Validation of crack-like defect fatigue analysis

Abstract French nuclear construction codes, RCC-M [1] for Pressurized-Water Reactors (PWRs) and RCC-MR [2] for Liquid-Metal Reactors, propose design methodologies to appraise the fatigue damage risk in the sharp notches tip or crack-like defects. This paper continues those presented by Autrusson et al. (1988) and R. Roche (1990), to follow the experimental validation of these analysis methods, aiming to predict initiation of cracks in crack-like defects existing on start-up pressure vessel components. The comparison of analysis results with experimental tests permits to propose some modifications in order to obtain a satisfactory correlation with the classical fatigue curve relevant for the material.