he辐照烧结SiC的高温损伤行为

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials Pub Date : 2025-04-21 DOI:10.1016/j.jnucmat.2025.155839

Jintao Zhang , Zhongzheng Wu , Jiale Huang , Jun Li , Haiyuan Wei , Tongmin Zhang , Tomas Polcar , Nabil Daghbouj , Bingsheng Li

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引用次数: 0

摘要

了解高温氦（He）辐照对烧结碳化硅（SiC）损伤行为的影响对于评估这种材料在先进核反应堆中的稳定性至关重要。在这项研究中，我们利用三种不同的辐照通量，研究了 230 keV He 离子在 800 °C 和 1000 °C 温度下对碳化硅的影响：2 × 1016/cm2、4 × 1016/cm2 和 1.6 × 1017/cm2。拉曼光谱和透射电子显微镜用于分析各种损伤特征，包括辐照诱发的晶格应变、血小板形成、位错环和氦气泡。我们的研究结果表明，过压血小板主要在 (0001) 平面上形成，附近检测到的位错环数量有限。与此相反，在没有小板的晶粒边界附近观察到大量黑点缺陷。缺陷分布的这种变化强调了与高温氦辐照相关的独特损伤行为。从这项研究中获得的见解对于理解碳化硅材料在核反应堆应用相关条件下的结构变化和完整性至关重要。

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Damage behavior of He-irradiated sintered SiC at high temperatures

Understanding the effects of high-temperature helium (He) irradiation on the damage behavior of sintered silicon carbide (SiC) is crucial for assessing the material's stability in advanced nuclear reactors. In this study, we investigate the impact of 230 keV He ions on SiC at temperatures of 800 °C and 1000 °C, utilizing three different irradiation fluences: 2 × 10¹⁶/cm², 4 × 10¹⁶/cm², and 1.6 × 10¹⁷/cm². Raman spectroscopy and transmission electron microscopy were employed to analyze various damage features, including irradiation-induced lattice strain, platelet formation, dislocation loops, and helium bubbles. Our findings indicate that over-pressurized platelets predominantly formed on the (0001) plane, with a limited number of dislocation loops detected nearby. In contrast, numerous black spot defects were observed near grain boundaries, where platelets were absent. This variation in defect distribution underscores the unique damage behavior associated with high-temperature He irradiation. The insights gained from this study are essential for understanding the structural changes and integrity of SiC materials under conditions relevant to nuclear reactor applications.

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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.