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Report date | Site | Event description | |
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05000325/LER-2017-002 | 12 June 2017 | Brunswick | On April 17, 2017, at 0004 Eastern Daylight Time, Unit 1 was in Mode 1 at approximately 100 percent of rated power, and Unit 2 was in Mode 1 at approximately 22 percent of rated power and was starting up from a refueling outage. Operators manually tripped the Unit 2 main turbine to halt increasing bearing vibration. The power circuit breakers (PCBs) for the Unit 2 main generator did not open as expected on the turbine trip, but subsequently opened when main generator reverse power relays actuated. This resulted in the automatic start of all four emergency diesel generators (EDGs). The EDGs did not tie to emergency busses because offsite power was still available. This event resulted from a component which failed due to foreign material intrusion. A limit switch associated with a main turbine stop valve failed to change states when the turbine was tripped. The limit switch configuration, together with other logic, satisfied the conditions to start the EDGs. The limit switch failure resulted from foreign material in the switch. The failed limit switch was replaced. Planned corrective actions include inspecting similar switches on both units, and sealing the wire entrances of the switch bodies to improve foreign material exclusion. |
05000325/LER-2017-001 | 22 March 2017 | Brunswick | On February 7, 2016, at 1312 Eastern Standard Time (EST), Unit 1 was in Mode 1 (i.e., Run) at 88 percent of rated power in end-of-cycle coastdown. At that time, an event occurred which resulted in a loss of offsite power (LOOP) on Unit 1. Emergency diesel generators EDG-1 and EDG-2 started and tied to their respective Unit 1 emergency buses. During diesel operation, EDG-1 exhibited oscillations in engine speed and bus frequency. These oscillations had no adverse effect on equipment supplied by the bus, and all supplied loads continued to perform their safety functions without interruption and without need for operator intervention. However, due to the speed and frequency oscillations, EDG-1 was deemed after the fact to have been inoperable. Following extensive testing and evaluation, the cause of the oscillations has not been determined. The governor system for EDG-1 was replaced on March 6, 2016, as part of a planned upgrade to EDG-1. Since that time, the oscillations have not recurred. Based on the fact that the governor replacement eliminated the oscillations, it's concluded that the oscillations resulted from a deficiency in the governor system. Since the cause of the inoperability has been eliminated, no further corrective actions are planned. |
05000325/LER-2016-006 | 13 February 2017 | Brunswick | On December 13, 2016, Units 1 and 2 were in Mode 1 (i.e., Run mode) at 98 percent and 97 percent of rated thermal power, respectively. At that time, shift personnel were notified that structural supports for the 2D Control Room air conditioning system condenser were corroded. An operability assessment found the affected air conditioning system inoperable due to the effect on its seismic qualification. Technical Specifications (TS) 3.7.4, Condition A, was entered. On January 30, 2017, the 1D air conditioning system was found with similar conditions. Since the conditions were determined to have existed for longer than the TS allowable out of service time, the plant was in a condition prohibited by the TS. Since more than one air conditioner had been inoperable concurrently, the safety function of maintaining Control Room habitability could have been prevented from being fulfilled during a seismic event. This event resulted from trapped moisture in contact with the steel supports and exposure to the local marine environment which corroded support steel to the point that seismic qualifications were compromised. All affected support steel was replaced by February 2, 2017. The remaining air conditioner was inspected and found acceptable. Guidance for initiating work requests for corroded support members will be enhanced. |
05000325/LER-2016-005 | 5 December 2016 | Brunswick | On October 3, 2016, Units 1 and 2 were in Mode 1 (i.e., Run mode) at 100 percent of rated thermal power. At that time, Engineering personnel were reviewing the plant response to NRC Information Notice 97-45, Supplement 1, on signal cables for the Drywell High Range Radiation Monitors (DWHRRMs). These cables are susceptible to thermally induced current (TIC), which can degrade the accuracy of DWHRRMs. The review resulted in DWHRRMs being declared inoperable on both units. The DWHRRMs were included in Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.3.3.1 beginning in 1984. As a result of the TIC effect on the DWHRRM cables, BSEP Unit 1 and Unit 2 have operated longer than the TS allowed completion times for inoperable DWHRRMs. This event is reportable per 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the plant TSs. The event resulted from the inherent characteristics of the cables and DWHRRMs. An existing site procedure directs the use of alternate indications for assessment of drywell or fuel cladding conditions, and this procedure will remain in place as a compensatory action. Corrective actions will include replacing the cables on a schedule to be developed after assessment of available options. |
