核聚变堆用钨及钨合金辐照损伤研究进展  

Research Progress in Irradiation Damage of Tungsten and Tungsten Alloys for Nuclear Fusion Reactor

作　　者：丁孝禹[1] 罗来马[1,2] 黄丽枚[1] 罗广南[3] 李萍[1,2] 吴玉程[1,2] 

机构地区：[1]合肥工业大学材料科学与工程学院,安徽合肥230009 [2]安徽省有色金属材料与加工工程实验室,安徽合肥230009 [3]中国科学院等离子体物理研究所,安徽合肥230031

出　　处：《稀有金属》2015年第12期1139-1147,共9页Chinese Journal of Rare Metals

基　　金：国际热核聚变实验堆(ITER)计划专项(2014GB121001;2010GB109004);中央高校基本科研业务费专项资金(2012HGQC0032)资助

摘　　要：随着人们对能源的需求持续迅速增长,化石燃料等传统能源在可预见的未来即将枯竭,并带来严峻的环境问题。轻原子核核聚变反应产生的聚变能是解决人类能源问题的重要潜在途径。近年来对聚变堆的研究取得了显著进展,随之而来的材料问题逐渐成为一个现实难题,这是由于反应堆中的材料将面临苛刻的工作环境。钨(W)具有高熔点、高导热率、高密度、低的热膨胀系数、低蒸气压、低氚滞留、低溅射产额和高自溅射阀值等优异性能,被认为是今后核聚变装置最有前途的面对等离子体第一壁材料,但钨及钨合金由于脆性问题在未来聚变装置中的应用也面临着巨大的挑战,其中辐照缺陷的产生往往导致材料脆化,缩短部件服役寿命。这些缺陷还会与珍贵的聚变燃料(如氚)产生相互作用,导致严重的滞留和渗透问题。因此,研究钨及钨基材料的辐照损伤就显得非常有必要,通过材料成分/结构/组织的设计来延缓辐照缺陷的产生将具有非常重要的意义。本文对商用钨材料及先进钨合金材料(超细晶钨、W-Ta合金、弥散增强钨)的辐照损伤现状及最新研究进展进行了综合评述。With the demand for energy growing rapidly, fossil fuel and other traditional energy resources which bring about serious environmental problems will soon dry up in the foreseeable future. Producing fusion energy by light nuclear fusion reaction is an important potential way to solve the energy problem of human. Recent researches on fusion reactor have made significant progress, and the resuhing material problem has become a realistic problem due to that the materials in the reactor will face harsh working environment. Tungsten （W） is considered to be the primary candidate for plasma facing materials like first wall in future fusion reactors owing to its superiority to other materials including high melting point, high thermal conductivity, high density, low thermal expansion coefficient, low vapor pressure, low tritium inventory, low sputtering yield and high energy threshold for physical sputtering, etc. However, there are still serious challenges of brittleness for W and W alloys in the future application for fusion reactors. Defects induced by irradiation often lead to embrittlement of the material, thus shortening the service life of components. The defects will also interact with precious fusion fuel （ such as tritium） , leading to serious retention and permeation. Therefore, it is very necessary to study the radiation damage of W and W-based materials and it will be of great significance to delay irradiation defects through designing material composition/ structure/organization. In order to provide a reference to researchers devoted to irradiation damage, irradiation damage status and the latest research progress of commercial W and advanced W alloys were reviewed in the paper.

关 键 词：钨 核聚变堆 辐照损伤 面对等离子体材料 

分 类 号：TL627[核科学技术—核技术及应用] TG146.411[一般工业技术—材料科学与工程]