Abstract

The neutron diffracttion studies of the samples of fuel element claddings made of EK-164 and ChS-600 austenitic steels in the initial state and after operation in a BN-600 fast neutron reactor are reviewed. The samples were investigated in a wide range of fast neutron doses and irradiation temperatures, up to ∼80 displacements per atoms (dpa) and 628°C, respectively. Various microstructural characteristics, such as microstresses, dislocation density, and crystallographic texture are determined using full-profile analysis of neutron diffraction data. At high doses studied in this work, the irradiation temperature is the dominant factor responsible for the dislocation density. The procedures of analyzing and revealing defect types were preliminary developed on specially prepared nickel samples and austenitic alloys upon irradiation in the IVV-2M reactor core at a temperature near 80°С, which provided a low thermal mobility of lattice defects. The results obtained in some first diffraction studies on samples after their operation in the reactor core show that neutron data can be used to characterize the microstructures formed due to irradiation and, at least, estimate the mechanical properties of irradiated materials without their destruction and with no radiation risk for personnel.

中文翻译：

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辐照反应堆奥氏体钢的中子学

摘要

回顾了在初始状态下和在BN-600快速中子反应堆中运行后，由EK-164和ChS-600奥氏体钢制成的燃料元件包壳样品的中子衍射研究。在大范围的快速中子剂量和辐照温度下对样品进行了研究，分别高达每原子约80个位移（dpa）和628°C。使用中子衍射数据的全轮廓分析可以确定各种微结构特征，例如微应力，位错密度和晶体学织构。在这项工作中研究的高剂量下，辐照温度是造成位错密度的主要因素。在IVV-2M反应堆堆芯中在接近80°C的温度下进行辐照后，针对特殊制备的镍样品和奥氏体合金初步开发了分析和揭示缺陷类型的程序，这降低了晶格缺陷的热迁移率。在样品在反应堆堆芯中运行后，对样品进行的一些首次衍射研究得出的结果表明，中子数据可用于表征由于辐照形成的微观结构，至少可以估算辐照材料的机械性能，而不会破坏辐照材料，并且不会破坏辐照材料。人员的辐射风险。