Ab-initio Investigation of Hydrogen and Helium Behavior Near W/ZrC Interfaces

Efforts to improve the thermomechanical properties of plasma-facing materials when exposed to high-energy particle irradiation led to a new class of materials; dispersoids strengthened W. For example, carbide-dispersion strengthened W has been shown to improve ductility, crack resistance, and radiation tolerance. Here we discuss various aspects of W-ZrC interfaces leading to their stabilities and focus on how these features impact the H/He behavior at and near the interface. Our results indicate that ZrC (111) –W (110) exhibits the highest adhesive energy compared with the other investigated interfaces and hence forms the most stable interface. The impurities hydrogen and helium tend to segregate to the interfaces, which can both embrittle the interface and potentially increase gas retention. In addition, the interface could facilitate the diffusion of hydrogen and helium by providing a low-barrier channel. The present work can provide key mechanistic insights towards interpreting recent experimental studies of the interface structure and H/He retention in W-ZrC interfaces and guide the design of future W-based materials for improving plasma-facing component performance in fusion energy applications.