Deuterium permeation studies through bare and Er2O3 coated SS 316L

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Sudhir Rai , P.A. Rayjada , P.B. Dhorajiya , R.B. Patel , S.K. Sharma , A. Sircar , R. Bhattacharyay
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Abstract

Hydrogen isotope permeation through the structural walls of a fusion reactor poses concern for both fuel loss and safety. A primary focus in the International Thermonuclear Experimental Reactor (ITER) is the reduction of hydrogen isotopes, particularly tritium permeation through structural materials. In this work, an experimental setup has been developed to study the permeation of deuterium through thin samples such as bare and erbia (Er2O3) coated 100 µm thick stainless steel (SS 316 L). Permeation results through a bare SS 316 L sample yielded deuterium diffusivity and permeability of SS 316 L. Erbia coating on SS 316 L was developed using dip coating technique and its performance was tested as a permeation reduction barrier (PRB). A permeation reduction factor of ∼87 and 359 have been achieved with coating thickness ∼ 100 and ∼200 nm respectively. No measureable permeation flux was obtained for coating thickness of 492 nm. These results indicate the effectiveness of erbia coating in reducing deuterium permeation through SS 316 L.

裸 SS 316L 和涂有 Er2O3 的 SS 316L 的氘渗透研究
氢同位素通过聚变反应堆结构壁的渗透对燃料损耗和安全都构成了威胁。国际热核实验反应堆(ITER)的一个主要重点是减少氢同位素,特别是氚在结构材料中的渗透。在这项工作中,开发了一套实验装置来研究氘通过薄样品(如裸不锈钢和涂有铒(Er2O3)的 100 微米厚不锈钢(SS 316 L))的渗透情况。通过裸 SS 316 L 样品的渗透结果得出了 SS 316 L 的氘扩散率和渗透率。采用浸涂技术在 SS 316 L 上形成了厄尔比亚涂层,并测试了其作为减少渗透屏障(PRB)的性能。涂层厚度分别为 100 纳米和 200 纳米时,渗透降低系数分别为 ∼87 和 359。涂层厚度为 492 nm 时,没有测得渗透流量。这些结果表明,埃尔比亚涂层能有效减少 SS 316 L 的氘渗透。
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来源期刊
Fusion Engineering and Design
Fusion Engineering and Design 工程技术-核科学技术
CiteScore
3.50
自引率
23.50%
发文量
275
审稿时长
3.8 months
期刊介绍: The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.
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