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| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 54091 | 26 May 2019 10:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a Reactor Coolant Pump Trip on Ground Fault | This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Plant Protection System (PPS) actuation. Arkansas Nuclear One, Unit 2, automatically tripped from 100 percent power at 0512 CDT. The reactor automatically tripped due to 2P-32B Reactor Coolant Pump tripping as a result of grounding. No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. The EFW actuation meets the 8-hour Non-Emergency Immediate Notification Criteria of 10 CFR 50.72(b)(3)(iv)(A). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident Inspector has been notified. Decay heat is being removed via the steam dump valves to the main condenser. Unit 2 is in a normal shutdown electrical lineup. Unit 1 was not affected by the transient on Unit 2. The licensee notified the State of Arkansas. |
| ENS 52571 | 23 February 2017 15:40:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Tornado Missile Vulnerabilities | In order to address the concerns outlined in RIS (Regulatory Information Summary) 2015-06 'TORNADO MISSILE PROTECTION', an evaluation of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment was conducted. This evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The BVPS Unit 1 (BV-1) and BVPS Unit 2 (BV-2) Main Steam Safety Valve (MSSVs) discharge flow paths to atmosphere (reference TS 3.7.1) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the MSSVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all MSSV's could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-1 and BV-2 Atmospheric Steam Dumps (ADVs) discharge flow paths to atmosphere (reference TS 3.7.4) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the ADVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all ADVs could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-2 Auxiliary Building tornado missile shield door (A-35-5A), credited for tornado missile protection of the Primary Component Cooling Water (PCCW) system, was found to not be fully closed and latched. A tornado could generate missiles capable of striking the PCCW system with the missile door open rendering both trains of the PCCW system inoperable. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The BV-2 Spent Fuel Building tornado missile shield door (F-66-3), credited for tornado missile protection of the irradiated fuel assemblies in the Spent Fuel Pool including the Cask Pit, was found to not be fully closed and latched. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The potential tornado missile vulnerabilities for the MSSVs and ADVs (discussed above) are being addressed in accordance with EGM-15-002 Rev 1 and DSS-ISG-2016-01 (NRC enforcement discretion and interim guidance documents). Immediate compensatory measures were taken to reduce the likelihood and mitigate the potential consequences of an onsite tornado generated missiles. The NRC Resident inspector has been notified. |
| ENS 52072 | 7 July 2016 16:10:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Accident Mitigation - Main Steam Safety Valves Not Adequately Protected from Tornado Missiles | During an evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Harris Nuclear Plant personnel identified conditions in the plant design such that specific TS equipment is considered not adequately protected from tornado missiles. A tornado could generate multiple missiles capable of striking exhaust piping on multiple main steam safety valves (MSSVs), resulting in crimping of the piping that could impact flow capacity and render the MSSVs inoperable. If the tornado caused a loss of offsite power, the MSSVs are credited to remove decay heat to achieve cold shutdown. Compensatory measures have been implemented to ensure safety in the event of a tornado. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 and Interim Staff Guidance DSS-ISG-2016-01 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector. |
