|Start date||Reporting criterion||Title||Event description||System||LER|
|ENS 50269||10 July 2014 13:59:00||10 CFR 26.719||Non-Licensed Employee Supervisor Found in Violation of Fitness-For-Duty Policy||A non-licensed employee supervisor has been found in violation of the Duke Energy Fitness for Duty Policy. The individual's access to the plant has been suspended. The licensee has notified the NRC Region 1 (Hammann).|
|ENS 49086||2 June 2013 01:29:00||10 CFR 50.72(b)(3)(iv)(A), System Actuation||Valid Actuation of Emergency Diesel Generator||On June 1, 2013 at 2129 (EDT), a valid actuation of the 'B' Emergency Diesel Generator occurred when power was lost to the 'B' 4160V engineered safeguards bus due to failure of the 'B' Unit 6900 V bus feeder breaker 3104. There were no interruptions of spent fuel pool cooling during this event. This event is reportable as an 8-hour notification per 10 CFR 50.72(b)(3)(iv)(B)(8). This condition has no adverse affect on the public's or employees' health and safety. NRC Region II has been notified.||Emergency Diesel Generator|
|ENS 48716||5 February 2013 13:00:00||10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release||Offsite Notification Due to Media/Press Release||At 0800 Eastern Standard Time (EST) on 02/05/2013, Progress Energy Florida, Inc., a subsidiary of Duke Energy, announced its plan to permanently shutdown and decommission the Crystal River Unit 3 (CR3) Nuclear Plant. A media release was issued at 0800 EST on 02/05/2013. CR3 Security personnel are stationed at the Emergency Operations Facility/Training Facility located outside the Owner Controlled Area to monitor potential media coverage and public assembly. At this time, Security reports no security issues. No other media releases are planned at this time. This event is being reported under 10 CFR 50.72(b)(2)(xi) for offsite notifications based on the media release of the planned shutdown and decommissioning of CR3. The NRC Resident Inspector has been notified.|
|ENS 48556||4 December 2012 17:00:00||10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness||Temporary Emergency Operating Facility Established for Planned Outage|
In support of the planned upgrades to Crystal River Unit 3's Emergency Operations Facility's (EOF) heating, ventilation and air conditioning (HVAC) system, on Dec. 4, 2012, at 1200 hours Eastern Standard Time a temporary EOF has been established and declared operational. The temporary EOF is located adjacent to the primary EOF and remains outside the 10-mile Emergency Planning Zone. The temporary EOF meets the functional requirements of the primary EOF. During the establishment of the temporary EOF, there was no loss in the functionality of the EOF. If an emergency requiring EOF activation occurs, the temporary EOF will be staffed and activated using emergency planning procedures. The Emergency Response Organization has been briefed on the use of the temporary EOF. Readiness of the temporary EOF has been confirmed by a facility walkdown using existing procedures. This condition has no adverse affect on the public's or employees' health and safety. The EOF HVAC system is scheduled to be out of service for approximately four months. The NRC Resident Inspector has been notified.
Primary Emergency Operating Facility Outage completed. This is a courtesy notification and provides an update to the information provided in Event Notification Number 48556 on December 4, 2012, Eastern Standard Time (EDT). The primary Emergency Operations Facility (EOF) at the Crystal River Nuclear Plant has been restored on April 24, 2013, with the completion of the planned maintenance activity on the EOF heating, ventilation and air conditioning (HVAC) system that commenced on December 4, 2012, EDT. The temporary EOF established to support this planned upgrade, previously identified in Event Notification Number 48556, is no longer in use. The primary EOF is currently operational and experienced no loss in functionality during the restoration activities. The Emergency Response Organization has been briefed on the restoration of the primary EOF. This condition has no adverse affect on the public's or employees' health and safety. NRC Region II has been notified. The Licensee has also notified the NRC Resident Inspector.
|ENS 48391||9 October 2012 14:57:00||10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness||Loss of Assessment Capability at the Emergency Operations Facility|
At approximately 0100 hours EDT, on October 9, 2012, a low voltage circuit breaker in the Crystal River Unit 3 (CR-3) Emergency Operations Facility (EOF) tripped due to an electrical fault. The associated loads were powered by backup batteries until they depleted at approximately 1057 EDT. This condition resulted in the loss of dose assessment (RASCAL) and facility cooling capability. Dose assessment capability at the EOF was restored by 1335 EDT. During this event, RASCAL could be run from any plant computer outside of the EOF and the results conveyed back to the EOF. EOF ambient temperature is being monitored. An estimated completion time for chiller restoration has not been established. A follow-up notification will be provided when the chiller operation has been restored. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii). At no time was the public's or plant employees' health or safety at risk. Due to numerous backup systems, alternate methods were available to calculate dose projections, if required, and communicate the information to the EOF. The NRC Resident Inspector has been notified.
