Integral Data Test of HENDL1.0/MG with Neutronics Shielding Experiments(Ⅱ)

The multi-group working nuclear data library HENDL1.0/MG is numerically tested with a series of existent benchmark spherical shell experiments (Si, Cr, Fe, Cu, Zr and Nb) by calculations using the multi-functional neutronics code VisualBUS. The ratio of calculated/measured neutron leakage rates and the neutron leakage spectra are presented in tabular and figural forms.The results from the calculations with the code ANISN and IAEA data library FENDL2.0/MGwere also included for comparison, where the origination of the data used is different from that of HENDL1.0/MG.     

Integral Data Test of HENDL1.0／MG and VisualBUS withNeutronics Shielding Experiments （Ⅰ）

HENDL1.0/MG, a multi-group working library of the Hybrid Evaluated Nuclear Data Library, was home-developed by the FDS Team of ASIPP (Institute of Plasma Physics,Chinese Academy of Sciences) on the basis of several national data libraries. To validate and qualify the process of producing HENDL1.0/MG, simulating calculations of a series of existent spherical shell benchmark experiments (A1, Mo, Co, Ti, Mn, W, Be and V) have been performed with HENDL1.0/MG and the multifunctional neutronics code system named VisualBUS homedeveloped also by FDS Team.     

混合评价核数据库HENDL1.0的研制与基准检验

为满足聚变 裂变次临界混合堆设计和其他相关研究的需要 ,以世界几个主要基本评价核数据库为数据来源 ,通过优选建立了名为HENDL1 .0 /E的多用途核数据库 ,采用国际通行的核数据库处理程序系统NJOY和TRANSX等程序制作了相应的工作数据库 ,其中包括多能群输运截面库HENDL1 .0 /MG、连续能量点状输运截面库HENDL1 .0 /MC、燃耗数据库HENDL1 .0 /BU和响应函数库HENDL1 .0 /RF ,利用世界上流行的中子输运程序对已有的一系列基准检验实验进行模拟计算和比较分析以检验混合库HENDL1 .0的正确性和有效性。     

Integral Data Test of HENDL1.0／MG with Neutronics Shielding Experiments （Ⅱ）

The multi-group working nuclear data library HENDL1.0/MG is numerically tested with a series of existent benchmark spherical shell experiments (Si, Cr, Fe, Cu, Zr and Nb) by calculations using the multi-functional neutronics code VisualBUS. The ratio of calculated/measured neutron leakage rates and the neutron leakage spectra are presented in tabular and figural forms. The results from the calculations with the code ANISN and IAEA data library FENDL2.0/MG were also included for comparison, where the origination of the data used is different from that of HENDL1.0/MG.     

聚变驱动次临界堆概念设计研究

在深入分析相关领域研究发展状况的基础上,提出一个可实现核废料嬗变、可裂变重金属核燃料和氚核燃料增殖、产能等多种功能的次临界反应堆概念。概念有托卡马克实验装置参数适量外推的等离子体物理和技术水平的聚变堆芯、高压He气/液态LiPb双冷却系统嬗变包层。介绍了系统基本概况和参考结构形式,阐述了系统聚变等离子体物理学、包层中子物理学和包层热工水力学设计的基本原则和主要工程约束限制条件,指出系统相关的安全和潜在环境影响问题,给出典型的参考设计参数。     

聚变驱动次临界堆双冷嬗变包层中子学设计与分析

对聚变驱动次临界堆的多功能双冷核废料嬗变包层进行了中子学设计和分析,设计目标是:①氚和钚燃料自持;②较少的初装料得到较高的废料嬗变率。使用的程序是自主开发的多功能中子输运/燃耗/优化程序VisuaIBUs1.0,相应的数据库是175群中子/42群光子的多群数据库HENDL1.0/MG。