Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5. Probabilistic fracture mechanics analysis. Load Combination Program Project I final report
The primary purpose of the Load Combination Program covered in this report is to estimate the probability of a seismic induced LOCA in the primary piping of a commercial pressurized water reactor (PWR). Best estimates, rather than upper bound results are desired. This was accomplished by use of a fracture mechanics model that employs a random distribution of initial cracks in the piping welds. Estimates of the probability of cracks of various sizes initially existing in the welds are combined with fracture mechanics calculations of how these cracks would grow during service. This then leads to direct estimates of the probability of failure as a function of time and location within the piping system. The influence of varying the stress history to which the piping is subjected is easily determined. Seismic events enter into the analysis through the stresses they impose on the pipes. Hence, the influence of various seismic events on the piping failure probability can be determined, thereby providing the desired information.
- Publication Date:
- OSTI Identifier:
- Report Number(s):
- NUREG/CR-2189-Vol.5; UCID-18967-Vol.5
- DOE Contract Number:
- Resource Type:
- Technical Report
- Research Org:
- Lawrence Livermore National Lab., CA (USA); Science Applications International Corp., La Jolla, CA (USA)
- Country of Publication:
- United States
- 22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; LOSS OF COOLANT; DYNAMIC LOADS; PIPES; FAILURES; FRACTURE MECHANICS; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; SEISMIC EFFECTS; CRACKS; REACTOR SAFETY; STRESS ANALYSIS; WELDED JOINTS; ACCIDENTS; COOLING SYSTEMS; ENERGY SYSTEMS; JOINTS; MECHANICS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; SAFETY; WATER COOLED REACTORS; WATER MODERATED REACTORS 220900* -- Nuclear Reactor Technology-- Reactor Safety; 210200 -- Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled
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