05000390/LER-2015-001
| Watts Bar Nuclear Plant, Unit 1 | |
| Event date: | 02-21-2015 |
|---|---|
| Report date: | 04-22-2015 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| ENS 50839 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation, 10 CFR 50.72(b)(3)(iv)(A), System Actuation |
| 3902015001R00 - NRC Website | |
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I. PLANT OPERATING CONDITIONS BEFORE THE EVENT
Watts Bar Nuclear Plant (WBN) Unit 1 was in Mode 1 at 100 percent rated thermal power (RTP). There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.
II. DESCRIPTION OF EVENT
A. Event
On February 21, 2015 at approximately 10:31 am Eastern Standard Time (EST), Watts Bar Nuclear Plant Unit 1 reactor [EIIS: AC] was operating at normal operating temperature and pressure when control room operators observed a rapid decrease in main condenser vacuum [EIIS: SG]. Due to the loss of main condenser vacuum and rising levels in the condenser hotwell, control room operators entered the appropriate response procedures and initiated a manual reactor trip at approximately 10:32 am EST. Subsequent to the reactor trip, the Auxiliary Feedwater [EIIS: BA] system actuated as designed, in response to isolation of main feedwater flow [EllS: SJ]. Control and Shutdown rods fully inserted, and required safety systems responded as designed. The unit was stabilized in Mode 3, with decay heat removal via Auxiliary Feedwater and the Atmospheric Dump Valves [EllS: JI] (ADVs) . The Main Steam Isolation Valves [EIIS: SB] were closed and remained closed during the event, and the station was maintained in a normal shutdown electrical alignment.
Main Control Room personnel responded appropriately to the plant transient using abnormal operating instructions which address loss of main condenser vacuum and rapid load reduction.
Operations entered: 1-A0I-11, "Loss of Condenser Vacuum," 1-E-0, "Reactor Trip or Safety Injection, ES-01, "Reactor Trip Response, and 1-GO-5, "Unit Shutdown From 30% Power to Hot Standby." Emergency and abnormal procedures were correctly followed, and the plant was placed in a stable condition in Mode 3.
Operations personnel confirmed that the plant response post trip was uncomplicated. Operations personnel, consistent with an uncomplicated shutdown, secured equipment including the following:
Condensate Circulating Pumps [EIIS: SD] and Raw Water Cooling Pumps [EllS: NN].
This event is reportable under 10 CFR 50.73(a)(2)(iv)(A).
B. Inoperable Structures, Components, or Systems that Contributed to the Event No inoperable structures, components, or systems contributed to this event.
C. Dates and Approximate Times of Occurrences Date Time Event Watts Bar Nuclear Plant, Unit 1 05000390 Replaced "C" condenser zone dog bone seal during scheduled refueling outage.
Operations entered 1-A0I-11 for loss of main condenser vacuum.
Turbine backpressure begins increasing in "C" zone of condenser, followed closely by "B" and "A" zones. Condenser Vacuum Lo/Lo-Lo alarm received in the main control room.
2006 N/A 2/21/15 10:31 am EST 2/21/15 10:31 am EST Date Time Event 2/21/15 10:32 am EST The plant was manually tripped due to loss of main condenser vacuum.
2/21/15 12:46 pm EST Event Notification 50839 was made to the NRC.
D. Manufacturer and Model Number of Components that Failed.
The "C" zone main condenser expansion joint boot seal failed at the splice joint. This seal was manufactured and installed by Keystone Rubber.
E. Other Systems or Secondary Functions Affected
There were no systems or secondary functions associated with this event.
F. Method of discovery of each Component or System Failure or Procedural Error The failure of the main condenser boot seal resulted in a loss of main condenser vacuum, requiring a manual reactor trip. The Tennessee Valley Authority (WA) investigation of the event revealed that the boot seal had failed on the "C" expansion zone. There were no procedural errors associated with this event.
G. Failure Mode and Effect of Each Failed Component Other than the main condenser expansion joint boot seal failure, there were no failed components associated with this event.
H. Operator Actions
This was an uncomplicated reactor trip. No special operator actions were required.
