05000328/LER-2007-001
| Sequoyah Nuclear Plant (Sqn) Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3282007001R00 - NRC Website | |
I. PLANT CONDITION(S)
Unit 2 was operating at 100 percent power when the reactor trip occurred.
II. DESCRIPTION OF EVENT
A. Event:
On January 23, 2007, at 1244 Eastern standard time (EST) with Unit 2 operating at 100 percent power, the reactor tripped as a result of low-low steam generator (EIIS code AB) level on Loop 2. The immediate cause was closure of the Loop 2 main feedwater regulating valve (EIIS code SJ) because of a failed control air line. The feedwater regulating valve's control air line was damaged resulting from improper routing of field tubing during a recent outage modification. During an attempt to place the bypass feedwater regulating valve in control in order to allow repair to the damaged control air line, the control air line to the main feedwater regulating valve broke. The main feedwater regulating valve failed closed as designed upon loss of control air.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Major Occurrences:
January 23, 2007 Operations was notified by Maintenance personnel of a at 10:30 EST significant air leak at the Loop 2 feedwater regulating valve.
January 23, 2007 Operations conducted a briefing for "dogging" Loop 2 at 11:30 EST feedwater regulating valve in accordance with system operating instructions to allow repair of the damaged control air line.
January 23, 2007 Operations placed the Loop 2 bypass regulating valve level at —12:35 EST controller in manual and began opening.
January 23, 2007 Entered Limiting Condition for Operation 3.7.1.6 while at 12:43 EST attempting to dog valve Loop 2 feedwater regulating valve.
January 23, 2007 Operations placed the Loop 2 bypass regulating valve level at —12:43 EST controller in automatic with bypass valve open.
January 23, 2007 Steam generator Loop 2 low level alarm was received.
at 12:43 EST January 23, 2007 The SRO directed that a manual reactor trip be initiated.
at 12:44 EST Steam generator Loop 2 low-low level automatic reactor trip occurred.
D. Other Systems or Secondary Functions Affected:
No other systems or secondary functions were affected by this event.
E. Method of Discovery:
Prior to the reactor trip, a Maintenance engineer performing a walk down noticed a control air leak on the Loop 2 feedwater regulating valve. Operations immediately began preparations to dog the valve so that the air leak could be repaired.
During the evolution to open and place the bypass regulating valve in automatic, a swing in control signal was seen and feedwater flow decreased rapidly.
F. Operator Actions:
After the SRO directed that a manual reactor trip be initiated, an automatic trip occurred.
Control Room personnel responded as prescribed by emergency procedures. They promptly diagnosed the plant condition and 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.
CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause of the event was closure of the Loop 2 main feedwater regulating valve as a result of a failed control air line. The closure of the feedwater regulating valve resulted in a reactor trip from low steam generator level.
B. Root Cause:
The feedwater regulating valve's control air line was damaged as a result of improper routing of field tubing during a recent outage modification. The routing of the control air line did not sufficiently account for movement of the valve due to thermal growth.
During an attempt to place the bypass feedwater regulating valve in control in order to allow repair to the damaged control air line, the control air line to the main feedwater regulating valve broke. The main feedwater regulating valve failed closed as designed upon loss of control air.
C. Contributing Factor:
There were no contributing factors.
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.
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 535 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 10 minutes after the reactor trip. RCS temperature then started to increase. 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. The Unit 2 Loop 2 feedwater regulating valve control air tubing was repaired and rerouted to ensure no interferences would result from thermal growth. All other feedwater regulation valves control air tubing was inspected and one other interference issue was resolved.
B. Corrective Actions to Prevent Recurrence:
Corrective actions include revisions to the conduct of modifications procedures to strengthen the constructability walk down process and thermal growth considerations in piping specifications and installation instructions.
VII.�ADDITIONAL INFORMATION
A. Failed Components:
Unit 2 Loop 2 feedwater regulating valve failed closed as a result of a control air tubing break.
B. Previous LERs on Similar Events:
A review of previous reportable events identified a similar event that resulted in a reactor trip that was initiated from a failed feedwater regulating valve control air line.
from a lack of programmatic controls for maintenance activities that affect vibration through system configuration changes. The January 23, 2007, Unit 2 event is similar in that the thermal movement of the valve was not adequately considered during the planning and implementation of the control air line configuration changes on the Loop 2 feedwater event would not have prevented this air line/fitting failure.
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. Loss of Normal Heat Removal Consideration:
This condition did not result in a loss of normal heat removal.
VIII. COMMITMENTS
None.