核爆发电问题中核燃料循环的中子学研究

Neutronic Performance in a Peaceful Nuclear Explosion Reactor

研究核爆聚变电站中的一类中子学问题———如何利用核爆炸产生的大量中子生产核燃料.首先,根据核爆中子场的特点,确定可以利用的中子能量范围.用慢化理论对慢化材料的厚度进行估计,并用MCNP程序进行数值计算.研究如何利用中子反照效应减少造材料层的厚度,最后提出一个较合理的造材料技术路线.

The neutronic behavior in a peaceful nuclear explosion reactor is analyzed, ^238 U and ^232 Th are used in the reactor to produce fission materials, ^239 Pu and ^233 U. MCNP is used to calculate thickness of the neutron slowing down material C2H4 . The thickness of ^232 Th is predicted in which the neutron capturing ability is above 50% . Considering the effect of reflected saving, the thickness of ^232Th is reduced.