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For use as fuel cladding of liquid metal fast reactors, Fe-0.12C-9Cr-2W ODS martensitic steel claddings were developed by cold-rolling under the softened ferrite phase induced by slow cooling from austenite phase, subsequently by ferrite to austenite phase transformation to break up substantially elongated grains produced by cold-rolling at the final heat-treatment. The produced claddings showed noticeable improvement in tensile and creep rupture strength that are considerably superior to PNC-FMS and even austenitic PNC316 at higher temperature and extended time to rupture. The strength improvement is mainly attributed to titanium addition in ODS martensitic steels through its reduction of Y 2 O 3 particle size and shortening inter-particles spacing. The behavior of oxide particle size reduction is associated with stoichiometry between Y 2 O 3 and TiO 2 .

For use as fuel cladding of liquid metal fast reactors, Fe-0.12C-9Cr-2W ODS martensitic steel claddings were developed by cold-rolling under the softened ferrite phase induced by slow cooling from austenite phase, subsequently by ferrite to austenite phase transformation to break up substantially elongated grains produced by cold-rolling at the final heat-treatment. The produced claddings showed noticeable improvement in tensile and creep rupture strength that are considerably superior to PNC-FMS and even austenitic PNC316 at higher temperature and extended time to rupture. The strength improvement is mainly attributed to titanium addition in ODS martensitic steels through its reduction of Y 2 O 3 particle size and shortening inter-particles spacing. The behavior of oxide particle size reduction is associated with stoichiometry between Y 2 O 3 and TiO 2 .

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Swelling behavior and microstructural evolution of 12% cold-worked 316 SS hexagonal ducts following irradiation in the outer rows of EBR-II is described. Immersion density measurements and transmission electron microscopy (TEM) examination were performed on a total of seven irradiation conditions. The samples were irradiated to temperatures between 375 and 430°C to doses between 23 and 51 dpa and at dose-rates ranging from 1.3 × 10-7 to 5.8 × 10-7 dpa/s. Dose-rates and temperatures approach conditions experienced by a variety of components in pressurized water reactors (PWR's) and those which may be present in future advanced reactors designs. TEM analysis was employed to elucidate the effect of radiation on the dislocation, void and precipitate structures as a function of irradiation conditions. A moderate dose-rate effect was observed for samples which were irradiated at dose-rates differing by a factor of two. Lower dose-rate samples contained voids of larger diameter and typically swelled more in the bulk. The dislocation and precipitate structure was not visibly influenced by a dose-rate decrease.