钨
微观结构
材料科学
氘
辐照
氢
核反应分析
聚变能
空隙(复合材料)
离子
托卡马克
核聚变
放射化学
分析化学(期刊)
原子物理学
等离子体
复合材料
核物理学
化学
冶金
物理
有机化学
色谱法
作者
Daniel R. Mason,F. Granberg,Max Boleininger,T. Schwarz‐Selinger,K. Nordlund,S. L. Dudarev
出处
期刊:Physical Review Materials
[American Physical Society]
日期:2021-09-17
卷期号:5 (9)
被引量:22
标识
DOI:10.1103/physrevmaterials.5.095403
摘要
Hydrogen isotopes are retained in materials for fusion power applications, changing both hydrogen embrittlement and tritium inventory as the microstructure undergoes irradiation damage. But modelling of highly damaged materials - exposed to over 0.1 displacements per atom (dpa) - where asymptotic saturation is observed, for example tungsten facing the plasma in a fusion tokamak reactor, is difficult because a highly damaged microstructure cannot be treated as weakly interacting isolated defect traps. In this paper we develop computational techniques to find the defect content in highly irradiated materials without adjustable parameters. First we show how to generate converged high dose (>1 dpa) microstructures using a combination of the creation-relaxation algorithm and molecular dynamics simulations of collision cascades. Then we make robust estimates of point defects and void regions with simple developments of the Wigner-Seitz decomposition of lattice sites. We use our estimates of the void surface area to predict the deuterium retention capacity of tungsten as a function of dose. This is then compared to 3He nuclear reaction analysis (NRA) measurements of tungsten samples self-irradiated at 290 K to different damage doses and exposed to deuterium plasma at low energy at 370 K. We show that our simulated microstructures give an excellent match to the experimental data, with both model and experiment showing 1.5-2.0 at.% deuterium retained in tungsten in the limit of high dose.
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