|
In reprocessor, solution tanks, sometimes containing rasching rings, are always composed of anomalous cells. Based on AMPX code system and by analysing this kind of cells, the present work was done to find a rational approximate method for resonance self-shieldling analysis and cell homogenization. A reliable multigroup cross section library for transportation calculations is provided. It gives an useful method for critical evaluation of anomalous cells system. 本文以后处理厂溶解槽、含拉西环溶液为例,通过分析其栅元特性,采取一系列合理近似,在现有的AMPX群截面库计算系统的基础上,寻找较为理想的栅元共振自屏及截面权重平均的近似处理方法,为最终迁移计算(或蒙特-卡洛模拟计算)提供较为满意的群截面库。从而为这类非常规栅元系统的临界安全分析开辟一条可靠途径。 This paper introduces the direction biasing, a special Monte Carlo variance reduction technique. Exponential transform method is used to eliminate the underirable fluctuations of direction biasing. Results from γ ray transportation calculations indicate that direction biasing in conjunction with the exponential transform greatly enhances the efficiency of Monte Carlo transport simulation. ( 本文介绍了一种非常独特的减方差技巧——方向偏移。通过结合使用指数变换方法消除了方向偏移过程中粒子权重起伏大的问题,从而大大提高了计算效率。将此法用于光子输运的计算中,取得了很好的效果。 For the neutron transportation calculation, accurate fission neutron source is the prerequisite and fundament ensuring that the calculation results are reliable. As the fuel assemblies burn-up is increased, nuclides such as 235U gradually consumed, and nuclides such as 239Pu、240Pu、241Pu gradually accumulated, resulting in the variation of average number of fission neutrons (ν) and average releasing energy of each fission(E), and the value of ν/E raised along with the increasing of fuel assemblies... For the neutron transportation calculation, accurate fission neutron source is the prerequisite and fundament ensuring that the calculation results are reliable. As the fuel assemblies burn-up is increased, nuclides such as 235U gradually consumed, and nuclides such as 239Pu、240Pu、241Pu gradually accumulated, resulting in the variation of average number of fission neutrons (ν) and average releasing energy of each fission(E), and the value of ν/E raised along with the increasing of fuel assemblies burn-up. So the impact of fuel assemblies burn-up on core fission neutron source parameters shall be considered in the neutron transportation calculation. 在中子输运计算中,准确的裂变中子源是保证计算结果可靠性的前提和基础。随着组件燃耗的加深,235U等核素不断消耗,239Pu、240Pu、241Pu等核素不断积累,导致每次裂变产生的平均中子数ν和裂变释放的平均能量E也随之变化,裂变中子源的归一化因子随燃耗增加而增加。因此,在中子输运计算中,必须考虑组件燃耗对堆芯的裂变中子源参数的影响。
|