共查询到13条相似文献,搜索用时 187 毫秒
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在蒙特卡罗程序中统计共轭通量并计算中子动力学参数的方法 总被引:1,自引:1,他引:0
文章介绍了在蒙特卡罗程序中,使用反复裂变几率的统计结果作为共轭通量的估计,并作为权重函数计算动力学参数βeff和Λ的方法,阐释了在连续能量蒙特卡罗程序MCNP和多群蒙特卡罗程序MCMG中实现这种方法的过程。数值校验结果表明:在几乎不带来附加计算量的同时,在MCMG中使用该方法统计得到的共轭通量与ANISN的共轭通量计算结果符合较好,在MCNP中使用该方法计算得到的中子动力学参数与基准测量结果符合较好。在蒙特卡罗程序中实现了高效率计算中子动力学参数的功能,为蒙特卡罗程序进一步用于反应堆动态行为的分析奠定了基础。 相似文献
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基于离散角方法,开发了蒙特卡罗多群数据库生成程序MGXSMC,该程序可以实现从输入文件读取截面数据或者从指定格式的截面库中读取截面,产生可供蒙特卡罗程序MCNP或RMC计算的数据库,并且可自动生成相应的索引文件列表。采用二维两群不带反射层的国际原子能机构(IAEA)压水堆(PWR)基准题和铅基快堆(RBEC-M)基准题对MGXSMC程序加工产生的核数据进行验证,计算结果表明,采用P5阶近似多群截面与连续点截面计算的有效增殖系数(keff)结果相差24 pcm(1pcm=10-5),而采用P0阶近似多群截面与连续点截面计算的keff结果相差较大。由此说明蒙特卡罗多群数据库的制作方法和所开发的程序是正确的,同时,中子各向异性散射对铅基快堆计算结果影响较大,故制作蒙特卡罗多群数据库时应加入中子散射角数据。 相似文献
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Based on the discrete angle method, a Monte Carlo multi-group cross section generation program MGXSMC was developed. This program can read the cross section data from an input file or read the cross section from a library in a specified format to generate the multi-group cross section for MCNP or RMC. The corresponding index file list can be automatically generated. The two-dimensional two-group IAEA pressurized water reactor (PWR) benchmark and lead-based fast reactor (RBEC-M) benchmark were used to verify the cross section library generated by the MGXSMC program. The calculation results show that the difference between the calculated result of the P5 order approximate multigroup section and the continuous point cross section is 24 pcm (1pcm = 10-5), and the difference of the keff result calculated by the P0 order approximate multigroup section and the continuous point section is large. This shows that the method and the program developed for the Monte Carlo Group Section Library are correct. At the same time, the neutron anisotropic scattering has a large impact on the calculation results of the lead-based fast reactor. Therefore, when the Monte Carlo Group Section library is produced, the neutron scattering angle data should be added. 相似文献
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In the nuclear reactor design, a code for automatically generated multi-temperature continuous-energy neu- tron cross section data library, which is called AMTND for short, was designed and developed to meet the need of the reactor core design coupled with thermal-hydraulic design. The code can provide a point-wise cross- section at any temperature for a Monte Carlo neutron transport program, such as MCNE In ensuring that the nuclear data produced by AMTND meets the testing of critical benchmark experiments, the time-consumed by the nuclear data generating of AMTND compared with NJOY's was carried out and the result shows the code's excellence. In order to test the accuracy of the code, out and the results verified the code preliminarily. the Doppler coefficient test benchmark was also carried 相似文献
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OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。 相似文献