Preliminary Structural Study of Chromium Coatings for Nuclear Applications

Following the Fukushima Daiichi disaster, an increasing number of studies concentrate on the development of accident tolerant fuel cladding materials for nuclear fuel, aiming to prevent the oxidation of zirconium during incidents such as loss of coolant accident and to effectively lower the amount of heat and hydrogen released during emergency core cooling. Zirconium alloy cladding with a protective chromium (Cr) coating is considered one of the promising candidates, largely due to its relatively short timeline for deployment in nuclear power plants. In this study, Cr‐coated and uncoated Zircaloy‐4 claddings were evaluated using high‐temperature X‐ray diffraction (HT‐XRD) in vacuum over a temperature range from RT to 1100°C. The temperatures corresponding to the formation of oxide phases are > 200°C and > 600°C for the uncoated and Cr‐coated samples, respectively. scanning electron microscopy (SEM) and transmission electron microscopy (TEM) characterisation of the sub‐surface in Cr‐coated specimen after HT‐XRD, revealed Fe segregation, formation of Zr(Fe, Cr) 2 Laves phase and nano‐bubbles at the former Cr/Z4 interface.