. The decision was made to trip the main generator which resulted in an automatic reactor trip. The plant entered EAL SU.1 as a result of the loss of offsite power for greater than fifteen minutes. Plant safety busses are being supplied by both emergency diesel generators while the licensee inspects the electrical system to determine any damage prior to bringing offsite power back into the facility. Offsite power is available to the facility. No offsite assistance was requested by the licensee. During the trip, all rods inserted into the core. Decay heat is being removed via the atmospheric dump valves with emergency feedwater supplying the steam generators. The main steam isolation valves were manually closed to protect the main condenser. There were no safeties or relief valves that actuated during the plant transient. There is no known primary-to-secondary leakage. Reactor cooling is via natural circulation. All safety equipment is available for the safe shutdown of the plant. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies. Notified DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).
This notification is also made under 10 CFR 50.72(b)(3)(v)(D). This is a non-emergency notification from Waterford 3. On July 17, 2017 at 1606 CDT, the reactor automatically tripped due to a loss of Forced Circulation, which was the result of Loss of Offsite Power (LOOP) to the electrical (safety and non-safety) buses. Both 'A' and 'B' trains of Emergency Diesel Generators (EDGs) started as designed to reenergize the 'A' and 'B' safety buses. The LOOP caused a loss of feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater (EFW) system. Prior to the reactor trip, at 1600 CDT, personnel noticed the isophase bus duct to main transformer 'B' glowing orange due to an unknown reason. Due to this, the main turbine was manually tripped at 1606 CDT. Following the turbine trip, the electrical (safety and non-safety) buses did not transfer to the startup transformers as expected due to an unknown reason. The plant entered the Emergency Operating Procedure for LOOP/Loss of Forced Circulation Recovery. At 1617 CDT, an Unusual Event was declared due to Initiating Condition (IC) SU1 - Loss of all offsite AC power to safety buses (greater than) 15 minutes. All safety systems responded as expected. The plant is currently in mode 3 and stable with the EDGs supplying both safety buses and with EFW feeding and maintaining both steam generators. Offsite power is in the process of being restored. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies.
The licensee terminated the Notification of Unusual Event at 2056 CDT. The basis for terminating was that offsite power was restored to the safety busses. The licensee has notified Louisiana Department of Environmental Quality, St. John and St. Charles Parishes, Louisiana Homeland Security Emergency Preparedness, and will be notifying the NRC Resident Inspector. Notified IRD (Stapleton), NRR (King), R4DO (Hipschman), DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).
This update is being reported under 10 CFR 50.72(b)(3)(v)(B). During the event discussed in EN# 52863, at 1642 CDT (on July 17, 2017), Condensate Storage Pool (CSP) level lowered to less than 92% resulting in entry to Technical Specification (TS) 3.7.1.3. Level in the CSP was lowered due to feeding from both Steam Generators with EFW. Normal makeup to the CSP was temporarily unavailable due to the LOOP. Filling the CSP commenced at 1815 CDT (on July 17, 2017), and TS 3.7.1.3 was exited on July 18, 2017 at 0039 CDT. The licensee notified the NRC Resident Inspector. Notified R4DO (Hipschman).