| ENS 51607 | 15 December 2015 11:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Feed to One Steam Generator | This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Reactor Protection System (RPS) actuation. Arkansas Nuclear One, Unit 1, was manually tripped from 43 percent power at 0544 CST. The reactor was manually tripped due to operator judgement during control issues with the Integrated Control System (ICS) during a planned downpower for Electro-Hydraulic Control (EHC) system maintenance. CV-2672 B, low load control valve, failed to close. Subsequently, CV-2674 B, low load block valve, began to close and caused a loss of feed to E-24B Steam Generator. No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. This (EFW actuation) meets the 8 hour Non-Emergency Immediate Notification Criteria ((10CFR50.72(b)(3)(iv)(A)). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident (Inspector) has been notified. Decay heat is being removed via the steam dump valves to the main condenser. Unit 1 is in a normal shutdown electrical lineup. Unit 2 was not affected by the transient on Unit 1. The licensee notified the State of Arkansas. |
| ENS 50835 | 19 February 2015 22:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Normally Closed Watertight Doors | During Main Steam Safety Valve testing conducted prior to refueling outages, normally closed watertight doors are opened in support of the testing. If a postulated one square foot non-mechanistic crack were to occur within the Break Exclusion Area during the test, safety related equipment located just outside of these doors could be adversely affected. With these watertight doors open, compliance with the Comanche Peak licensing basis may not be assured. This condition has been conservatively determined to be reportable as an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Currently, the watertight doors on both Units 1 and 2 are closed, therefore, all safety related equipment is currently operable. Comanche Peak Engineering is performing a review of the original Comanche Peak licensing basis regarding the non-mechanistic crack event to determine what equipment impacts are required to be assessed. The NRC Resident Inspector has been notified. |
| ENS 50002 | 5 April 2014 14:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Loss of Enclosure Building | In preparation for a scheduled outage, maintenance personnel removed the upper and lower boots of the main steam safety valves. Upon discovery, Operations personnel declared the Enclosure Building inoperable. Maintenance re-installed the boots and the integrity of the Enclosure Building was restored and the building returned to service. The licensee notified the NRC Resident Inspector, the State of Connecticut, and the town of Waterford. |
| ENS 49708 | 10 January 2014 03:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 2234 hours EST on 01/09/2014, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. At the time of the event, Steam Generator Water level Protection Channel Testing was in progress. While testing was in progress with the 'C' Steam Generator Channel 1 Water Level Protection channel in trip for testing, a Turbine Trip occurred. The cause of the Turbine Trip is under investigation. The (Turbine Driven and Motor Driven) Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. Currently, the unit is stable in Hot Standby (Mode 3). This event is being reported as a 4 hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation. This event is being reported as an 8 hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of Auxiliary Feedwater System. At no time during this occurrence was the public or plant staff at risk as a result of this event. The NRC Resident Inspector has been notified. State and local authorities will be notified. Estimated restart date is 1/12/2014 |
| 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 47874 | 30 April 2012 14:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Unusual Event Due to a Potential Fire in Containment | Performing an I&C functional (test) caused an inadvertent Safety Injection signal resulting in a reactor trip/safety injection. All safety systems responded as designed for a safety injection. Electrical systems are aligned to normal offsite power sources. All fire alarms have been validated by the Fire Protection Department as invalid alarms and confirmed that no fire event in the protected area. The reactor trip was successful and all rods (fully) inserted. Decay heat removal is via auxiliary feedwater through the atmospheric (steam) dumps. Unknown at this time is the cause of the inadvertent safety injection signal. No injuries occurred as a result of this event. The licensee believes that the trip/safety injection may have caused piping to shake resulting in dust near the fire detection equipment resulting in the invalid fire indication. The instrument being tested was the high steam flow channel-1 bistable for PT505. The maximum pressurizer level during this event was 95%. The licensee notified the NRC Resident Inspector.
