Methods for determining atypical gate valve thrust requirements
- Idaho National Engineering Lab., Idaho Falls, ID (United States); and others
Evaluating the performance of rising stem, wedge type, gate valves used in nuclear power plant 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 plant system limitations or when they can be tested only at some less-than-design-basis condition. To evaluate the performance of these valves requires various analytical and/or extrapolation methods by which the design-basis stem thrust requirement can be determined. This has been typically accomplished with valve stem thrust models used to calculate the requirements or by extrapolating the results from a less-than-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. Opening responses were not a priority in our early research, so that phenomenon was set aside for later evaluation.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab. (BNL), Upton, NY (United States)
- OSTI ID:
- 93882
- Report Number(s):
- NUREG/CP-0140-Vol.3; ON: TI95011768; TRN: 95:005142-0011
- Resource Relation:
- Other Information: PBD: Apr 1995; Related Information: Is Part Of Twenty-second water reactor safety information meeting: Proceedings. Volume 3: Primary systems integrity; Structural and seismic engineering; Aging research, products and applications; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 269 p.
- Country of Publication:
- United States
- Language:
- English
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