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On December 11, 2017 at 0857 Eastern Standard Time (EST), the Watts Bar Nuclear Plant Unit 2 reactor was manually tripped after Operators observed multiple dropped control rods. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant was stabilized with decay heat removal through Auxiliary Feedwater and the Steam Dump System.

An intermittent electrical connection between a rod control power cabinet card and the power cabinet backplane power supply caused a rod control malfunction with control bank A group 2 control rods. The malfunction resulted in four control rods dropping into the reactor core. As a corrective action, all five control rod power cabinets had a 100 percent inspection of backplane connectors on the card cages. Backplane connectors were reformed and aligned as necessary to attain suitable electrical and mechanical connection. Additionally, all associated circuit cards for the power cabinets had their connectors re-formed with a precision tool.

This event is being reported under 10 CFR 50.73(a)(2)(iv)(A) as safety system actuations of the Reactor Protection System and Auxiliary Feed Water System.

On July 25, 2017, at 0428 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 3.

commencing a Reactor Startup. While in the initial phase of withdrawing the first of four Control Banks, the two associated group demand position indicators deviated greater than 2 steps from each other. In accordance with Technical Requirement 3.1.7, Position Indication System, Shutdown, with one or more group demand position indicators inoperable, the reactor trip breakers are to be opened immediately. Operations personnel opened the reactor trip breakers immediately by initiating a manual trip of the Reactor Protection System. The Auxiliary Feedwater system was in service and controlling Steam Generator water levels at the time of the event and did not receive any valid actuation signals. No other system actuations occurred as a result of this reactor trip and all systems operated as designed.

The rod demand indication deviation was determined to be caused by a failed logic card, which was replaced.

On March 20, 2017 at 0813 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) Unit 2 operations personnel manually tripped the plant from approximately 91 percent power based on lowering steam generator levels. Prior to the plant trip, the 2A Hotwell pump tripped at 0759 EDT and the 2C Condensate Booster Pump subsequently tripped at 0803 EDT. Operations personnel commenced to lower plant power after the 2A Hotwell pump trip in an attempt to maintain steam generator levels, but were unable to recover level and manually tripped the unit.

All control rods fully inserted and all automatically actuated safety related equipment operated as designed. At 0905 EDT, operations personnel exited the emergency operating instructions after the plant was stabilized.

This event resulted when scaffold crews inadvertently depressed the local trip button for the 2A Hotwell pump, which resulted in the secondary system transient. Bump guard covers were subsequently installed on local pushbuttons for selected pumps in the turbine building.

NRC I ORM TEE :36'01 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/3112018 comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to NEOB-10202. (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

On August 30, 2016, at 2110 Eastern Daylight Time (EDT), the Watts Bar Nuclear Plant (WBN) Unit 2 reactor tripped on turbine trip as a result of an electrical fault. All control rods fully inserted and no safety or relief valves lifted. The Auxiliary Feedwater system actuated as designed.

The electrical fault was caused by an internal fault on the low voltage side of the 2B Main Bank Transformer (MBT) which resulted in a fire. The electrical fault was cleared by the 2B MBT sudden pressure and phase differential relays.

Automatic fire suppression operated as expected and a fire fighting team was established by the fire brigade with assistance from local fire departments. The fire was extinguished at 2230 EDT.

The failed 2B MBT was removed from the plant and the spare MBT was connected in its place. The unit was returned to power and replacement transformers are being procured by the Tennessee Valley Authority for long term reliability.

On August 13, 2016 Watts Bar Nuclear Plant Unit 2 (WBN2) was being stabilized following a pre-planned reactor trip.

Both motor driven auxiliary feed water pumps and the turbine driven auxiliary feed water pump (TDAFW pump) were in operation maintaining steam generator (SG) water level between 6 - 50 percent in accordance with the Reactor Trip Response Procedure.

At 0333 Eastern Daylight Time (EDT) the TDAFW pump was secured by procedure and SG water level lowered to the Lo-Lo Alarm setpoint (17 percent). With the Unit at 0 percent power, a trip time delay of 3 minutes is present for auxiliary feedwater actuation. At 0337 EDT, the TDAFW pump automatically started with SG water levels less than the Lo-Lo alarm setpoint (lowest level reached was 15 percent).

The cause of the event was a failure to brief the auto start feature of the TDAFW pump at Lo-Lo SG water level of 17 percent when briefing the control band for the SGs is between 6 to 50 percent.

