Thermal fatigue of a 304 L type steel

Abstract Various components of nuclear reactors are submitted to very sharp thermo-mechanical loadings. Thermal fatigue cracking has been clearly detected in auxiliary loops of the primary cooling circuit of Pressurized Water Reactors (PWRs). The study presented here is focused on the 304 L type stainless steel used in PWRs. The thermal fatigue behaviour of this steel has been investigated using a specific thermal fatigue test equipment called SPLASH. This test equipment allows the reproduction of multiple cracking networks similar to those detected during inspections. The present study deals with two points : i) the experimental determination of crack initiation conditions and the morphological description of the growing crack networks; ii) the multiple crack growth numerical simulation, using a Skelton model, and a generalized Paris law. This modelling, in spite of simplified assumptions, gives predictions in good agreement with observations, as far as the evolution of the mean and deepest cracks during cycling are concerned.