At 1500 EDT on 10/10/12 Chiller Operations for the EOF has been restored. At 1700 EDT on 10/10/12 the EOF returned to normal office temperatures. At no time did temperatures affect the habitability or functionality of the EOF. The NRC Resident Inspector has been informed. Notified R2DO (Seymour).
|ENS 48173||9 August 2012 22:00:00||10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release||Offsite Notification - Non Work Related On-Site Fatality||This is a non-emergency 4 hour informational notification to the NRC in accordance with the reporting requirements of 10 CFR 50.72(b)(2)(xi). On August 9, 2012 at approximately 1510 hours (EDT), a contract employee suffered a non-work related personal medical event while in an office environment that was located outside of the protected area. The individual was transported offsite and was pronounced deceased at a local hospital. OSHA is being notified pursuant to the requirements of 29 CFR 1904.39. There was no radioactive contamination involved in this event. Duke Energy has not observed any heightened media interest as a result of the fatality. No other notifications to government agencies are expected and no press releases are intended to be made at this time. The NRC Resident Inspector has been notified.|
|ENS 47166||17 August 2011 22:45:00||10 CFR 26.719||Fitness for Duty - Presence of Alcohol in the Protected Area||A non-licensed contract employee was determined to have alcohol in his possession while in the Protected Area. The contractor's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector.|
|ENS 47056||16 July 2011 09:00:00||10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness||Technical Support Center Non-Functional Due to Planned Maintenance|
At 0500 on Saturday, July 16th and 0700 on Saturday, July 23rd, a sequence of activities are planned that will render the Technical Support Center (TSC) non-functional by removing all normal and emergency power. These activities are being performed in support of planned TSC facility upgrades. In preparation for this power outage(s), the TSC emergency responders were notified that use of the alternate TSC will be required in accordance with station procedures. Each TSC emergency response function has performed a walkdown and verification that required functions can be established in the alternate location. The duration for each of these TSC power outages is expected to be less than 24 hours. The NRC Operations Center will be provided an update to this notification when power has been removed and restored during this time period. The licensee has notified the NRC Resident Inspector.
On 7/16/2011 at 0707 power was removed from the TSC. If necessary, the alternate TSC is ready for use.
On 7/16/11 at 1729 EDT, power was restored to the TSC. The TSC has been restored to a normal operational status and can be utilized for emergency response. The licensee has notified the NRC Resident Inspector. R2DO (Freeman) notified.
On 7/23/11 at 0645 power was removed from the TSC. If necessary, the alternate TSC is ready for use. The licensee has notified the NRC Resident Inspector. R2DO (Sykes) notified.
(At 1613), "power has been restored to the TSC. The normal TSC can be utilized for ERO response. There are no further planned TSC outages at this time. Notified the R2DO (Sykes).
|ENS 47026||5 July 2011 21:14:00||10 CFR 26.719||Supervisory Employee Tested Positive for Alcohol||A non-licensed employee supervisor had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's unescorted access has been terminated. Contact the NRC Headquarters Operations Officer for additional details. The plant is currently defueled and not in refueling mode. The licensee notified the NRC Resident Inspector.|
|ENS 46673||14 March 2011 18:11:00||Other Unspec Reqmnt||Additional Delaminated Containment Concrete Discovered||Crystal River Unit 3 is currently shutdown. Final tendon re-tensioning activities were being conducted following repair of delaminated (separated) concrete at the periphery of the containment wall. At approximately 1311 (EDT) on March 14, 2011, an indication was received from acoustic monitoring instrumentation located on the containment wall outside the previously repaired area. Re-tensioning activities were stopped and Plant Operations verified that there were no changes in plant parameters. Non-destructive examinations were initiated and preliminary indications are that there is delaminated concrete in the area identified by the acoustic monitoring. An assessment concluded that the containment continues to maintain the closure-pressure retaining capability required for Mode 5. There continues to be no threat to the public heath and safety. Delaminated concrete at the periphery of the containment wall was created in late 2009 during the process of creating an opening in the structure to remove and replace the steam generators inside (Reference Event Report 45416). The unit was already shut down for refueling and maintenance at the time the damage was found and has remained shut down. The extent of newly discovered delaminated has not yet been determined. The licensee plans to make a press release. The licensee has notified the NRC Resident Inspector and the Florida Public Service Commission.||Steam Generator|
|ENS 46437||24 November 2010 20:50:00||10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release||Offsite Notification Due to Release of an Oily Mixture to Discharge Canal||During maintenance activities at Crystal River Unit 3, approximately 25 to 30 gallons of an oily water mixture was pumped from a manway/vault. A small amount of this mixture was inadvertently discharged to a storm drain that leads to the Crystal River Unit 3 outfall (discharge canal). There was a slight sheen observed at the outfall as approximately 1/2 cup of the mixture entered the outfall area. This is reportable under 40CFR110 due to a light sheen observed on the site discharge canal. Crystal River Environmental Specialists responded to the spill with a combination of sand bags, floating sorbent boom, as well as sorbent pads. After deployment of the sand bags, floating sorbent booms, and securing the pumping activities, the spill status was confirmed as ceased and contained. Notifications will be made to the Florida Department of Environmental Protection State Warning Point and the National Response Center. This is 4-hour reportable per 10 CFR 50.72(b)(2)(xi) as a situation related to protection of environment, for which government agencies will be notified. The substance discharged appears to be some sort of motor oil. Prior to pumping, a sample of the liquid was analyzed and did not contain any radioactive nuclides. The licensee has notified the NRC Resident Inspector.|
