|Report date||Site||Event description|
|05000302/LER-2008-003||10 December 2008||Crystal River||At 15:53 on August 24, 2008, Progress Energy. Florida, Inc. (PEF), Crystal River Unit 3 (CR-3) was operating in MODE 1 (POWER OPERATION) at approximately 60 percent RATED THERMAL POWER, when the reactor was manually tripped. At 15:37, the Condensate Pump CDP-1A magnetic-coupling became uncoupled. Delays in rapid power reduction led to a low Deaerator (FWHE-1) level and eventual cavitation of the Main Feedwater (FW) System booster pumps and Main FW pumps. The cavitation caused a loss of FW flow control to the Once-Through Steam Generators. This loss of flow control led to the decision to manually trip the reactor. The root cause for this event was original plant design failing to provide adequate alarms to the operating crews to promptly identify Condensate Pump failures. Required equipment operated as designed during the manual reactor trip. Loss of FW is an analyzed event bounded by the CR-3 Final Safety Analysis Report accident analysis. This condition does not represent a reduction in the public health and safety. Operating crews were trained on this event. Compensatory measures to alert the operating crew of a loss of a Condensate Pump were established. Design options are being evaluated to alert the operating crew of Condensate Pump failures. A previous similar occurrence has not occurred or been reported to the NRC.|
|05000302/LER-2007-001||9 March 2007||Crystal River||At 19:01, on January 11, and at 13:16, on January 23; 2007, Progress Energy Florida, Inc., Crystal River Unit 3 (CR-3) was operating in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER when conditions not meeting 10CFR50, Appendix R, Section III.G.2 cable separation criteria were identified for high/low pressure interface valves. During performance of the CR-3 Safe Shutdown Analysis Revalidation Project Fire Area Assessment for the Intermediate and Reactor Buildings, respectively, reviews revealed that the power cable for Decay Heat Removal (DH) System valve DHV-3 and the power and control cables for DHV-4 were routed with other energized cables. Low probability three-phase external cable hot shorts of the proper voltage due to a hypothetical fire could cause spurious opening of both valves, resulting in an unanalyzed loss of coolant condition in the Auxiliary Building. The cause for this event was a misunderstanding of 10CFR50, Appendix R cable separation criteria in 1997 and 2002 pertaining to high/low pressure interfaces resulting in missed opportunities to correct the identified conditions. Appropriate compensatory measures have been put into place. This report is being submitted pursuant to 10CFR50.73(a)(2)(ii)(B). This condition does not represent a reduction in the public health and safety. Previous similar occurrences have been reported to the NRC.|
|05000302/LER-2005-005||11 January 2006||Crystal River||At 05:34, on November 14, 2005, Progress Energy Florida, Inc., Crystal River Unit 3 was operating in MODE 6 (REFUELING) at zero (0) percent RATED THERMAL POWER when an inadvertent Engineered Safeguards actuation occurred while restoring power following a planned electrical bus outage. The actuation caused the automatic start of the B Train Emergency Diesel Generator, Emergency Nuclear Service Seawater Pump, Emergency Nuclear Services Closed Cycle Cooling Pump and Decay Heat Closed Cycle Cooling Fan. All remaining equipment that would normally be actuated from this signal was removed from service due to the current refueling outage conditions and was unaffected. The cause for this event was inadequate guidance in the shutdown procedures for removal of the Engineered Safeguards System from service. The guidance only required the system to be placed in "bypass" which did not disable the system actuation in all conditions. Successful recovery from the actuation was achieved. This report is being submitted pursuant to 10CFR50.73(a)(2)(iv)(A). The equipment that was available operated as designed. There were no equipment failures or damage as a result of the inadvertent actuation. This condition does not represent a reduction in the public health and safety. No previous similar occurrences have been reported.|
|05000302/LER-2005-001||23 March 2005||Crystal River||At 18:30, on January 27, 2005, Progress Energy Florida, Inc., Crystal River Unit 3, was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. A non emergency eight-hour notification was made to the NRC Operations Center under 10CFR50.72(b)(3)(ii)(B) to report a design configuration subject to a single failure that could prevent both onsite and both offsite power sources from supplying power to their respective 4160 volt Engineered Safeguards buses. This condition was identified by NRC inspection personnel during the NRC Triennial Fire Protection Inspection. No failure modes effects analysis was performed during the design change process in effect at the time the Offsite Power Transformer and Back-up Engineered Safeguards Transformer were installed in 1990 and 1993, respectively. Also, inadequate technical rigor was exercised during the design, verification, and acceptance of the modification packages developed by the Architect Engineer. Modifications to remove the single failure vulnerability have been implemented. This report is being submitted pursuant to 10CFR50.73(a)(2)(ii)(B) and 10CFR50.73(a)(2)(ix)(A). This condition does not represent a reduction in the public health and safety. No previous similar occurrences have been reported.|
|05000302/LER-2004-004||22 November 2004||Crystal River|
At 14:00, on September 24, 2004, Progress Energy Florida, Inc., Crystal River Unit 3, was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. The NRC published a Notice of Clarification to Steam Generator Tube Integrity Event Reporting Guidelines in NUREG-1022, "Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73." The Notice of Clarification revised the guidelines used to determine reporting of steam generator inspection results. The NRC directed licensees to consider the results of the previous steam generator tube inspections, including the structural integrity criteria and leakage criteria, against the new reporting guidelines. Crystal River Unit 3 determined that the as-found steam generator projected leakage value for Steam Line Break (SLB), which exceeded the leak rate limit for Refueling Outage 13 (13R), is reportable per the revised NUREG-1022 guidance, under 10 CFR 50.73(a)(2)(ii)(A), as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The cause for exceeding the leakage criterion was finding new indications not accounted for in the previous cycle Operational Assessment, and human error in the leakage calculation. Repairs performed during 13R reduced the SLB projected leakage to well below the limit prior to plant startup. This condition does not represent a reduction in the public health and safety. No similar occurrences have been previously reported.
NRC ) PRINTED ON RECYCLED PAPER
|05000302/LER-2004-003||29 October 2004||Crystal River|
At 11:51, on September 6, 2004, Progress Energy Florida, Inc., Crystal River Unit 3, was in MODE 1 (POWER OPERATION) at 97 percent RATED THERMAL POWER. During Tropical Storm Frances, phase-to-ground faults occurred concurrently on a 230 kilovolt transmission line and a 230 kilovolt switchyard south bus breaker. Loss of power to the Startup transformer resulted in opening the reactor trip breakers, tripping of the reactor coolant pumps and main feedwater pumps, and loss of power to the Main Turbine Lube Oil System. A main turbine trip and subsequent actuation of the Reactor Protection System occurred. The Emergency Feedwater Initiation and Control System actuated to feed the Once Through Steam Generators.
The transmission line fault was caused by mechanical failure of a carbon steel pin in a vertical string of insulators due to high wind conditions. The breaker fault was caused by flashover due to contamination from wind and salt spray. The insulator string was replaced. The breaker was tested, inspected, washed down and returned to service. Actuation of the Reactor Protection System and Emergency Feedwater System are reportable to the NRC. This report is being submitted pursuant to 10CFR50.73(a)(2)(iv)(A). This condition does not represent a reduction in the public health and safety. No previous similar occurrences have been reported.
NRC FORM 366 (6-2004) PRINTED ON RECYCLED PAPER