|Report date||Site||Event description|
|05000346/LER-2017-002||27 November 2017||Davis Besse|
On September 13, 2017, with the Davis-Besse Nuclear Power Station operating at approximately 100 percent power, Auxiliary Feed Water (AFW) Pump Turbine 1 experienced high inboard bearing temperature during performance of quarterly Surveillance Testing. The turbine was tripped, and disassembly revealed damage to the journal bearing. The bearirig was replaced, and following successful post maintenance testing, AFW Train 1 was declared Operable on September 16. The cause of the bearing damage was an improperly marked oil sight glass, which allowed operation with improper bearing lubrication. The improper markings were due to the maintenance work instruction for replacing the sight glass not including dimensions or guidance for setting required operational bands.
On September 26, 2017, it was identified that low inboard bearing oil level had likely existed since completion of the previous quarterly surveillance test on June 21, when an oil sample was taken following testing but the bearing was not refilled due to the improperly marked sight glass. This issue is being reported in accordance with 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of the safety function, and in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
|05000346/LER-2017-001||18 September 2017||Davis Besse|
On July 20, 2017, with the Davis-Besse Nuclear Power Station (DBNPS) operating at approximately 100 percent power, it was identified that the Emergency Diesel Generator (EDG) fuel oil storage tank vents were not adequately protected from potential tornado-generated missiles. If a missile crimped the vent it could disable the transfer pump or tank, potentially impacting the seven-day fuel supply for the affected train(s) of EDG. While the storage tanks were protected from tornado missiles when installed, the vents were not provided with any such protection. Compensatory measures were established to ensure a vent path remained following a tornado event, and actions will be taken to ensure the vents for each EDG fuel oil storage tank are adequately protected from tornado missiles.
This issue is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety, in accordance with 10 CFR 50.73(a)(2)(v) as a condition that could have prevented the fulfillment of the safety function, in accordance with 10 CFR 50.73(a)(2)(vii) as an event where a single cause or condition caused two independent trains to become inoperable in a single system, and in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
|05000346/LER-2016-009||9 November 2016||Davis Besse|
On September 10, 2016, with the Davis-Besse Nuclear Power Station (DBNPS) operating at approximately 100 percent power, rainwater intrusion into the Main Generator Automatic Voltage Regulator (AVR) cabinet due to an open roof vent caused a lockout of the Main Generator, resulting in a trip of the Main Turbine and Reactor. Following the Reactor trip, the Steam Feedwater Rupture Control System (SFRCS) actuated due to high Steam Generator 1 level and initiated the Auxiliary Feedwater System. The most probable cause of the SFRCS actuation was a failed operational amplifier in the Integrated Control System (ICS), causing the ICS to not reduce Feedwater flow to Steam Generator 1 following the Reactor trip. , Completed corrective actions include closing the roof vents, sealing the top of the AVR cabinet, improved configuration control of the vents, and replacement of the failed ICS module. Scheduled corrective actions include presenting a case study to improve recognition of elevated risk issues, and review of the ICS by a multi-functional team to address system performance concerns.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of the Reactor Protection System, and an automatic actuation of the Auxiliary Feedwater System.
|05000346/LER-2016-004||30 August 2016||Davis Besse|
On April 5, 2016, at 0243, with the Davis-Besse Nuclear Power Station (DBNPS) in Mode 5, it was discovered that the six dual-element Resistance Temperature Detectors .(RTD) on both Reactor Coolant SyStem (RCS) Hot Legs had varying degrees of wire insulation degradation. The cause of the RTD insulation degradation was accelerated aging due to high temperatures as a result of improper configuration of piping insulation on the RTD during the previous refueling outage. Corrective actions taken include replacing five of the six RCS Hot Leg RTDs. The sixth RTD was determined to be suitable for continued operation and is planned to be replaced during the next Refueling Outage.
