【摘要】 三维多群P3 近似MonteCarlo中子输运程序MCMGP3是为反应堆临界安全计算设计的 ,它是从连续点截面中子—光子耦合输运MonteCarlo程序MCNP发展而来 ,程序用多群截面代替了MCNP程序的连续点截面 ,但保留了MCNP程序的几何处理能力 ,计数能力和降低方差技巧及图形功能。能群数可扩展 ,使用宏观截面或微观截面均可 ,中子角分布采用P3 近似和广义Gaussian求积。多个基准问题结果显示 ,MCMGP3程序结果与其它方法计算结果符合很好 ,计算还表明在同样计算精度下 ,MCMGP3程序的计算时间较MCNP程序少得多。此外 ,MCMGP3程序还实现了与WIMS程序的连算 ,可作反应堆全堆数值模拟。更多还原

【Abstract】 It outlines MCMGP3A 3D multigroup P3 Monte Carlo neutron transport code, which is designed for computation of reactor critical safety. It is developed from the coupled neutron and photon transport Monte Carlo code MCNP with continual point cross sections and the continual point cross section module of MCNP code has been replaced by multigroup cross section module. MCMGP3 code is with the capablilities of MCNP geometry treatment, counting, deviation reducing techniques and plot. Either macroscopic or microscopic cross sections can be used.The neutron scattering angular distribution adopts P\-3 approximation and generalized Gaussian quadrature scheme. The sample problem results prove the computational correctness of the code. MCMGP3 code can treat steady or unsteady porblems with exterior source and critical problems. It suits various cross section formats and the energy groups can be expanded easily from one group to multigroup. In addition, the MCMGP3 code has successfully realized the coupled with the reactor lattic code WIMS. It can be used to simulate the full reactor problems.更多还原