05000325/LER-2016-002 | 8 August 2016 | Brunswick | On March 4, 2016, at 1235 Eastern Standard Time (EST), Emergency Diesel Generator (EDG) 3 was declared inoperable. At this time, EDG 1, Emergency bus El, and balance of plant (BOP) bus 1D were inoperable due to planned maintenance. Two inoperable EDGs represents a loss of safety function, for the onsite standby power source. Therefore, this condition is being reported in accordance with 10 CFR 50.73(a)(2)(v)(D), as an event or condition that could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. In addition, it was determined that EDG 3 was inoperable for greater than Technical Specifications (TSs) Completion Times. Therefore, this condition is also being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as operation prohibited by the TSs. Additionally, on March 3, work was ongoing to restore power to BOP bus 1D when an error in the restoration sequence resulted in an invalid auto-start of EDGs 2 and 4. Because the invalid auto-starts of EDGs 2 and 4 are directly related to the events associated with this LER, it is being reported, herein, per 10 CFR
The root cause of the EDG 3 inoperability is a design vulnerability associated with relaxation of the EDG 3 fuse holder fingers which was not properly mitigated. The corrective action to prevent recurrence will be to implement a design change to address the vulnerability. |
05000325/LER-2014-004 | 16 May 2014 | Brunswick | On March 20, 2014, as a result of the transition process from 10 CFR 50, Appendix R, to NFPA 805, a review of the Brunswick Steam Electric Plant Safe Shutdown Analysis determined that a postulated fire in specific fire areas could disable critical components, potentially resulting in equipment required for safe shutdown being inoperable. The safety significance of this event is minimal. Deterministic analysis methods used to comply with Appendix R require every possible fire scenario to be addressed; however, the risk posed by these hypothetical events has been determined by analysis to be minimal. Corrective actions for this event include maintaining fire watches in affected areas and revising fire response procedures to mitigate the consequences of a potential fire in these areas in order to establish compliance with Appendix R and the current licensing basis. |
05000325/LER-2012-007 | 11 April 2013 | Brunswick | On December 14, 2012, at approximately 1306 hours Eastern Standard Time (EST), inoperability of both subsystems of the Control Room Emergency Ventilation (CREV) system occurred. Because Brunswick has a shared control room, this placed Unit 1 and Unit 2 in Technical Specification (TS) 3.7.3, Required Action C.1, for two CREV subsystems inoperable (i.e., be in Mode 3 within 12 hours). At the time of the event, a modification to upgrade the Control Building fire detection system was in progress. The 2A CREV subsystem was placed in the radiation/smoke protection mode in compliance with the Technical Requirements Manual. This action prevented an auto-start of the 2B CREV subsystem and, as such, TS 3.7.3 Condition A was entered to restore 2B CREV subsystem to operable status within 7 days. During work to electrically isolate one of the fire detectors associated with the 2A CREV subsystem, electrical continuity was lost resulting in a charcoal fire signal being sent to the 2A CREV subsystem circuitry and shutting it down. With the 2A CREV subsystem shut down due to the signal, TS 3.7.3 Required Action C.1 applied for both CREV subsystems being inoperable. Actions were taken to re-start the 2A CREV subsystem, and TS 3.7.3 Required Action C.1 was exited within approximately two minutes. The safety consequences of this event were minimal. The condition existed for approximately two minutes, and plant staff took immediate action to return the equipment to service. The apparent cause of the event was inadequate documentation and communication of the required system alignment to support the ongoing modification. |
05000325/LER-2012-005 | 29 October 2012 | Brunswick | On August 28, 2012, during planned maintenance on Emergency Diesel Generator No.2 (EDG 2), a post-maintenance continuity test associated with the Alternate Safe Shutdown (ASSD) switch on EDG 2 revealed unexpected results when the switch was taken to the LOCAL position. Troubleshooting activities determined the switch to be operating properly. However, a current path preventing isolation of the control room circuit remained. EIt was determined that a wire, not identified in EDG wiring diagrams, created a short between two ASSD switch contacts. At 2134 hours Eastern Daylight Time (EDT) on August 29, 2012, it was concluded that the condition may impact the ability of EDG 2 to perform its intended ASSD function. In the event of a fire, an induced fault could potentially affect the ability to locally control EDG 2. Local control of EDG 2 is credited in the safe shutdown analysis. This condition did not affect the Technical Specification operability of EDG 2 and it remained fully capable of performing its intended safety function. The direct cause of this event was a wiring error associated with the local control circuitry for EDG 2. This was a historical error which was likely introduced during the original installation. Therefore, no root cause was determined. The error was limited to EDG 2 and has been corrected. |
