Microstructure analysis of Kr+ irradiation and post-irradiation corrosion of modified N36 zirconium alloy

The irradiation behaviors and corrosion properties of a modified N36 zirconium alloy with the composition of Zr-0.8Sn-1Nb-0.3Fe, developed by Nuclear Power Institute of China, were investigated by transmission electron microscopy and focused ion beam. The polished samples were irradiated by 400 keV Kr+ ions up to 25 dpa at 360 °C using a NEC 400 kV ion implanter. The as-received and irradiated samples were corroded for 14 days at the water-vapor environment with 10.3 MPa and 400 °C. The krypton gas bubbles were formed in zirconium matrix and their size was increased with increasing ion dose. Meanwhile, a model that related with gas bubble size and displacement damage had been established. After the corrosion, a layer composed of zircona with different stoichiometric composition was formed on the sample surface. The higher the displacement damage was, the thicker the corrosion layer would be. An empirical equation between oxide thickness and displacement damage was provided. From sample surface to matrix inner, the oxygen content was decreased with increasing corrosion depth. Correspondingly, the zircona was changed from ZrO2 with monoclinic structure on the sample surface to the mixtures of ZrO2 with tetragonal structure and ZrO2 with monoclinic structure in the middle of corrosion layer, and then to ZrO2 with tetragonal structure near alloy matrix.