Waterford 3 is retracting a follow up notification made on July 19, 2017 for EN# 52863, concerning the loss of safety function associated with the Condensate Storage Pool (CSP) per 10 CFR 50.72(b)(3)(v)(B). The Condensate Storage Pool was performing its required safety function by providing inventory to the Emergency Feed Water pumps when the required Tech Spec level (T.S. 3.7.1.3) dropped below 92%. The Technical Specification was entered at 1624 (CDT) on July 17, 2017 and exited after filling at 0039 on July 18, 2017. The total allowed outage time allowed by Tech Spec 3.7.1.3 is 10 hours to be in Hot Shutdown if not restored. The Condensate Storage Pool level was restored to greater than 92% prior to exceeding the allowed outage time. Based on level being restored and the Condensate Storage Pool performing its required safety function, 10 CFR 50.72(b)(3)(v)(B) does not apply. Prior to the automatic reactor trip, Condensate Storage Pool level was greater than 92%. The NRC Resident Inspector has been notified of the retraction. Notified R4DO (Groom). |
| ENS 52710 | 26 April 2017 14:49:00 | Arkansas Nuclear | Automatic Scram | NRC Region 4 | CE B&W-L-LP | At 1004 CDT, Arkansas Nuclear One, Unit 1 (ANO-1) reactor automatically tripped due to the partial loss of offsite power. At the time of the trip, the site was in a Tornado Warning and a Severe Thunderstorm Warning. The Emergency Feedwater (EFW) system auto-actuated due to the loss of main feedwater pumps and the loss of the Reactor Coolant pumps. Both Emergency Diesel Generators started as expected with only one loading as expected. All control rods fully inserted. Currently, ANO-1 has stabilized in Hot Standby via natural circulation. ANO-1 also lost Spent Fuel Pool cooling for approximately 69 minutes. The temperature of the spent fuel pool at the beginning of the event was approximately 102 (degrees) F. The spent fuel pool saw a heatup of 1 (degree) F during the loss of spent fuel pool cooling. The Spent Fuel Pool cooling has been restored. ANO-2 is currently in a refueling outage with all fuel in the spent fuel pool. ANO-2 completed a full core off load to the spent fuel pool and this was completed on April 12, 2017. Spent Fuel Pool cooling was lost for approximately 10 minutes. The Spent Fuel Pool temperature was 91 (degrees) F prior to the event. No heat up of the pool was identified during the event. Cooling has subsequently been restored. The #1 Emergency Diesel Generator auto-started as designed but did not supply the safety bus due to availability of offsite power. No radiological releases have occurred from either unit due to this event. The licensee has notified the NRC Resident Inspector. |
| ENS 52406 | 4 December 2016 01:48:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | On 12/3/16 at 2224 EST, Calvert Cliffs Unit-2 experienced an automatic reactor trip from full power due to a leak in the Unit-2 Main Turbine Electro-Hydraulic Control (EHC) system. The EHC leak caused the Unit-2 Main Turbine governor valves to close, resulting in a turbine trip and automatic reactor trip. The site Outage Control Center is manned, and investigation into the cause of the leak is underway. Unit-2 remains stable in Mode 3 with normal heat removal. Unit-1 remains at full power and was not affected by the trip. The plant is in a normal shutdown electrical lineup. All Control rods fully inserted and no primary or secondary safety relief valves lifted during the trip. The licensee has notified the NRC Resident Inspector. The licensee will be notifying Calvert County. |
| ENS 52226 | 8 September 2016 04:27:00 | Palo Verde | Manual Scram | NRC Region 4 | CE | On September 7th, 2016 at approximately 2131 Mountain Standard Time (MST), Palo Verde Unit 1 was manually tripped due to a stuck open main spray valve. Unit 1 was operating at 100 percent power at normal operating temperature and pressure prior to the event. A 120 VAC non-class instrument distribution panel was being transferred to its alternate power supply to establish maintenance conditions. The distribution panel failed to transfer. The panel remained energized from its normal power supply; however, multiple components powered from the distribution panel began to exhibit uncharacteristic behavior. At this time, it was noted that a reactor coolant system main spray valve was open. The alarm response procedure was followed; however, the actions taken were unsuccessful at closing the main spray valve. The plant was then manually tripped due to pressurizer pressure continuing to lower. The reactor coolant pumps were turned off to terminate main pressurizer spray flow to control pressurizer pressure due to the inability to close the main spray valve. No ESF (Engineered Safety Features) actuations occurred and none were required. No emergency classification was required per the emergency plan. 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. Limiting condition for operation 3.4.1 was entered due to low pressurizer pressure. No major equipment was inoperable prior to the event that contributed to the event. Unit 1 is stable at normal operating temperature and pressure in Mode 3. Reactor coolant pumps are secured and natural circulation has been verified. Primary pressure is being maintained at its normal operating pressure manually with pressurizer heaters and auxiliary spray, from the charging 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 minimum RCS pressure was approximately 2070 psia (normal 2250). The event did not adversely affect the safe operation of the plant or the health and safety of the public. All rods inserted and the trip was uncomplicated. Units 2 and 3 were not affected and continue to run at full power. The NRC Resident Inspector has been notified. |
