Semantic search

On 5/27/09, TVA identified that the Watts Bar Nuclear Plant (WBN) Unit 1 Auxiliary Building (AB) Gas Treatment System (ABGTS) pressure test surveillance instruction was inadequate, as closed nonsafety related dampers could mask leakage through credited safety related dampers in the AB Secondary Containment Enclosure (ABSCE).

AB General Ventilation manipulation to place WBN in a known tested condition created a pressure differential that caused failure of temporary ABSCE boundary doors installed to facilitate Unit 2 construction. WBN entered LCO 3.7.12 Condition B for 2 trains of ABGTS inoperable. WA repaired the boundary, and the LCO Condition was exited in approximately 3.5 hours.

Subsequently, WA retested ABGTS, and both trains were verified as operable. From initial licensing until 5/27/09, WBN operated in noncompliance with TS because of the inadequate surveillance instruction.

On 6/27/09, another AB General Ventilation manipulation created a pressure differential that caused the temporary doors to fail once again. One temporary door and a permanent steel door are now closed to ensure boundary operability.

Failures of the temporary doors were due to an inadequate design and insufficient interim actions after the first event to prevent another failure. This event is reported in accordance with 10 CFR 50.73 (a)(2)(i)(B) and (a)(2)(v)(C) and (D).

During normal plant operation on September 20, 2008, unexpected annunciator alarms were received in the control room indicating an automatic reactor trip based on a loss of electrical load. Subsequently, the control room was informed by a Nuclear Assistant Unit Operator that he had tripped open the Exciter Field Breaker, leading to the turbine trip and successive reactor trip.

The cause of this event has been determined to be human performance error in that the NAUO failed to recognize the need to utilize error reduction techniques when opening the exciter field breaker cabinet door. Corrective action has been taken to add the applicable information to operator requalification programs.

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)(B) and 10 CFR 50.72(b)(3)(iv)(A), respectively, as ENS notification 44506. This event is also being reported as this Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(iv)(A).

On January 27, 2006, engineering personnel identified a scenario involving a potential loss of cooling water to the Chemical and Volume Control System (CVCS) Seal Water Heat Exchanger during an Appendix R fire event. The scenario involves a loss of Component Cooling System (CCS) flow to the Seal Water Heat Exchanger due to fire damage. The loss of CCS flow results in a potential high suction temperature on the running CVCS Centrifugal Charging Pump (CCP) causing a loss of adequate suction head. During an Appendix R fire event, CCP suction is aligned to the Refueling Water Storage Tank (RWST), normal charging and letdown are isolated and the only makeup flow to Reactor COolant System (RCS) is via the Reactor Coolant Pump (RCP) seal injection flow path. If the Seal Water Heat Exchanger cooling is lost, the CCP recirculation flow and RCP seal return flow are not being cooled. The outlet of the heat exchanger combines with cool water from the RWST. The net result is that the CCP suction temperature could reach saturation temperature leading to pump cavitation. The temperature increase could be h gh enough to potentially damage both the CCP and the RCP seals, which would result in increased seal leakage and a potential loss of RCS inventory.

The cause of this event is a latent error in the WBN Fire Safe Shutdown Analysis. The original analysis did not evaluate the ramifications of not protecting cooling water flow to the CVCS Seal Water Heat Exchanger. Corrective actions include: 1) posting of roving fire watches in the areas affected, 2) procedure revisions to provide operator actions and 3) to issue a design change to reroute andor protect the identified vulnerable cables.

0NRC FORM 366 (6-2004) PRINTED ON REC"CLED PAPER

On October 3, 2003, Watts Bar Nuclear Plant Unit 1 was in Mode 6 during a refueling outage with core re-load in progress. At 1516 hrs (EDT), main control room personnel became aware that an activity involved with an upcoming test had inappropriately placed the B-Train Auxiliary Building Gas Treatment System (ABGTS) Exhaust Fan 460v breaker in the OPEN position. An operator was immediately dispatched and the breaker was closed at 1521 hrs, restoring the train to OPERABLE status. The opening of this breaker at about 1324 hrs caused the B ABGTS train to be INOPERABLE at a time when the A-train ABGTS was available to start but technically inoperable due to an unavailable emergency power supply (2A-A Emergency Diesel Generator). The ABGTS system is required for the mitigation of a postulated fuel handling accident. The immediate cause of the event was human error by a shift test director who did not realize the A-Train ABGTS was inoperable when he directed opening the subject breaker in support of equipment alignments for the upcoming test. Corrective actions include counseling of involved individuals on their awareness of plant conditions and need for adequate communications and adding a separate item to the Outage Schedule for preliminary alignments for 18 Month Blackout tests.

The safety significance of this event was low. � in an actual FHA, performance of site emergency procedures would have detected and quickly restored ABGTS Train B. Further, with off site power available, ABGTS Train A would have immediately responded to an event. Dose consequences assuming loss of all ABGTS were determined to remain with regulatory limits.

NRC FORM 368 (7

• 200il

On May 21, 2003, WBN Unit 1 was operating at 100 percent reactor power. At approximately 2153 EDT, the WBN operators were alerted to potential instrument channel III trouble when control room alarms were received. � This equipment provides channel III process instrumentation of the engineered safety features actuation subsystem portion of the Reactor Protection System. The operators were able to clear the alarms and plant continued operation. However, throughout the early morning of May 22, 2003, alarms indicating the same trouble continued.

