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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.

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.

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).

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.

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.

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.

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.

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

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.

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.

On August 11, 2002, with the plant defueled, a deficiency was identified where the last six feet of the Emergency Diesel Generators (EDG) exhaust piping is not protected from tornado-generated missiles.

Additionally, an exterior door to a Main Steam Line Room was identified as not adequately protecting the Main Steam Safety Valves (MSSVs) from tornado-generated missiles. While the probability of a missile striking this equipment and damaging it to the point that it can no longer perform its designated safety function is remote, the licensing basis at the time of discovery was that systems vital to safe shutdown in the event of a tornado are enclosed in structures designed to withstand such missiles.

dThe affected equipment was declared inoperable.

Since these design deficiencies for the protection of the safety equipment existed since initial plant operation, the operation of the plant in this condition was prohibited by the Technical Specifications, and therefore these events were reported in accordance with 10CFR50.73(a)(2)(i)(B).

A probabilistic methodology for tornado missile risk analysis of unprotected plant features was incorporated into the plant's licensing basis; no other facility modifications were determined to be necessary.

NRC FORM366 (7-2001)