05000328/LER-2003-001
| Sequoyah Nuclear Plant (Son) Unit 2 | |
| Event date: | 01-01-2003 |
|---|---|
| Report date: | 02-26-2003 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3282003001R00 - NRC Website | |
I. � PLANT CONDITION(S) Unit 2 was in Mode 5, cold shutdown.
IL � DESCRIPTION OF EVENT
A. Event
On January 1, 2003, at 0626 Eastern Standard Time, a steam generator (SG) [EIIS Code SB] Loop No. 2 low-low level initiated a reactor trip signal. The reactor trip signal did not initiate any pump or valve operation because the unit was already in Mode 5, cold shutdown, and related equipment had been secured or was already in the required position. Before the event, it was determined that the secondary side of the SGs needed to be drained to support chemistry. in accordance with plant procedures, instrument maintenance technicians were dispatched to bypass the steam generator level reactor trip set points of the reactor protection system [E11S Code SJ]. However, the technicians inadvertently bypassed the SG low level alarms rather than the low-low level trips. As a result, a trip signal was generated upon draining the 5G.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Major Occurrences:
December 31, 2002, at 2200 EST December 31, 2002, at 2245 EST January 1, 2003, at 0513
EST
January 1, 2003, at 0626
EST
Operations requests maintenance personnel to ensure the high-high and low-low level functions are bypassed for each of the four SGs.
Maintenance personnel notify Operations that SG level trips are in bypass.
Operations begins to drain the Loop 2 SG.
A reactor trip signal was annunciated in the main control room.
Other Systems or Secondary Functions Affected:
None.
E. Method of Discovery:
The reactor trip signal annunciated on the main control room panels.
F. Operator Actions:
No operator action was required for the trip signal, control room operators promptly diagnosed the condition as being related to draining of the SG.
G. Safety System Responses:
The reactor trip signal did not initiate any pump or valve operation because the unit was already in Mode 5, cold shutdown, and related equipment had been secured or was already in the required position.
III. CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause of the event was the draining of the Loop 2 SG. Once the water level passed SG level set points, the reactor trip signal was generated.
B. Root Cause:
The root cause of the event was inadequate procedure; the information contained in the procedure was too generic. The procedure did not contain noun descriptions of the set point functions and the layout of the procedure suggested that SG trip set points were grouped together for each SG where actually the low-low level SG trip was separate from the other trip set points. This led to misinterpretation by the performers.
C. Contributing Factor:
Contributing to the event was that the supervisors incorrectly selected portions of the procedure for performance and did not verify and validate the selection. Additionally, inattention to detail and unawareness led the performers to miss opportunities to identify the error during procedure performance.
IV. ANALYSIS OF THE EVENT
The reactor protection system automatically initiates a reactor trip when any monitored variable or combination of variables exceeds a predetermined setpoint value. Set points are provided to envelope safe operating conditions with adequate margin for uncertainties to system, associated with the steam generator, is the low-low level setpoint. This function is associated with the water level necessary to preserve the steam generator heat sink for removal of long term residual heat from the reactor coolant system, including the fuel rods.
Should a complete loss of feedwater occur, the reactor would be tripped on low-low steam generator narrow range water level and the auxiliary feedwater pumps would start. This reactor trip acts before the steam generators are dry to reduce the required capacity and starting time requirements of auxiliary feedwater pumps and to minimize thermal transients on the reactor coolant system and steam generators. The reactor protection system and the low-low steam generator level trip functioned as described in the Final Safety Analysis Report.
V ASSESSMENT OF SAFETY CONSEQUENCES At the time of the event, the unit was in Mode 5 (cold shutdown) operation with the auxiliary feedwater pumps secured. Additionally, the low-low steam generator trip variable of the reactor protection system was not required for that mode of operation. Therefore, based on plant conditions and the above Analysis of The Event, this event did not adversely affect the health and safety of plant personnel or the general public.
VI. � CORRECTIVE ACTIONS
A. Immediate Corrective Actions:
The correct SG level set points were bypassed and work was completed without incident.
B. Corrective Actions to Prevent Recurrence:
The event was documented in the corrective action program and an action is in place for development of a procedure revision to add more detailed information. A review was performed of other similar instrument maintenance procedures for layout and description content problems. The review did not identify any other procedures with layout or description content problems.
The involved individuals were counseled on the use of error reduction tools and the requirements for following procedures. Briefings were conducted with appropriate instrument maintenance personnel to communicate event lessons learned.
VII. � ADDITIONAL INFORMATION
A. Failed Components:
None.
B. Previous LERs on Similar Events:
A review of previous reportable events for the past three years did not identify any similar events.
C. Additional Information:
None D. Safety System Functional Failure:
This event did not result in a safety system functional failure in accordance with 10 CFR 50.73(a)(2)(v),
VIII. COMMITMENTS
None.