|
Development of Oxide Dispersion Strengthened Ferritic Steel Claddings for Sodium-Cooled Fast Breeder ReactorSep. 19, 2006Superior radiation resistance and high temperature capability are essential to fuel pin cladding material, which are exposed to peak burn-up of 250 GWd/t and peak neutron dose of 250 dpa at high temperature up to 973 K for sodium-cooled fast reactor fuels in severe. Oxide dispersion strengthened (ODS) ferritic steel is the most prospective candidate cladding material for the sodium-cooled fast reactor. Although precipitation hardened ferritic steels are well known to have excellent radiation resistance, they have inferior high temperature strength compared with the conventional austenitic stainless steels. So their utilization is limited to around 600C. In order to overcome such disadvantage, ODS steels has been developed since 1987 in Japan Atomic Energy Agency (JAEA). The ODS steels strengthened by the Y2O3 particles tend to be too hard to manufacture cladding by the cold-rolling. Also, the manufactured ODS steel claddings will have inferior creep strength and ductility loss in the cladding hoop direction if their grains extensively grows toward a rolling direction. These technical problems have been successfully solved through a grain structure control by means of α to γ phase transformation with appropriate intermediate heat treatments. In order to confirm and demonstrate the ODS fuel pins integrity to high burnup conditions, the irradiation tests in BOR-60 have been conducted under the collaborative works between JAEA and Research Institute of Atomic Reactors (RIAR) in Russia. The first irradiation test had been already completed to the burnup of 50GWd/t and neutron dose of 21 dpa. In addition, the ODS fuel pins irradiation tests in JOYO MK-III core are also scheduled. These irradiation tests data will be applied to the licensing of ODS driver fuels in MONJU up-grade core.
•Target of ODS Ferritic Steel Cladding Development |