The Electrical Conductor Seal Assemblies (ECSA) taken from the removed RTDs were evaluated for reuse on replacement RTDs. It was discovered that the Midlock Ferrules were installed backwards on two RTDs during the past refueling outage. The cause was determined to be less than adequate installation instructions. Corrective actions are to revise the Maintenance procedure and implement further training.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(v)(A), and 10 CFR 50.73(a)(2)(vii).
|05000346/LER-2016-007||22 August 2016||Davis Besse|
The two pressurizer safety valves (PSVs) at the Davis-Besse Nuclear Power Station (DBNPS) were replaced during an outage in Spring, 2016 with tested spares. The removed PSVs were sent to an offsite vendor for testing and refurbishment. In June, 2016 the test results were received showing both PSVs lifted higher than the allowed one percent tolerance above the 2500 psig setpoint (2525 psig) for As-Found testing. Because both valves had As- Found setpoints above the Technical Specifications (TS) allowed value, a past operability evaluation was .
performed, which concluded that both valves were inoperable during their time in service.
Based on the as-found lift setting pressures (2559 psig and 2554 psig), there was no adverse effect on transients described in the Updated Safety Analysis Report that can produce a Reactor Coolant System (RCS) overpressurization. The cause of this event was due to setpoint drift and narrow allowable setpoint range.
Procedures will be revised to establish more restrictive testing requirements. The PSVs were replaced and a TS change will be submitted to provide for current ASME acceptance test criteria for the PSV setpoint. This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(D) and 10 CFR 50.73(a)(2)(i)(B).
|05000346/LER-2016-005||11 July 2016||Davis Besse|
On May 10, 2016, the Davis-Besse Nuclear Power Station (DBNPS) was in Mode 1 and increasing power following refueling outage activities. At 0528 hours, it was identified that all four of the Anticipatory Reactor Trip System (ARTS) channel switches were in the bypass position for the Main Turbine function while at approximately 53 percent power.
These switches are required by the Technical Specifications to be in the normal/enabled position when above 45 percent power. The switches were restored to Normal and the applicable Technical Specifications exited at 0552 hours.
An Operations Standing Order was issued to require paired periodic walk downs of all Control Room panels to ensure a comprehensive understanding of plant status awareness. Walk downs were also performed to independently identify any additional concerns or omissions in plant startup activities. The root cause of this event is the operators failed to effectively work as a team to ensure a safety system was in an operable condition when required. Corrective actions include implementation of the Operations Section Continuous Improvement Plan and revision of applicable procedures.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
|05000346/LER-2016-003||31 May 2016||Davis Besse|
On March 30, 2016, with the Davis-Besse Nuclear Power Station shutdown for a scheduled refueling outage in Mode 6 with the Reactor Coolant System depressurized, approximately one half teaspoon of dry boric acid was identified on the Reactor Coolant Pump (RCP) 1-1 first stage seal cavity vent line flexible braided piping connection, which was determined to be reactor coolant pressure boundary leakage. This leak was from the welded end connection of the small bore ASME Section III Class 2 flexible braided piping assembly between the RCP seal and the first isolation valve.
The most probable cause of this leak was a weld solidification through-wall crack at the flange to hose / bellows tube pressure boundary weld that occurred during manufacture. The post manufacture testing was not adequate to detect this extremely small pressure boundary defect. Corrective actions include inspecting the other RCP seal vent line flexible hoses, replacement of RCP 1-1 seal vent line flexible piping assembly with one that passed a more stringent leak test, and revising procurement requirements to incorporate this more stringent leak test.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(ii)(A) as degradation of a principal safety barrier and 10 CFR 50.73(a)(2)(i)(B) as operation or condition prohibited by Technical. Specifications.
|05000346/LER-2016-002||29 March 2016||Davis Besse|
On January 30, 2016, with the plant in Mode 3, with the Auxiliary Feedwater System (AFW) in operation and proceeding with Post Trip recovery actions from the January 29, 2016 Steam Feedwater Rupture Control System (SFRCS) High Level Trip (reference Licensee Event Report (LER) 2016-001) when a SFRCS Steam Generator (SG) 1 High Reverse Differential Pressure (D/P) Trip was unexpectedly received. The trip was a result of a feedwater isolation valve being opened to align the Motor Driven Feed Pump's (MDFP) discharge to SG 1. The unexpected SFRCS SG 1 Reverse D/P Trip resulted in the closure of the appropriate Main Steam and Main Feedwater (MFW) valves, as designed. AFW Pumps 1 and 2 remained in operation, as expected and the plant was verified to be stable.
