Prediction of stress corrosion cracking (SCC) in nuclear power systems

Abstract: This chapter addresses the phenomenon of stress corrosion cracking in light water reactors, and specifically the prediction of the crack propagation rate in stainless steel components in boiling water reactors (BWR). Attention is focused on the various approaches that may be used for life prediction. These vary from analyses of plant incidents, to empirical correlations between the propagation rate and various engineering parameters in laboratory experiments, to algorithms based on knowledge of the processes and mechanism of crack propagation. The chapter concludes with a comparison between prediction and observation for cracking of stainless steels in both unirradiated and irradiated BWR components.