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| Site | Start date | Title | Description | |
|---|---|---|---|---|
| 05000296/LER-2012-001 | 26 January 2012 | Annunciator Panel Power Supply Fire in Unit 3 Control Room | On January 26, 2012, at 1908 hours Central Standard Time (CST) Browns Ferry Nuclear Plant (BFN), Unit 3, was in Mode 1 at 100 percent power. Operations personnel in the BFN, Unit 3, Control Room, smelled smoke and observed a flame coming from the bottom of a power supply located in an annunciator panel. Operations personnel opened a breaker, which resulted in the fire being extinguished at 1918 CST. The loss of some annunciator alarms and indications resulted when the breaker was opened. Operations personnel were able to monitor compensatory indications. There was no loss of assessment capability. The event did not result in any entry to a required emergency action level, such as declaring a Notification of Unusual Event (NOUE). The cause of the event was a failed power supply. An overcurrent was caused by an aged capacitor that had not received preventative maintenance to address its service life. A corrective action was initiated to replace the affected power supplies. Previous similar events addressed the failure of an annunciator module in March, 2008; the failure of a power supply in May 2008; and a failed power supply in an annunciator panel in July 2009. | |
| 05000366/LER-2009-004 | Docket Number | 23 June 2009 | Turbine Trip On High Reactor Water Level Due To Failed Circuit Board Results in Reactor Scram | On June 23, 2009 at 03:51 EDT, Unit 2 was in mode 1 with an approximate reactor power of 1710 CMWTh. At this time an automatic reactor scram occurred as a result of a turbine trip due to high reactor water level. Prior to the event the reactor was increasing in power during startup from a recent outage. An instrument which controls reactor water level failed resulting in an increase in reactor water level. The water level increased to a point where the turbine control valve fast closure trip signal was initiated due to high reactor water level. All control rods fully inserted and Reactor Feed pumps tripped. Reactor water level initially decreased to approximately negative 25 inches, due to void collapse. The Primary Containment isolation Valve Group 2 isolation setpoint was reached, and the Group 2 valves isolated. Both the 'A' and 'B' Reactor Feed Pumps initially tripped, and the 'A' feed pump was restarted and used to restore and maintain reactor water level. Reactor pressure reached an upper value of approximately 964 psig. No Safety Relief Valves opened, nor were they required to open, based on the maximum pressure reached. Reactor water level increased to a maximum value of approximately 60 inches above instrument zero, but was restored to normal range. The cause of this event was the failure of an internal power supply electrolytic capacitor which caused a failure of the DC power supply for the Yokogawa level controller 2C32-K648. The failed power supply card containing the capacitor was replaced following the event and repetitive tasks have been createclace this and similar power supply cards at a prescribed interval. |
| 05000302/LER-2008-001 | 23 January 2008 | Software Change Causes Inoperability of Redundant Core Subcooling Monitors for Longer Than TS Allowable | On January 23, 2008, while operating in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER, at Progress Energy Florida, Inc. Crystal River Unit 3 (CR-3), during the performance of Surveillance Procedure SP-144C, "Core Exit Thermocouple Calibration," the Core Subcooling Monitors would indicate out of range above a simulated Core Exit Thermocouple (CET) input temperature of 1250 degrees Fahrenheit (F). As part of the extent of condition investigation, SP-144C was performed for the Channel B recorder and the same failure mechanism was identified. At 1717 on January 25, 2008, both Core Subcooling Monitors were determined to be inoperable. These instruments are OPERABLE when accurately indicating between 0 to 2500 degrees F. Improved Technical Specification (ITS) 3.3.17, "Post Accident Monitoring (PAM) Instrumentation," Condition C, allows 7 days for restoration if one or more functions with two required channels inoperable. The condition was determined to have existed since the new Safety Parameter Display System (SPDS) multiplexer server Modbus driver was installed on August 13, 2007. This exceeded the ITS completion time and as such, this report is being submitted under 10 CFR 50.73(a)(2)(i)(B). The cause for this condition was that the responsible engineer did not identify the impact of the software change to the system and require full range, post modification testing for this function following implementation of the Engineering Change. The condition was corrected on January 26, 2008 through a software change. This condition does not represent a reduction in the public health and safety. A previous similar occurrence has not been reported to the NRC. | |
