Corrosion characteristics of zirconia coated zirconium alloy in lithiated water

摘要

After the Fukushima-Daiichi accident, the de velopment of accident tolerant fuel cladding materials to improve reactor safety has become a hot topic in the field of nuclear industry. ZrO 2 has a satisfactory neutron economy and can guarantee the fission chain reaction process, which enables it to be a promising coating for zirconium alloy cladding. Maintaining a good corrosion resistance in primary coolant loop during normal operations of Pressurized Water Reactors is a prerequisite for ZrO 2 as a protective coating on zirconium alloy cladding. Research on corrosion performance of ZrO2 coating in nuclear water chemistry is relatively scarce, and existing reports failed to provide an in-depth explanation for failure causes of ZrO2 coating. Herein, a detailed corrosion process of ZrO 2 coating in lithiated water at 360 ° C and 18.5 MPa was proposed based on experimental research and molecular dynamics simulation. The protective effect and failure mechanism of ZrO 2 coating on Zry -4 under varied LiOH concentrations was further revealed. ZrO 2 coating provided a favorable corrosion protection with the occurrence of localized corrosion at low LiOH concentrations. Factors influencing corrosion resistance mainly include pitting corrosion extension, enhanced Li+ permeation, short-circuit diffusion of O2- and ZrO2 phase transformation. In highly-concentrated LiOH solutions, intergranular corrosion, internal oxidation and perforation resulted in coating failure. Zr ions were released to coating surface to form flocculent ZrO2 and ZrO2 clusters due to the strong diffusion and dissolution tendency of α-Zr in the Zry-4 substrate. This work can provides some references to understand service behavior of nuclear coatings under variable water chemistry conditions and promote in-pile application of ZrO2 coating.