Fabrication and validation of multigroup cross section library based on the OpenMC code

Shuang HONG; Yongwei YANG; Lu ZHANG; Yucui GAO

doi:10.11889/j.0253-3219.2017.hjs.40.040502

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Fabrication and validation of multigroup cross section library based on the OpenMC code

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- Fabrication and validation of multigroup cross section library based on the OpenMC code
- NUCLEAR TECHNIQUES Vol. 40, Issue 4, Article number: 040502(2017)
- 作者机构：
  
  1.中国科学技术大学核科学与技术学院合肥 230027
  2.中国科学院近代物理研究所兰州 730000
  3.中国科学院大学北京 100049
- 作者简介：
- 基金信息：
  
  Strategic Priority Research Program of Chinese Academy of Sciences
- DOI：10.11889/j.0253-3219.2017.hjs.40.040502
  CLC： TL329.3
- Received：08 December 2016，
  
  Revised：24 January 2017，
  
  Published：2017-04
- 稿件说明：
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Shuang HONG, Yongwei YANG, Lu ZHANG, et al. Fabrication and validation of multigroup cross section library based on the OpenMC code[J]. Nuclear techniques, 2017, 40(4): 040502
DOI：

Shuang HONG, Yongwei YANG, Lu ZHANG, et al. Fabrication and validation of multigroup cross section library based on the OpenMC code[J]. Nuclear techniques, 2017, 40(4): 040502 DOI： 10.11889/j.0253-3219.2017.hjs.40.040502.

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

Background

OpenMC is an open source Monte Carlo code developed by the Computational Reactor Physics Group (CRPG) of Massachusetts Institute of Technology (MIT). It is convenient to use OpenMC to generate the multigroup cross sections and high order Legendre scattering cross sections based on specific core neutron spectrum

which could be applied to the widely used discrete ordinate transport code ANISN.

Purpose

This study aims at producing the ANISN multigroup cross section library based on the ENDF/B-Ⅶ.1 and CENDL-3.1 evaluated neutron database using the OpenMC code and validating the accuracy of the calculation results through the benchmark calculation.

Methods

Since the output of OpenMC is a text file containing the 0-

scattering moments

absorption rate

scattering rate

total reaction rate

fission neutron production rate and neutron flux

we wrote a cross section convert code to match the output data with ANISN cross section library format.

Results

To validate the correction of the cross section libraries

we perform a critical benchmark and calculate the neutron effective multiplication factor

eff

and the neutron flux

. It shows that the results given by ANISN using the library generated by OpenMC are in good agreement with Monte Carlo calculation.

Conclusion

The OpenMC code can be used to provide the multigroup cross sections and high order Legendre scattering cross sections for the ANISN code effectively and this can be applied to the two-dimensional and three-dimensional neutron transport calculation in the future.

关键词

Keywords

references

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