| ENS 52169 | 11 August 2016 11:09:00 | Millstone | Manual Scram | NRC Region 1 | CE | Reactor operators manually tripped the reactor due to the loss of two out of four circulating water pumps which caused a drop in condenser vacuum. The trip was uncomplicated. The reactor is shutdown and stable with decay heat removal via steam dumps to the condenser. The cause of the circulating water pump trips is currently unknown, but initial indications are that the pumps tripped due to a lightning strike that caused an electrical perturbation. The reactor will remain shutdown while the licensee investigates the cause. Unit 3 was not affected. The licensee notified the NRC Resident Inspector and the State and Local governments. |
| ENS 51683 | 25 January 2016 05:45:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | At 0315 EST on 1/25/16, Calvert Cliffs Unit 1 was manually tripped from 10 percent power due to elevated condenser sodium levels. All systems responded per design. Main Feed was secured and auxiliary feed water was initiated. The elevated sodium levels are believed to be due to a condenser tube leak. The reactor is currently shutdown and stable in Mode 3 and will remain in Mode 3 until repairs are effected. Unit 2 was not affected and remains at full power. The Licensee has notified the NRC Resident Inspector. |
| ENS 51577 | 1 December 2015 20:25:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | On 12/01/2015 at 1820 EST, the Main Control Room received a 22 Steam Generator Feed Pump trip. The 22 Steam Generator Feed Pump was not able to be reset and the Main Control Room manually tripped the Unit 2 Reactor. The licensee entered Emergency Operating Procedure (EOP)-0, 'Post Trip Immediate Actions' and all safety functions were met. At 1833, Unit 2 transitioned into EOP-1, 'Uncomplicated Reactor Trip.' At 1841, Unit 2 transitioned into Operating Procedure #4 , 'Plant Shutdown from Power to Hot Stand-by.' The plant is stable in Mode 3. All control rods inserted fully on the reactor trip. No primary or secondary safety relief valves lifted. The steam generators are being fed by the 21 steam generator feed pump and decay heat is being dumped to the condenser via the steam dumps. The electric plant is in a normal shutdown electrical lineup and there was no impact on Unit 1. Unit 1 continues to operate at 100 percent power. The cause of the 22 steam generator feed pump trip is still under investigation. The licensee notified the NRC Resident Inspector. |
| ENS 51521 | 8 November 2015 02:55:00 | Millstone | Manual Scram | NRC Region 1 | CE | During power ascension following refueling outage, a decreasing oil level in the 'C' Reactor Coolant Pump was noted. When the oil level reached 69 percent, with the reactor at approximately 56 percent rated thermal power, per plant procedure, a rapid downpower was initiated which brought the plant to approximately 15 percent power and a manual reactor trip was initiated at that point. The reactor trip was uncomplicated and all plant equipment responded as expected. The licensee notified the NRC Resident Inspector. |
| ENS 51447 | 4 October 2015 03:51:00 | Waterford | Automatic Scram | NRC Region 4 | CE | At 2307 CDT Waterford 3 experienced an automatic reactor trip and all Control Element Assemblies (CEAs) inserted into the core. The cause of the automatic reactor trip is currently under investigation. The plant is currently in Mode 3 (Hot Standby) and stable with Main Feedwater feeding and maintaining both Steam Generators. Main Feedwater Pump 'A' tripped subsequent to the reactor trip. Emergency Feedwater actuated following the plant trip as expected, but was not required to maintain Steam Generator level. The plant entered the Emergency Operating Procedure for an uncomplicated reactor trip and has now transitioned to the normal operating shutdown procedure. Unit 3 is in a normal post trip electrical lineup. The Main Condenser is in-service removing decay heat.. The licensee informed the NRC Resident Inspector. |
