05000327/LER-2008-001
| Sequoyah Nuclear Plant Unit 1 | |
| Event date: | 01-16-2008 |
|---|---|
| Report date: | 03-04-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3272008001R00 - NRC Website | |
I. PLANT CONDITION(S)
Unit 1 was operating at 100 percent power when the reactor trip was initiated.
II. DESCRIPTION OF EVENT
A. Event:
On January 16, 2008, at 1925 Eastern standard time (EST), the Unit 1 reactor was manually tripped as a result of a low steam generator (El IS code AB) level on Loop 3.
Maintenance personnel were performing a transmitter calibration on a Loop 3 steam generator pressure transmitter. During performance of the calibration, Maintenance personnel made an incorrect assumption that a place-keeping aid placed in the procedure was intended to be the procedure start point when in fact they should have started two pages previous to the place-keeping aid location. The two missed pages resulted in Operations not being notified to remove the channel being tested as a controlling channel. This resulted in performing testing on the pressure transmitter on a controlling channel, causing closure of the Loop 3 main feedwater regulating valve (EllS code SJ). Operations placed the feedwater regulating valve in manual to open the valve. However, before steam generator water level began to increase, the pre established low steam generator level trigger-point was reached and the operating crew initiated a manual reactor trip.
Following the reactor trip, the plant systems responded as designed. The rods fully inserted as required. Immediate operator actions were performed and auxiliary feedwater (AFW) flows were adjusted to control the reactor coolant system (RCS) cooldown. The unit entered Mode 3 and an event investigation was initiated.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Major Occurrences:
January 16, 2008 Operations authorized performance of a calibration on steam at 1649 EST generator Loop 3 pressure transmitter.
January 16, 2008 Operations enter LCO 3.3.2 action 17 and LCO 3.3.3.7 action at 1734 EST 1 for performance of the calibration on steam generator Loop 3 pressure transmitter.
January 16, 2008 Operations initiated a manual reactor trip as a result of a low at 1925 EST steam generator level on Loop 3.
D. Other Systems or Secondary Functions Affected:
No other systems or secondary functions were affected by this event.
E. Method of Discovery:
The steam generator low-level alarm was annunciated in the Main Control Room.
F. Operator Actions:
Operations placed the feedwater regulating valve in manual to open the valve. However, before steam generator water level began to increase, the pre-established low steam generator level trigger-point was reached and the operating crew initiated a manual reactor trip. After the reactor trip, the operating crew took actions necessary to stabilize the unit in a safe condition and maintained the unit in hot standby, Mode 3.
G. Safety System Responses:
The plant responded to the reactor trip as designed.
Ill. CAUSE OF THE EVENT A. Immediate Cause:
The immediate cause of the event was closure of the Loop 3 main feedwater regulating valve as a result of an error during performance of a calibration procedure. The closure of the feedwater regulating valve resulted in a low steam generator level.
B. Root Cause:
The root cause was determined to be a failure to follow procedure because of personnel not performing proper place keeping during performance of a calibration procedure.
C. Contributing Factor:
Contributing to the event was an inadequate pre-job brief by Maintenance and Operations personnel and Operations allowed Maintenance to perform work on a controlling channel.
IV. ANALYSIS OF THE EVENT
The plant systems responded to the reactor trip as designed. The reactor coolant system (RCS) average temperature was near 578.2 degrees F prior to the loss of main feedwater to Loop 3 steam generator. When feedwater was lost, RCS average temperature made a slight increase before the reactor trip. Following the reactor trip, the loss of nuclear heat generation resulted in a rapid decrease in RCS average temperature to 548 degrees F. As heat removal in the steam generators decreased as a result of increased steam pressure, the decrease in RCS temperature slowed. The introduction of cold auxiliary feedwater (AFW) resulted in a slower, but continued reduction in RCS temperature until AFW flow was reduced about 5 minutes after the reactor trip. RCS temperature then started to increase.
The minimal RCS temperature was 543 degrees F. RCS temperature remained within Technical Specification limits and bounded by the Safety Analysis Report (SAR) analysis.
The plant responded as expected for the conditions of the trip. No Technical Specification limits were exceeded and the SAR analysis of this event remained bounding.
V. ASSESSMENT OF SAFETY CONSEQUENCES
Based on 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:
Control Room personnel responded as prescribed by emergency procedures. They diagnosed the plant condition and took action necessary to stabilize the unit in a safe condition.
B. Corrective Actions to Prevent Recurrence:
Corrective actions include establishing and proceduralizing standard place-keeping requirements, strengthening and proceduralizing standard pre-job brief requirements, and training on these standard requirements.
VII. ADDITIONAL INFORMATION
A. Failed Components:
None.
B. Previous LERs on Similar Events:
A review of previous reportable events did not identify any previous 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).
E. Unplanned Scram with Complications:
This condition did not result in an unplanned scram with complications.
VIII. COMMITMENTS
None.