Abstract
Reliable operation of existing and new light water reactors over extended lifetimes of 60 - 100 years may be best achieved with improved materials, and this program is designed to evaluate candidate materials for use both in fuel cladding and structural applications. The rapid corrosion of zircaloy fuel cladding at the Fukushima Daiichi plants produced large quantities of hydrogen, which was responsible for explosions that damaged the plants and further hampered the efforts to re-establish adequate cooling to the reactors. Structural materials in current use, including standard austenitic stainless steels for the core shroud and piping in boiling water reactors, have experienced on-going issues with stress corrosion cracking, causing increased inspections, repairs and replacements. Some of the same alloys that are strong candidates for fuel cladding are also good candidates in structural applications. The reduced nickel content of ferritic “stainless steels” translates to lower cost and lower activation, and these alloys also possess superior resistance to radiation damage and stress corrosion cracking. This paper will summarize the data obtained to quantify the SCC growth rate response of a range of ferritic steels being considered as candidate alloys for fuel cladding and structural components.