At 1003 hours on April 30, 2012, Salem Unit 1 experienced a reactor trip and safety injection (SI) signal due to a high steam flow coincident with a low steam pressure signal. At the time of the safety injection signal, function testing of the 1PT505 turbine inlet pressure channel was in progress. This testing required the tripping of the high steam flow bistables. As a result of the reactor trip and safety injection signal, the Emergency Diesel Generators started but did not load, the ECCS system (high head safety injection pumps actuated and injected into the reactor vessel, intermediate head safety injection pumps and low head (RHR) safety injection pumps) actuated. All 4 main steam isolation valves closed along with feedwater isolation and start of the auxiliary feedwater pumps. All control rods fully inserted following the reactor trip. Following the main steam line isolation, the atmospheric relief valves opened along with the lifting of several main steam safety valves. The unit is currently in Mode 3 and will be cooling down to Mode 4. Train A SSPS (Solid State Protection System) is currently out of service and suspected of causing the safety injection signal. Train B SSPS has not been reset due to the standing safety injection signal. With Train A SSPS inoperable and Train B SSPS not reset, TS 3.0.3 was entered and a shutdown required by TS 3.0.3 was commenced at 1345 hours. This report is being made in accordance with 10CFR50.72(b )(2)(iv)(B), 50.72(b)(3)(iv)(A), 50.72(b)(2)(i) and 50.72(b)(2)(iv)(A). The licensee exited the Unusual Event at 1249 EDT. The licensee notified the NRC Resident Inspector. Notified the R1DO (Conte). The NRC Operations Center notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via e-mail). |
| ENS 47781 | 28 March 2012 19:03:00 | 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 Steam Generator High Water Level Condition | At 1503 hours EDT on March 28, 2012, with the unit in Mode 1 at 55% power, an automatic reactor trip occurred. The reactor trip was the result of a turbine trip from a 'B' Steam Generator Hi Level. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feeedwater (AFW) System automatically actuated due to both main feedwater pump breakers opening from a valid feedwater isolation signal. Steam Generator Levels were then controlled by Auxiliary Feedwater pumps. Steam Generator Blowdown was automatically isolated with the AFW actuation. The RCS Code Safety valves, Pressurizer Power Operated Relief Valves (PORVs), Steam Generator PORVs or the Main Steam Safety valves (MSSVs) did not open during the event. All control rods fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently in Mode 3 and stable. There were no radiological consequences or releases as a result of this event. The cause of the Steam Generator Hi Level is under investigation. The Resident NRC Inspector has been informed. |
| ENS 47293 | 26 September 2011 15:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to One Loop Low Flow Signal | At 1145 hours EDT on September 26, 2011, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The steam generator and pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The cause of the reactor trip is under investigation. |
| ENS 46313 | 7 October 2010 04:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a One Loop Reactor Coolant Low Flow | At 0013 hours EDT on October 7, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. Reactor Coolant Pump (RCP) 'C' tripped during the event. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. Steam generator and pressurizer Power Operated Relief Valves (PORVs) and Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The Main Turbine Lube Oil Deluge System actuated without a fire during the event. A fire hose station pipe ruptured after the deluge system actuation. The fire hose station was isolated. In addition, the 'B' Main Feedwater (MFW) Pump tripped during the event and the 'A' Main Feedwater pump subsequently tripped due to high steam generator level in the 'C' Steam Generator at about 0023 hours. There is currently indication of RCP 'C' Number 2 seal leakage of approximately 2.5 gpm. At approximately 0405 EDT the Auxiliary Feedwater (AFW) system actuated due to a trip of MFW Pump 'A' while attempting to start the pump in accordance with procedure GP-004, Post Trip Stabilization. The AFW system actuation signal caused motor-driven AFW Pump 'B' to start. The motor-driven AFW Pump 'A' was already in operation due to the post-trip condition. The cause of the MFW Pump 'A' trip is under investigation. This AFW system actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector. |
