Interaction among dislocation and complex oxide particles in ODS steels heavily-irradiated at high temperature

Yamashita, Shinichiro   ; Akasaka, Naoaki; Onose, Shoji 

Oxide Dispersion Strengthened (ODS) ferritic steel dealt with this study was a MA957 (Fe-0.015C-14Cr-0.3Mo-1.0Ti-0.25Y2O3). The objectives of this study were to understand oxide particle stability of ODS steel during irradiation and interaction among dislocation and oxide particles, reflecting to advanced nuclear reactor design of next generation. Development of some nuclear energy generating systems has been proposed and supported intensively under several international collaboration programs (Generation IV International Forum (GIF), Advanced Fuel Cycle Initiative (AFCI), International Nuclear Energy Research Initiative (I-NERI) etc).Current research issue on ODS ferritic steels is considered to be poverty of experience and understanding on their practical neutron-irradiation behaviors at the temperature higher than 600C.In this research, a MA957, most familiar but primitive 14CrODS ferritic steel contained the highly textured-anisotropic grain structures, was irradiated at 500-700C to fast fluences ranging from 19.8 to 20.8 1026 n/m2 (E 0.1MeV) in the experimental fast reactor JOYO. The dose achieved varied from 99 to 104 dpa. TEM observation and micro-hardness measurement were carried out to clarify the irradiation effects on microstructural evolution of 14CrODS ferritic steel at elevated temperature and high dose. Microstructural examination revealed that all of the highly textured- anisotropic grain structures, following heavy irradiation at the temperature above 600C, have not changed. In addition, large regions in all specimens have retained high dislocation density, contained negligible cavitation.