基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

Burnup Analysis of Molten Salt Fast Reactor Based on MCNP and ORIGEN

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。

Molten salt reactor is the only one of liquid fuel reactor among the six candi-date reactors chosen by the Generation ]V International Forum （GIF） , and the molten salt circled in the primary loop is nuclear fuel as well as coolant. This promises reactor concept features peculiar characteristics such as no manufacture of fuel assembles, online fuel reprocessing and refueling, and therefore the burnup calculation of molten salt reactor differs from conventional solid fuel nuclear reactors. MCORE is a reactor physics analysis code based the MCNP and ORIGEN used for conventional solid fuel nuclear reactors. A new burnup analysis code （MCORE-MS） which considered the online reprocessing and refueling was developed based on MCORE. The preliminary study indicates that under 233 U-started mode, molten salt on-line processing can reduce fission products in the reactor core effectively, resulting in better neutron economy. And during MSFR operation, the temperature reactivity coefficient is always negative.