05000390/LER-2006-005
| Watts Bar Nuclear Plant | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation |
| 3902006005R00 - NRC Website | |
I. Plant Conditions:
On July 31, 2006, WBN Unit 1 was in Mode 1 at approximately 100 percent reactor thermal power. The operating temperature was 588 degrees F and Reactor Coolant System (RCS) (Energy Industry Identification System (EIIS) Code AB) pressure was 2235 psig.
II. Description of Event:
A. Event:
On July 31, 2006 at approximately 12:13 EDT, the control room unexpectedly received an exciter field overcurrent alarm, followed immediately thereafter ( rods (EIIS Code ROD) inserted and the Auxiliary Feedwater System (EIIS code BA) auto started as designed. The plant was stabilized in Mode 3. The plant trip was normal except for Reactor Coolant Pump (RCP) No. 2 (EIIS Code AB/P) which did not auto transfer back to its start bus (EIIS Code BU).
As a result of the plant trip, the actuation of the Reactor Protection (EIIS code JC) 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).
B. Inoperable Structures, Components, or Systms that Contributed to the Event There were no additional structures, components or systems inoperable at the start of the event that contributed to the event.
C. Dates and Approximate Times of Major Occurrences Date � Time � Event � July 31, 2006 - 12:12 � Unit 1 Reactor at Full Power - All conditions normal � - 12:13 � Exciter Field Overcurrent Alarm � - 12:13 � Reactor Trip/Turbine Trip � - 12:13 � Generator Trip � - 12:13 � Entered in Mode 3 D. Other Systems or SeCondary Functions Affected The plant trip was normal except for Reactor Coolant Pump (RCP) No. 2 which did not auto transfer back to its start bus.
A manual attempt at reclosure by the operator was unsuccessful. Investigation found, based on interview of the craft .
performing the repair work, that the "anti pumping" relay (52Y) on the RCP supply breaker (EIIS Code BKR) did not have the minimum spacing between the armature tail piece and the molded posts to assure free movement of the armature. The craft corrected this problem by providing more clearance between the contact mounting posts and the armature. Contact wear may also be a contributing factor. Subsequently, RCP No. 2 was returned to service.
E. Method of Discovery
As described above, this condition was first identifed when the Exciter Field Overcurrent Alarm sounded.
F. Operator Actions
Crew response to the event was timely and met Operations and Training management expectations. There were no human performance issues.
G. Safety System Responses
Upon reactor trip, Auxiliary Feedwater System started as designed.
CAUSE OF EVENT
The Kepner-Tregoe (KT) analysis indicated that the most likely cause of the trip was an intermittent failure in the automatic controls of the excitation system. The most likely cause was determined to be a failed Maximum Excitation Limiter (MXL) circuit board in the main generator automatic excitation control circuitry. Data obtained from the plant annunciator system and the main generator protective devices from the time of the trip indicate that the plant tripped due to a loss of the main generator exciter field.
IV. ANALYSIS OF THE EVENT
Plant safety systems functioned normally in response to the reactor trip. All rods inserted fully. Steam generator level was maintained initially via the Auxiliary Feedwater system. See Section V, 'Assessment of Safety Consequences," below for further discussion.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The automatic reactor trip on July 31, 2006 can be compared to the FSAR Loss of External Electrical Load and/or Turbine Trip in Update Final Safety Analysis Report section 15.2.7 (page 15.2-21). The automatic trip at approximately 12:13 occurred as a result of secondary side electrical problems. The alarm printout indicated exciter overcurrent which may have been due to either problems in the voltage regulator circuit or a fault in the generator bus to the main transformers. The plant trip was normal except for RCP No. 2 did not auto transfer to the start bus. A manual attempt at reclosure was unsuccessful and the RCP No. 2 breaker was subsequently repaired. The plant was stabilized using steam dumps. The reactor coolant system responded to the initial transient as expected with no pressurizer power operated relief valve (PORV) (EllS Code PCV) actuation, no safety injection initiation, and no steam generator atomspheric dump valve actuation.
The UFSAR 15.2.7 analysis contains several analysis conservatisms which were not characteristic of the actual event. The UFSAR analysis assumes the reactor trip is based on a reactor protection system trip setpoint exceedance and in this case was initiated by a secondary side turbine trip. In addition, reactor control is assumed to be in manual, no credit is taken for the steam dump system, and no credit is taken for the steam generator atmospheric relief valves (only steam generator safeties are credited). The actual event had automatic rod control and steam dumps available. The UFSAR analysis demonstrates for two cases (DNB case where credit is taken for the pressurizer PORVs and spray, and RCS overpressure case where no credit is taken for the pressurizer spray or PORVs) that the minimum DNBR is well above the limiting value and that the RCS pressure safety analysis limits are met.
Therefore based upon the above, the actual event is bounded by the UFSAR safety analysis assumptions.
V. CORRECTIVE ACTIONS
A. Immediate Corrective Actions
Operators responded to the plant transient in accordance with appropriate plant procedures. An event team was assembled to investigate the cause of the event.
B. Corrective Actions to Prevent Recurrence (TVA does not consider these items to constitute regulatory commitments.
TVA'scorrective action program tracks completion ofthese actions.) Initial actions taken to identify the cause included: (1) a complete checkout of the voltage regulator per vendor instructions, (2) a crawl through inspection of the isophase bus, (3) megger testing of the main generator and exciter, (4) inspection and testing of the main generator potential transformer and associated fuses, (5) testing of the exciter ground detector circuit, (6) electromagnetic and radio frequency interference diagnostic testing to detect intermittent arcing, and (7) various informal walk downs. None of these actions identified a failed component.
With the most likely cause determined to be in the main generator automatic excitation control circuitry, a decision was made to return the plant to service and to monitor 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 full buck which would ultimately result in a loss of the exciter field which was the case during the plant trip. 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 wring caused the failure.
VI. ADDITIONAL INFORMATION
A. Failed Components
As discussed previously, the most likely cause was determined to be a failed Maximum Excitation Limiter (MXL) circuit board originally manufactured by Siemens-Westinghouse part of voltage regulator model WTA.
B. Previous LERs on Similar Events A review was performed of the previous WBN Licensee Event Reports (LERs) for any events associated with generator trips involving failed electronic cards. Although WBN has experienced events associated with failed circuit cards, there were no previous LERs identified which were associated with main generator automatic excitation control circuitry which caused a turbine/reactor trip.
C. Additional Information:
None.
D. Safety System Functional Failure This event did not involve a safety system functional failure as defined in NEI 99 02, Revision 4.
E. Loss of Normal Heat Removal Consideration This event is not considered a scram with loss of normal heat removal.
VII. COMMITMENTS
None