Self healing of early stage radiation damage by site selective solute segregation is a promising approach to extend the lifetime of nuclear reactor components. In the present study, the creation and autonomous healing of irradiation-induced damage is investigated in pure Fe and high purity Fe–Au and Fe–Cu model alloys. To create radiation damage samples are irradiated at 550 °C by 120 keV He+ ions with fluences of 5.0 × 1015, 1.0 × 1016 and 5.0 × 1016 ions/cm2. The observed increase in the S and W parameters determined in the variable energy positron annihilation spectroscopy measurements indicates the formation of vacancy-like defects, precipitates and vacancy-solute complexes. The presence of substitutionally dissolved Au is found to reduce the formation of radiation defects more efficiently than solute Cu. Site-specific Au precipitation at defect sites is indicated, which results in damage healing with a reduced swelling, whereas Cu precipitates and radiation damage only show weak interaction. Ab initio calculations show that the binding energies of Au solutes to vacancy clusters (Au-Vn) are significantly larger than those of Cu solutes (Cu-Vn) whereas the binding energies of helium filled vacancy clusters Au-HenVn and Cu-HenVn are comparable.@en

High temperature radiation damage in binary bcc Fe alloys containing 1 atomic % Au or Cu due to Fe ion irradiation at 550 °C to a peak dose of 2.8 and 8.3 dpa is studied. The precipitation behavior of gold and copper and its correlation to the irradiation-induced defects is studied by transmission electron microscopy and variable energy positron annihilation spectroscopy (VEPAS). The increase of S parameters from VEPAS indicates the formation of open volume defects upon irradiation. Disc-shaped Au precipitates, grown from the irradiation induced dislocations, are observed in the Fe-Au alloy. In the Fe-Cu alloy, spherical Cu particles are formed but no direct connection between Cu precipitates and radiation damage is detected. For the Fe-Au alloy, the surface hardness dramatically increases for a dose of 2.8 dpa, with a slight decrease as the irradiation dose is enhanced to 8.3 dpa. In the Fe-Cu alloy, radiation hardening increases continuously.@en

Al-Mg alloys are normally prone to lose part of their yield and tensile strength at high temperatures due to insufficient thermal stability of the microstructure. Here, we present a Ca-modified Al–Mg–Sc alloy demonstrating high strength at elevated temperatures. The microstructure contains Al₄Ca phases distributed as a network along the grain boundary and Al₃(Sc,Zr) nano-particles dispersed within the grains. The microstructure evolution and age-hardening analysis indicate that the combination of an Al₄Ca network and Sc-rich nano-particles leads to excellent thermal stability even upon aging at 300 °C. The tensile strength of the alloy for temperatures up to 250 °C is significantly improved by an aging treatment and is comparable with the commercial heat-resistant aluminum alloys, i.e., A356 and A319. At a high temperature of 300 °C, the tensile strength is superior to the above-mentioned commercial alloys, even more so when expressed as the specific strength due to the low density of Ca-modified Al–Mg–Sc alloy. The excellent high-temperature strength results from a synergistic effect of solid solution strengthening, grain boundary strengthening and nanoparticle order strengthening.

@en