On June 5, 2016 at 1227 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant Unit 2 was in MODE 1 at approximately 12.5 percent power when a safety injection (SI) actuation occurred, followed by an automatic reactor protection system (RPS) trip. No primary safety barriers (Reactor Coolant System, containment and fuel clad) were challenged and no primary or secondary safety or relief valves actuated during the event. The Unit 2 plant trip was considered a complicated trip due to SI actuation. Safety equipment operated as expected and SI was promptly terminated.

The reactor trip and SI were caused by a turbine governor valve failing open, causing a steam header pressure rate of decrease SI actuation signal.

On April 14, 2016, during performance of Surveillance Requirement (SR) 3.7.5.2, the Turbine Driven Auxiliary Feedwater pump (TDAFWP) failed to achieve required rated speed of 3950 rpm ± 25 rpm due to an equipment failure. The TDAFWP was declared inoperable, and Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.5, Condition B, was entered. The equipment was repaired, the TDAFWP was re-tested successfully and returned to service. TS LCO 3.7.5 was exited on May 4, 2016.

On May 11, 2016, a past operability evaluation concluded that the TDAFWP had been inoperable from March 30 through April 17, 2016, during periods of time when the TDAFWP was required for Mode 3 operations. This is reportable as a condition prohibited by TS.

During the same time period, the 2A-A Motor Driven Auxiliary Feedwater Pump (MDAFWP) experienced an oil leak through the inboard bearing housing vent cap that resulted in the need to add approximately 4 ounces of oil on a daily basis. The MDAFWP was determined to be operable.

On November 9, 2017, an issue was identified where Technical Specification (TS) Limiting Conditions of Operation (LCOs) were not entered when non-TS Engineered Safety Feature (ESF) area coolers were removed from service for maintenance. The Watts Bar Nuclear Plant (WBN) had been performing maintenance on ESF coolers serving Auxiliary Building areas without entering the TS LCO Action Statements associated with equipment present in those areas. Specific areas of concern identified were the general areas of the 713 foot and 737 foot elevations of the Auxiliary Building. These coolers were taken out of service for time periods longer than allowed for ESF equipment (typically 72 hours), which would represent a condition prohibited by the TS.

At this time, WBN has not confirmed that for those cases where a cooler was taken out of service without entering a TS LCO if an actual adverse impact on safety function would have occurred if an accident with a single failure had occurred during those time periods. Those details, and the cause and corrective actions related to a 2010 guidance change, will be provided in a supplement to this report.

On August 23, 2017, Watts Bar Nuclear Plant (WBN) personnel identified Technical Specification (TS) 3.6.3, Containment Isolation Valves, was not entered for on-going work related to 1-FCV-31-330, Incore Instrument Room Air Handler Unit 1B Chilled Water System Isolation Valve. A clearance was placed on 1-FCV-31-330 by Operations Work Control for scheduled work on May 17, 2017 rendering the valve inoperable. Work was completed May 19, 2017, however, the clearance remained in place pending post maintenance testing after other related system work was complete. Due to a human performance error, the appropriate TS tracking program was not activated and no narrative log entry was made to signify entry into the TS as required by procedure. The in-place clearance satisfactorily met the required actions of TS 3.6.3 condition A.1. to isolate the affected containment penetration flow path by use of at least one closed and de-activated automatic valve. However, without the required TS tracking program activated, personnel failed to comply with TS 3.6.3 condition A.2. to verify the affected penetration flow path is isolated every 31 days.

Failure to enter the TS tracking program in accordance with procedure was a human performance error. Corrective actions included coaching and department operating experience communication.

On July 12, 2017, at 1238 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows adequate Essential Raw Cooling Water (ERCW) flow may not be available during dual unit limiting design basis conditions of one unit in Hot Shutdown on Residual Heat Removal (RHR) cooling when the other unit experiences a Loss of Coolant Accident (LOCA). Based on preliminary analysis, during a Unit 1 LOCA, Unit 1 receives adequate flow when following existing procedural guidance. However, Unit 2 may not receive adequate flow to meet cool-down requirements with design basis maximum temperatures. During a Unit 2 LOCA, however, current procedural guidance is not adequate to ensure the proper system alignment to establish correct ERCW Component Cooling Water (CCS) Heat Exchanger A and B flow rates for either unit's cool down requirements.

At the time of the event, Unit 2 had been shutdown for an extended period of time such that the flow delivered by ERCW was adequate to serve both Unit 1 in a LOCA and Unit 2 in Mode 4 or 5. Immediate corrective actions included procedure changes to ensure adequate ERCW flow for all possible plant situations in both units. The causes of this issue are a failure to address cross train elements of the ERCW system in the design analysis and a failure to address procedural impacts when the plant transitioned to dual unit operation. The procedure errors were corrected.

