Study of the stress corrosion cracking of zirconium alloys in iodine vapor and its relation with nuclear fuel element failures for pellet-cladding interaction

Fission product induced Stress Corrosion Cracking (SCC) susceptibility of Zirconium alloy cladding is a relevant attribute to be considered when evaluating nuclear fuel performance in power reactors. It is normally introduced in the codes through a typical parameter, namely the threshold stress needed to start a brittle fissure in normal reactor operating conditions. Iodine induced SCC of Zircaloy-4 cladding was measured at 350 and 400 C degrees by the pressurised-tube technique, method which permits an accurate control of the tensile hoop stress applied on the material. Materials with two different thermo-mechanical treatments were used, essentially differing in the number of roll steps used to arrive to tube finale dimensions. Threshold stress values of about 180-200 MPa were obtained in both cases at 400 C degrees, whereas the results at 350 C degrees have been strongly affected by the different fabrication routes. From the microscopic failure analysis carried out using optical and electronic microscopy, it was concluded that improper treatment may produce surface defects which can act as SCC initiators, giving way to failure at much lower applied stresses than in the case of well finished tubes