|ENS 45703||16 February 2010 15:30:00||10 CFR 26.719||Fitness for Duty Event - Confirmed Positive Ffd Test||A licensed, non-supervisory employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's unescorted access to the facility has been suspended. Contact the Headquarters Operations Officer for additional details.|
|ENS 45350||14 September 2009 20:15:00||10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release||Fish Kill Reported to the Florida Fish and Wildlife Conservation Commission||On September 14, 2009, Progress Energy Environmental Services determined that approximately 200 deceased or stressed Ladyfish (Elops saurus) were observed near the banks of the Crystal River Energy Complex (CREC) discharge canal. Only Ladyfish were found stressed or deceased. Three of the stressed fish were sampled and examined by the CREC Mariculture Center to evaluate potential causes of these fish mortalities. Of the three fish examined, the only item of note was that the stomach and intestinal tracks were empty. It is likely that the runoff from recent heavy precipitation, coupled with summertime warm water temperatures and rapid salinity changes, resulted in a localized area of low dissolved oxygen and physiologic stress that affected this population of Ladyfish. There were no unusual conditions or factors associated with CREC operations that would have contributed to this event. Notification to the Florida Fish and Wildlife Conservation Commission will be made. This is reportable per 10CFR50.72(b)(2)(xi) as an event for which notification to other government agencies will be made. The licensee notified the NRC Resident Inspector.|
|ENS 45317||31 August 2009 18:45:00||10 CFR 21.21||Part 21 Report - Failure of a Spare Limitorque Actuator Motor||On July 31, 2009, (Crystal River Unit 3) personnel performed a video probe inspection of a spare safety-related Limitorque SB-3/SMB-3 actuator motor magnesium rotor and end rings. Visual indications were observed in the outboard end of the motor which were cause for rejection based on specified acceptance criteria. The specific indication was a cracked weld with dislodged metal in the outboard end of the motor. In 2007, the spare safety-related Limitorque SB-3/SMB-3 actuator motor (Motor Serial Number 7497004-001T1AL: Limitorque Part No. R-403-F04-0821) was purchased as safety-related Quality level 1 (QL-1) from AREVA under Purchase Order No. 337566. Limitorque purchased the motor from the Baldor Electric Company, doing business as the Reliance Electric Company, as commercial grade and dedicated the motor to safety related. The above evaluation was completed on August 28, 2009. The FPC director/responsible officer was notified of the above determination on August 31, 2009. The vendor (Limitorque) and the NRC Senior Resident Inspector have been notified of FPC's intent to report this issue under 10CFR21.21. The NRC Resident Inspector has been notified.|
|ENS 45286||24 August 2009 15:00:00||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation||Manual Trip Due to Loss of Power to Control Rods||Following completion of surveillance testing for the electrical checks of the Control Rod Drive power train, the operating crew observed that Group 7 regulating control rods unexpectedly lost power and inserted into the core. The Operators manually tripped the reactor prior to exceeding any RPS trip set point. There were no other safety system actuations and the plant is stable at normal post-trip temperature and pressure. There are seven rods in Group 7. Operators tripped the unit within 7 seconds. All rods fully inserted on the trip. The plant is removing decay heat through the main condenser and feeding generators with auxiliary feed. The licensee notified the NRC Resident Inspector.|
|ENS 44807||27 January 2009 15:17:00||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation||Manual Reactor Trip Due to Loss of 4160V Bus||Calibrations of the 'A' Unit 4160V Switchgear metering were in progress when the 'A' Unit 4160V Bus tripped. This resulted in the loss of the 'A' Feedwater Booster Pump (FWBP) and 'A' Condensate Pump (CDP). The Operating crew identified the loss of the 'A' FWBP with increasing RCS pressure and manually tripped the reactor. There were no other safety system actuations and the plant is stable at normal post-trip temperature and pressure. All rods inserted during the trip. Decay heat is being removed via steam dumps to the condenser. The electrical grid is stable with plant loads being supplied by offsite power via the startup transformer. Both vital busses are being powered from offsite. During the transient, main steam relief valves did lift but have been reseated. The NRC Resident Inspector has been notified.||Feedwater||05000302/LER-2009-001|
|ENS 44438||24 August 2008 19:57:00||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation||Manual Reactor Trip Due to Feedwater Flow Oscillations||The 'A' Condensate Pump became uncoupled, lowering Condensate flow. Operators began to manually lower reactor power to maintain deaerator level. Reactor power was lowered to approximately 62 percent. At this power, (feedwater) flow oscillations began and were excessive. With these flow oscillations increasing the decision was made to manually trip the Reactor. The Reactor was manually tripped at 1557 hours. There were no safety system actuations other than RPS (Manual). The plant is stable in a normal post trip configuration. All control rods inserted into the core during the reactor trip. Offsite power is available and powering safety loads. The steam generator safeties lifted during the transient and reseated. There is no known primary to secondary leakage. Decay heat is being removed via steam dumps to the condenser using normal feedwater to the steam generator. The emergency feedwater system was not initiated during the reactor trip. The licensee notified the NRC Resident Inspector.||Steam Generator|
