文章信息/Info

Title:

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

作者:

張濤; 吳學(xué)邦; 謝卓明; 李祥艷; 劉瑞; 王先平; 方前鋒; 劉長松

(中國科學(xué)院固體物理研究所，安徽，合肥，230031)

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)

關(guān)鍵詞:

核聚變第一壁材料; W-ZrC;  力學(xué)性能; 抗輻照性能; 抗高熱負(fù)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.

DOI:

10.7502/j.issn.1674-3962.2018.05.01

文獻(xiàn)標(biāo)志碼:

A

摘要:

核聚變能是利用氫同位素氘與氚進(jìn)行聚變反應(yīng)而釋放出的巨大能量來發(fā)電。作為人類最理想的清潔能源之一，實(shí)現(xiàn)核聚變能源不僅是中國的夢想，也是全人類的夢想。而材料問題尤其是面向等離子體材料(PFMs)一直是核聚變能發(fā)展面臨的主要挑戰(zhàn)之一。由于PFMs直接包圍高溫等離子體，不但要承受高熱負(fù)荷（5-20MW/m2穩(wěn)態(tài)熱流，～GW/m2瞬態(tài)熱流），而且還經(jīng)受高通量的高能中子輻照、等離子體燃料粒子等的轟擊等。鎢(W)具有高熔點(diǎn)、高濺射閾值/低濺射率和高熱導(dǎo)率等優(yōu)點(diǎn)，而被認(rèn)為是最有希望的面向等離子體第一壁材料,目前ITER及EAST已經(jīng)選用純W作為第一壁及偏濾器材料。而對于下一代聚變堆如中國聚變工程實(shí)驗(yàn)堆（CFETR），其設(shè)計(jì)參數(shù)更高，PFMs服役環(huán)境比ITER及EAST更加嚴(yán)峻。因此純W由于一些自身的弱點(diǎn)如低溫脆性（DBTT ~ 400 oC）、再結(jié)晶脆化以及輻照脆化、高熱負(fù)荷開裂熔化、等離子刻蝕嚴(yán)重等不足將無法滿足未來需求。因此研究材料的輻照損傷與氫氦效應(yīng)機(jī)理，揭示輻照引起材料微觀結(jié)構(gòu)與性能的變化以探索開發(fā)新型抗輻照W基第一壁材料變得十分迫切。近年來，國內(nèi)外研究組針對上述問題開展了系統(tǒng)的研究工作，研發(fā)了不同種類的W基復(fù)合材料如W-Y2O3，W-La2O3，W-TiC及W-ZrC等，性能及工藝均取得了一定的進(jìn)展。其中基于計(jì)算模擬結(jié)果發(fā)展的W-ZrC材料具有較好的綜合性能，是未來聚變裝置第一壁候選材料之一。本文系統(tǒng)介紹了W-ZrC材料研究進(jìn)展包括設(shè)計(jì)：鎢中輻照損傷/氫氦效應(yīng)機(jī)理、界面耦合強(qiáng)化的計(jì)算模擬結(jié)果、微結(jié)構(gòu)分析測試及服役性能評估的研究，通過全面的總結(jié)分析，提出W基第一壁材料其后的主要研究方向，以及研發(fā)高性能面向未來聚變堆第一壁材料應(yīng)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 .