| ENS 51397 | 16 September 2015 03:59:00 | Palisades | Automatic Scram | NRC Region 3 | CE | At 0117 (EDT) on 9/16/2015 a reactor trip occurred (4-hr non-emergency). The plant was at approximately 85% power performing a coastdown in preparation for a refueling outage when a Digital Electro-Hydraulic (DEH) alarm was received in the control room. Shortly following receipt of the alarm the turbine tripped. This resulted in an RPS actuation and a reactor trip on Loss of Load. The crew entered EOP-1 Standard Post Trip Actions and completed all required actions. The crew subsequently entered EOP-2 Reactor Trip Recovery. All full-length control rods inserted fully. Auxiliary Feedwater System actuated in response to low steam generator water levels (8-hr non-emergency). Steam generator water levels are in progress of being returned to normal operating levels. No known primary to secondary leakage. Atmospheric Steam Dump Valves lifted after the trip and subsequently reseated. The plant is currently stable in Mode 3 at NOP/NOT being maintained by the Turbine Bypass Valve. Initial investigation into the cause of the turbine trip appears to be from a DEH power supply failure. The NRC Resident Inspector was notified of the reactor trip at 0139 on 9/16/2015. |
| ENS 51302 | 10 August 2015 00:23:00 | Saint Lucie | Automatic Scram | NRC Region 2 | CE | 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. |
| ENS 51116 | 3 June 2015 21:36:00 | Waterford | Manual Scram Automatic Scram | NRC Region 4 | CE | This is a non-emergency notification from Waterford 3. At 1705 (CDT) the reactor was manually tripped in anticipation of an automatic trip due to loss of main feedwater pump 'A'. The plant is currently in mode 3 and stable with emergency feedwater feeding and maintaining both steam generators due to an automatic emergency feed actuation signal. During the trip, the 'B' electrical safety and non safety busses did not automatically transfer from the unit auxiliary transformer to the startup transformer causing a loss of off-site power to the 'B' electrical busses. This resulted in a loss of main feedwater pump 'B'. The 'B' emergency diesel generator started as designed and reenergized the 'B' safety related buses. The plant entered the emergency operating procedure for loss of main feedwater. Off-site power has been restored to the 'B' safety and non safety busses, and the emergency diesel generator 'B' is secured. All control rods fully inserted into the core following the trip. Decay heat is being removed by the main condenser using the turbine bypass valves. The electric plant is in a normal shutdown lineup. The licensee has notified the NRC Resident Inspector. |
| ENS 50961 | 7 April 2015 15:45:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | A loss of Main Generator Load which caused a Reactor Trip on Units 1 & 2. A switchyard voltage transient from a highline occurred, which caused an undervoltage condition on both units' safety related 4KV buses. Unit 1 is on normal heat removal to the condenser. Unit 2 is on auxiliary feedwater and normal condenser bypass valves for temperature control. An Auxiliary Feedwater Actuation System (AFAS) actuation occurred on Unit 2. The (Unit 2) 2B emergency diesel generator did not start and load on its respective 24-4 KV bus. The 24-4KV Bus was repowered from the alternate feeder breaker. Cause of the emergency diesel failure to start is under investigation. All safety functions are met for both units. All control rods fully inserted. The site is in a normal shutdown electrical configuration powered from offsite. The site plans to stay in Mode 3 pending restart. The licensee notified the NRC Resident Inspector, State and local authorities. A press release is planned.
During post trip review, it was determined that the 21 saltwater pump had to be manually started. With the failure of 2B emergency diesel generator, there were no saltwater pumps running for approximately 12 minutes. Additional troubleshooting determined the 2A emergency diesel generator sequencer did not automatically start 21 saltwater pump. The 2B emergency diesel generator was returned to service on 4/8/2015 at 1730 (EDT). The loss of saltwater (pump) and emergency diesel generator is reportable as an event that could have prevented fulfillment of a safety function and is also an unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R1DO (Ferdas), IRD MOC (Grant), NRR EO (Morris). |
| ENS 50607 | 12 November 2014 18:31:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 50063 | 27 April 2014 23:57:00 | Arkansas Nuclear | Automatic Scram | NRC Region 4 | CE | At approximately 1932 (CDT) on 4/27/2014, the System Operations Center (SOC - Dispatcher) informed Unit 2 of a system wide grid emergency and ordered both Unit 1 and Unit 2 to come off line as soon as possible. At approximately 2012 (CDT), Unit 2 automatically tripped from 51% power due to an Auxiliary Trip on CPCs (Core Protection Calculator) due to Axial Shape Index (ASI) trip. All Control Element Assemblies inserted into the core. Both vital and non-vital 4160V and 6900V buses remain powered from Startup #3 Transformer. All Systems responded as designed. At 1932 (CDT), Unit 1 commenced a Rapid Plant Shutdown at a rate 5-7% per min with the intention to take the turbine offline and leave the reactor critical at 10-12% power on the Turbine Bypass Valves. When the Unit 2 reactor tripped, Unit 1 stopped the power reduction and stabilized the plant at approximately 19% Reactor Power and 125 Generated Megawatts. With SOC concurrence, Unit 1 stabilized power and was told to limit site output to <200 MWe. At 1932 CDT, Unit 1 began a down power from 100% power and Unit 2 began a down power from 95% power. On Unit 2, decay heat is being removed by the main condenser using the turbine bypass valves. Unit 2 is stable in Mode 3 with stable offsite power. The system wide grid emergency is believed to be caused by tornados in the region. The licensee has notified the NRC Resident Inspector and the State. |
