Radiation-induced effects in self-passivating W-Cr-Y alloy

Irradiation of the promising self-passivating W-10wt%Cr-0.5wt%Y alloy suggested as the first wall material for demonstration power plant, DEMO, was performed with 5.9 MeV Co ²⁺ ions at 300 °C and 5.6 MeV Fe ²⁺ ions at 500 °C to a dose peak of 12 dpa (displacements per atom). Atom probe tomography and transmission electron microscopy were used to study the chemical compositions of this alloy at the atomic scale before and after irradiation. Before irradiation, the microstructure consists of a supersaturated bcc solid solution (αW,Cr) phase with an average grain size of about 1 µm. Cr-enriched nanoscale clusters with an average size of (2.0 ± 0.4) with a density of (6.3 ± 0.5) ×10 ²⁴ m ⁻³ and (1.8 ± 0.4) nm with a density of (5.6 ± 0.5) ×10 ²⁴ m ⁻³ were found in the damage range after irradiation at 300 and 500 °C, respectively. The Cr concentration in the Cr-enriched clusters increased from 48 at.% to 78 at.% with increasing the irradiation temperature from 300 to 500 °C, but the average size slightly decreased. This indicates that Cr clusters are embryos of a Cr-rich (αCr,W) phase. It was found that the Cr-enriched clusters play a dominant role in the radiation-induced hardening of the alloy. Beyond the damage range, no Cr clusters were observed. It was found that the formation of the Cr-enriched clusters in the W-10Cr-0.5Y alloy mitigates the formation of dislocation-type defects. A segregation of yttria particles in the form of platelets was observed at the grain boundaries under irradiation at 500 °C. Although the average yttria particle size remains unchanged after irradiation, the size distribution becomes more uniform and shifts towards smaller sizes. Therefore, the beneficial effect of nanoscale yttria particles on the oxidation resistance and alloy strengthening is expected to remain unchanged after irradiation under current experimental conditions.