Semantic search

On March 7, 2002, while the plant was in Mode 6, an on-shift senior reactor operator discovered two inspection covers missing from the guard pipe which encapsulates the Unit 1 four inch Auxiliary Feedwater Pump Turbine (AFWPT) steam supply line where this line passes through the Unit 1 Auxiliary Building elevation 692.0 penetration room and enters into the AFWPT room. The guard pipe is used as a barrier to isolate a steam leak in case of a steam supply line rupture and vent steam into the AFWPT room where redundant temperature sensors will initiate isolation by closure of redundant valves located in the south main steam valve vault room. This guard pipe was installed primarily to prevent unacceptable damage to essential equipment located in the auxiliary building. The encapsulation serves to restrain the pipe thus preventing pipe whip and confines the steam jet. Upon initial evaluation, it was postulated that a circumferential rupture at the 90° elbow (near the area where these inspection covers were found missing), could have resulted in potentially unacceptable impact on environmental qualification of safety related equipment needed to mitigate the event. In addition to the missing covers, a penetration seal was found to be inappropriately installed in the mechanical sleeve which connects the guard pipe to the AFWPT room which may have delayed automatic isolation of the postulated rupture of the steam supply line. However, since that time, TVA has performed an extensive evaluation and determined that equipment required to mitigate the event would perform required functions.

This LER is being provided as a voluntary report.

The cause of the condition has not been identified. The condition appears to have existed since prior to initial fuel load in November of 1995. Corrective actions included fabrication and replacement of the pipe covers, removal of the penetration seal, and a confirmation review of similar configurations in the plant.

On March 10, 2003, at 0012 Eastern Standard Time, with Watts Bar 1 at 100 percent power, a Generator Backup Relay unexpectedly actuated causing an automatic turbine/reactor trip. The relay actuated due to a ground fault caused by a broken o-ring in the C phase main transformer's high side bushing capacitance tap connector. Plant safety equipment performed as designed which included the auto-start of the auxiliary feedwater system.

The root cause of this event was inadequate preventive maintenance procedure. Corrective actions include repairing the connector, revising preventative maintenance (PM) procedures, and a design change to the affected single point vulnerability relay scheme.

On August 25, 2003, Watts Bar (WBN) Unit 1 was operating at 100 percent power when there was an operation of the 2 out of 3 logic for the Sudden Pressure Relays (SPRs) for Main Transformer Bank 1C. The actuation of the relays resulted in a turbine trip and a subsequent reactor trip at approximately 0945 EDT. All control rods inserted as required and the safety systems actuated as designed including the motor and turbine driven pumps for the Auxiliary Feedwater (AFW) System. AFW pump 1B-8 was available but not operable at the time of the trip due to work on an associated penetration room cooler. The pump started as required. There was no loss of safety function. Unit 1 was stabilized in Mode 3.

The immediate cause of the trip was the actuation of the SPRs which were initiated by a worker bumping into the junction box that houses the relays in the switchyard. Subsequent to this, it was identified that the design of the SPA configuration was sensitive to actuation when force was applied to the relay housing junction box.

Corrective actions include a modification to improve the vibration isolation of the SPRs and the installation of a permanent protective barrier fence with access gate around each of the four SPRs, pipe, hydraulic hose, junction box, and support installations to prevent accidental impact during normal plant operations. Trip hazard signs have been placed on the fence barrier.

ARC FORM 355 (7.200 r) i NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

On October 21, 2003, WBN Unit I was returning to service after completion of the Cycle 5 refueling outage. The unit was in Mode 1 at approximately 36% reactor power, when it was established that 6.9KV Shutdown Board breaker, 1-BKR-72-10, was not connected and was not capable of supplying power to the Containment Spray System (CSS) Pump 1B-B. In accordance with LCO 3.6.6, "CSS," Pump 1B-9 is required to be operable In Mode 4.

The restoration of the breaker should have been performed in accordance with Section 5.2.8 of General Operating (GO) Instruction 1, "Unit Startup from Cold Shutdown to Hot Standby," and verified in accordance with Step 6 of Appendix B, 'Wade 5-to-Mode 4 Review and Approval," of GO-1. Due to the CSS pump not being operable as the unit transitioned from Mode 5 to Mode i, the mode change restrictions of LCO 3.0.4 were not met. The total time CSS Pump 1B-B was inoperable was approximately 113.6 hours. Considering this, Action A of LCO 3.6.6, "CSS," requires that an inoperable CSS train be restored within 72 hours. When this action is not met, Action C.1 requires that the Unit be in Mode 3 in 6 hours. Neither of these actions were met. The failure to comply with the requirements of LCO 3.0.4 and LCO 3.6.6 is being reported as a violation of the Technical Specifications in accordance with 10 CFR 50.73 (a)(2)(i)(B).

