Hydrogen retention in fusion reactor materials

与固体表面与高温等离子体相互作用有关的材料问题--侵蚀、熔化、再沉积和氚保留--是当前和未来聚变反应堆设计中的一些最大挑战。 在过去的35年里，我们在德克萨斯大学的聚变材料小组一直是与上述问题相关的关键领域的全球研究活动的一部分，即聚变反应堆中面向等离子体的部件的侵蚀和氢保留。特别是，我们在捕获和传输钨中的氢以及去除含氢沉积层方面的专业知识与一些最大的关切密切相关。在等离子体-材料相互作用领域，我计划在未来五年内的研究将集中在两个关键领域：1)钨中氢的捕获和传输；2)碳基托卡马克沉积物的表征和去除。 1)钨中的氢：一个令人感兴趣的关键领域，也是我们唯一有能力进行研究的领域，是钨与氢和其他物种(如He、N、Ne、Ar等)的共同轰击。以前的研究表明，钨与氦和氢的共同轰击大大减少了氢在钨材料中的传输。最终，这可能导致运行中的反应堆中钨组件的氚库存减少数量级。人们对其他物种对氢运输的影响知之甚少。使用我们的双束高通量低能粒子加速器，我们能够将样品同时暴露在两种离子物种(例如，D+和He+)中，并独立控制能量和通量，使我们能够系统地研究这种相互作用的基本方面。 2)碳基沉积：材料侵蚀是聚变等离子体和面向等离子体的固体表面相互作用的必然结果。被侵蚀的原子进入等离子体，最终沉积在可能远离其起始点的表面上。在目前的托卡马克中，可以观察到许多沉积发生在特定的区域，这些表面可能积累厚厚的(毫米)沉积物，这可能严重扰乱等离子体的运行。虽然我们已经研究了此类沉积物的许多方面，但我们仅限于当前托卡马克(DIII-D、ASDEX-U和JET)的沉积物，所有这些沉积物都在底壁温度&lt；300℃下运行，然而，未来的聚变反应堆很可能在高于700摄氏度的基础温度下运行，导致沉积具有非常不同的特征。虽然这样的高温沉积目前还不能供我们测试，但我们可以通过加热电流沉积来模拟沉积的某些方面。DIII-D托卡马克计划引入加热的内表面，这将在未来几年内提供适当的样品。

Materials issues related to the interaction of solid surfaces with high temperature plasmas - erosion, melting, redeposition, and tritium retention - represent some of the greatest challenges in the design of current and future fusion reactors. For the past 35 years, our fusion materials group at U of T has been part of world-wide research activities in key areas related to the concerns above; namely, the erosion of, and hydrogen retention in, plasma-facing components in fusion reactors. In particular, our expertise in the trapping and transport of hydrogen in tungsten and in the removal of hydrogen-containing deposited layers closely match some of the greatest concerns. Within the area of plasma-materials interactions, my planned research for the next five years will focus on two key areas: 1) hydrogen trapping and transport in tungsten and 2) characterization and removal of carbon-based tokamak deposits. 1) Hydrogen in Tungsten: A key area of interest, and one in which we are uniquely equipped to investigate, is the co-bombardment of tungsten with hydrogen and other species; e.g., He, N, Ne, Ar., etc. Previous studies have shown that co-bombardment of tungsten with helium and hydrogen dramatically reduces the transport of hydrogen in tungsten materials. Ultimately, this could lead to orders of magnitude smaller tritium inventories in tungsten components in an operating reactor. Less is known about the effect of other species on hydrogen transport. With our dual-beam high-flux low-energy particle accelerator we are able to simultaneously expose specimens to two ion species (e.g., D+ and He+) with independently controlled energies and fluxes, allowing us to systematically investigate the fundamental aspects of such interactions. 2) Carbon-based Deposits: Material erosion is an inevitable consequence of the interaction between a fusion plasma and the solid surfaces which face the plasma. Eroded atoms enter the plasma and are, eventually, deposited on surfaces which may be far from their point of origin. In current tokamaks it is observed that much of the deposition occurs in specific areas, and these surfaces may accumulate thick (mm’s) deposits which can seriously disrupt plasma operation. While we have studied many aspects of such deposits, we have been limited to deposits from current tokamaks (DIII-D, ASDEX-U and JET), all of which operate with base wall temperatures < 300 C. Future fusion reactors, however, are likely to operate at base temperatures above 700 C, leading to deposits with very different characteristics. While such high-temperature deposits are not currently available for us to test, we can simulate some aspects of the deposits through heating current deposits. The DIII-D tokamak plans to introduce heated internal surfaces which will provide appropriate specimens within the next few years.