基于OpenMC的多群截面库制作及有效性验证
发布时间：2019-05-13 23:13
【摘要】：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。
[Abstract]:OpenMC is an open source Monte Carlo program developed by the MIT Computing reactor Physics Group. It is convenient to fabricate multi-group reaction cross sections and high-order Legendre scattering cross sections suitable for the distribution of neutrons in specific cores for the calculation of discrete coordinate transport program ANISN. In this paper, based on ENDF/B- VII. 1 and CENDL-3.1 evaluation database, the multi-group cross section of ANSIN scheme is fabricated by OpenMC calculation, and the accuracy of the calculation result is verified by the calculation of reference questions. Through the programming of the cross section conversion program, the information given by OpenMC, such as the Legendre scattering components of the core, the energy spectrum distribution of the core neutrons, the scattering, the absorption reaction rate and the fission neutron generation rate, are converted into the section library format which can be read by the ANISN program. The effective neutron proliferation factor and the neutron flux distribution in the core are calculated by ANINS. The results show that the calculated results of ANISN determinism are in good agreement with the Monte Carlo calculation results given by OpenMC. It is verified that this method can effectively provide cross section data for ANISN and can be extended to two-dimensional and three-dimensional deterministic neutron transport calculation in the future.
【作者单位】： 中国科学技术大学核科学与技术学院;中国科学院近代物理研究所;中国科学院大学;
【基金】：中国科学院战略性先导科技专项(No.XDA03030102)资助~~
【分类号】：TL329.2
，
本文编号：2476255
[Abstract]:OpenMC is an open source Monte Carlo program developed by the MIT Computing reactor Physics Group. It is convenient to fabricate multi-group reaction cross sections and high-order Legendre scattering cross sections suitable for the distribution of neutrons in specific cores for the calculation of discrete coordinate transport program ANISN. In this paper, based on ENDF/B- VII. 1 and CENDL-3.1 evaluation database, the multi-group cross section of ANSIN scheme is fabricated by OpenMC calculation, and the accuracy of the calculation result is verified by the calculation of reference questions. Through the programming of the cross section conversion program, the information given by OpenMC, such as the Legendre scattering components of the core, the energy spectrum distribution of the core neutrons, the scattering, the absorption reaction rate and the fission neutron generation rate, are converted into the section library format which can be read by the ANISN program. The effective neutron proliferation factor and the neutron flux distribution in the core are calculated by ANINS. The results show that the calculated results of ANISN determinism are in good agreement with the Monte Carlo calculation results given by OpenMC. It is verified that this method can effectively provide cross section data for ANISN and can be extended to two-dimensional and three-dimensional deterministic neutron transport calculation in the future.
【作者单位】： 中国科学技术大学核科学与技术学院;中国科学院近代物理研究所;中国科学院大学;
【基金】：中国科学院战略性先导科技专项(No.XDA03030102)资助~~
【分类号】：TL329.2
，
本文编号：2476255
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