Methods for evaluating nontypical gate valve thrust requirements
- Idaho National Engineering Lab., Idaho Falls, ID (United States)
Evaluating rising stem wedge type gate valves is not a problem when the valves can be design-basis tested and their operability margins determined diagnostically. The problem occurs when they cannot be tested because of system limitations or when they can be tested only at some less-than-design-basis condition. To evaluate these valves requires various analytical methods by which the design-basis stem thrust requirement can be determined. This is typically done with valve stem thrust force models that are used to calculate the requirements or to extrapolate the results from less than a design-basis test. The stem thrust models used by the nuclear industry to determine the opening or closing stem thrust requirements for these gate valves have generally assumed that the highest load the valve experiences during closure (but before seating) is at flow isolation and during unwedging or before flow initiation in the opening direction. However, during full-scale valve testing conducted for the USNRC, several of the valves produced stem thrust histories that showed peak closing stem forces occurring before flow isolation in the closing direction and after flow initiation in the opening direction. All of the valves that exhibited this behavior in the closing direction also showed signs of internal damage. Initially, we dismissed the early peak in the closing stem thrust requirement as damage-induced and labeled it nonpredictable behavior.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research
- OSTI ID:
- 54592
- Report Number(s):
- NUREG/CP-0139; CONF-9410216-; ON: TI95001469; TRN: 95:012626
- Resource Relation:
- Conference: 22. water reactor safety information meeting, Bethesda, MD (United States), 24-26 Oct 1994; Other Information: PBD: Oct 1994; Related Information: Is Part Of Transactions of the twenty-second water reactor safety information meeting; PB: 129 p.
- Country of Publication:
- United States
- Language:
- English
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