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作 者:廖业宏 戴龚颖 严俊 林晓冬 彭振驯 梁雪[3] 李毅丰 薛佳祥 李强[3] Liao Yehong;Dai Gongying;Yan Jun;Lin Xiaodong;Peng Zhenxun;Liang Xue;Li Yifeng;Xue Jiaxiang;Li Qiang(Nuclear Fuel and Materials Department,China Nuclear Power Technology Research Institute Co.,Ltd,Shenzhen 518026,China;Institute of Materials,Shanghai University,Shanghai 200072,China;Instrumental Analysis&Research Center,Shanghai University,Shanghai 200444,China)
机构地区:[1]中广核研究院有限公司核燃料与材料研究所,广东深圳518026 [2]上海大学材料研究所,上海200072 [3]上海大学分析测试中心,上海200444
出 处:《稀有金属材料与工程》2024年第10期2843-2851,共9页Rare Metal Materials and Engineering
基 金:国家自然科学基金(52001192,52201079)。
摘 要:采用3D光学表面轮廓仪、扫描电子显微镜、电子背散射衍射和能谱仪等测试技术研究了290和310℃下Cr涂层Zr-1Nb合金包壳管与格架在模拟压水堆一回路水环境中的微动磨损行为。结果表明,当Cr涂层Zr-1Nb合金包壳管与对磨副材料(Zr-4刚凸或Inconel 718弹簧)组成摩擦副时,微动磨损机制均以粘着磨损为主,伴随着对磨副材料向Cr涂层包壳管的转移。随温度升高,Cr涂层包壳管表面磨损量增加,其抗微动磨损性能下降,但在试验温度范围内微动磨损机制未发生变化。此外,当对磨副为刚凸时Cr涂层包壳管的磨损程度大于对磨副为弹簧时的磨损程度,这与对磨副的硬度和接触方式有关。The fretting wear behaviors of Cr-coated Zr-1Nb cladding tube with different griding pairs in simulated pressurized water reactor primary water environment at 290 and 310℃were studied by three-dimensional optical surface profilometer,scanning electron microscope,electron backscattered diffraction and energy spectrometer.The results show that the fretting wear behaviors between the Cr-coated Zr-1Nb cladding and griding pairs(i.e.,Zr-4 dimple or Inconel 718 spring)are both dominated by the adhesive wear mechanism,accompanied by material transfer from the grinding pair to the Cr-coated cladding.With the increase in temperature,the fretting wear resistance of the Cr-coated cladding is decreased,as manifested by increased surface wear depth and volume.However,the fretting wear mechanism still remains unchanged within the temperature range tested in this work.In addition,the wear degree of the Cr-coated cladding with Zr-4 dimple is greater than that with Inconel 718 spring,which is related to the hardness and contact mode of the grinding pair.
关 键 词:核燃料包壳 Cr涂层锆合金 高温高压水 微动磨损 对磨副
分 类 号:TG174.4[金属学及工艺—金属表面处理] TL352.22[金属学及工艺—金属学]
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