Deterioration of irradiation resistance of ODS-F/M steel under high concentration of helium

Abstract

Simultaneous Fe+He ion irradiations of China Low Activation Martensitic (CLAM) steel and oxide-dispersion-strengthened ferritic/martensitic (ODS-F/M) steel were carried out at 350 °C-550 °C to the damage dose of 11 dpa with different He concentrations to investigate the effect of oxide particles on the evolution of He bubbles and dislocation loops. The microstructures of all irradiated samples were observed by transmission electron microscopy (TEM). The average size of oxide particles decreased and the density increased after irradiation. The aggregation of He bubbles was observed at the edge of oxide particles. As the He concentration increased, the aggregation of helium bubbles appeared at the edge of more oxide particles. Furthermore, He bubbles were not observed inside ODS-F/M steel grains or at grain boundaries, which suggests that the oxide particle interface is more capable of capturing helium atoms than other sinks. The average dislocation loop size of ODS-F/M steels is only 58% of that in CLAM steel at 450 °C, showing that ODS-F/M steel has excellent inhibition on the growth of dislocation loops. Nevertheless, with the increase of temperature and He concentration, the number density and size of dislocation loops in ODS-F/M steels gradually approach those in CLAM steels. The ability of ODS-F/M steels to inhibit irradiation hardening relative to CLAM steels increases with increasing temperature and decreases with increasing He concentration. The result of first-principles calculations shows that the formation energy of interstitial Fe atom at the interface increases with the increasing of the number of He atoms, which means that the aggregation of excessive He atoms at the interface will inhibit superior irradiation resistance of ODS-F/M steel.