| ENS 46238 | 9 September 2010 18:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Overtemperature Delta-T Signal | At 1437 hours EDT on September 9, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the Overtemperature Delta-T reactor protection function. During the event, the steam generator power operated relief valves (PORVs) and one pressurizer PORV briefly opened and re-closed, in response to pressure changes in the steam generators and pressurizer due to the plant transient condition. The Auxiliary Feedwater System automatically actuated, as expected, and provided feedwater to the steam generators. The main steam safety valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. There was an indication of an approximate 0.65 gpm leak to the pressurizer relief tank following the reactor trip. The isolation valve to the pressurizer PORV that opened during the reactor trip was closed and the leak indication stopped. The indicated leakage was within Technical Specification leakage rate limits. The cause of the reactor trip and indication of pressurizer PORV leakage is under investigation. The licensee notified the NRC Resident Inspector. |
| ENS 45809 | 2 April 2010 04:00:00 | 10 CFR 21.21, Notification of failure to comply or existence of a defect and its evaluation | Possible Failure of Solenoid Valves with Dual Function Main Steam Safety Valves | The following information was received via facsimile: Tyco Valves & Controls LP (TVC), a business unit of Tyco Flow Control, is a contractor to certain NRC licensees, and on behalf of itself and certain of its affiliates, hereby submits information in response to notification requirements of 10 CFR Part 21. TVC manufactures, sells, and services Pressure Relief Valves (PRVs) to NRC licensees for use in NRC licensed facilities. On February 4, 2010, EPRI issued a notification of a potential 10 CFR Part 21 defect for Main Steam Safety Valves manufactured by Crosby Valve & Gage (Crosby). Crosby became a brand of TVC when they were purchased by them in 1998. The reported problem stems from an issue with the solenoid valve that controls the pneumatic actuator on certain main steam safety valves that were supplied during the early 1980's. Following is a list of U.S. plants that may have been affected by this issue: Utility Unit Start Up PA Power & Light Susquehanna 1 & 2 1983 & 1985 Exelon LaSalle I & 2 1984 Energy NW Columbia 1984 Entergy Riverbend 1986 TVC continues to investigate the reported problem, but is requesting a 30 day extension to complete the evaluation due to the complexity of the situation. |
| ENS 44276 | 7 June 2008 12:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following the Trip of a Condensate Pump | On 6/7/08 at 0818 hours, an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 100% power due to a trip of the 2B Condensate Pump, which led to a trip of the 2B Main Feedwater Pump (MFP) and decreasing Steam Generator (S/G) levels. The reactor was manually tripped due to decreasing S/G levels. Following the reactor trip, EOP-1, Standard Post Trip Actions and EOP-2, Reactor Trip Recovery procedures were completed and Unit 2 was stabilized in Mode 3. All control rods fully inserted. The Main Steam Safety Valves lifted as expected. Feedwater to the S/Gs was initially supplied by the 2A (MFP) until Auxiliary Feedwater Actuation System (AFAS) actuated as expected on low S/G level. Subsequently, the Auxiliary Feedwater Pumps restored S/G levels. Unit 2 electrical requirements were provided from offsite power. Other than the trip of the 2B Condensate Pump (initiating event) there were no major equipment failures. Unit 1 was not affected by this event. The grid is stable. Decay heat is being removed by the Auxiliary Feedwater Pumps feeding the S/Gs steaming to the bypass valves in the Condenser. The licensee notified the NRC Resident Inspector. |
| ENS 43199 | 28 February 2007 11:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Malfunction of Pressure Transmitter | At approximately 0633 AM on February 28, 2007, control room operators manually initiated a reactor trip after observing decreasing steam generator levels as a result of low feed water flow. Control room alarms indicated low main feed water pump suction pressure which was subsequently, attributed to a malfunction of the main feed water pump low suction pressure cutback pressure transmitter (PT-408B) on the common main feed pump water supply header. The malfunction of the main feed pump water low suction cutback pressure transmitter resulted in a cutback of both main feed water pumps reducing main feed water flow to the steam generators and causing decreasing steam generator levels. All control rods fully inserted and all safety systems responded as expected. The auxiliary feedwater system actuated as expected from low steam generator levels which occurs as a result of a full power reactor trip. The emergency diesel generators (EDGs) did not start as offsite power remained available. The plant was stabilized in hot shutdown with the (Auxiliary Feedwater System) providing decay heat removal via the main condenser. The main feed water pumps are shutoff and secured. The event is under investigation and a post trip review is being conducted. A courtesy call to stakeholders will be made. All EDGs remain available in standby. No steam generator Power Operated Relief Valves or Main Steam Safety Valves lifted. The licensee notified state and will notify local authorities. The licensee plans to issue a press release. The licensee notified the NRC Resident Inspector. |
