First-Principles Study of Hydrogen Behaviors at Oxide/Ferrite Interface in ODS Steels

Ferritic oxide dispersion strengthened (ODS) steels, which usually contain a very high density of nano-sized Y-Ti-O particles and oxide precipitates (Y2Ti2O7 or/and Y2TiO5), have been demonstrated to be a leading candidate for promising structural materials in advanced fission and fusion energy applications. By means of first-principles calculations, the defect formation energies and preference sites of hydrogen (H) and helium (He) atoms trapped in Y2Ti2O7, Y2TiO5 and Y2Ti2O7/bcc-Fe interface, were investigated. The calculations uncover that (1) H atoms prefer to occupy the interstitial sites with high preexsiting charge densities of Y2Ti2O7, and Y2TiO5, (2) the Y2Ti2O7/bcc-Fe interface trends to attract vacancies in bcc-Fe matrix because of its lower vacancy formation energies, (3) at the Y2Ti2O7/bcc-Fe interface, H atom prefers to occupy the interstitial sites around the bcc-Fe side while He atom prefers to occupy the interstitial sites around Y2Ti2O7 side. All these results demonstrate that both H and He atoms produced by nuclear transmutation reactions would be trapped by oxides precipitates and Y2Ti2O7/bcc-Fe interface in case of the formation of large bubbles. This implies that high density of nanometer-sized oxide precipitates and Y2Ti2O7/bcc-Fe interfaces in ODS steels effectively disperse H atoms and inhibit H clusters in finer size. Besides that, during the growth process of the finer H clusters at interfaces they trap a large number of both H atoms and vacancies, acting as self-healing sites for irradiation damage. These facts potentially corresponds to the excellent capability of ODS steels to resist irradiation damage. Moreover, the calculation results may also interpret the synergistic effect of irradiation damage produced by both H and He to ODS steels.