On June 9, 2017. Watts Bar Nuclear Plant (WBN) personnel determined that the reporting requirements of 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v), as clarified by guidance in NUREG-1022, Revision 3. were being incorrectly applied for certain events associated with single train safety systems. When events occurred that resulted in these systems not meeting Technical Specification (TS) Limiting Conditions for Operation (LCO). the short duration of these events relative to their required action completion time, coupled with prompt return to allowable values, were not considered a loss of safety function by Operations and Licensing personnel. As a result, multiple potential loss of safety function events were not reported as required. These events were related to Refueling Water Storage Tank (RVVST) level, Containment and Shield Building pressure, and Control Room Envelope integrity.

A review of these events indicate, when considering the actual system capability and the response of equipment and personnel. a loss of safety function capability impacting public health and safety did not occur for events associated with the RWST, Containment. Shield Building, or Control Room. Corrective actions include briefing personnel on the regulatory impact of these events, and the importance of the control room boundary.

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On May 10, 2017, at 0907 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) Unit 1 operations personnel discovered the 1B-B Safety Injection pump discharge isolation valve (1-ISV-63-527) closed. Technical Specification (TS) 3.5.2, ECCS - Operating, Condition A was immediately entered for one or more trains of the Emergency Core Cooling System (ECCS) inoperable. TS 3.5.2 Condition A was exited at 0913 EDT when 1-ISV-63-527 was opened.

Investigation determined that the 1 B-B SI pump discharge isolation valve had been closed prior to Unit 1 entering Mode 3 on April 26, 2017, representing a condition prohibited by TS. During this time period, the 1A-A SI pump was inoperable for 21 minutes, representing a condition that could have prevented fulfillment of a safety function.

The cause of the mispositioned valve was the result of an individual failing to follow procedure use and adherence requirements during the performance of Emergency Diesel Generator (EDG) Blackout testing. The safety injection pump discharge valve was closed to support the test but was not reopened following the testing. Corrective actions for this event include personal accountability actions, revision of the EDG blackout procedures to ensure the SI pump discharge valves are reopened, and additional station focus on procedure use, particularly use of Not Applicable (N/A) in performing procedures.

On January 4, 2017 at 1010 Eastern Standard Time (EST), Watts Bar Nuclear Plant Operations personnel declared Essential Raw Cooling Water (ERCW) strainer flush valve 2-FCV-67-9B inoperable due to having a through-wall leak.

The valve was replaced and the ERCW strainer was returned to service on January 5, 2017 at 0952 EST. This event is reportable because the valve had had a through-wall leak since January 31, 2016 and had not been declared inoperable. With a through-wall leak, a flaw evaluation is required to be performed to demonstrate the through-wall leak was stable. The failure to perform an adequate operability evaluation allowed the valve to remain in service for a period of time longer than allowed by Technical Specification (TS) 3.7.8, Essential Raw Cooling Water, Limiting Condition for Operation (LCO) Condition A. This represents a condition prohibited by the TS. Subsequent analysis of the valve demonstrated that it remained structurally sound with the leak, and would not have impacted the operability of the ERCW system.

The cause of the failure to perform an adequate operability evaluation has been determined to be human performance errors on the part of both operations and engineering personnel. Additional training of operations and engineering personnel is planned as corrective action to address the potential for recurrence of this issue.

On November 10, 2016, Watts Bar Nuclear Plant personnel identified a failure of the non-reverse clutch key on Emergency Raw Cooling Water (ERCW) motor B-A. While performing a lubrication work order, it was discovered that the clutch key was sheared. Subsequent investigation identified that other clutch key failures had occurred in the recent past. The non-reverse clutch prevents the ERCW pump from rotating in the reverse direction after pump trip, which could cause the motor to develop a higher than normal in-rush current if the motor was subsequently started, such as following an accident.

Based on the potential common mode failure of the non-reverse clutch, immediate corrective actions were put in place to ensure that the safety function of the ERCW pumps to start following an accident would not be impaired. The cause of the failure is under investigation.

