Preliminary Structural Study of Chromium Coatings for Nuclear Applications

TL;DR

This study investigates the high-temperature structural behavior of chromium-coated Zircaloy-4 claddings, highlighting oxide formation temperatures and microstructural changes relevant for nuclear safety applications.

Contribution

It provides the first detailed high-temperature analysis of Cr-coated Zircaloy-4 claddings, revealing oxide formation thresholds and microstructural evolution.

Findings

Oxide phases form above 200°C in uncoated samples

Cr-coated samples form oxides above 600°C

Fe segregation and Zr(Fe,Cr)2 phase observed after high-temperature exposure

Abstract

Following the Fukushima Daiichi disaster, an increasing number of studies concentrate on the development of Accident Tolerant Fuel (ATF) cladding materials for nuclear fuel, aiming to prevent the oxidation of zirconium during incidents such as Loss of Coolant Accident (LOCA) and to effectively lower the amount of heat and hydrogen released during emergency core cooling (ECC). 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 1100oC. The temperatures corresponding to the formation of oxide phases are >200oC and >600oC for the uncoated and Cr-coated samples, respectively. SEM and 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.