“太脆”有多脆?:碳化硅和淬火后蒸汽氧化锆合金包层的力学性能比较

IF 3.2 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY
Hailemariam M. Gebrelibanos , SungHoon Joung , Farhad Mohammadi-Koumleh , Youho Lee
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引用次数: 0

摘要

本研究在设计基础事故条件下,比较了三相SiC复合材料包层的塑性与锆合金的淬火后塑性极限。在目前的设计基础事故(DBA)安全标准下,SiC包层的脆性超过了明显氧化锆合金允许的极限。虽然三层碳化硅包层只有最小的氧化引起的延性损失,并且具有很强的可冷却几何形状保持能力,但这种优势被其固有的脆性所掩盖。这意味着SiC包层的部署需要对当前的监管框架进行实质性修订,以适应SiC复合材料包层的未经历脆性,即使在稳态反应堆运行下也是如此。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
How brittle is 'too brittle'?: Comparative mechanical assessment of silicon carbide and post-quench steam-oxidized zirconium alloy cladding
This study compares the ductility of Triplex SiC composite cladding to established post-quench ductility (PQD) limits of zirconium alloys under Design Basis Accident conditions. As-received SiC cladding exhibits brittleness exceeding the limits permitted for significantly oxidized zirconium alloys under current design basis accident (DBA) safety standards. Although Triplex SiC cladding experiences only minimal oxidation-induced ductility loss with great capability for coolable geometry retention, this advantage is overshadowed by its inherent brittleness. It implies that deployment of SiC cladding requires substantial revisions to current regulatory frameworks to accommodate the unexperienced brittleness of SiC composite cladding, even under steady-state reactor operations.
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来源期刊
Journal of Nuclear Materials
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.
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