Tungsten (W) is regarded as the most promising candidates for plasma-facing materials (PFMs) in future fusion reactors. However, W will be exposed to high fluxes of helium (He) plasma in a nuclear fusion environment, which will degrade the thermal and mechanical properties of W. We have investigated the behavior of He in W using a first-principles method based on density functional theory. It has demonstrated that the quantitative stress indicator and the qualitative hard-sphere model can be used to characterize the effective volume and the stability of interstitial He in W. 

The hydrostatic tensile (compressive) strain will enhance (suppress) the dissolution and diffusion of He in W. The existing of He will have negative effect on the ideal strength of W. Vacancies and grain boundaries defects in W can serve as trapping centers for He, because the defects can provide large space for He. He will spontaneously form He clusters at the near surface by self-trapping. 

The pre-existing impurity/alloying element can significantly reduce the solution energy of He in W due to the electron density redistribution induced by them, such as C and Nb. The results will provide good reference for developing W materials as PFMs. This paper reviews our recent findings regarding the behavior of He in W.

This paper was originally published in Computational Materials 112, Part B, (2016) 487-491.

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