|ENS 44037||6 March 2008 02:00:00||10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded||Circumferential Piping Crack|
On March 5, 2008, a volumetric ultrasonic examination of a Reactor Coolant to Decay Heat System pipe weld was analyzed and determined to identify an unacceptable indication. This dissimilar metal weld is located on the nozzle to the Decay Heat System drop line piping, which is the common suction line from the Reactor Coolant System and is a 12 inch outer diameter pipe. The indication is circumferential, is 15 inches in length, and reaches a maximum localized depth of 65 percent through-wall in one location. The weld was previously partially inspected during the November 2007 refueling outage using manual ultrasonic examination and no indications were identified. The current inspection was done using newly qualified phased array ultrasonic examination techniques in response to industry operating experience regarding dissimilar metal weld flaws. The indication has been found unacceptable per paragraph IWB 3514.4 of the 1989 Addenda of ASME Section XI without further fracture mechanics analysis and is therefore considered reportable. Preparations for a weld overlay repair and further confirmatory manual ultrasonic testing examinations are in progress. The licensee notified the NRC Resident Inspector.
On March 8, 2008, further fracture mechanics evaluation of the circumferential indication on the Decay Heat System Drop Line determined that the requirements of the ASME Section Xl pipe code were maintained. Specifically, the acceptance criteria of the 1989 Edition Section XI, Table IWB-3641-1, -2, were met with an allowable value for flaw depth / wall thickness (a/t) of 0.75. Consequently, the 65% through wall indication would be considered acceptable for operation and the Degraded Condition Reporting Criteria would not be exceeded. Therefore, this event is not reportable under any 10CFR50.72 criterion. However, due to industry operating experience with dissimilar metals welds, this notification is being made voluntarily. Confirmatory manual ultrasonic testing examinations were completed which validated the presence of the indication originally found via phased array UT examination techniques. A full structural weld overlay repair is in progress which will be completed before returning Crystal River 3 to power operation. The repair effort has already successfully deposited the first weld layer over the location of the flaw. No Licensee Event Report will be submitted for this event. The licensee notified the NRC Resident Inspector. Notified R2DO(Hopper).
|Reactor Coolant System|
|ENS 43706||10 October 2007 17:12:00||Information Only|
Other Unspec Reqmnt
10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release
|Pressurizer Porv Opened Momentarily with Rcs Discharge of Approximately 10 Gallons||On 10/10/2007 at 1312, Crystal River Unit 3 was performing maintenance on its Pressurizer Pilot Operated Relief Valve (PORV). The I&C technicians were replacing a circuit board card in the circuitry for the PORV. The PORV had been selected to close during the replacement of the card. When the PORV was selected to its auto position, the PORV opened and was immediately closed by the reactor operator. It was estimated that the PORV was opened for 2 second with a total flow of 10 gallons. This rate is in excess of the 25 gpm as found in the EAL. The event was quickly terminated with the closing of the PORV and PORV block valve. The plant remains stable with the PORV and PORV block valve closed. CR3 briefly met the conditions of an Unusual Event, however declaration was not required. The licensee is still trouble-shooting the cause of the event. The reactor is stable at full power and the technical specification action statement requirement for an inoperable PORV valve (TS 3.4.10) have been satisfied. The licensee reported this event for information based on the guidance of NUREG 1022 concerning reporting an event that was rapidly terminated that may have met the criteria of an emergency based on an after the fact review. The licensee notified the State and local authorities and the NRC Resident Inspector.|
|ENS 43475||7 July 2007 15:03:00||10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness||Loss of Safety Parameter Display System (Spds) Due to Planned Modification|
At 11:03 hours EST, on July 7, 2007, the Plant Information Computer System (PICS) was removed from service to perform a planned modification to install and test a Standard Digital I&C System Platform (DICSP) infrastructure and to update the network configuration of PICS. The expected duration of PICS inoperability is approximately 24 to 36 hours. PICS provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Plant Process Computer System (PPCS), Emergency Response Computer Logger (ERCO-1), Meteorological Data link system and the Inadequate Core Cooling Monitor (ICCM). The loss of PICS requires alternate methods, as described in plant procedures, to be used for the above described functions. Therefore, appropriate assessment of plant conditions, notifications and communications can still be made, if required, during the time that PICS is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii) which is any event that results in a major loss of emergency assessment capability, offsite response capability or offsite communications capability. As previously stated, alternate means remain available to assess plant conditions, make notifications and accomplish required communications, as necessary. An additional notification will be provided when PICS operability is restored. The NRC Resident Inspector has been notified.