The cause of this event was inadequate procedural guidance contained in the Trip Recovery Procedure with a corrective action to revise the procedure. This report is being submitted as an event that resulted in an automatic actuation of the SFRCS, therefore, reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A).
|05000346/LER-2015-004||12 October 2015||Davis Besse|
On August 12, 2015, with the Davis-Besse Nuclear Power Station (DBNPS) operating in Mode 1 at approximately 100 percent power, it was determined that the plant operated the previous operating cycle (Cycle 18) with an Axial Power Shaping Rod (APSR) fully inserted. This issue was identified by the DBNPS fuel vendor while reviewing the previous cycle's quadrant power tilt graph and comparing it to the current cycle's (Cycle 19) expected end of cycle graph, since the plant is currently operating with a known APSR fully inserted. The previous cycle operations started with initial criticality on June 11, 2012, and ended on February 1, 2014.
The most likely cause of the APSR being disconnected during the previous operating cycle is an inadvertent work practice error during the coupling process. Corrective actions include a revision to the APSR coupling procedure to require positive verification of APSR coupling (such as verifying the weight addition of the APSR to the lead screw after the coupling is completed) since visual verification is not practicable.
This issue is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as operation of the plant in a condition prohibited by the Technical Specifications (TS) because the necessary TS actions were not taken last cycle with the APSR fully inserted.
|05000346/LER-2015-002||8 July 2015||Davis Besse|
On May 9, 2015, with the Davis-Besse Nuclear Power Station (DBNPS) operating in Mode 1 at approximately 100 percent power, a steam leak was identified in the Turbine Building. A rapid shutdown was initiated, and the reactor was manually tripped at 1909 hours from approxjrnately 30 percent power.
The Steam Feedwater Rupture Control System was manually initiated to isolate the leak and start the Auxiliary Feedwater System. The cause of the leak was failure of a four-inch pipe in the Moisture Separator Reheater System due to Flow Accelerated Corrosion (FAC). An incorrect data input caused the FAC software model to underestimate the predicted wear rate, so inspections were not performed to identify the piping wall thinning prior to failure. Additionally, a previous event was not evaluated to ensure the proposed corrective actions would encompass a validation of all critical data inputs. Corrective Actions include improvements in the fidelity of the data in the FAC Software model, and improvements in the Corrective Action Program with respect to Root Cause Evaluations.
This issue is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as a manual actuation of the Reactor Protection System and the Auxiliary Feedwater System.
|05000346/LER-2014-004||23 January 2015||Davis Besse|
On November 25, 2014, with the Davis-Besse Nuclear Power Station (DBNPS) operating in Mode 1 at approximately 100 percent power, the fuel vendor for the DBNPS notified the FirstEnergy Nuclear Operating Company (FENOC) of the final evaluation results for a deficiency discovered in the thermal conductivity model computer codes. Accounting for the deficiency resulted in the predicted analytical Peak Cladding Temperature (PCT) for Large Break Loss of Coolant Accident Conditions increasing to an estimated 2513 degrees F, which is in excess of 2200 degrees F specified in 10 CFR 50.46(b)(1).
Because the station had sufficient margin in the reactor core reload analysis and had previously implemented compensatory measures recommended by the fuel vendor when the potential issue was first identified, the deficiency had no impact on current plant operation or public health and safety. The cause of this event was related to the evolution of the fuel vendor's modeling.
This issue is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety.
|05000346/LER-2014-003||17 October 2014||Davis Besse|
On August 18, 2014, with the Davis-Besse Nuclear Power Station operating in Mode 1 approximately 100 percent full power, a door to a mechanical penetration room could not be secured following normal usage.
