首页> 外文OA文献 >Interactions mécanique-oxydation à haute température dans l'alliage 600 : application à la fissuration dans le milieu primaire des réacteurs nucléaires à eau sous pression
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Interactions mécanique-oxydation à haute température dans l'alliage 600 : application à la fissuration dans le milieu primaire des réacteurs nucléaires à eau sous pression

机译:合金600在高温下的机械氧化相互作用:在压水核反应堆主要环境中的裂化中的应用

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摘要

Since the early 1970s certain component parts of primary loops of nuclear pressurised water reactors, such as steam generator tubing and pressure vessel head penetrations, have been affected by intergranular stress corrosion cracking. This form of cracking, which occurs after several thousand hours, initiates on the component side exposed to primary water. The purpose of this study is to gain a better understanding of the cracking mechanism. The focus is on the interactions between the local mechanical loading conditions and the effects of an oxidising environment. Fatigue and creep fatigue crack propagation tests were carried out using CT specimens. the local mechanical loading conditions were determined by finite element calculations. The oxidation behaviour was studied by transmission electron microscopy. Creep tests were conducted on thin sheet samples to investigate the effects of environment on the creep properties. Intergranular crack propagation, as determined at 550°C in air and in vacuum, has been related to a creep damage mechanism. Dislocation creep deformation mechanisms, which appear to induce intergranular damage in this material at high temperature, are enhanced when the growing oxide scale is nickel- or iron-rich, compared to the results obtained in high vacuum with the protective chromium oxide scale. Thus, intergranular fracture, which occurs in all environmental conditions under creep fatigue cycles, occurs even under fatigue cycles when creep is enhanced. Fatigue crack propagation tests at 400°C gave essentially the same results as those obtained at 550°C. Crack propagation tests in deaerated water, either in low pressure vapour at 400°C or in primary water at 320°C, lead to the same cracking results. The fracture surfaces were identical to those observed after conventional stress corrosion cracking tests. For these environmental conditions, the enhanced creep model was also proposed to account for the experimental results.
机译:自1970年代初以来,核压水反应堆主回路的某些组成部分,例如蒸汽发生器的管道和压力容器的头部穿透,一直受到晶间应力腐蚀开裂的影响。几千小时后发生的这种开裂形式是在暴露于一次水的组件一侧开始的。这项研究的目的是为了更好地了解破裂机理。重点是局部机械负载条件与氧化环境的影响之间的相互作用。使用CT试样进行了疲劳和蠕变疲劳裂纹扩展测试。局部机械载荷条件是通过有限元计算确定的。通过透射电子显微镜研究了氧化行为。对薄板样品进行了蠕变测试,以研究环境对蠕变性能的影响。 550℃在空气和真空中确定的晶间裂纹扩展与蠕变损伤机制有关。与在保护性氧化铬水垢下在高真空下获得的结果相比,当生长的氧化皮富含镍或铁时,位错蠕变变形机制似乎会在高温下引起这种材料的晶间损伤。因此,在蠕变疲劳循环下的所有环境条件下均会发生晶间断裂,即使在增强蠕变的疲劳循环下也会发生。在400°C下进行的疲劳裂纹扩展测试与550°C下获得的结果基本相同。在脱气的水中(在400°C的低压蒸气中或在320°C的一次水中)的裂纹扩展测试得出相同的裂纹结果。断裂表面与常规应力腐蚀开裂试验后观察到的相同。对于这些环境条件,还提出了增强的蠕变模型来说明实验结果。

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