Abstract
Canadian Nuclear Laboratories has an on-going Research & Development program to support the development of a scaled–down 300 MWe version of the Canadian Super-Critical Water Reactor concept. The 300 MWe and 170–channel reactor core concept uses low enriched uranium fuel and features a maximum cladding temperature of 500 °C. Our goal is to test surface-modified zirconium alloys for use as fuel cladding. Zirconium alloys are attractive as they offer low neutron cross section thereby allowing the use of low enriched fuel. In this paper, we report on the results of general corrosion experiments used to evaluate chromium-coated zirconium-based alloys in the two chemistries (630 μg/kg O2 in both de-aerated and lithiated supercritical water). These experiments were conducted in a refreshed autoclave at 500 °C and 23.5 MPa. After exposure, the weight gain and the hydrogen absorption were examined. At adequate coating thickness, longitudinal and transverse coupons show similar corrosion behavior with improved corrosion resistance compared to uncoated coupons. The measured concentrations of hydrogen absorption are higher for the transverse coupons. Alkaline treatment resulted in higher weight gains than was found in pure oxygenated supercritical water.