This door is required to be latched closed to maintain the shield building negative pressure boundary except when open under administrative control. With the door unable to be latched closed, the Station Emergency Ventilation System could not perform its required safety function of maintaining a negative pressure in the affected area. The door was restored to the latched status in ten minutes. A similar failure to latch on August 20, 2014, was also experienced, and the door was re-latched in four minutes.
The cause of this event was a design flaw that could cause the latch fingers to stick, preventing the door from latching; and an infantile failure of the door closer. A vendor modified version of the latch was installed along with a new door closer to correct the problem.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(v)(C) and (D) as an event that could have prevented fulfillment of the safety function of the Station Emergency Ventilation System.
|05000346/LER-2014-002||7 July 2014||Davis Besse|
On May 5, 2014, with the Davis-Besse Nuclear Power Station in Mode 3 and the Reactor Coolant System at normal operating temperature and pressure, workers were replacing a Control Rod Drive (CRD) Position Indication Tube due to incorrect indications identified during startup testing. Because of the tight working conditions, a flexible cooling line was moved aside to replace the Position Indication Tube, which caused inadvertent disengagement of a quick disconnect fitting, resulting in inadvertent isolation of cooling water to another CRD mechanism. When the CRD mechanism reached a procedural temperature limit, the reactor trip breakers were manually opened from the Control Room.
The cause of this event was not completely identifying and assessing all risks and consequences before conducting the tube replacement. The cooling line was reconnected, and the preventive maintenance activity for performing this work will be revised to add a specific precaution regarding the susceptibility of the cooling hose quick disconnects to become disengaged.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual actuation of the Reactor Protection System with the reactor not critical.
|05000346/LER-2014-001||27 June 2014||Davis Besse|
On May 4, 2014, with the Davis-Besse Nuclear Power Station in Mode 3 and the Reactor Coolant System at normal operating temperature and pressure, testing was in progress on the Control Rod Drive (CRD) System. When a Group 8 Axial Power Shaping Rod was withdrawn, indication was received of movement of a Group 4 Control Rod. The Group 4 Control Rod indication was manually driven to zero percent indication and the reactor trip breakers were opened from the Control Room.
The erroneous indication was due to improper connection of a Containment Electrical Penetration for CRD System position indication during the refueling outage. The penetration was rewired to permit startup activities to continue. The Post-Maintenance Test Manual will be updated to add guidance for testing Containment Electrical Penetrations associated with the CRD System.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual actuation of the Reactor Protection System with the reactor not critical.
|05000346/LER-2013-002||27 August 2013||Davis Besse|
On July 1, 2013, with the Davis-Besse Nuclear Power Station in Mode 3 and the Reactor Coolant System at normal operating temperature and pressure, a leak from Reactor Coolant Pump (RCP) 1-2 first stage seal vent cavity line was identified. The 8 to 9 drops per minute leak was from a flange socket weld of the small bore ASME Section III Class 2 piping between the RCP seal and the first isolation valve. The station cooled down to Mode 5 and replaced the spool piece containing the weld.
Laboratory failure analysis concluded the cause of this leak to be high cycle fatigue of the seal cavity vent line socket weld. The high cycle fatigue was due to a less than adequate design modification to minimize vibration of the line. Planned corrective actions, based upon a weld leak identified in 2012 on the same line, include either replacement of this and similar seal piping for all four RCPs with flexible hoses or increasing the socket weld size during the next refueling outage, and performing walkdowns of risk significant socket welded piping susceptible to vibration fatigue failures.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(ii)(A) as degradation of a principal safety barrier.
Davis-Besse Unit Number 1 05000346
|05000346/LER-2013-001||27 August 2013||Davis Besse||ecify|
|05000346/LER-2011-004||6 March 2012||Davis Besse|
On July 26, 2011, with the Davis-Besse Nuclear Power Station in Mode 1 at approximately 100 percent power, information was received from the NRC regarding design issues with the Direct Current (DC) System. The first issue was that non-essential equipment in containment, powered by the DC System, was not environmentally qualified as required. This could challenge the adequacy of electrical separation between potentially grounded equipment and the safety related batteries. The second was that automatic transfer switches supplying power to non-essential instrumentation could transfer a fault to the redundant power source, potentially impacting both safety related DC power sources.
