In-situ irradiation of uranium carbide

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials Pub Date : 2025-07-30 DOI:10.1016/j.jnucmat.2025.156082

Rashed Almasri , Adrian R. Wagner , Laura Hawkins , Wei-Ying Chen , Jennifer K. Watkins , Jian Gan , Lingfeng He

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Abstract

Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe⁺ and 1 MeV Kr²⁺ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO₂ or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO₂ and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

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碳化铀的原位辐照

碳化铀（UC）由于其高铀密度和导热性，是第四代反应堆的主要候选燃料。然而，它的辐照性能，特别是气泡膨胀和缺陷动力学，仍然缺乏表征。利用原位透射电子显微镜（TEM），我们用300 keV Xe+和1 MeV Kr2+离子在900°C的温度下照射UC，以量化膨胀行为和位错环的演变。在所有温度下，溶胀率都保持在0.6%以下，表明UC在这些温度下辐照的尺寸稳定性。位错环在UC中的生长速度快于UO2或UN，这与UC的同源温度较低有关。值得注意的是，在片层的薄区出现了纳米颗粒结构，这与之前在UO2和ZrC中观察到的现象相类似。这些结果解决了UC辐射耐受性方面的关键知识空白，并为其在先进反应堆系统中的适用性提供了见解。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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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.