Abstract
The use of zirconium alloys such as Zircaloy-4 in nuclear reactors is limited by brittle hydride formation, for example in harsh reactor operating conditions such as those in the pressurized water reactors. Even though much research has studied this phenomenon, a more systematic, in-depth study of the effects of hydrides on the properties of Zircaloy-4, using various analytical techniques, can add more details to the understanding of these effects. In the current study, hydrogen charging was performed on Zircaloy-4 at different levels compatible with commercial reactor operating conditions and conditions in accident scenarios. Effects of hydriding on the Zircaloy-4 (α-(hcp)-Zr) tensile properties, together with an in-depth study of any relationship between tensile and the microstructural properties of the material, are discussed. While a moderate pinning effect on the equiaxed α-Zr grain size was observed, no significant change in bulk texture of the Zry-4 was revealed as the hydrogen level increased up to nominally 1230 wppmH. At a nominal value of 490 wppmH, Zircaloy-4 fractured at a low strain value just above the yielding point, and brittle tensile fracture was observed in a sample H-charged to nominally 1230 wppmH. The low ductility and brittle fracture of the samples were attributed to the presence of brittle inter-grain boundary hydrides. It was observed that brittle fracturing was also supported by intra-grain boundary hydrides that were affected by the tensile deformation of the sample.