| ENS 42957 | 2 November 2006 18:34:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Turbine Trip/ Reactor Trip Due to Invalid Low Condenser Vacuum Signal | At 1334 on 11-2-06 an Automatic Reactor Trip occurred from 100% power. All systems functioned as required. One safety valve stuck open on both OTSGs. They subsequently re-seated. An employee working on the roof at the time of the trip fell off a ladder and injured his leg. Emergency medical was contacted to assist with the injured worker. Two fire trucks and an ambulance was dispatched to the site to remove the injured worker. The worker was not contaminated. There is no indication of any OTSG tube leaks. Initial investigation indicates the reactor tripped, due to a turbine trip due to an invalid low vacuum signal. State and local officials will be notified of this event by the licensee. I&C Techs were performing maintenance on one of the low pressure vacuum switches. An electrical fault fed to the other two low pressure vacuum switches causing a 2 out of 3 signal which resulted in a turbine trip followed by a reactor trip signal, as expected. All rods fully inserted into the core. One safety valve (9 safety valves on each OTSG) on each Once Through Steam Generator stuck open. OTSG "B" safety relief valve was open less than one minute. There are no leaking OTSG tubes. A condensate relief valve located in the turbine building opened/shut - nobody injured. The ICS (Integrated Control System) operated as expected. All emergency core cooling systems and the emergency diesel generators are fully operable plus the electrical grid is stable. A licensee working on the industrial coolers on top of the industrial building, standing on a ladder, fell off the ladder when OTSG relief valve opened. Licensee either broke or badly sprained his leg. The NRC Resident Inspector was informed of this event by the licensee.
Post trip evaluation determined that the Main Steam safety valves were not stuck open. The safety valves were operating within their tolerance band. The "B" OTSG Main Steam safety valve reseated with no operator action as steam pressure decreased. The "A" Main Steam safety valve was reseated when operators lowered OTSG pressure in accordance with Plant Operating Procedures. TMI-1 issued a press release on this event at 15:13 on 11/2/06." R1DO (John White) notified. The NRC Resident Inspector was notified of this update by the licensee. |
| ENS 42828 | 6 September 2006 06:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip on Loss of Condenser Vacuum | Observed degrading condenser pressure. Entered abnormal procedure DB-OP-02518, High Condenser Pressure and reduced reactor power. At <280 Mwe and > 5 inches Hg (mercury) A (absolute) , manually tripped the reactor at approximately 45% power in accordance with procedure. Normal post-trip response. Condenser pressure is slowly recovering. Still trying to determine the source of the condenser air in-leakage. Notified Ottawa County Sheriff of main steam safety / atmospheric vent valve operation at 0231 hours per procedure. All control rods fully inserted on the trip. Decay heat is being removed using the turbine bypass valves and the motor driven feed pump. There is no steam generator tube leakage. The atmospheric vent valves / main steam safety valves lifted for a few seconds following the trip and fully reseated after the initial lifting. Plant electrical power if from the grid backfeeding to the station. The electric grid is stable. The NRC Resident Inspector was notified of this event by the licensee. |
| ENS 42260 | 15 January 2006 17:04:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Ts Shutdown Due to Electromatic Relief Valves (Ervs) Declared Inoperable | At 1014 hours on January 15, 2006, the four Unit 1 Electromatic Relief Valves (ERVs) were declared inoperable due to concerns related to the ERV actuators. These operability concerns originated from inspections and testing performed on the Unit 2 ERVs which indicate evidence of mechanical interference issues that could prevent proper ERV operation (note that Unit 2 was shutdown on January 14, 2006 to perform ERV actuator inspections). In accordance with Technical Specification 3.4.3, Condition B, 3.5.1 Condition H, and 3.6.1.6 Condition B, Unit 1 commenced a shutdown on January 15, 2006 at 1104 hours. This notification is being made in accordance with 10 CFR 50.72(b)(2)(i) as commencement of any nuclear plant shutdown required by the plant's Technical Specifications. During the upcoming Unit 1 outage, actuator inspections and ERV testing is planned to confirm the operational ability of the ERVs. Note that the Unit 1 Target Rock Relief Valve and the Main Steam Safety Valves are operable and available to support reactor overpressure protection. The plant is conducting a normal shutdown per Tech Spec requirements. At 16:00 CST the unit will going off line and by 24:00 CST the plant is expected to be in Mode 4 (Cold Shutdown). The licensee notified the NRC Resident Inspector. |
