Nuclear Techniques ›› 2017, Vol. 40 ›› Issue (4): 40502-040502.doi: 10.11889/j.0253-3219.2017.hjs.40.040502

• NUCLEAR PHYSICS, INTERDISCIPLINARY RESEARCH • Previous Articles     Next Articles

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

HONG Shuang1,2, YANG Yongwei2, ZHANG Lu2,3, GAO Yucui2   

  1. 1 School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, China;
    2 Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000, China;
    3 University of Chinese Academy of Sciences, Beijing 100049, China
  • Received:2016-12-08 Revised:2017-01-24 Online:2017-04-10 Published:2017-04-07
  • Supported by:

    Supported by Strategic Priority Research Program of Chinese Academy of Sciences (No.XDA03030102)

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-Nth 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 keff and the neutron flux F. 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.

Key words: OpenMC, Multigroup cross section, ANISN, ENDF/B-VII.1, CENDL-3.1

CLC Number: 

  • TL329+.3