原子能科学技术2017，Vol.51Issue(12)：2230-2234,5.DOI:10.7538/yzk.2017.51.12.2230
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
摘要
Abstract
Molten salt reactor is the only one of liquid fuel reactor among the six candi-date reactors chosen by the Generation Ⅳ 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 .关键词
熔盐快堆/燃耗计算/MCNP/ORIGENKey words
molten salt fast reactor/burnup calculation/MCNP/ORIGEN分类
能源科技引用本文复制引用
张俊,张大林,王成龙,田文喜,秋穗正,苏光辉..基于MCNP和ORIGEN的熔盐快堆燃耗分析计算[J].原子能科学技术,2017,51(12):2230-2234,5.基金项目
国家自然科学基金资助项目(91326201 ) （91326201 ）
