Deuterium gas-driven permeation and retention in ZrC dispersion-strengthened W and pure W

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2024-11-19 DOI:10.1016/j.jnucmat.2024.155529

Long Li , Ze Chen , Zeshi Gao , Yangyang Li , Zhe Liu , Wenyu Lin , Chao Yin , Shifeng Mao , Xudi Wang , Minyou Ye

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Abstract

ZrC dispersion-strengthened W (WZrC) exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. However, the hydrogen isotope behaviors in WZrC are not well understood. In this study, experiments of deuterium gas-driven permeation (GDP) through 0.5 wt.% ZrC dispersion-strengthened W and pure W in the temperature range of 923–1123 K have been performed. The deuterium permeability of WZrC is close to pure W. In comparison, the deuterium diffusivity of WZrC is significantly lower than pure W in the experimental temperature range. Besides, thermal desorption spectroscopy (TDS) was used to investigate the deuterium retention properties of both W materials after static deuterium gas charging. The results reveal two deuterium desorption peaks with detrapping energies of 0.91 ± 0.09 eV and 1.07 ± 0.28 eV for WZrC, while one peak of 1.12 ± 0.31 eV for pure W. The deuterium retention of WZrC is approximately one order of magnitude higher than pure W. The study further discusses the impact of microstructural features on the transport and retention properties of deuterium in these materials.

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ZrC 分散强化 W 和纯 W 中氘气驱动的渗透和保留

ZrC分散强化W（WZrC）具有高强度/韧性、低韧性-脆性转变温度和优异的抗热震性，使其成为未来核聚变装置中一种很有前途的等离子体面材料。然而，人们对 WZrC 中的氢同位素行为还不甚了解。本研究在 923-1123 K 的温度范围内进行了 0.5 wt.% ZrC 分散强化 W 和纯 W 的氘气驱动渗透（GDP）实验。相比之下，在实验温度范围内，WZrC 的氘扩散率明显低于纯 W。此外，还利用热解吸光谱（TDS）研究了两种 W 材料在静态氘气充填后的氘保留特性。结果显示，WZrC 有两个氘解吸峰，解吸能量分别为 0.91 ± 0.09 eV 和 1.07 ± 0.28 eV，而纯 W 只有一个 1.12 ± 0.31 eV 的峰值。

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来源期刊

Journal of Nuclear Materials 工程技术-材料科学：综合

CiteScore

5.70

自引率

25.80%

发文量

601

审稿时长

63 days

期刊介绍： The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome. The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example. Topics covered by JNM Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior. Materials aspects of the entire fuel cycle. Materials aspects of the actinides and their compounds. Performance of nuclear waste materials; materials aspects of the immobilization of wastes. Fusion reactor materials, including first walls, blankets, insulators and magnets. Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties. Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.