NRC FORM 26t3 (7.2001)

On July 31, 2006, at approximately 12:13 EDT, the control room unexpectedly received an exciter field overcurrent alarm, followed immediately thereafter ( An event team was assembled subsequent to the plant trip. The most likely cause was determined to be in the main generator automatic excitation control circuitry. Corrective action taken was to retum the plant to service while monitoring several input and output signals associated with the automatic controls so that any subsequent events could be captured and analyzed. The excitation system would also be operated in TEST mode instead of AUTOMATIC to preclude the possibility of an additional plant trip.

Since restarting the plant, there have been several instances where the output of the Maximum Excitation Limiter (MXL) circuit board in the automatic excitation control circuitry has changed significantly (+14 volts dc to -15 volts dc) with no corresponding change on any of the inputs. Had the automatic circuit been in service, the MXL output change would have driven the excitation controls into a loss of the exciter field which was the case during the plant trip. Based on this data, during the current refueling outage, the MXL circuit board will be removed to determine if a discreet component on the device has failed and an inspection performed of the interface wiring to the MXL circuit board to determine if degraded wiring caused the failure.

As a result of the plant trip, the actuation of the Reactor Protection and the Auxiliary Feedwater Systems were reported in accordance with 10 CFR 50.72(b)(2)(iv) and 10 CFR 50.72(b)(3)(iv), respectively. This event is also being reported as this Licensee Event Report in accordance with 10 CFR 50.73 (a)(2)(iv).

NRC Inspection Report 05000390 identified in a noncited violation (NCV) that WA started Watts Bar Nuclear Plant (WBN) Unit 1 since initial plant startup without an operable channel of auxiliary feedwater (AFW) automatic start on a trip of all main feedwater pumps as required by Technical Specification (TS) 3.3.2 Function 6.e. The NCV finding was determined to be of very low safety significance because the finding did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time since other initiation signals were available to automatically start the auxiliary feedwater pumps if needed.

With this inspection report, NRC clarified that the instrumentation channels must not only be capable of transmitting a trip signal, but must also reflect the actual operating condition of the main feedwater pump.

WA has submitted a TS change to permit operation without the AFW autostart on trip of all main feedwater pumps until a main feedwater pump is actually providing feedwater flow to the steam generators. Until operation in accordance with the revised TS is approved, WA will startup using all three AFW pumps, which will eliminate the need for the autostart, since the signal would be to start the pumps that are already running.

On October 29, 2008, a discrepancy in the setpoint of the particulate channel of the radiation monitor being credited for meeting Technical Specification (TS) 3.4.15 was identified and the appropriate Limiting Condition for Operation (LCO) was entered. On October 14, the radiation monitor had been calibrated to a setpoint that was no longer within the specified tolerance as a result of a design change. From October 14 to October 29, the Reactor Coolant System Leakage Detection System had been inoperable due to this incorrect setpoint. Consequently, WBN had been operating in a condition prohibited by Technical Specifications.

The cause of this event was determined to be a human performance error during the preparation of design change impact forms. An insufficient level of Question, Validate, and Verify (QV&V) was used, and self-checking was flawed by a wrong assumption regarding design change scope. The setpoint was corrected October 30.

On May 9, 2011 at 15:17 Eastern Standard Time (EST) with Watts Bar Nuclear Plant, Unit 1 in Mode 5, and the Reactor Coolant System (RCS) in a near water solid condition, the licensed operator started Safety Injection Pump 1A-A (SIP 1A-A) to fill and vent the Cold Leg Accumulators (CLAs) in accordance with System Operating Instruction SOI- 63.01. Following startup of SIP 1A-A, RCS pressure immediately began to rise and reached a maximum pressure of 328 psig before the operators secured the pump. The RCS pressure transient during this event did not exceed the Cold Overpressure Mitigation System (COMS) setpoint. The unexpected pressure transient was due to improper alignment of the Safety Injection System (SIS) when used to fill and vent the CLAs. Specifically, SIP 1A-A Crosstie Valve (1-FCV-63-152) was opened when it should have been closed. Misalignment of the SIS was due to a failure to follow procedures for a temporary clearance lift.

LCO 3.4.12 was not met because a SIP was capable of injecting into the RCS in Mode 5, which is reportable as a condition prohibited by Technical Specifications in accordance with 10 CFR 50.73(a)(2)(i)(B).

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.

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 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.

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 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).

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 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 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).

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 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.

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 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 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.

.._ _ NRr, FORM Kri 2017:

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.