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CARR堆芯容器材料采用6061-T6铝。ASME CodeCase N-519阐明,由于中子辐照导致6061-T6铝脆化和降低延性,因此,对于使用这种材料的堆芯容器必须制定辐照监督大纲,用于监测材料力学性能变化,特别是断裂韧性的降低程度,以便进行安全评审和经济效益的权衡,从而合理地确定堆芯容器的使用寿命。为保证今后辐照监督试验的进行, 相似文献
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中国聚变工程试验堆(CFETR)先进材料辐照考验样品所在胶囊结构较为复杂,其内部填充氦气,胶囊肋条尺寸、位置以及胶囊内部填充材料对样品温度影响大。基于STAR-CCM+程序建立CFETR先进小样品辐照装置内胶囊全尺寸模型,针对样品的目标温度,对胶囊的肋条和填充材料进行了调整。对于胶囊内整体样品释热率较低的情况,采用释热率较大的钨材料作为填充材料,可以明显提高整体样品温度;对于局部样品释热率差别较大的情况,调整局部肋条的尺寸和位置,能够很好控制样品间的温度,使样品计算温度满足目标温度范围。结果表明:采用上述方法进行优化后,样品中心温度能够满足目标温度范围,且满足入高通量工程试验堆(HFETR)辐照的热工安全,保证整个辐照任务能够顺利开展。 相似文献
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碳化硅(SiC)晶体可以用作无源监控器测量反应堆的中子辐照温度,在未来高温强辐射的先进核反应堆中具有重要的应用前景。SiC测温技术自20世纪60年代被首次提出以来,基于SiC结构、热学和电学性质的中子辐照效应,人们建立了宏观尺寸法、质量密度法、热导率法和电阻率法等各种SiC测温方法。本文首先综述了这些SiC测温方法的基本原理和工作特点,然后着重介绍了中国原子能科学研究院(China Institute of Atomic Energy,CIAE)SiC测温系统的研究进展,通过中子辐照诱导SiC晶格肿胀的理论计算,分析和验证了该系统测温结果的可靠性,探讨了进一步提高SiC测温效率的实验方法。 相似文献
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小冲杆试验方法以其所需测试样品尺寸小而带来的样品感生放射性小等优势,越来越多地应用于核材料力学性能评价领域。本文设计了一套利用光栅尺直接测量样品变形的小冲杆试验装置,较传统装置精度有明显提高。利用该套装置对注量为10×1019 cm-2(E≥1 MeV)快中子辐照的国产A508 3钢材料进行了小冲杆测试研究,探索了针对放射性样品从制备到测试的试验方法,并获得了国产A508 3钢材料的小冲杆屈服特征值、抗拉特征值和韧脆转变温度与标准试验之间的关系式。 相似文献
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核监测用断裂韧性Charpy尺寸试样的合理设计 总被引:1,自引:1,他引:0
预制疲劳裂纹侧槽Charpy尺寸试样是一种经济、方便的评价核压力容器用钢弹塑性断裂韧性的单试样方法。本文就几种常用压力容器用钢详细研究了侧槽相对深度对断裂韧性及相应的稳定裂纹扩展量的影响,并和满足GB2038要求的大尺寸试样的试验结果进行了对比。研究结果表明,采用预制疲劳裂纹、侧槽相对深度为30%的Charpy尺寸试样及三点弯曲试验曲线上最大载荷前的能量,可以偏安全地评价裂纹开始扩展时材料的弹塑性断裂韧性,建立了核监测用断裂韧性试验Charpy尺寸试样的合理设计。此外,还研究了侧槽的拘束效应和对试样的加厚作用,对试验结果进行了理论解释。 相似文献
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对不同厚度国产A508-3钢小尺寸拉伸样品进行了室温拉伸试验,分析了拉伸性能及颈缩段参数,并基于有限元逆运算构建了小尺寸拉伸样品拉伸过程的GTN(Gurson-Tvergaard-Needleman)细观损伤模型,研究了厚度对小尺寸拉伸样品拉伸颈缩行为的影响规律与机理。试验结果表明,小尺寸拉伸样品在变形过程中发生了弹性变形、均匀塑性变形和颈缩变形;随着样品厚度由0.75 mm降低至0.30 mm,屈服强度、抗拉强度和均匀延伸率无明显变化,非均匀延伸率及总延伸率逐渐降低,颈缩角逐渐增大,断裂角在厚度降低至0.50 mm后逐渐增大。GTN细观损伤模型中用于表征空洞形核和融合率的参数在0.30 mm样品中明显降低,此结果与小尺寸拉伸样品颈缩行为规律相互印证。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(4):356-368
There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed. This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):848-856
Both transverse and longitudinal Zircaloy-2 specimens irradiated up to 1.2 × 1020 n/cm2 (E> 1 MeV) were tested in tension with strain rates ranging 1.1 × 10-4~1.1 × 10-2 s-1 in the temperature range 200~400°C. Detailed observations of the specimen wall surface and microstructure were also made on samples deformed to various amounts of plastic strain, with a projector and an optical microscope. It was found that localized plastic deformation bands occurred in the temperature range approximately 280~330°C during straining to the ultimate tensile stress. Results also showed that the strain rate dependence of tensile properties, particularly the strain to the ultimate tensile stress, was associated with changes in the number and width of the localized deformation band with strain rates at a temperature of 300°C at which localized bands occurred. From a break of the straight line tracing the true stress-true plastic strain relationship, it was established that the onset stress and strain of the localized deformation band could be estimated. The effect of specimen orientation on localized deformation band was also discussed on the basis of differences in the onset stress and strain between the transverse and longitudinal specimens. 相似文献