At 1526 Eastern Daylight Time on 6/8/2016, a determination was made involving the potential impact of a tornado on the Emergency Diesel Generators (EDGs). The EDGs are required to be operable to provide power to ensure that acceptable fuel design limits, reactor coolant system pressure boundary limits, and containment integrity are not exceeded during abnormal transients. Further, the EDGs are designed with a crankcase pressure trip (setpoint of 1 inch water), which is bypassed following an emergency start. Engineering has determined that a tornado could potentially cause actuation of the crankcase pressure trip due to a low barometric condition. If an emergency start signal has NOT previously occurred, then during a tornado, actuation of the crankcase pressure trip would energize the shutdown relay causing an EDG lockout condition. The EDG lockout condition prevents subsequent EDG starts (normal or emergency) until operators manually reset the lockout condition locally at the EDG. This condition could potentially affect all four EDGs simultaneously.

A compensatory measure has been established, that upon notification of a Tornado Warning, the EDGs would be 'emergency started' and run during the time the Tornado Warning was in effect. This action bypasses the crankcase pressure trip function and allows the EDGs to perform their required safety function.

On May 17, 2016, at 1630 hours while restoring from a plant modification related to installation of new protective relays designed to detect open phase conditions on the 6.9kV shutdown boards, the feeder breakers for the 6.9kV Shutdown Board 1B-B tripped resulting in a loss of bus voltage. The feeder breakers tripped due to actuation of the loss of voltage relays in the shutdown board protective relay trip logic circuit resulting in separation of offsite power from the 6.9kV Shutdown Board 1B-B. The 1B-B emergency diesel generator did not auto start during this event because it was out of service due to planned maintenance.

In response to the loss of power on the 6.9kV Shutdown Board 1B-B, the operators immediately entered Abnormal Operating Instruction, 0-A01-43.02, Loss of Unit 1 Train B Shutdown Boards, and manually started emergency diesel generators 1A-A, 2A-A, and 2B-B. All equipment operated properly. The emergency diesel generators were not required to be paralleled to their respective boards because offsite power was available.

Offsite power was restored to the 6.9kV Shutdown Board 1B-B at 1802 hours on May 17, 2016. Event Notification 51940 was issued May 17, 2016. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A).

On May 13, 2016, Watts Bar Nuclear Plant Unit 1 (WBN1) determined that a condition prohibited by Technical Specifications had previously occurred. During the Fall 2015 WBN1 outage, maintenance performed on the 1B-B centrifugal charging pump (CCP) room cooling fan introduced a condition that resulted in a subsequent bearing failure of the room cooling fan on December 4, 2015. This condition would have prevented the 1B-B CCP pump from performing its specified function for its designed mission time. Based on the reduced reliability of the fan, the 1B-B CCP was considered to be inoperable from October 7, 2015 until the fan was repaired and returned to service on December 6, 2015. During this time period, there were several short time periods when the 1A-A CCP was inoperable.

An investigation into the cause of the failure was completed on April 21, 2016. The cause of the fan bearing failure was an undersized fan shaft, resulting in the 1B-B CCP fan having excess shaft to bearing clearance which caused the bearing inner ring to loosen from the eccentric locking collar. These excessive clearances allowed the fan bearing inner ring to slide on the shaft. The sliding rotation of the inner ring on the shaft resulted in excessive heat being generated within the bearing leading to catastrophic failure.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(D).

On March 22, 2016, at 1131 Eastern Daylight Time, the Watts Bar Nuclear Plant Unit 1 (WBN1) reactor tripped due to the actuation of the Over Temperature Delta Temperature bistables. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated. All control rods inserted upon the reactor trip and safety systems functioned as expected.

An investigation into the cause of the trip determined that a failure of a Valve Position Limit up/down counter circuit card in the Analog Electro-Hydraulic Turbine Control System resulted in the closure of the turbine high pressure governor valves, resulting in an automatic reactor trip and turbine trip on WBN1. The failed card was replaced and WBN Unit 1 was returned to service.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A).

On March 5, 2016, at 1512 Eastern Standard Time (EST), Watts Bar Nuclear Plant (WBN) Unit 1 entered Technical Specification (TS) 3.6.3, Containment Isolation Valves, Condition A for a containment isolation valve being inoperable.

During a containment walkdown, leakage was found on valve 1-FCV-61-122, Glycol Cooled Floor Return Header Isolation and the valve was declared inoperable. TS 3.6.3 Condition A requires that a penetration flow path with one containment isolation valve inoperable to be isolated by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve within 4 hours. The penetration associated with this containment isolation valve was not isolated until 2113 EST on March 5, 2016. The cause of this event was operations staff misunderstanding the applicability of the Note associated with TS 3.6.3, which allows administrative controls under certain conditions.