| ENS 49754 | 22 January 2014 00:33:00 | Calvert Cliffs | Automatic Scram | NRC Region 1 | CE | Dual Unit Trip due to loss of '21' 13 KV bus . All safety functions are met for both units. Unit 1 remained with normal heat removal. Unit 2 lost power to its normal heat sink and is stable on Auxiliary Feed water and Atmospheric Dump Valves for temperature control. Both trips were automatic trips. Due to loss of power a Under Voltage actuation occurred on both units ('14' and '24' 4Kv bus). Due to loss of main feed on Unit 2 a Auxiliary Feed water Actuation System (AFW) actuation occurred on Unit 2. Cause is under investigation. All control rods fully inserted on the loss of power to the Control Rod Drive Mechanisms (CRDMs). Both Units Reactor Coolant Pumps (RCPs) remained running during the transient. The normal Unit 2 heat sink was unavailable due to the loss of the operating circulating water pumps resulting in a loss of condenser vacuum. The Unit 2 AFW actuation included one of two steam-driven pumps and the motor-driven pump. Both Units Emergency Diesel Generators started and loaded and have since been secured. Both Units are stable and will remain in mode 3 (Hot Standby) pending the results of the investigation. The licensee will inform the NRC Resident Inspector. |
| ENS 49600 | 2 December 2013 23:26:00 | Palo Verde | Automatic Scram | NRC Region 4 | CE | 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 December 2, 2013, at approximately 1758 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 control room received reactor protection system alarms for low departure from nucleate boiling ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 2 was operating normally at 100% power. Plant operators entered the emergency operations procedures and diagnosed an uncomplicated reactor trip but noted the 1A reactor coolant pump (RCP) was not running. All CEAs fully inserted into the core. Following the reactor trip, indications on the train A logarithmic (log) power nuclear instrument initially responded normally but then did not trend as expected. All other nuclear instruments responded normally and the train A log power channel was declared inoperable and technical specification limiting conditions for operation 3.3.10 and 3.3.11 were entered. No emergency classification was required per the PVNGS Emergency Plan. The Unit 2 safety related electrical busses remained energized from normal offsite power during the event. Due to planned maintenance on one switchyard breaker, the Ruud offsite power line was disconnected from the PVNGS switchyard when the Unit 2 main generator output breakers opened. There was no impact to the required circuits between the offsite transmission network and the onsite Class 1E Electrical Power Distribution System; the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 2 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. Preliminary information indicates the reactor trip resulted from an electrical protection trip of the power supply circuit breaker for the 1A RCP. 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 Resident Inspector has been informed of the Unit 2 reactor trip. There was no impact on either Unit 1 or Unit 3. |
| ENS 49536 | 14 November 2013 14:57:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 49528 | 12 November 2013 01:23:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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 49525 | 9 November 2013 16:09:00 | Millstone | Automatic Scram | NRC Region 1 | CE | Millstone Unit 2 automatically tripped following a turbine trip due to a loss of condenser vacuum. The loss of vacuum was caused by the trip of the "C" circ water pump with the "D" circ water pump out of service. The licensee is still investigating the trip of the "C" circ water pump. The MSIVs are open with steam generators discharging steam to the main condenser. Auxiliary feedwater automatically started as expected following the reactor trip. All rods fully inserted and there were no complications following the reactor trip. All systems functioned as required and the unit is stable in Mode 3. There was no impact on Unit 3. The licensee has notified state and local authorities and the NRC Resident Inspector. |