Engineering change (EC) of PICS continues. Testing of the installed modifications is now in progress. As stated above, alternate means remain available to assess plant conditions, make notifications, and accomplish required communications, as necessary. Testing of the EC and EC turnover is expected to be completed on 07/09/2007. The plant computers, including SPDS, are presently running providing plant data thus are available but not operable. The licensee notified the NRC Resident Inspector. Notified R2DO(Henson)
Crystal River Unit 3 initially reported removal of its Plant computer systems from service for planned modification on 07/07/07 at 11:51 and made the required notification in accordance with 10 CFR 50.72(b)(3)(xiii). The Engineering change associated with the Safety Parameter Display System (SPDS) and the Plant Information Computer System (PICS) has been completed and the systems are currently running providing data. Modification testing has been completed and the systems are considered operable. The compensatory measures put in place to assure appropriate assessment of plant conditions, notifications and communications have been terminated. Licensee also stated that the ERDS system is now operable. The licensee notified the NRC Resident Inspector. Notified R2DO (Hopper)
|Emergency Response Data System|
Safety Parameter Display System
|ENS 43179||22 February 2007 00:15:00||10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation|
10 CFR 50.72(b)(3)(iv)(A), System Actuation
|Automatic Reactor Trip on High Rcs Pressure Following a Feedwater Transient||At 1915 on February 21, 2007, Crystal River Unit 3 was at reduced power (71%) for planned maintenance on one of four condenser waterboxes when the Integrated Control System (ICS) for Main Feedwater became erratic causing a feedwater transient that underfed the once thru steam generators. The reactor protection system (RPS) actuated on high Reactor Coolant System pressure causing a reactor trip. CR3 also received an Emergency Feedwater Actuation (EFIC) on low steam generator levels. This event is reportable as a 4-Hour Non-Emergency Notification per 10CFR50.72 (b)(2)(iv)(B) for Reactor Protection System Actuation and as an 8-Hour Non-Emergency Notification per 10CFR50.72 (b)(3)(iv)(A) for Emergency Feedwater Actuation and for Reactor Protection System actuation. All systems responded as designed and the plant remains stable with Emergency Feedwater in MODE 3. The licensee informed the NRC Resident Inspector.||Steam Generator|
Reactor Coolant System
Reactor Protection System
|ENS 43116||23 January 2007 20:15:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition||Crystal River 3 has determined that an existing condition in the Decay Heat (DH) System is reportable under 10 CFR 50.72(b)(3)(ii)(B), an Unanalyzed Condition. Two valves in question are in series in the DH System drop leg from the Reactor Coolant System (RCS). They are Appendix R High-Low pressure interface valves, where 3 phase hot shorts from another power cable must be considered. A review of cable routing concluded that one of the valve's power cables and one of the valve's control power cables do not meet required separation criteria. There is a low probability that both could open due to hot shorts under normal operating RCS pressure and temperature conditions, therefore a rupture of the downstream piping in the Reactor Building or Auxiliary Building is possible. The postulated event is a Loss of Coolant Accident (LOCA) in the Auxiliary Building, which is an unanalyzed condition. This condition is similar to the one reported in EN# 43098 on 01/12/2007. Cables in question are for the same valves, however, they are in a different area of the plant. The licensee will notify the NRC Resident Inspector.||Reactor Coolant System||05000302/LER-2007-001|
|ENS 43098||12 January 2007 21:31:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Unanalyzed Condition of Decay Heat System||Crystal River Unit 3 has determined that an existing condition in the Decay Heat (DH) System is reportable under 10 CFR.50.72 (b)(3)(ii)(B), an Unanalyzed Condition. Two valves in question are in series in the Decay Heat (DH) System drop leg from the Reactor Coolant System (RCS). These valves are High-Low pressure interface valves, where three phase hot shorts from another power cable must be considered under Appendix R. A review of cable routing concluded that the two valve's power cables do not meet required Appendix R separation or protection criteria. As such, there is a low probability that both valves could open due to hot shorts under normal operating RCS pressure and temperature conditions, resulting in a rupture of the downstream piping in the Reactor Building or Auxiliary Building. This event could result in a Loss of Coolant Accident (LOCA) in the Auxiliary Building, which is an unanalyzed condition. The licensee will notify the NRC Resident Inspector.||Reactor Coolant System|
|ENS 42632||12 June 2006 14:56:00||10 CFR 50.72(a)(1)(i), Emergency Class Declaration||Unusual Event Due to a Posting of Hurricane Warning at the Site|
Crystal River 3 is in a Hurricane Warning area. The site is evaluating the need for a plant shutdown at this time. The declaration of NOUE is per the licensee's Emergency Action Guidelines due to posting of the hurricane warning. There are currently no significant equipment problems or LCOs that might be impacted by the impending weather conditions. The licensee will provide updates as appropriate if weather conditions change or a decision to shutdown is made. The licensee notified the NRC Resident Inspector, State, and Local authorities.
