Corrosion Wall Loss and Oxide Film Thickness on SCWR Fuel Cladding

Abstract

The Canadian SCWR is a pressure-tube reactor concept with collapsible fuel cladding and a zirconium-alloy pressure boundary held at low temperature by a zirconia insulator. The only components exposed to high-temperature, high-pressure water and irradiated conditions are the metal-alloy fuel assembly and liner tube, which reside in the core for 3 fuelling cycles, for a total of 3.5 years before replacement. This design mitigates several materials performance challenges, including in particular, the conflicting needs for both metal alloy strength and neutron economy. While the metal alloys may experience irradiation-induced microstructural changes, their short time in the core limits these effects. As an example, recent neutron physics analyses peg fuel cladding damage at 10 dpa (displacements per atom) and helium generation at 50 ppm atomic for high-nickel alloys over 3.5 years, values within the realm of experience of Light Water Reactors (LWRs), but at higher temperature. Metal alloys used to fabricate the fuel assembly are therefore held to three key materials performance metrics: the lifetime wall loss due to corrosion, the likelihood of cracking, and the oxide thickness developed on the cladding between fuel cycles. The present paper describes the weight change data from static autoclave corrosion tests of Alloy 800H, Alloy 214 and AL6XN at 450 ºC and 25 MPa and evaluates them in terms of acceptability for the fuel cladding of the Canadian SCWR.