| ENS 49082 | 31 May 2013 10:26:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 49054 | 21 May 2013 06:20:00 | Calvert Cliffs | Manual Scram | NRC Region 1 | CE | Reactor trip. All safety functions met with normal heat removal. 22 SGFP (Steam Generator Feedpump) exhibited high vibrations and signs of coupling damage. Further investigation will be performed. All control rods fully inserted on the trip. Steam Generator level is being maintained with the remaining feedpump. Decay heat is being dumped to the main condenser. Electrical power is in the normal shutdown lineup. No relief or safety valves lifted during the trip. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector. |
| ENS 48818 | 12 March 2013 18:01:00 | Saint Lucie | Automatic Scram | NRC Region 2 | CE | 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 48687 | 21 January 2013 19:55:00 | Waterford | Automatic Scram | NRC Region 4 | CE | At 15:51 CST, Waterford 3 experienced an uncomplicated automatic reactor trip from 84.5% reactor power. The actuations of the Reactor Protection System (RPS) and the Emergency Feedwater Actuation System (EFAS) resulted from Steam Generator #1 Low Level, which is at a nominal 27.4% narrow range setpoint. Safety systems responded as expected. All three (3) Emergency Feedwater Pumps started and injected into Steam Generator #1. Auxiliary Feedwater pump has been started, feeding both Steam Generators (#1 and #2) at levels above the EFAS low level setpoint. All control rods inserted by the automatic RPS actuation. Electrical power is being supplied from normal off-site power and condenser vacuum is available for Steam Generator heat removal via the Steam Dump Bypass Control system. There are no safety systems out of service or inoperable, nor any safety system TS (Technical Specification) LCO (Limiting Condition for Operation) actions entered. The cause of the Steam Generator #1 Low Level condition, and associated Reactor Trip, is under investigation. This event occurred during the initial power escalation from refuel outage RF18, after attempting to place C Heater Drain Pump (HDP), the first of three, into service. After starting, C HDP tripped for a reason not yet verified. Subsequently, based on initial Control Room operator observations, the Steam Generator #1 Main Feedwater control valve position was observed to be at 10-20% open, but with an open position demand signal of 100%. Main Feedwater response to the reactor trip (Reactor Trip Override) was as expected. The NRC Resident Inspector has been informed. * * * UPDATE FROM WILLIAM HARDIN TO PETE SNYDER AT 1645 EST ON 3/7/13 * * * The original reactor power level stated in the report should be 91% in lieu of 84.5%. This information has been changed in the event heading. The licensee notified the NRC Resident Inspector. Notified R4DO (Werner). |
| ENS 48169 | 8 August 2012 13:15:00 | Arkansas Nuclear | Automatic Scram | NRC Region 4 | CE | At 0823 hours on August 8, 2012, Arkansas Nuclear One, Unit 2 (ANO-2) experienced an automatic reactor trip. The reactor automatically tripped due to High Reactor Coolant System Pressurizer Pressure that was caused by a Main Turbine trip due to high condenser back pressure from a degraded vacuum condition. The Reactor Protection System (RPS) performed as designed in response to the High Reactor Coolant System Pressurizer Pressure condition resulting in automatic shutdown of the reactor from approximately 100 percent power. All Control Element Assemblies (CEAs) fully inserted on the trip. The Emergency Feedwater Actuation System (EFAS) actuated for the 'A' Steam Generator only due to level trending slightly below the setpoint. The plant has transitioned to supplying the steam generators using the Auxiliary Feedwater (AFW) system. The unit is currently in Mode 3 and implementing the transient response process. The investigation into the cause of the trip is ongoing and the local NRC Resident Inspectors have been notified. The unit is in a normal electrical lineup, and the decay heat is being removed by the main condenser via the turbine bypass valves. The State Department of Health was notified and ANO-2 will be issuing a press release. |
| ENS 47915 | 11 May 2012 07:07:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 47837 | 15 April 2012 17:13:00 | Palo Verde | Manual Scram | NRC Region 4 | CE | 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 April 15, 2012 at approximately 1220 Mountain Standard Time (MST), Palo Verde Unit 3 was manually tripped during low power physics testing. While conducting low power physics testing following a refueling outage, Regulating Group 1 rods were being inserted while simultaneously diluting to maintain a constant power level below the Point of Adding Heat. While inserting rods one rod deviated from its subgroup when it stopped moving. The Reactor Operator immediately ceased rod motion and the dilution was stopped. The residual positive reactivity in the core caused a corresponding reactor power increase that approached procedural power limits set forth in the low power physics testing procedure. Based on these indications, operators initiated a manual reactor trip. Following the reactor trip, all CEAs inserted fully into the core. All systems operated as expected and this event was diagnosed as an uncomplicated reactor trip. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 3 is stable and in Mode 3 feeding Steam Generators with Auxiliary Feedwater Pump 'N'. 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 Resident Inspector was informed of the Unit 3 reactor trip. The electrical lineup remained normal. Decay heat is being removed via the steam bypass to the main condenser. |
