Development of fuel cladding materials is crucial for high burnup (more than 100GWd/t) operation of advanced nuclear energy systems, such as advanced light water reactors and supercritical pressurized water (SCPW) reactors. In order to overcome the requirements for the fuel claddings, materials R&D have been performed for high-Cr oxide dispersion strengthening (ODS) steels. Corrosion tests were performed in a SCPW (783 K, 25 MPa) environment. The weight gains of all high-Cr ODS steels are smaller than that of austenitic stainless steel (SUS316L). More uniform and thinner oxidation layers were observed in the ODS steels after corrosion test rather than in 9Cr martensitic steel and SUS316L The effects of neutron irradiation on the mechanical properties of the ODS steels have been investigated. High-Cr ODS steels showed a significant hardening after the irradiation at 290 and 400°C, while no effect was observed after the irradiation at 600°C. The irradiation hardening, however, was not accompanied by the reduction of total elongation. The nano-oxides of the 19Cr-ODS steel were cubic pyrochlore Y2Ti2O7 while those of 19Cr-4Al-ODS steel were mainly perovskite AlYO3 of which the difference can account for the difference in the tensile strength between the steels. The microstructure observation after heavy ion irradiation revealed that the dispersed oxides were stable up to a dose of 150 dpa at 973K. The average size and number density of cavities formed in the ODS steels were twice as small and two orders of magnitude higher density than those in the reduced activation ferritic (RAF) steel, resulting that the ODS steels had superior resistance to swelling. The particle diameter and its size distribution range decreased gradually with increasing mechanical milling time up to 12 h and then increased drastically thereafter.