| ENS 40816 | 14 June 2004 14:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Due to Reactor Trip Due to Loss of Off Site Power | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 3 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated at 12:07 MST. Unit 3 received an automatic Main Steam Isolation System ESF actuation. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The unit was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with the unit started as expected in response to the loss of offsite power to their safety buses. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. The Unit 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 NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) Note: see related events 40815 , 40814 and 40818 |
| ENS 40815 | 14 June 2004 14:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to Reactor Trip Due to Loss of Off Site Power | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 1 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated at 12:07 MST. Unit 1 manually initiated a Main Steam Isolation System ESF actuation by procedure. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The unit was at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with the unit started as expected in response to the loss of offsite power to their safety buses. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. The Unit 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 NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) Note: see related events # 40814, 40816 and 40818 |
| ENS 40814 | 14 June 2004 14:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Alert Declared - Reactor Trip Due to Loss of Off Site Power | On June 14, 2004, at approximately 07:44 Mountain Standard Time (MST) all three units at the Palo Verde Nuclear Generating Station experienced automatic reactor trips coincident with a grid disturbance and loss of offsite power in the Palo Verde Switchyard. Unit 2 declared an ALERT Emergency Plan classification at approximately 07:54 due to a loss of AC power to essential buses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in a station blackout. Subsequently, at 09:51 Unit 2 downgraded the Emergency Plan classification to a NOTIFICATION OF UNUSUAL EVENT when AC power was restored from a single essential bus to both essential buses. Units 1 and 3 declared a NOTIFICATION OF UNUSUAL EVENT at 07:53 MST due to a loss of offsite power to essential buses for greater than 15 minutes. The NOTIFICATION OF UNUSUAL EVENT was terminated for all 3 units at 12:07 MST. Unit 1 and 2 manually initiated a Main Steam Isolation System ESF actuation by procedure. Unit 3 received an automatic Main Steam Isolation System ESF actuation. Due to the loss of offsite power, the Emergency Plan Technical Support Center (TSC) was unavailable. The Unit 2 Satellite TSC was to be staffed by the Emergency Response Organization in response to the loss of assessment capability. Power to the TSC has since been restored. The Emergency Plan ALERT declaration includes staffing of the Joint Emergency New Center to address expected media interest. All three units were at normal operating temperature and pressure prior to the trip. All CEAs inserted fully into the reactor cores. All Emergency Diesel Generators (EDGs) (2 per unit) associated with each of the 3 units started as expected in response to the loss of offsite power to their safety buses. Unit 2's train "A" EDG started, but did not indicate volts or amps and was manually shutdown. The offsite power grid had several perturbations for approximately one hour following the event but has been stable since. LCO 3.8.1, AC Sources - Operating, was entered in each unit as a result of this event. Heat removal is to atmosphere via atmospheric dump valves in natural circulation. Main steam safety valves may have lifted for a brief time. Restoration of forced reactor coolant circulation is pending assurance that the offsite power grid can reliably support the load. No major equipment was inoperable prior to the event that contributed to the event. All 3 units are 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 NRC Resident Inspector was notified. FBI Jeff Muller and Mr. Rosales (Mexican National Commission of Nuclear Safety and Safeguards (CNSNS)) were notified.
Notified Reg 4 RDO (Graves), NRR (Bateman), DHS (Lee), FEMA (Canupp), DOE (Sal Moroni), EPA (Stalcup), EPA (Crews), HSS (Davidson), and Mexico (Rosales) NOTE: See events 40815, 40816 and 40818 |