05000390/LER-2008-004
| Watts Bar Nuclear Plant | |
| Event date: | 09-20-2008 |
|---|---|
| Report date: | 11-19-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3902008004R00 - NRC Website | |
I. PLANT CONDITIONS:
The events discussed in LER 2008-004 occurred on September 20, 2008 when the unit was at 100% power.
II. DESCRIPTION OF EVENT:
A. Event:
On September 20, 2008, Watts Bar Nuclear (WBN) Unit 1 was operating at 100% power. During rounds, the NAUO stopped and opened the exciter field breaker panel to show a trainee the breaker. While explaining that the breaker had to be manually aligned and pushed into the cubicle with a second party to push the de latching bar, the NAUO pushed the de-latching bar. This resulted in the breaker opening, a turbine trip, and a reactor trip. All systems responded as designed to their intended safety functions in response to the trip.
This event is addressed in TVA's Corrective Action Program as Problem Evaluation Reports (PERs) 152951, 152954, and 152955.
B. Inoperable Structures, Components, or Systems 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 (EDT)� Event September 20, 2008�0900 NAUO opened cabinet on Exciter Field Breaker and began showing trainee how to operate.
0903�NAUO depressed latching mechanism, causing breaker to open. This caused a turbine trip and subsequent automatic reactor trip.
D. Other Systems or Secondary Functions Affected The plant trip was normal except that the operators took manual control of the auxiliary feedwater (AFW) level control valves (EllS Code BA LCV) to control steam generator (EllS Code SG) level and reactor coolant system (El IS Code AB) cooldown. The valves controlled level properly in the manual state. The need for the operations staff to manually control AFW was previously documented in PER 150508.
E. Method of Discovery Control room personnel received annunciator alarms consistent with a reactor trip due to an open Exciter Field Breaker. The control room was also notified of the NAUO's actions by the NAUO.
F. Operator Actions The Operations staff (licensed personnel) responded to the event in accordance with all operations management and operations training management expectations. No human performance issues were identified during the response to the automatic trip.
II. DESCRIPTION OF EVENT (continued):
G. Safety System Responses
All systems responded as designed and as previously stated, the operations staff took manual control of AFW to regulate steam generator level.
III. CAUSE OF EVENT
The NAUO had lapses in judgment and violated sound operator fundamentals and procedures. The NAUO failed to meet the expectation to notify Shift Manager or use error reduction tools when opening the exciter field breaker panel for inspection. No time or schedule pressures existed that contributed to this event.
IV. ANALYSIS OF THE EVENT
Plant safety systems performed their intended safety functions in response to the automatic reactor trip. All control rods fully inserted into the core, and plant decay heat removal functioned properly. During the trip, the C feedwater heater string (EllS Code SJ HX) isolated. This was expected, per a previous Engineering Design Change (EDC) 52270. See Section V, "Assessment of Safety Consequences," below for further discussion.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The automatic reactor trip on 09/20/2008 can be compared to the Final Safety Analysis Report (FSAR) "LOSS OF ELECTRICAL LOAD AND/OR TURBINE TRIP", Updated FSAR (UFSAR) section 15.2.7. The reactor (EllS Code RCT) was automatically tripped at approximately 09:03 due to a loss of electrical load on the main generator (EllS Code TB) when the exciter breaker tripped open. Subsequently, a hi-hi level in heater C2 resulted in the C-heater string isolation. Therefore, the FSAR "LOSS OF NORMAL FEEDWATER" UFSAR section 15.2.8 is also applicable to this event. Main feedwater flow was isolated due to the low Tavg with Rx trip signal. The hi-hi C-2 heater isolation was caused by the bypass to condenser level control valves (LCVs) being closed during high turbine load operation in accordance with plant design to minimize water hammer.
The plant was stabilized using Auxiliary Feedwater and the Main Steam dump valves (El IS Code SB). The secondary side steam generator atmospheric relief valves (EllS Code RV) and safety valves did not operate during the transient. The Reactor Coolant System responded to the initial transient as expected with no pressurizer PORV relief, no safety injection initiation, and no steam generator PORV relief.
Therefore, the 09/20/2008 trip is bounded by the FSAR safety analysis assumptions.
VI. CORRECTIVE ACTIONS- The corrective actions for this condition are being managed within TVA's Corrective Action Program (PERs 152951, 152954, and 152955) and therefore are not considered to be regulatory commitments. An overview of the corrective action plan is provided below:
A. Immediate Corrective Actions
1. The NAUO was restricted from duty.
2. A crew stand-down was held to discuss expectations for use of Human Performance Tools.
3. A training brief was issued by the Operations Superintendent.
VI.�CORRECTIVE ACTIONS (continued) B.�Corrective Actions to Prevent Recurrence 1. A stand-down brief was held during the Initial Turnover Meeting upon each crew's first shift back to direct the rules for inspecting normal plant equipment for training purpose and ensure no hands-on manipulations occur without the proper procedures in hand, notifications of training being in progress, and Main Control Room concurrence.
2. A training brief on TI-12.10, "Control of Sensitive Equipment," was completed on each crew's first shift back.
3. The TI-12.10 procedure will be included in the Cycle 8 Non-Licensed Operator training.
4. Reactor trip warning labels will be revised to reference the TI.
5. The area around the sensitive equipment listed in TI-12.10 will be marked with a noticeable caution.
VII.�ADDITIONAL INFORMATION
A. Failed Components
None.
B. Previous LERs on Similar Events No similar LERs have occurred at Watts Bar.
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 5.
E. Loss of Normal Heat Removal Consideration There was no loss of normal heat removal due to this condition.
VIII. COMMITMENTS
None.