Hailemariam M. Gebrelibanos , SungHoon Joung , Farhad Mohammadi-Koumleh , Youho Lee
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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.
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.