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Title:: The progress and outlook of W-ZrC alloy as plasma facing component

Author(s):: ZHANG Tao; WU Xuebang; XIE Zhuoming; LI Xiangyan; LIU Rui; WANG Xianping; FANG Qianfeng; LIU Changsong; (Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031)

Keywords:: Nuclear Fusion; plasma facing materials; W-ZrC alloy; mechanical properties; thermal shock resistance; plasma irradiation resistance; hydrogen retention; outlook.

CLC:

PACS:: -

DOI:: 10.7502/j.issn.1674-3962.2018.05.01

DocumentCode:

Abstract:: Nuclear fusion can generate electricity by using enormous energy released by hydrogen isotopes, deuterium and tritium fusion reactions. Nuclear fusion energy is not only the Chinese dream but also is the dream of mankind. And one of the key problems is the development of plasma facing materials（PFMs). Due to the plasma facing components (PFCs) face extreme conditions, such as high levels of neutron irradiation, a high heat flux of energetic particles, sputtering erosion, and transient events like plasma disruptions, the comprehensive servicing performance is closely related to the safety operation of fusion devices. The serving condition of PFMs in CFETR is much more harsh than that in current fusion device, so it is a big challenge for future PFMs. Tungsten is selected as the prime candidate for PFCs due to its melting temperature, high thermal conductivity, high neutron load capacity, low tritium retention and low sputtering yield. However, pure W exhibits the serious embrittlement in several regimes, i.e., low-temperature embrittlement (relatively high ductile-brittle transition temperature DBTT), recrystallization embrittlement and radiation embrittlement, as well as the surface crack and melting under high thermal load, which will be not suitable for the future advanced fusion devices. Therefore, it is very important to reveal the irradiation damage mechanism, hydrogen/helium effect and boundary strengthening by simulation as well develops new W based PFMs. Focusing on the above issues, in recent decades, many efforts have been devoted to improving the strengthen and ductility of W alloys by developing new W alloys such oxide dispersion strengthen and carbide dispersion strengthen W alloy, whose some properties were improved compared with pure W. In which, bulk W-0.5w%ZrC materials developed by Institute of Solid State Physics, Chinese Academy of Sciences, has better performance, which is a very promising candidate PFCs for the future fusion devices. Therefore, this paper reviews the R＆D experience: including simulation for designing, microstructure and servicing performance of W-0.5w%ZrC alloy. The further research direction of this material as well as the strategies for R＆D of other PFCs with high performance are proposed by comprehensive analysis and summary .

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The progress and outlook of W-ZrC alloy as plasma facing component(PDF)

MATERIALS CHINA[ISSN:1674-3962/CN:61-1473/TG]

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