高温气冷堆在950℃石墨包覆环境下的渗碳行为和拉伸性能

IF 2.1 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Bin Du, Jiahao Chang, Penghui Xiao, Lei Peng, Huaqiang Yin, Xuedong He, Tao Ma
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引用次数: 0

摘要

本文研究了镍铬合金617和镍铬合金800H在950℃的石墨粉尘环境中腐蚀100小时后的渗碳机理和拉伸性能的变化。通过组织分析和力学性能测试发现,Inconel 617发生碳原子向内扩散,形成碳化铬/碳化钼析出,导致极限抗拉强度下降25.6%,总伸长率下降64.3%,导致脆性断裂。相比之下,由于表面Si元素抑制碳扩散,incoly 800H的机械性能下降最小。进一步研究表明,预氧化处理可在Inconel 617表面形成致密的Cr2O3层,使其渗碳率降低63.4%,使其力学性能恢复到接近原始水平。这些发现为VHTR应用中的材料选择和防腐策略提供了重要指导。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
Carburization behavior and tensile properties of superalloys Inconel 617 and Incoloy 800H for very-high-temperature gas-cooled reactors under graphite dust-coated environment at 950 °C
The carburization mechanisms and changes in tensile properties of Inconel 617 and Incoloy 800H following a 100-hour corrosion exposure in an environment coated with graphite dust at 950 °C was studied in this paper. Through the microstructural analysis and mechanical property testing, it was discovered that Inconel 617 undergoes carbon atom inward diffusion, forming chromium/molybdenum carbide precipitates, which result in a reduction in ultimate tensile strength of 25.6 % and a decrease in total elongation of 64.3 %, leading to brittle fracture. In contrast, Incoloy 800H exhibits minimal degradation in mechanical properties due to the surface Si element suppressing carbon diffusion. Further research demonstrates that pre-oxidation treatment can form a dense Cr2O3 layer on the surface of Inconel 617, reducing its carburization by 63.4 % and restoring its mechanical properties to levels close to the original. These findings provide critical guidance for material selection and corrosion prevention strategies in VHTR applications.
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来源期刊
Nuclear Engineering and Design
Nuclear Engineering and Design 工程技术-核科学技术
CiteScore
3.40
自引率
11.80%
发文量
377
审稿时长
5 months
期刊介绍: Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.
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