* * * UPDATE FROM M. WOLF TO M. ABRAMOVITZ AT 1530 ON 06/13/06 * * *
The hurricane warning necessitating the entrance by the site into the Unusual Event was exited by the state this morning at 11 am. The site waited until high tide passed and exited the Unusual Event at 1511. The licensee notified the NRC Resident Inspector. Notified R2DO (Ayers), NRR EO (Jung), IRD (Blount), DHS (S. York) , and FEMA (E. Casto).
|ENS 42122||6 November 2005 14:35:00||10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness||Temporary Loss of Offsite Notification Sirens||At 0935 on November 6, 2005, all offsite notification sirens for Crystal River Unit 3 were lost. One siren failed and sent a feedback signal which in turn disabled all remaining sirens. The failed siren was bypassed at 1124, thereby restoring the remaining 39 sirens to service. The failed siren will be repaired promptly. During the loss of siren function, backup means of notifying the public were available, including route alerting using local law enforcement and use of the Code Red autodialing system. This is reportable as an immediate notification (eight-hour report) in accordance with 10CFR50.72(b)(3)(xiii). The siren system is maintained by the county. Compensatory measures for the remaining area is either: dispatching a police car, or fire truck, or using the Code Red autodialing system for the affected area. The licensee notified the NRC Resident Inspector.|
|ENS 41440||24 February 2005 20:30:00||10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release||Offsite Notification Related to Heart Attack Fatality||At 1423 the control room received an emergency phone call of an employee who was experiencing intermittent unconsciousness and in cardiac arrest. The individual was transported to a local hospital where he was declared deceased at 1530. 29CFR1904.39(b)(5) requires to report a fatality to OSHA caused by a heart attack at work. This ENS report is being made in accordance with 50.72(b)(2)(xi) contact of off site agency. There was no radioactive contamination involved in this event. The licensee does not plan any media or press release and has not notified any other government agencies besides OSHA. The NRC Resident Inspector will be notified.|
|ENS 41362||27 January 2005 23:30:00||10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition||Single Failure Identified That Could Prevent Re-Energizing Both Es Busses|
On January 27, 2005, Crystal River - Unit 3 (CR-3) discovered an installation subject to a single failure that could prevent both Emergency Diesel Generators (EDGS) and both offsite power sources from supplying power to their respective Engineered Safeguards (ES) Busses. This is a condition contrary to 10CFR50.72 (b)(3)(ii)(B). The installation involves the 4.16 Kv supply breakers from the offsite power transformer (OPT) and the Back-up Engineered Safeguards Transformer (BEST) which are the two required offsite power sources. From the OPT, one breaker supplies the 'A' ES Bus and a second breaker supplies the 'B' ES Bus. From the BEST, one breaker supplies the 'A' ES Bus and a second breaker supplies the 'B' ES Bus. A traditional (non-nuclear) design for such pairs of supply breakers includes current transformers that are connected to each other for each phase and a watt hour meter from those connections to monitor power. These circuits were installed around 1990 to support the coordination of the new 230 Kv breakers to supply the OPT and the coordination with the 230 Kv supply breakers for the new BEST. Should a failure of the wire connecting the current transformers or the watthour meter occur such that the ES Bus supply breakers' lockout relays are actuated, all breakers supplying and receiving power via the ES Busses would be opened and locked out. The result is that neither EDG nor the offsite power source would be able to automatically supply power to its respective ES Bus. CR-3 is currently disconnecting and isolating the watthour meters and removing the connection between the respective breakers' current transformers. Licensee entered Tech Spec 188.8.131.52 at 1830 hours based on single failure but considers both ES busses operable. The NRC Resident Inspector was notified of this event by the licensee.
On January 28, 2005, Crystal River - Unit 3 (CR-3) corrected the single failure vulnerability associated with the Engineered Safeguards (ES) Busses. A permanent modification which removed the single failure vulnerability for the offsite power transformer (OPT) was completed at (0152), and the modification for the Back-up Engineered Safeguards Transformer (BEST) was completed at (0644). R2DO (T. Decker) notified. The NRC Resident Inspector was notified of this event by the licensee.
|Emergency Diesel Generator||05000237/LER-2005-001|
|ENS 41074||26 September 2004 09:05:00||10 CFR 50.72(a)(1)(i), Emergency Class Declaration||Entered Unusual Event Classification Due to Hurricane Warning in Effect|
Crystal River Unit 3 site has declared an Unusual Event due to the declaration of a hurricane warning for Citrus County including the Crystal River Unit 3 site. Current wind speed on-site is 20 mph with 40 mph gusts. The licensee's current plan is to remain in power operation. The licensee notified both state and local agencies and the NRC Resident Inspector.
