摘要
开发了一套MCNP与ORIGENS耦合的接口程序MCORGS。其在绝对通量计算与截面更新的处理上都比MCCOOR程序严格,且可以处理外源问题、快中子谱问题,对复杂几何问题适应性好;整套程序(MCNP、ORIGENS、MCORGS)均用Visual FORTRAN开发,可在WindowsXP操作系统上运行。接口程序自动化程度较高,用户输入简单。2个VVER带可燃毒物Gd的组件燃耗基准题计算结果表明,MCORGS的精度和速度都优于MCCOOR;通过计算加速器驱动的次临界系统(ADS)燃耗基准题,验证了MCORGS程序处理外源及快中子谱问题的能力。
A new code MCORGS which links MCNP and ORIGENS is developed to do the transport-burnup calculation.MCORGS can be used to compute both criticality problem and subcriticality problem with an external neutron source.It has no limitation on neutron spectrum or geometry shape.MCORGS is more versatile than MCCOOR which is dedicated to compute thermal reactors with regular shape.Visual FORTRAN is used to develop MCNP,ORIGENS and MCORGS,so it can be run in any WINDOWS operation system.MCORGS is easy to use for it automatically produces lots of information in input files.There are different treatments in absolute flux calculation and cross section update between MCORGS and MCCOOR.Results of a VVER-1000 benchmark show that MCORGS can get more accurate results in less time than MCCOOR.ADS benchmark is also calculated to validate its ability to deal with fast sub critical reactor with external source.
出处
《核动力工程》
EI
CAS
CSCD
北大核心
2010年第3期1-4,共4页
Nuclear Power Engineering
