高温堆乏燃料贮罐中子屏蔽性能计算

球床模块式高温气冷堆采用包覆颗粒球形燃料元件，在反应堆运行过程中，不断排出的乏燃料球将被装入乏燃料贮罐。乏燃料贮罐应选取适当的材料和厚度，对光子和中子进行有效屏蔽，使罐外的剂量率满足相应的限值要求。为此，使用张弛长度法和蒙特卡罗模拟法研究乏燃料贮罐的中子屏蔽性能。屏蔽材料为铁和含硼聚乙烯，计算了铁和不同B4C含量聚乙烯的屏蔽性能，并给出了乏燃料贮罐装满乏燃料球后，乏燃料球自吸收对贮罐外剂量率的影响。两种方法计算结果吻合很好，可以为实际工程中的屏蔽设计提供参考意见。

Neutron shielding effects of spent fuel tank of high temperature reactor

High temperature gas cooled reactor-pebble bed module (HTR-PM) adopts the coated particle spherical fuel elements, during the reactor's running, the constantly discharged spent fuel spheres will be loaded into the spent fuel tank. The spent fuel tank should use proper materials and thicknesses to shield gammas and neutrons effectively, and guarantee the dose limit not to be exceeded outside the tanks. Both relaxation length method and Monte Carlo simulation method were employed to study the neutrons' shielding capabilities of the spent fuel tank. Iron and borated polyethylene were chosen to be the shielding materials. The shielding capabilities of iron and borated polyethylene with different B4C contents (mass fraction 0, 5%, 10% and 15%) were calculated. The effect of the spent fuel spheres’ self-absorption to the dose rate outside the tank was also considered, when the tank was full of the spent fuel spheres. The calculation results of these two methods are in good agreement, and provide important guiding suggestions for the shielding design in the practical engineering.