I. Automatically and Manually Initiated Safety System Responses The reactor was tripped manually on decreasing main condenser vacuum. All automatic and manually initiated safety systems responded as expected
III. CAUSE OF THE EVENT
A. The cause of each component or system failure or personnel error, if known.
The expansion joint boot seal on the "C" condenser to low pressure turbine failed, resulting in loss of condenser vacuum which led to a manual trip. The failure of the seal was due to a non-optimal vulcanization process and inadequately overlapped application which significantly weakened the seal at the splice joint area. Additionally, seal water is supplied to the boot seals to minimize air inleakage through the expansion joint. During the failure analysis conducted to determine the seal failure, it was noted that water had wicked into the polyester fibers of the boot seal, further weakening the splice joint area.
B. The cause(s) and circumstances for each human performance related root cause.
An organizational driver for this event was an inadequate risk assessment process for critical maintenance. Specifically, in 2006 (prior refueling outage where the boot seal was replaced) the procedure for oversight of supplemental personnel screening criteria/process was inadequate to ensure proper risk categorization for critical maintenance. The work document used for installation of the expansion joint boot seals in 2006 did not provide adequate site reviews or inspection points. This resulted in inadequate performance of the boot seal splice by the vendor.
Additionally, inspection of the splice area was not part of the condenser expansion joint boot seal inspection process, it would not have identified external degradation in this inspection.
IV. ASSESSMENT OF SAFETY CONSEQUENCES
A. Safety Significance
During the event and subsequent recovery actions, there was no loss of safety systems, structures or components. The Auxiliary Feedwater system started as expected and remained available to remove decay heat after the reactor trip. Other plant systems functioned as required following the manual reactor trip. Control room operators responded appropriately by manually actuating the reactor protection system to shut down the reactor. All control rods fully inserted into the core as designed to control reactivity and temperature of the core. The reactivity effects during this event had no impact on the safety of the core and thus, the event was determined to be of very low safety significance. There was no impact to the health and safety of the public or plant employees as a result of this event.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident Not applicable.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service Not applicable.
V. CORRECTIVE ACTIONS
This event was entered into the TVA Corrective Action Program (CAP) and is being tracked under problem evaluation report (PER) 991403.
A. Immediate Corrective Actions
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.
B. Corrective Actions to Prevent Recurrence
In August 2014, TVA implemented a new procedure, NPG-SPP-07.2.16, "Critical Maintenance Identification and Oversight," for outage-related work. This is a new process to add work controls to reduce the risk of introducing latent issues or operational failures due to improperly performed work. Since this is a new process for outage work, to prevent recurrence, outage work packages for the upcoming refueling outage (U1R13) for critical maintenance will be reviewed.
Additionally, Watts Bar will revise the inspection preventative maintenance document to add inspection of the splice along with guidance for inspection criteria in 0-MI-2.002, "Opening and Closing of the Main Condenser Shell and Hotwell," to include necessary inspection details and signoffs for the boot seal inspection by incorporating internal and external benchmarking.
Based on operating experience and the TVA's failure analysis, seal water supply to the boot seals to minimize air inleakage is not necessary and the system has been secured.
VI. ADDITIONAL INFORMATION
A. Previous similar events at the same plant
On August 17, 2001, TVA submitted LER 50-390/2001-001, "Manual Reactor Trip Due to Reduced Circulating Water Flow." This LER describes an event where during normal operation at 100 percent power, the unit was manually tripped due to a rise in back-pressure in the condenser.
This was caused by reduced condenser circulating water flow, which resulted from cooling tower fill material obstructing the intake flume screens to the circulating water pumps. However, the root cause was different; no approval was needed to deviate from design output documents, which allowed cooling tower fill repairs and/or replacement other than those issued in the design documents. This resulted in installed supports which failed, leading to this event.
B. Additional Information
None.
C. Safety System Functional Failure Consideration
This condition did not result in a safety system functional failure.
D. Scrams with Complications Consideration This reactor trip was not complicated.
VII. COMMITMENTS
None.