【摘要】：OpenMC是麻省理工大學計算反應堆物理組開發(fā)的開源蒙特卡羅程序,能夠方便地制作適用于特定堆芯中子能譜分布的多群反應截面及高階勒讓德散射截面以用于離散坐標輸運程序ANISN的計算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1評價數(shù)據(jù)庫,利用OpenMC計算制作了ANSIN格式的多群截面并通過基準題的計算驗證計算結果的準確性。通過截面轉換程序的編寫,將OpenMC給出的堆芯各階勒讓德散射分量,堆芯中子能譜分布,散射、吸收反應率以及裂變中子產(chǎn)生速率等信息轉換為ANISN程序可讀取的截面庫格式。采用制作的截面庫利用ANINS計算有效中子增殖因子及堆芯中子通量分布。結果表明,ANISN確定論的計算結果與OpenMC給出的蒙特卡羅計算結果相吻合,驗證了這種方法可有效地為ANISN提供截面數(shù)據(jù),將來可推廣應用于二維、三維確定論中子輸運計算。
[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.
【作者單位】： 中國科學技術大學核科學與技術學院;中國科學院近代物理研究所;中國科學院大學;
【基金】：中國科學院戰(zhàn)略性先導科技專項(No.XDA03030102)資助~~
【分類號】：TL329.2
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本文編號：2476255