本论文立足于氧化铍（BeO）陶瓷材料应用于微型核反应堆中这一重大工程需求，其用途主要有两种：一种是BeO陶瓷作为中子反射层材料；另一种是作为新型核燃料芯块的基体材料。目前国内应用在电子产业商业领域纯BeO产品已经比较成熟，但对于应用在核领域尚缺少有关研究。同时，弥散包覆颗粒（TRISO）的BeO基燃料芯块制备研究目前尚属空白。同时，有关BeO的研究大多数来自上世纪50~60年代，面临新的应用需求，需要重新掌握目前商业BeO产品的性能。因此本文开展了BeO陶瓷的热力学性能、BeO基复合燃料芯块制备及BeO辐照效应的研究。本文的主要研究内容与结果如下：1、BeO陶瓷的热性能、力学性能研究是针对目前商业化BeO陶瓷开展的，目的是为微型核反应堆中子反射层的设计提供有关参数。首先对"Φ" 200 mm×200 mm和112 mm×112 mm×30 mm大尺寸BeO试样的均匀性进行分析测试。同时对BeO高温力学性能进行了研究，研究认为BeO晶粒间的玻璃相是影响其高温强度的主要因素。最后对BeO在室温水环境中的应力腐蚀现象进行了研究。2、BeO基复合燃料芯块的制备研究表明，三层包覆层、最外层为SiC的TRISO颗粒适合弥散在BeO基体中，并且在SiC/BeO之间存在非晶结构，这种非晶结构将SiC和BeO粘结在一起。同时，采用最优热压烧结工艺参数，成功地制备了BeO基复合燃料芯块，尺寸为"Φ" 50 mm×10 mm和"Φ" 20 mm×36 mm，TRISO颗粒所占体积分数为10%~24%。3、对于BeO的辐照效应研究，目前正在西安脉冲堆开展中子辐照实验，但中子辐照实验周期过长，暂时无法获得相关辐照实验数据。因此本文采用分子动力学方法对BeO在不同辐照温度下的辐照损伤累积过程进行研究。结果表明BeO在600 ℃及以下温度辐照时，辐照产生的晶体缺陷是缺陷簇，而在800 ℃及以上温度辐照时，会产生伯氏矢量分别为1/3<1"1" ?00>和1/3<11"2" ?0>的两种位错。辐照BeO退火过程分子动力学研究表明，在较低温度辐照的BeO在高温退火时，辐照产生的非晶结构会转变为α-BeO和β-BeO。同时，退火过程还会产生伯氏矢量为1/3<1"1" ?00>的位错环，该位错环位于基面上，同时垂直于c轴，这与以往文献中的实验结果一致。
This dissertation is based on the major engineering demand of beryllium oxide (BeO) ceramic material used in micro nuclear reactors, which is designed for deep space or deep sea exploration. BeO ceramic materials used in micro nuclear reactors serve two purposes, one is BeO ceramics as neutron reflector material located in the periphery of the reactor core; the other is as a new nuclear fuel element matrix material. Most of the research on BeO comes from the 1950s and 1960s, but in the face of new application requirements, it is necessary to re-master the basic performance of commercial BeO products. Therefore, this paper carries out the research on the thermal and mechanical performance, preparation of the composite fuel pellet and the irradiation effects of BeO ceramics.Firstly, in cooperation with China Minmetals, the large-size BeO samples of Φ 200 mm×200 mm and 112 mm×112 mm×30 mm were prepared by the commercial process, and then the uniformity of the samples was analyzed and tested. At the same time, the high temperature mechanical properties of BeO were studied, and the glass phase between BeO grains was the main factor affecting its strength. Finally, the stress corrosion phenomenon of BeO in water environment under the action of stress was studied at room temperature, and the mechanism of stress corrosion was clarified that the original Be-O bond in BeO was broken, and Be atoms and O atoms from water formed a new Be-O bond.Secondly, the TRISO particles with three layers of coating and the outermost layer of SiC are suitable for dispersion in the BeO matrix, and there is an amorphous glass phase between SiC and BeO, which binds SiC and BeO together. However, the conventional TRISO particles with four coating layers are not suitable for dispersion in BeO matrix, because the C atoms will diffuse into the BeO matrix, and there are obvious gaps between TRISO particles and BeO matrix. At the same time, the new BeO-based nuclear fuel elements with the sizes of Φ 50 mm×10 mm and Φ 20 mm×36 mm were prepared by using the optimal hot pressing sintering parameters, and the volume fraction of TRISO particles was 10%~24%.Thirdly, the accumulation process of irradiation damage of BeO at different irradiation temperatures was studied by molecular dynamics method. The results show that when BeO is irradiated at 600 ℃ and below, the main crystal defects produced are defect clusters, and when irradiated at 800 ℃ and above, two kinds of dislocations with the Burgers vector of 1/3<1"1" ?00>and 1/3<11"2" ?0> are produced. The molecular dynamics study of the annealing process of irradiated BeO shows that when BeO is irradiated at a lower temperature and annealed at a high temperature, the amorphous structure produced by irradiation will change into α-BeO and β-BeO. At the same time, the annealing process will also produce a dislocation ring with the Burgers vector of 1/3<1"1" ?00>, which is located on the basal plane and perpendicular to the c-axis, which is consistent with the experimental results in previous literature.