05000325/LER-2012-003 | 7 June 2012 | Brunswick | On April 9, 2012, at 0529 hours Eastern Daylight Time (EDT), electrical power was lost to the 4160 V emergency bus E1. Activities to support performance of procedure OMST-DG11R, "DG-1 Loading Test," were in progress when technicians connected a recorder to the incorrect terminals of an under voltage relay on emergency bus El and caused the normal supply breakers for emergency bus El to open. Emergency Diesel Generator (EDG) 1 automatically started and re-energized emergency bus El per plant design. This condition is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of a system specified in 10 CFR 50.73(a)(2)(iv)(B). The root cause of this event is inadequate use of human performance tools when connecting recorders in preparation for performing OMST-DG11R. Corrective actions include revising the EDG loading test procedures to provide instructions on labeling cables and to incorporate a method to record cable assignments to respective procedural steps. |
05000325/LER-2011-003 | 30 January 2012 | Brunswick | On December 1, 2011, at 1344 hours Eastern Standard Time (EST), the Control Building (CB) instrument air dryer failed resulting in loss of control air. As a result, the three Control Room Air Conditioning subsystems required by Technical Specification (TS) 3.7.4, "Control Room Air Conditioning (AC) System," and the two Control Room Emergency Ventilation subsystems required by TS 3.7.3, "Control Room Emergency Ventilation (CREV) System," became inoperable. Because Brunswick has a shared control room, Unit 1 and Unit 2 entered TS 3.7.3 Required Action B.1, for two CREV subsystems inoperable (i.e., be in Mode 3 within 12 hours) and TS 3.7.4, Required Action E.1, for three Control Room (CR) AC subsystems inoperable (i.e., enter LCO 3.0.3 immediately). At 1410 hours, operability of two Control Room AC subsystems and one CREV subsystem was restored, and LCO 3.0.3 was exited, when the CB instrument air dryer was bypassed. No power reduction took place as a result of the LCO 3.0.3 entry. The failure of the CB instrument air dryer was due to low refrigerant pressure leading to ice blockage of the instrument air supply line. The cause was inadequate monitoring to detect the low refrigerant pressure. Corrective actions include replacing the instrument air dryer and a procedure revision to bypass the dryer when low refrigerant pressure conditions exist. |
05000325/LER-2011-002 | 8 December 2011 | Brunswick | On October 13, 2011, in preparation for converting from 10 CFR 50, Appendix R, to NFPA 805, a review of the Brunswick Steam Electric Plant (BSEP) Safe Shutdown Analysis identified conditions that may not ensure a protected train of equipment remains available under certain postulated fire scenarios. The analysis determined that a postulated fire in specific fire areas could cause spurious actuation of critical components, potentially resulting in loss of equipment required for safe shutdown. A fire in one of the specified fire areas could potentially adversely affect the following: Suppression Pool level instrument 2-CAC-LT-2602, Residual Heat Removal net positive suction head (i.e., drywell containment overpressure), Reactor Core Isolation Cooling (RCIC), Emergency Bus E-1, and Emergency Bus E-3. The safety significance of this event is minimal. Fire watches were established for the affected portions of fire areas RB1-1, RB2-1, TB1, CB-2, CB-13, and CB-23. Additionally, fire detection and suppression equipment in the affected areas were fully functional. This was determined to be a historical condition and no root cause could be identified. Corrective actions include establishing an hourly fire watch in the affected fire areas, revision of alternative safe shutdown procedures, and completion of a new safe shutdown analysis. |
05000325/LER-2009-002 | 8 September 2009 | Brunswick | On July 8, 2009, at 1013 hours Eastern Daylight Time (EDT), during planned preventive maintenance activities, electrical power was lost to the 4160V emergency bus E2. Emergency Diesel Generator 2 automatically started and re-energized the E2 bus. The loss of power to E2 resulted in Unit 1 Primary Containment Isolation System Groups 2, 3, 6, and 10 isolations. Per design, no Unit 2 safety system group isolations or actuations occurred. Other Unit 1 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), automatic start of both trains of the Standby Gas Treatment System and automatic start of both trains of the Control Room Emergency Ventilation System. The affected equipment responded as designed. This event occurred during activities associated with instrument calibration of an emergency bus E2 voltage transducer. Technicians performing the activity opened the wrong test switch. As a result, arcing occurred when test equipment was connected to an energized circuit. This caused the blown fuse in the C phase of emergency bus E2, which in turn caused a loss of power to the emergency bus and the E2 master/slave breaker to trip. The root cause of this event is inadequacies associated with procedure OPIC-CNV023 and the associated work order used to perform the preventive maintenance task. Corrective actions to prevent recurrence will correct identified problems with these documents. |
