Neutron Irradiated Mechanical Properties of Some Rapidly Solidified Austenitic Stainless Steels

The development of several rapidly solidified Type 316 stainless steel alloys for possible application in the first wall of fusion reactors is described. One series is based on titanium additions leading to fine titanium carbide (TiC) precipitates, and another on oxide dispersion strengthening through addition of yttrium oxide. These materials were irradiated in the High Flux Isotope Reactor of the Oak Ridge National Laboratory at doses up to 34 displacements per atom (dpa) with associated high helium production of 3100 atomic parts per million (appm). Post-irradiation mechanical properties derived from the MIT Miniaturized Disk Bend Test are reported. It is concluded that post-irradiation yield strength is not a problem for this class of alloy, but helium embrittlement leading to intergranular fracture is a critical issue. This first generation of rapidly solidified alloys is found to behave similarly to ingot processed material; nevertheless, rapid solidification is expected to be a highly useful tool in future efforts to improve grain and grain boundary microstructures. Such improvements may lead to mitigation of radiation-induced helium embrittlement effects and other negative structural factors.