[1]张涛,吴学邦,谢卓明,等.核聚变第一壁用W-ZrC材料研究进展与展望[J].中国材料进展,2018,(05):001-5.
 ZHANG Tao,WU Xuebang,XIE Zhuoming,et al.The progress and outlook of W-ZrC alloy as plasma facing component[J].MATERIALS CHINA,2018,(05):001-5.
点击复制

核聚变第一壁用W-ZrC材料研究进展与展望()
分享到:

中国材料进展[ISSN:1674-3962/CN:61-1473/TG]

卷:
期数:
2018年第05期
页码:
001-5
栏目:
出版日期:
2018-05-31

文章信息/Info

Title:
The progress and outlook of W-ZrC alloy as plasma facing component
作者:
张涛吴学邦谢卓明李祥艳刘瑞王先平方前锋刘长松
(中国科学院固体物理研究所,安徽,合肥,230031)
Author(s):
ZHANG TaoWU Xuebang XIE ZhuomingLI Xiangyan LIU Rui WANG XianpingFANG QianfengLIU Changsong
(Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031)
关键词:
核聚变第一壁材料W-ZrC 力学性能抗辐照性能抗高热负荷性能展望
Keywords:
Nuclear Fusion plasma facing materials W-ZrC alloy mechanical properties thermal shock resistance plasma irradiation resistance hydrogen retention outlook.
文献标志码:
A
摘要:
核聚变能是利用氢同位素氘与氚进行聚变反应而释放出的巨大能量来发电。作为人类最理想的清洁能源之一,实现核聚变能源不仅是中国的梦想,也是全人类的梦想。而材料问题尤其是面向等离子体材料(PFMs)一直是核聚变能发展面临的主要挑战之一。由于PFMs直接包围高温等离子体,不但要承受高热负荷(5-20MW/m2稳态热流,~GW/m2瞬态热流),而且还经受高通量的高能中子辐照、等离子体燃料粒子等的轰击等。钨(W)具有高熔点、高溅射阈值/低溅射率和高热导率等优点,而被认为是最有希望的面向等离子体第一壁材料,目前ITER及EAST已经选用纯W作为第一壁及偏滤器材料。而对于下一代聚变堆如中国聚变工程实验堆(CFETR),其设计参数更高,PFMs服役环境比ITER及EAST更加严峻。因此纯W由于一些自身的弱点如低温脆性(DBTT ~ 400 oC)、再结晶脆化以及辐照脆化、高热负荷开裂熔化、等离子刻蚀严重等不足将无法满足未来需求。因此研究材料的辐照损伤与氢氦效应机理,揭示辐照引起材料微观结构与性能的变化以探索开发新型抗辐照W基第一壁材料变得十分迫切。近年来,国内外研究组针对上述问题开展了系统的研究工作,研发了不同种类的W基复合材料如W-Y2O3,W-La2O3,W-TiC及W-ZrC等,性能及工艺均取得了一定的进展。其中基于计算模拟结果发展的W-ZrC材料具有较好的综合性能,是未来聚变装置第一壁候选材料之一。本文系统介绍了W-ZrC材料研究进展包括设计:钨中辐照损伤/氢氦效应机理、界面耦合强化的计算模拟结果、微结构分析测试及服役性能评估的研究,通过全面的总结分析,提出W基第一壁材料其后的主要研究方向,以及研发高性能面向未来聚变堆第一壁材料应采取的策略及措施。
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 .
更新日期/Last Update: 2018-04-27