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Various components of nuclear reactors are submitted to very sharp thermomechanicalloadings. Thermal fatigue cracking has been clearly detected in auxiliary loopsof the primary cooling circuit of Pressurized Water Reactors (PWRs). The study presentedhere is focused on the 304 L type stainless steel used in PWRs. The thermal fatiguebehavior of this steel has been investigated using a specific thermal fatigue facility called„SPLASH test equipment”. This test equipment allows the reproduction of multiplecracking networks similar to those detected during in-service inspections. The present studydeals with the three following points:i) the experimental determination of crack initiation conditions and the morphologicaldescription of the growing crack networks;ii) the growth and shielding effect between cracks of the networks tested under four pointbending fatigue crack growth conditions;iii) the multiple crack growth numerical simulation, using Skelton’s model, and ageneralized Paris’ law. In spite of simplified assumptions, this modelling, gives predictionsin good agreement with observations, as far as the evolution of the mean and deepestcracks during cycling are concerned.