Hurricane warning has been discontinued. Unusual Event exited at 0345 ET. The licensee notified the NRC Resident Inspector. Notified R2 Response Manager (S. Cahill), R2 DO (R. Haag), NRR EO (S. Richards), IRD Manager (S. Frant), FEMA (J. Canupp), DHS (Sr Watch Officer)
|ENS 41027||8 September 2004 18:45:00||10 CFR 50.72(b)(3)(iv)(A), System Actuation||Emergency Feedwater System Actuation|
At 1445 on September 8, 2004, while in Mode 3 after an unplanned reactor trip during tropical storm Frances, Crystal River Unit 3 experienced a re-actuation of the B-Train of the Emergency Feedwater System. The Emergency Feedwater System had previously actuated during a Loss of Offsite Power experienced during Hurricane Frances and the B-Train actuation had procedurally been bypassed as part of the plant recovery. While In the process of restoring a Main Feedwater Pump to an operating condition, the signal bypassing the B-Train Emergency Feedwater actuation was momentarily removed due to unusual plant conditions present during recovery from the loss of power This resulted in the re-actuation of the B-Train of Emergency Feedwater. This condition is 8-Hour reportable per 10CFR50.72(b)(3)(iv)(B)(6). The re-actuation of the B-Train of Emergency Feedwater did not have a significant impact on plant operation and the associated equipment has been returned to a standby condition. The licensee informed the NRC Resident Inspector.
The following information was provided by the licensee via facsimile: At 1823 (hrs. EDT) on 9/8/04, Crystal River Unit 3 made an 8-hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A) and 50.72(b)(3)(iv)(B)(6) concerning re-actuation of the B-train of the Emergency Feedwater System. The Emergency Feedwater System had previously actuated during a Loss of Offsite Power experienced during Hurricane Frances and the B-Train actuation had procedurally been bypassed as part of the plant recovery. While in the process of restoring a Main Feedwater Pump to an operating condition, the signal bypassing the B-Train Emergency Feedwater actuation was momentarily removed due to unusual plant conditions present during recovery from the loss of power. This allowed for a re-actuation of the B-Train of Emergency Feedwater based on a loss of Main Feedwater Pump signal. However, the two Main Feedwater Pumps had not been operating since the loss of offsite power that occurred on 9/6/04. The A-Train Emergency Feedwater System had been operating continuously since the initial actuation on the loss of offsite power and continued to operate during and after the re-actuation of the B-Train. Re-actuation of the B-Train has been determined to be an invalid actuation in accordance with the guidance of NUREG-1022, and therefore EN 41027 is retracted. The licensee has informed the NRC resident inspector. R2DO (Lesser) notified.
|ENS 41081||8 September 2004 18:45:00||10 CFR 50.73(a)(1), Submit an LER||Invalid Emergency Feedwater Actuation||The following information was provided by the licensee via facsimile: At 1445 (hrs. EDT) on September 8, 2004 Crystal River Unit 3 experienced an invalid actuation of the B-Train of the Emergency Feedwater System. This occurrence is being reported in accordance with 10CFR50.73(a)(2)(iv)(A) and 50.73(a)(2)(iv)(B)(6) using the optional process described in 50.73(a)(1). A complete actuation of the B-Train of Emergency Feedwater occurred and the train started and functioned successfully. The A-Train of Emergency Feedwater was operating at the time, and it continued to operate successfully during and after the B-Train actuation. The licensee has notified the NRC resident inspector. See retraction of Event #41027 for related information.||Feedwater|
|ENS 41024||8 September 2004 05:00:00||10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded||Potential Reactor Coolant System Pressure Boundary Leak Located|
At 0100 on September 8, 2004, while in Mode 3 conducting a Reactor Building walkdown after an unplanned reactor trip during Tropical Storm Frances, Crystal River Unit 3 identified a potential Reactor Coolant System pressure boundary leak located on a weld associated with a pressurizer level sensing line (upstream of reactor coolant isolation valve RCV-75). The cause of the leak is not known at this time, and the leak did not have a significant effect on plant operation. The last unidentified leak rates that were completed prior to the plant shutdown were 0.10 - 0.13 gpm. This defect in the primary coolant system is unacceptable per ASME Section XI. The condition is 8-Hour reportable per 10CFR 50.72 (b)(3)(ii)(A). The plant will be cooled down to Mode 5 and additional inspections and necessary repairs implemented. Dry boron crystal deposit are on the sensing line upstream of RCV-75 . See similar event reported by Crystal River Unit 3 on 10/04/03 (event # 40222) The NRC Resident Inspector was notified of this event by the licensee.
