Jianxi Deng, Shangle Huo, Donghui Geng, Qiaoyan Sun, Zhongxiao Song, Jun Sun
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引用次数: 0
Abstract
Mechanical properties, particularly ductility, are crucial for evaluating the safety threshold and limits of Cr-coated zirconium claddings at loss of coolant accident (LOCA). The mechanical properties of Cr-coated Zr-1Nb alloy plates after 1000–1200 °C steam oxidation are investigated through uniaxial tensile tests at 135 °C, with uncoated specimens as comparison. The results show that the ductile-to-brittle transition of Cr-coated samples is significantly affected by oxidation temperatures. After 2 h oxidation at 1000 °C, Cr-coated samples remain good ductile with 16.2 % elongation compared to 0.4 % for uncoated ones. While, Cr-coated samples exhibit brittle fracture with 0.2 % elongation after 2 h oxidation at 1150 °C, and the tensile strength diminishes to 321 MPa, a reduction of 18 % compared to before oxidation. The ductile-to-brittle transition is associated with the protective structure evolution of Cr coating, which essentially determines the formation of brittle α-Zr(O) in Zr substrate. At 1000 °C, good oxygen diffusion barrier of Cr coating significantly reduces α-Zr(O) layer formation. However, accelerated oxidation of Cr coating results in the generation of α-Zr(O) phase. The critical α-Zr(O) intrusion depth for brittle fracture is about 85 % of Zr substrate, and the content of O and Cr increased by 0.96 % (wt.) and 1.84 % (wt.), respectively. Formation of brittle α-Zr(O) dominates the ductility degradation, and O and Cr element diffusion into substrate also have a certain effect. The findings can provide insights for performance degradation modeling of Cr-coated Zr alloys under LOCA conditions.
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.