Abstract
Oxide dispersion strengthened (ODS) ferritic and martensitic steels are promising materials for fuel cladding in fast reactors and for structural materials in fusion reactors. ODS steels for fuel cladding in fast reactors have been developed by Japan Nuclear Cycle Development Institute (JNC). The first generation ODS steels had a serious problem from anisotropy in their mechanical properties. Recently, this problem has been solved by controlling grain shape as equiaxed grains. Three types of modified ODS steels were irradiated in the experimental fast reactor JOYO for investigation of the irradiation-induced microstructural changes. The dispersed oxide particles were very stable when irradiated. In addition, dislocation structure development was minor. The high resistance to irradiation-induced microstructural changes in the ODS steels was attributed to oxide particles which acted as sinks for point defects.
Sign-in/Register to access full text options 
Free) 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a call