* * * RETRACTED ON 9/9/04 AT 0424 EDT FROM LARRY MOFFATT TO GERRY WAIG * * *
At 0311 on 9/8/04, Crystal River Unit 3 made an 8-Hour ENS notification in accordance with 10CFR50.72 (b)(3)(ii)(A) concerning a potential Reactor Coolant System pressure boundary leak (Event Notification # 41024). The leak was presumed to exist based on the presence of white, crystalline material deposited on the pipe. The material has been sampled and does NOT contain boron. The presumed leak site has been cleaned and visually inspected with no evidence of a defect in the weld joint or pipe. These additional inspections have demonstrated that the material deposited on the pipe does NOT constitute pressure boundary leakage, and therefore EN 41024 is retracted. The licensee has notified the NRC resident inspector.
|Reactor Coolant System|
|ENS 41022||5 September 2004 21:10:00||10 CFR 50.72(a)(1)(i), Emergency Class Declaration||Unusual Event Declared Due to Hurricane Warning|
Crystal River declared a Notification of Unusual Event due to hurricane warning in effect for the plant and surrounding areas. All safety systems are operable. The licensee notified the NRC Resident Inspector and the State emergency response organization.
The licensee terminated the Notification of Unusual Event at 1017 today, 9/7/04. The basis of the termination is that the hurricane warning was lifted yesterday 9/6/04 and offsite power has been restored. Notified R2 (Pribish), NRR (Reis), IRD (Wessman), DHS (Akers), FEMA (Kuzia).
|ENS 40945||13 August 2004 03:00:00||10 CFR 50.72(a)(1)(i), Emergency Class Declaration||Unusual Event Declared Due to Hurricane Warning for Hurricane Charley|
Licensee declared an Unusual Event due to the notification of a Hurricane Warning for the site area from the National Weather Service. The unit will be required to shutdown if Category 3 hurricane winds are experienced on site. Presently Hurricane Charley is exhibiting Category 2 winds (105 MPH). The licensee has notified the NRC Resident Inspector, State and Local Emergency management organizations.
The Agency entering Monitoring Mode at 0800 on 8/13/04 and Region 2 is in the lead. Notified IRD management and the following federal agencies: DHS, FEMA, DOE, EPA (NRC), USDA and HHS.
The agency exited the Monitoring Mode at 2127 EDT on 8/13/04. Notified the following federal agencies: DHS (Akers), FEMA (Sullivan), DOE (Pauley), EPA (NRC) (Layman), USDA (Mahar), and HHS (Hogan). Additionally, the following NRC staff were notified: H. Nieh, R. Orchard, S. Frant, P. Wilson, M. Weber, C. Jackson, R. Hogan, W. Outlaw, S. Gagner, L. Smith (R4DO).
The licensee reported that hurricane Charley has cleared their service territory and that Crystal River 3 exited the Notice of Unusual Event (NOUE) on 8/13/04 at 2141 EDT. Notified the following federal agencies: DHS, FEMA. Additionally, the following NRC staff were notified: H. Nieh, R. Borchardt, S. Frant, P. Wilson, M. Weber, C. Jackson, R. Hogan, W. Outlaw, S. Gagner, L. Smith (R4DO).
|ENS 40299||5 November 2003 22:34:00||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 High Reactor Coolant System Pressure||Crystal River Unit 3 was at 37 percent power starting up from a refueling outage. During troubleshooting of control problems with the B Main Feedwater Pump, a feedwater transient occurred which underfed the steam generators. This resulted in a reactor trip on high Reactor Coolant System pressure and Emergency Feedwater Actuation on low steam generator levels. This event is reportable as a 4-Hour Non-Emergency Notification per 10CFR50.72 (b)(2)(iv)(B) for Reactor Protection System Actuation and as an 8-Hour Non Emergency Notification per 10CFR50.72 (b)(3)(iv)(A) for Emergency Feedwater Actuation. The licensee reported that all control rods fully inserted on the trip and that steam generator safety valves lifted and reseated as expected. The primary system is currently at 2155 psi, 549 degrees F with steam generator feedwater being supplied by auxiliary feedwater. The main condenser is available and is being used for primary system heat removal via steam dump. The station electrical grid is stable and in normal configuration; the emergency diesel generators are operable and in standby. The licensee has notified the NRC Resident Inspector.||Steam Generator|
Reactor Coolant System
Reactor Protection System
Emergency Diesel Generator
|ENS 40222||4 October 2003 14:32:00||10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded||Degraded Condition Due to Rcs Pressure Boundary Leakage|
At 1032 (EDT) on October 4, 2003 while in mode 4 conducting a planned refueling outage reactor coolant system inspection, Crystal River Unit 3 identified two potential RCS Pressure Boundary Leaks located in Pressurizer penetrations associated with upper level sensing lines. Upon closer inspection of the first one inch penetration, indications confirmed a Pressure Boundary Leak. A closer inspection of the second penetration is planned after obtaining access to the location. The last unidentified leak rate that was completed prior to the plant shutdown was 0.15 gpm. This defect in the primary coolant system is not acceptable per ASME XI. This condition is REPORTABLE per 10CFR 50.72 (b)(3)(ii)(A). Licensee reported that a small amount of boric acid residue and residual rust was identified during inspection in the vicinity of the upper level Pressurizer tap near the Pressurizer shell and the nozzle. A closer inspection is planned of the other two upper Pressurizer level taps.
The NRC resident has been notified by the Licensee.
|Reactor Coolant System||05000302/LER-2003-003|