| 05000313/LER-2004-001 | Docket Number | 29 September 2004 | or NRC Form 366A LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NUMBER (Include Area Code) Fred Van Buskirk, Nuclear Safety and Licensing Specialist 479-858-3155 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) REPORTABLE ,U ' REPORTABLECAUSE SYSTEM COMPONENT MANU- CAUSE SYSTEM COMPONENT MANU- TO EPIX , ,. ,UU
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| 05000321/LER-2004-004 | Edwin I. Hatch Nuclear Plant - Unit 1 | 9 March 2004 | Momentary False Low Reactor Water Level Signal Results in Actuations of Safety Systems | On 03/09/2004 at 0340 EST, Unit 1 was in the Cold Shutdown mode with reactor coolant temperature at approximately 189 degrees Fahrenheit to perform the vessel leakage test. Personnel had reduced vessel pressure from 869 psig to 129 psig in order to place the shutdown cooling mode of the Residual Heat Removal (RHR) system into service to facilitate repairs to a reactor vessel head vent line flange discovered leaking during the test. At 0340 EST, as Operations personnel were opening shutdown cooling suction valve 1E11F009, several water level transmitters generated a false low level signal resulting in the actuations of multiple safety systems, including Emergency Diesel Generators, Core Spray pumps, RHR pumps, the Main Control Room Environmental Control System, and all four trains of the Standby Gas Treatment systems. Operators determined that the indications of low level were false and terminated injection by the lA Core Spray pump within 36 seconds and the 1B Core Spray pump within 12 seconds. Personnel reset the actuation signals and returned the affected systems and components to normal by 0453 EST. This event was caused by personnel error and procedure omissions. Personnel had incorrectly assembled the head vent line flange. The location of the resulting leak was such that the vessel became water-solid during the leakage test. Opening valve 1E11F009 with the vessel water-solid caused a pressure transient that led to a false indication of low water level. The leakage test procedure did not contain instructions for lowering pressure and placing shutdown cooling into service with the vessel water-solid. The head vent line was assembled correctly and the leakage test was completed satisfactorily. The Unit 1 and Unit 2 leakage test procedures will be revised to provide information to help personnel determine whether the vessel is or may be water-solid and to include instructions to recover from such a condition. |
| 05000263/LER-2001-005 | Docket Number | 19 February 2001 | On February 19, 2001 with the reactor operating at 100% power, the Monticello Nuclear Generating Plant (MNGP) staff determined that there was reasonable doubt that plant operators could manually establish torus cooling following a DBA LOCA within the 10 minute design assumption. The containment cooling system was declared inoperable. To restore operability, a dedicated operator was stationed in the control room with the sole purpose of initiating torus cooling during a LOCA. Following an unrelated reactor shutdown on February 25, 2001, a solution team was assembled to study torus cooling issues and recommend corrective actions. The team determined that the plant procedures used to initiate torus cooling were not streamlined for emergency conditions and were not written with the purpose of satisfying the 10 minute design assumption. On March 15, 2001 with the reactor shutdown, the solution team was evaluating operator actions to support torus cooling. It was determined by a calculation of post-accident drywell conditions that the potential for flashing of the fuel zone reactor water level reference leg during a DBA LOCA, although slight, was more probable than had been previously thought. This flashing could further delay operator actions to initiate torus cooling. Prior to startup on April 2, 2001, a modification to relocate nearly all of the fuel zone instrument reference leg piping from the drywell to the reactor building was completed. In addition, changes were made to the torus cooling procedures to directly support the 10 minute requirement and the operators were trained on these changes. | |
| 05000461/LER-1998-013, Responds to J Carter Telephone Request for GE to Review Clinton LER 98-013-00 | Clinton | 9 July 1998 | Responds to J Carter Telephone Request for GE to Review Clinton LER 98-013-00 |