Because the action specified by TS 3.6.3 was not completed within four hours, this condition is reportable as an operation or condition prohibited by TS per 10 CFR 50.73(a)(2)(i)(B).

reactor was manually tripped by control room operators due to a decreasing main condenser vacuum. Subsequent to the reactor trip, the Auxiliary Feedwater system actuated. Control and Shutdown rods fully inserted into the reactor core, and safety systems responded as designed. The unit was stabilized in Mode 3, with decay heat removal via Auxiliary Feedwater and the Steam Generator Atmospheric Dump Valves. The Main Steam Isolation Valves were closed and remained closed during the event.

Tennessee Valley Authority (TVA) has determined that the decreasing condenser vacuum was due to a failure of an expansion joint boot seal in the "C" zone of the main condenser. This seal functions as the expansion joint between the condenser and low pressure turbines. The failure of the seal was due to a non-optimal vulcanization process and inadequate overlap in a joint splice, which significantly weakened the seal and allowed seal water to permeate the seal, further weakening the joint. The failed main condenser boot seal was replaced with a new boot seal on the "C" zone of the condenser. As a preventative measure, the boot seals on the "A" and "B" zones were also replaced.

On February 11, 2014, Watts Bar Nuclear Plant (WBN) engineering and operations personnel discovered that non- conservative operator manual action times were credited in Appendix R analyses. Preliminary Westinghouse transient analysis calculations of WBN Unit 1 fire protection features revealed that there was less time than previously credited to perform certain operator manual actions to prevent pressurizer overfill during certain Appendix R fire scenarios. The Westinghouse analysis assumes the time required to isolate the normal charging path, secure the second charging pump and isolate the emergency charging path is approximately 12.5 minutes. Watts Bar Unit 1 procedures are non- conservative in that they allow these actions to be completed in 18 minutes.

The Tennessee Valley Authority (TVA) has verified that potentially impacted Appendix R equipment remains functional; however, a compensatory fire watch has been established for the affected areas until plant modifications are completed.

This event was caused by an error in a fire protection program design calculation prior to commercial operation of Unit 1.

Modifications to address this issue will be completed during the Fall 2015 refueling outage.

On May 2, 2013 at 0845, B-train Emergency Gas Treatment System (EGTS) was removed from service for planned maintenance and Operations declared Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.9 not met and entered Condition A for one EGTS train inoperable. On May 3, 2013 at 0111, the Main Control Room was notified that the A-A Auxiliary Air Compressor Air Dryer was not purging due to failure of the Auxiliary Control Air System (ACAS) A-A dryer central timing unit. Operations declared A-train ACAS and supported Technical Specification systems inoperable, including A-train EGTS.

WBN operations entered LCO 3.0.3 due to the inoperability of two trains of EGTS and began preparations to initiate an orderly shutdown within one hour. Operations initiated actions to restore B-train EGTS to standby in accordance with System Operating Instruction (S01)-65.02, Emergency Gas Treatment System. At 0155, B-train EGTS was declared operable and the actions of LCO 3.0.3 exited. No action was taken to reduce reactor power while in LCO 3.0.3.

The A-A Auxiliary Air Compressor Air Dryer central timing unit motor was replaced. The apparent cause of this event was that there were missed opportunities to identify the need for replacement preventive maintenance (PM) for the central timing unit. Change requests have been initiated for periodic replacement of the ACAS dryer central timing unit. Components in other systems which could be subject to the same failure mechanism will be reviewed and PM activities initiated as necessary.

On October 16, 2012, at 2330 EDT, Watts Bar Nuclear Plant (WBN-1) licensed operators attempted a manual fast transfer of the 1B-B 6.9kV Shutdown Board (SDBD) from the normal feeder breaker to the alternate feeder breaker. The transfer was not successful, resulting in the automatic start of the four Emergency Diesel Generators (EDGs). After the 1B-B 6.9kV SDBD de-energized and the loads were shed, the alternate feeder breaker closed and re-energized the 1B-B 6.9kV SDBD. The loads supplied by the 1B-B 6.9kV SDBD were subsequently reconnected, and required tests were successfully completed to ensure operability of the 1B-B 6.9kV SDBD.

At the time of the event, WBN-1 was in MODE 5 following a refueling outage. Operations personnel promptly entered the appropriate response procedure and re-established power to required loads. Required safety systems functioned as designed. This condition did not adversely affect the safe operation of the plant or the health and safety of the public.

The cause of this event was that plant operators did not ensure the alternate feeder breaker hand-switch was held firmly in the "closed" position while initiating the fast board transfer.

This report is being submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A), a condition that resulted in automatic actuation of the EDGs.