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Preliminary Analysis of the Effect of Fatigue Loading and Crack Propagation on Crack Acceptance Criteria for Nuclear Power Plant Components

机译:浅析疲劳载荷和裂纹扩展对核电厂部件裂纹接收标准的影响

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A preliminary study was initiated to evaluate the conservatism of existing design methods regarding cyclic loadings and crack growth in nuclear power plant components. The study was based on the assumption that the application of the ASME Boiler and Pressure Vessel Code (the Code) should be consistent throughout the design, operation, and inspection phases. Specifically, any undetectable or allowable crack subjected to fatigue stress levels permitted by Section III of the Code should not be expected to grow larger than the size permitted by Section XI in-service inspection criteria. The objective of this analysis was to estimate the magnitude of acceptable initial crack sizes under various conditions and to identify some important parameters that may be used in the development of crack acceptance standards.

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