| ENS 47793 | 31 March 2012 03:42:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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 47752 | 19 March 2012 02:24:00 | Saint Lucie | Manual Scram | NRC Region 2 | CE | 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. |
| ENS 47628 | 31 January 2012 22:58:00 | San Onofre | Manual Scram | NRC Region 4 | CE | At 1505 PST, Unit 3 entered Abnormal Operation Instruction S023-13-14 'Reactor Coolant Leak' for a steam generator leak exceeding 5 gallons per day. At 1549 PST, the leak rate was determined to be 82 gallons per day. At 1610 PST, a leak rate greater than 75 gallons per day with an increasing rate of leakage exceeding 30 gallons per hour was established and entry into S023-13-28 'Rapid Power Reduction' was performed. At 1630 PST, commenced rapid power reduction per S023-13-28 'Rapid Power Reduction'. At 1731 PST, with reactor power at 35% the Unit was manually tripped. At 1738 PST, Unit 3 entered Emergency Operation Instruction S023-12-4 'Steam Generator Tube Rupture'. At 1800 PST the affected steam generator was isolated. All control rods fully inserted on the trip. Decay heat is being removed thru the main steam bypass valves into the main condenser. Main feedwater is maintaining steam generator level. No relief valves lifted during the manual trip. The plant is in normal shutdown electrical lineup. Unit 2 is presently in a refueling outage and was not affected by this event. The licensee has notified the NRC Resident Inspector. The licensee has issued a press release. |
| ENS 47523 | 14 December 2011 16:40:00 | Palisades | Manual Scram | NRC Region 3 | CE | The reactor was manually tripped at 1510 EST on 12/14/11 due to loss of both main feedpumps. Both feedpumps tripped on low suction pressure due to an apparent unplanned opening of the 'A' main feedpump recirculation valve. The cause of the main feedpump recirculation valve opening has not been determined. All full length control rods fully inserted. Auxiliary feed pump P-8A automatically started at 1511 EST on steam generator level as designed (10CFR50.72(b)(3)(iv)(A)). The turbine bypass valve is in service maintaining reactor coolant system temperature (by directing steam flow to the main condenser). The plant is stable in mode 3 (and the reactor trip was considered uncomplicated). The Van Buren County Sherriff was notified (per other plant requirements) concerning use of the atmospheric steam dump causing excessive noise in the vicinity of the plant (immediately following the plant trip). The plant electric power is in the normal shutdown configuration. There was no primary to secondary leakage. A press release is planned for the local media. The licensee notified the NRC Resident Inspector. |
| ENS 47472 | 22 November 2011 23:03:00 | Palo Verde | Manual Scram | NRC Region 4 | CE | 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. PVNGS (Palo Verde Nuclear Generating Station) (Unit 1 was in the process of Low Power Physics Testing following a refueling outage. The Unit was at 0.4 percent power and performing individual CEA (Control Element Assembly) group worth testing. Specifically Regulating Group (RG) 2 was being inserted while RG 4 was being withdrawn. During the CEA manipulations, it was identified that there was a CEA deviation on RG 2 subgroup 17 that exceeded 6 inches from the remainder of the RG. RG 2 subgroup 18 was at 134 inches withdrawn and RG 2 subgroup 17 had two CEAs at 124 inches, one CEA at 122 inches and one CEA at 118 inches withdrawn. The CEA Malfunctions abnormal operating procedure 40AO-9ZZ11 was entered and a manual reactor trip was directed by the Control Room Supervisor. The reactor was tripped at 1925 hours. Unit 1 was at normal temperature and pressure prior to the trip. All CEAs inserted fully into the reactor core. This was an uncomplicated reactor trip. No ESF actuations occurred and none were required. Safety related buses remained energized during and following the reactor trip. The offsite power grid is stable. No significant LCOs have been entered as a result of this event. There was no loss of normal heat removal capabilities, or loss of any safety functions associated with this event. No major equipment was inoperable prior to the event that contributed to the event. 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 Resident Inspector was informed of the Unit 1 reactor trip and this notification. |