Recent progress on the R&D of W-ZrC alloys for plasma facing components in fusion devices

• Published in: Nuclear materials and energy 2018-08, Vol.16, p.191-206
• Main Authors: Liu, R., Xie, Z.M., Yang, J.F., Zhang, T., Hao, T., Wang, X.P., Fang, Q.F., Liu, C.S.
• Format: Article
• Language: English
• Subjects: Mechanical property; Microstructure; Plasma-facing materials; Tungsten; ZrC
• ISSN: 2352-1791; 2352-1791
• DOI: 10.1016/j.nme.2018.07.002

•Recent progress on the R&D of W-ZrC alloys as PFMs was reviewed.•The performance of W-ZrC and some newly developed W alloys were compared.•Encouraging results of W-ZrC alloys as PFMs were achieved. Tungsten is considered as the most promising material for plasma facing components (PFCs) in the magnetic confinement fusion devices, due to its high melting temperature, high thermal conductivity, low swelling, low tritium retention and low sputtering yield. However, the brittleness, poor machinability and low strength at high-temperatures of tungsten limits its application. Focusing on this issue, various W alloys with enhanced mechanical properties have been developed over recent decades. Among them, the W-ZrC alloys exhibit high strength, high ductility, low ductile-to-brittle transition temperature, good high-temperature stability, excellent resistance to thermal shock and low H/He plasma etching, making it one of the most promising candidate plasma facing materials (PFMs) for the future fusion devices. Therefore, the R&D experience of W-ZrC materials for PFCs, including the design idea, optimization of composition, fabrication technology (from powder metallurgy processing to hot working, from small specimen to large-sized bulk material), microstructure and performance (mechanical properties, thermal conductivity, resistance to thermal shocks and to H/He irradiation, H isotope retention behaviors) were reviewed in this paper.