05000325/LER-2008-003 | 11 August 2008 | Brunswick | On June 11, 2008, as a result of a review of the Reactor Building crane structure by the original Reactor Building cranes did not ensure the crane structural integrity during a design basis seismic event. Specifically, the allowable design stresses for the design basis seismic event are exceeded in the end connector plates and bolted connections connecting the crane girders. This condition has existed since operation of the plants began. The direct cause of this event is that the crane girder end connection design was not adequately evaluated during the initial design of the crane by Whiting Corporation (i.e., the OEM). The crane design was performed by Whiting Corporation in the 1970's. Due to the historical nature of this condition, determining a plausible cause is not practical or feasible. Modifications have been implemented for the Unit 1 and Unit 2 Reactor Building cranes to allow continued restricted use of the cranes for loads of up to 40 tons. Engineering Changes will be developed and implemented to restore the cranes to their original seismic design requirements. |
05000325/LER-2006-003 | 7 June 2006 | Brunswick | At 1630 on April 10, 2006, plant personnel identified a potential detection system. The failure mode has the potential to render the (CREV) system inoperable following power restoration after a (LOOP/LOCA) event. Due to this design deficiency, the CREV Specification (TS) 3.7.3, "Control Room Emergency Ventilation Condition B of TS 3.7.3. This required the units to be in Mode 36 hours. At 2100 on April 10, 2006, the CREV system radiation operable after removal of the chlorine tank car from the exclusion prevent them from operating. This event is being reported in accordance event or condition that could have prevented the fulfillment of are needed to mitigate the consequences of an accident. The safety minimal. The root cause of this event was determined to be ineffective review modification prior to approval. Corrective actions include revising approval checklist, revision of the modification to eliminate the personnel training. failure mode of the recently modified chlorine Control Room Emergency Ventilation Loss of Offsite Power/Loss of Coolant Accident system was declared inoperable per Technical System," which placed Units 1 and 2 in 3 within 12 hours and in Mode 4 within and smoke detection mode was restored to area and disabling the chlorine detectors to with 10 CFR 50.73(a)(2)(v)(D), as an the safety function of structures or systems that significance of this event is considered of the chlorine detector replacement an engineering procedure to add a final unanticipated failure mode, and additional |
05000325/LER-2004-001 | 4 March 2004 | Brunswick | On January 4, 2004, while performing the Emergency Diesel Generator (EDG) No. 3 monthly load test, a jacket water system piping leak of sufficient quantity to render EDG No. 3 inoperable was identified. As part of the investigation into this condition a past operability review was performed. The results of this review indicate that (1) a jacket water system pipe coupling was improperly installed during a coupling gasket replacement activity performed on February 3, 2003, and (2) tightening of the improperly installed coupling on December 8, 2003, resulted in excessive further misalignment of the coupling which impacted EDG operability until the misalignment was identified on January 4, 2004. The cause of the condition is attributed to missing pipe supports which resulted in an inadequate pipe coupling alignment. Failure to perform a functional verification following coupling maintenance on December 8, 2003, is considered a contributing cause. This condition is reportable in accordance with the 10 CFR 50.73(a)(2)(i)(B), as operation prohibited by the plant's Technical Specifications (TS), in that the EDG was inoperable for a period of time greater than that allowed by the TS. By January 7, 2004, the EDG No. 3 jacket water piping configuration was restored to the as-designed condition, satisfactorily tested, and the EDG returned to service. Additional corrective actions include . inspection of the other EDGs for similar jacket water configuration concerns and reinforcement of minor maintenance functional verification requirements with maintenance personnel. Previous reportable occurrences involving either the inoperability of the EDGs or degraded conditions resulting from maintenance activities within the last two years were not identified. |
05000325/LER-2003-002 | 26 November 2003 | Brunswick | On October 5, 2003, Progress Energy Carolinas, Inc. (PEC) received notification from GE Nuclear Energy (GENE) of a reportable condition in accordance with 10 CFR 21.21(d) (i.e., SC03-20, "Stability Option III Period Based Detection Algorithm Allowable Settings," dated October 4, 2003). SC03-20 identified the potential for numerous, unexpected confirmation count resets in the event of an instability condition. These confirmation count resets could result in the inoperability of Technical Specification (TS) Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 2.f, "OPRM Upscale?' Therefore, at 1100 Eastern Daylight Time (EDT) on October 5, 2003, the Unit I and Unit 2 Oscillation Power Range Monitor (OPRM) channels were declared inoperable in accordance with TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," and an alternate method to detect and suppress thermal hydraulic instability oscillations was placed into effect as directed by Condition I of TS 3.3.1.1. PEC is preparing modifications to implement the recommendations of SC03-20 and return the OPRM Upscale function to operable status for both Units. The safety significance of this occurrence is considered minimal. The apparent cause of the event is an incomplete analysis, performed by GENE, when establishing the OPRM's period based detection algorithm (PBDA). NRC FORM 56617.2001)
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