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本工作用Monte Carlo方法计算了核燃料废包壳缓发裂变中子形成的热中子注量率。在MCNP程序现有能力的基础上做了一些修改与补充,通过对两个典型方案的各步计算结果所做的比较分析,找到了较好的解决办法。在核燃料废包壳处理装置中(如图1),加速器发射的14MeV中子,进入废包壳篮,引起其中的裂变元素U~(235)、U~(238)、Pu~(238)和Pu~(239)等产生缓发裂变中子,实验上通过
In this work, the thermal neutron fluence rate of fission neutron bursts was calculated by Monte Carlo method. Based on the existing capabilities of MCNP program, some modifications and additions were made. By comparing and analyzing the results of two typical schemes, we found a better solution. In the nuclear fuel scrap processing device (Figure 1), the 14MeV neutron emitted by the accelerator enters the waste shell basket, causing the fission elements U 235, U 238, Pu 238 and Pu ~ (239) and other slow-release fission neutrons, experimentally passed