The breakers for the four Reactor Coolant Pump backup oil lift pump motors and for the emergency power supply to the Containment Lighting Panel were opened, and one train of instrumentation power was placed on its alternate power source from the Alternating Current system. Modifications have been implemented to provide the alternate power supply from a regulated instrumentation distribution panel to restore compliance with NRC Bulletin 79-27 for the automatic transfer switch loads while maintaining separation of the DC power system trains. The cause of these issues was inadequate design during initial plant design and during installation of the automatic transfer switches in the 1980s.
|05000346/LER-2011-003||2 May 2011||Davis Besse|
On March 3, 2011, with the Davis-Besse Nuclear Power Station (DBNPS) operating at approximately 100 percent power, fire detector testing was being conducted at the Auxiliary Shutdown Panel (C3630). When a radio used by Instrumentation and Control personnel for communications was keyed twice while instrumentation cabinet doors were open, momentary reductions in the control signals to the Auxiliary Feedwater Pump and Motor Driven Feedwater Pump discharge valves occurred. These momentary signal reductions resulted in the rendering of all three trains of Emergency Feedwater inoperable for approximately two minutes.
The cause of this event was a station decision made in 1991 to remove the specific requirement to exclude radio usage in the vicinity of the Auxiliary Shutdown Panel. A sign restricting radio usage in the Auxiliary Shutdown Panel Room was installed shortly after this event occurred, and procedures will be revised to prohibit the use of portable radios within six feet of cabinets in the room when the cabinet doors are open. This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(A) and (B) as a loss of safety function.
|05000346/LER-2011-002||23 March 2011||Davis Besse|
On January 22, 2011, with the Davis-Besse Nuclear Power Station in Mode 1 at approximately 100 percent power, routine testing discovered the outlet temperature control valve for Containment Air Cooler (CAC) 3 could not be closed from the control room. This containment isolation valve had been inadvertently rendered inoperable on November 30, 2010, while electrically isolating the CAC 3 fan motor that had tripped its supply breaker on overload when the motor was started on October 24, 2010.
The root cause of this event was determined to be a lack of information on the elementary wiring diagram used to prepare the clearance for isolating the fan motor. This drawing did not reference an additional drawing showing that the fan motor breaker contacts impact the power supply to the temperature control valve controls. When a clearance for the fan motor was revised, this lack of information led to the electrical isolation of the valve control circuitry, failing the valve in the open position. Corrective actions include revision of these drawings.
This Issue is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation prohibited by the Technical Specifications for the inoperable containment isolation valve. Because an additional failure would be required for a potential containment bypass pathway to exist, this event had very low safety significance.
|05000346/LER-2007-002||28 February 2008||Davis Besse|
On December 30, 2007, with the plant in Mode 5, a step decrease of approximately six inches in the Reactor Coolant System Pressurizer was observed when aligning Decay Heat Removal (DHR) Pump 1 suction from Low Pressure Injection mode to DHR mode. The alignment was suspended, and a piping void was discovered in the discharge piping of DHR Pump 1. The void was vented from the high point vent valve inside Containment to restore train operability. It was determined that this void was formed during on-line maintenance on October 31, 2007, due to an inadequate procedure for recovering the DHR train following on-line maintenance. This void rendered DHR Train 1 inoperable for approximately 59 days due to the inability to successfully meet Technical Specification Surveillance Requirement 4.5.2.b, which requires the train to be full of water while in Modes 1 through 3.
An evaluation determined DHR Train 1 remained capable of performing all required safety functions as a result of the estimated 17 cubic feet of voided piping that existed while the plant was operating. This issue is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the Technical Specifications. The procedure will be revised to properly vent the system following maintenance, and procedures for other standby safety systems will be reviewed for similar issues.