Dispersion strengthened ferritic steels as fast reactor structural materials

Dispersion-strengthened ferritic steels are being studied for possible use as canning material for sodium-cooled fast reactors. The basic alloy chosen contains nominally Fe - 13% Cr - 1. 5% Mo - 3. 5% Ti to which 2% TiO//2 or 1% Y//2O//3 is added by a powder metallurgy technique. At 700 C, the alloys studied can favorably be compared in stress rupture tests (up to 12,000 h) to the best austenitic steels. Corrosion tests in dynamic sodium at 700 C showed that after 4000 h the affected zones remained narrow and had no significant influence on the mechanical resistance at high temperature. Neutron irradiation of these alloys demonstrated their remarkable resistance to embrittlement in mechanical tests at 700 C. Comparison with other alloys showed that they had the highest elongation to rupture after irradiation. Simulation tests by 1-MeV electrons gave almost zero swelling in the temperature range of 475 to 700 C. The combined properties of dispersion-strengthened ferritic alloys make them excellent candidates not only for canning material but also for shroud tubes for fast-reactor fuel elements.