Maintenance personnel performed troubleshooting and identified that a faulty power distribution panel located within Reactor Protection Set Channel III, Panel 1-R-9 was the source of the problem. The faulty power distribution panel was replaced and Panel 1-R-9 was returned to service.

Throughout this event, the operators did not identify Panel 1-R-9 as inoperable since after clearing the alarms, plant parameters were within range (with an exception of entry into LCO 3.4.1 "RCS Pressure, Temperature, and Flow DNB Limits due to reactor coolant pressure less than 2214 for a short period) and plant operation continued. � However, it was concluded after the event that the panel should have been declared inoperable and Technical Specifications actions completed within the 6 hour time requirements. This event is being reported under 10 CFR 50.73 (a)(2)(i)(B) as operation prohibited plant technical specifications.

The cause for this event was that there was no specific guidance to assist the operators In diagnosing a panel malfunction of the sort that would also affect operability of the Reactor Protection System panel. � Corrective actions include the development of an Abnormal Operating Procedure to address panel malfunctions. Additionally, the initial maintenance troubleshooting activities identified the problem to be in the data link handler portion of the rack which would not affect operability.

At 0824 EDT on September 27, 2002, Watts Bar Nuclear Plant (WBN) Unit 1 was operating at 100% power when 6.9 kV shutdown logic board panel 1A-A load stripping relay actuated. At 0842 EDT, the 6.9 kV shutdown logic board panel 1B-B load stripping relay actuated. These actuations occurred due to a loss of both offsite power lines which resulted in an automatic start and loading of both trains of Emergency Diesel Generators. At 0852 EDT, WBN declared a notification of unusual event (NOUE) due to the loss of both offsite power sources which resulted from a fire at the Watts Bar Hydroelectric Generating Plant (WBH). The WBN fire brigade was dispatched to fight the fire and remained there until the fire was extinguished. Since the brigade remained at the fire location and callout of additional brigade staffing took greater than two hours, 10 CFR 50.54(x) was invoked to address this departure from the minimum fire brigade staffing requirement of the WBN Fire Protection Report. WBN remained at 100% power with all four emergency diesel generators operating throughout the event. Offsite power was restored using an interim offsite configuration from Sequoyah and Rockwood lines. TVA evaluated and determined this configuration to be a Generic Letter (GL) 91-18 non-conforming condition that supports functionality and operability of both 161 kV preferred power sources per GDC 17. Condition D of LCO 3.8.1 was exited at 0125 EDT on September 28, 2002, when the first qualified offsite source was returned to service. WBN remained in Condition A of LCO 3.8.1 until the second qualified offsite source was returned to service at 0300 EDT. The NOUE was exited at 0308 EDT on September 28, 2002.

F RN1 "-2001

On July 13, 2002, at approximately 1622 EDT, while the plant was in Mode 1, at 100% power, Watts Bar Unit 1 experienced an automatic turbine/reactor trip when a C-Phase Main Transformer differential relay actuated This occurred because a bolted cable splice associated with a C-phase current transformer (CT) came into contact with the CT junction box; thereby shorting the differential relay protection circuit to ground.

The apparent factors contributing to this short circuit condition include temperature, cable splice material, vibration, and configuration of the splice inside the junction box.

All control rods inserted properly in response to the reactor trip. The Auxiliary Feedwater (AFW) System actuated in response to the trip, as designed. Plant response was in accordance with design with no complications. Operations shift personnel performance was in accordance with applicable procedures.

Subject to confirmatory laboratory testing, the root cause of this event was determined to be inadequate work instructions that allowed lower temperature rated tape to be used on a cable replacement and/or inadequate application of splice material. Corrective actions include revision and training on TVA's engineering and maintenance procedures for high temperature jacketing material, laboratory analysis of damaged splices, and reinspection and taping of similar vulnerable cable splices.

On March 1, 2002 at approximately 0258 hours, with Watts Bar Unit 1 in Mode 6 (Refueling) and RCS temperature at 100 degrees F, while attempting to realign the RHR system from RWST supply to RCS loop operation, operators isolated the common suction to the Residual Heat Removal (RHR) pumps on two occasions over a three minute span. At the time of the event, operators were performing full flow testing of ECCS lines in conjunction with reactor cavity fill.

The crew had not been briefed during shift turnover of other work activities which removed power from a rack which provided a permissive pressure switch signal to two valves which required manipulation during the realignment. The root cause of this event was inadequate work review and scheduling coupled with less than expected transfer of information, pre-job brief, and response to an emergent operating condition. Corrective actions to include: Counsel/coach individuals/groups involved, include this event in training, refine the scheduling and planning process by which work activities are tied to specific plant conditions or milestones and impacts are evaluated based on plant conditions and scheduled activities. (Voluntary Report)

On September 29, 2001, a blockage in an instrument sense line was discovered. A review of computer trend data for the Watts Bar Unit 1 #4 steam generator pressure transmitter loop, determined the channel may not have performed its design function, and thus had been inoperable without being placed in the "trip" position for about 8 hours and 40 minutes longer than the 6 hour period allowed by the ESFAS technical specifications.

Although the channel had been providing the correct pressure value to the ESFAS system and the control room, the blockage had slowed the time response to pressure changes beyond the time assumed in the FSAR.

The other two channels which were part of the two out of three logic circuit were operable during the time period to provide the required signal to the ESFAS system.

The clearing of the blockage corrected the immediate cause, and recurrence control actions included : Back filling of a sampling of the main steam pressure and flow transmitters during the upcoming refueling outage, and providing information to Operations and Engineering on this event and symptoms of a sense line blockage.