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开展了钍样品装置内钍核参数的积分中子学基础研究.参考混合堆概念设计搭建了内部放置了钍样品的一维贫铀/聚乙烯交替系统装置,采用加速器D-T中子源模拟聚变堆芯,利用前期开发的离线伽马测量方法测定了不同位置、不同中子谱情况下的232Th(n,γ)反应率,不确定度约为5%.结果显示,聚乙烯对14.1 MeV中子的慢化作用可有效提升钍俘获率,且贫铀对钍俘获率也有显著提升作用.实验结果与主流核数据库计算结果的对比显示,ENDF/B-VI.6和JENDL-3.3数据库的计算值比实验值平均约大6%,而较新的ENDF/B-VII.0数据库的计算值比实验值平均约大4%.因此,相比于之前数据库的钍核数据,ENDF/B-VII.0的计算值与实验结果匹配得较好,可作为相关概念设计的推荐核数据库.
The integral neutron basic research on the thorium nucleus parameters in the thorium sample device was carried out.According to the concept of the hybrid reactor, a one-dimensional depleted uranium / polyethylene alternating system with thorium sample was set up, and the accelerator DT neutron source was used to simulate the fusion reactor core , The reaction rate of 232Th (n, γ) at different positions and different neutron spectra was measured by the off-line Gamma-ray measurement method developed in earlier stage, and the uncertainty was about 5%. The results showed that the reaction rate of polyethylene to 14.1 MeV neutron The moderating effect can effectively increase the thorium capture rate, and the depleted uranium can also significantly improve the thorium capture rate.The comparison between the experimental results and the calculation results of the mainstream nuclear database shows that the calculated values of the ENDF / B-VI.6 and JENDL-3.3 databases are higher than that of the experimental While the newer ENDF / B-VII.0 database averages about 4% more than the experimental value, so compared to the thorium kernel data from the previous database, ENDF / B-VII.0 The calculated values match well with the experimental results, which can be used as the recommended nuclear database for the related concept design.