05000327/LER-2009-005
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. | |
| Event date: | 05-06-2009 |
|---|---|
| Report date: | 07-06-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3272009005R00 - NRC Website | |
I. PLANT CONDITION(S)
Unit 1 was in Mode 1 operating at approximately 100 percent power when the reactor trip was initiated.
II. DESCRIPTION OF EVENT
A. Event:
On May 6, 2009, at 2256 Eastern daylight time (EDT) with Unit 1 operating at 100 percent reactor power, the reactor was manually tripped because of imminent loss of steam generator (SG) [EIIS code AB] level induced by the closure of the Loop 1 main feedwater [EIIS code SJ] regulating valve. The control room received a steam flow/feedwater flow mismatch alarm. Operations placed the Loop 1 main feedwater regulating valve in manual and attempted to open the valve. Operations observed that there was still no flow to the Loop 1 SG. Operations initiated a manual trip. The unit was stabilized in 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:
Date Description May 6, 2009 Operations received various Loop 1 SG low feedwater and at 2256 EDT low level alarms. Loop 1 feedwater regulating valve was placed in manual in an attempt to control Loop 1 SG level.
May 6, 2009 Operations initiated a manual reactor trip because of a loss at 2256 EDT of feedwater flow to Loop 1 SG and imminent Loop 1 SG low level.
May 6, 2009 Unit 1 entered Mode 3.
at 2256 EDT
D. Other Systems or Secondary Functions Affected:
No other systems or secondary functions were affected by this event.
E. Method of Discovery:
The main control room received a steam flow/feedwater flow mismatch alarm.
F. Operator Actions:
Operations placed the Loop 1 feedwater regulating valve in manual and attempted to open the valve with no result. Loop 1 SG level was approximately 25-26 percent and decreasing rapidly. Operations performed a manual reactor trip because of the imminent loss of Loop 1 SG level. 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.
III.0 CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause was the failure of the Loop 1 feedwater regulating valve air operated diaphragm.
B. Root Cause:
The diaphragm failure is attributed to an improper clamping force of the diaphragm to the actuator stem. This insufficient clamping force was a result of insufficient torque applied to the fastening screw. The insufficient clamping force allowed stress concentrations at the diaphragm hole, which caused an initial tear in the diaphragm composite material and led to an instantaneous failure of the diaphragm.
The root cause of the equipment failure and subsequent reactor trip was determined to be that the governing vendor manual control procedure does not consider applicability to critical components. The vendor manual for the valves originally did not specify a torque value for diaphragm installation. The vendor updated the vendor manual to specify a torque value, but the TVA vendor manual program did not require checking for updates to non-safety related vendor manuals.
C. Contributing Factor:
Contributing factors include the absence of a torque specification in the vendor manual and maintenance instructions, which can lead to various interpretations of the instructions.
IV. ANALYSIS OF THE EVENT
The plant responded to the manual reactor trip as designed. The reactor coolant system (RCS) average temperature was at approximately 578 degrees Fahrenheit (°F) prior to the plant transient. Following the reactor trip, the loss of nuclear heat generation resulted in a decrease in RCS average temperature to approximately 535°F. As heat removal in the SG 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 after the reactor trip. RCS temperature then started to increase. RCS temperature remained within technical specification (TS) limits and was bounded by the Updated Final Safety Analysis Report (UFSAR) analysis.
The plant responded as expected for the conditions of the trip. No TS limits were exceeded and the UFSAR 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:
The valve was disassembled, the diaphragm replaced and returned to service.
B. Corrective Actions to Prevent Recurrence:
The vendor manual was revised to include the torque requirement for clamping the diaphragm to the actuator stem. Other corrective actions include revising the maintenance procedures to clearly define the torque requirements, revising the procedure governing the vendor manual to review vendor manual classifications, and replacing the diaphragms in the feedwater regulating and bypass valves on both Units.
VII. ADDITIONAL INFORMATION
A. Failed Components:
The failed component was a diaphragm in a Fisher Controls International Inc. Size 80 Type 667 actuator in the Loop 1 main feedwater regulating valve. The diaphragm failure was determined to be improper clamping force of the diaphragm plates that support and attach the diaphragm to the actuator stem. This insufficient clamping force was a result of insufficient torque applied to the fastening cap screw. The insufficient clamping force allowed stress concentrations at the diaphragm hole to be approximately three times nominal values, which caused an initial tear in the diaphragm composite material and led to the instantaneous failure of the diaphragm.
All equipment responded as required except the failed main feedwater regulating valve had a "not closed" indication in the main control room following a feedwater isolation signal.
B. Previous LERs on Similar Events:
A review of previous reportable events for the past 10 years indicated that there have been reactor trips regarding main feedwater regulating valves; however, there are no previous events that were a result of a diaphragm failure in a feedwater regulating valve.
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.