Solution of neutron-transport Multigroup equations system in subcritical systems

I. V. Shamanin, S. V. Bedenko, V. N. Nesterov, I. O. Lutsik, A. A. Prets

Research output: Contribution to journalArticle

4 Citations (Scopus)

Abstract

The iteration process of the neutron-transport equation solution in diffuse 28-group approximation was implemented. A PuO2-fueled thermoelectric generator providing thermal and electric power for space vehicles was researched as a system. The search for more true function of neutron distribution for zero iteration is considered an important simulation stage, which allows increasing accuracy of the neutron transport equation solution by the method of iteration in diffuse multigroup approximation. Moreover it makes it possible to improve neutron-physical characteristics of the researched system, when there is some small increase in the number of computed iterations. It should be noted that the required functions can be computed using such specialized programs as SOURCE-4C and NEDIS-2m. These programs allow calculating the intensity and spectrum of the neutrons in (a,n) reactions and at spontaneous fission, preparing files with the output data in the form usable for solving the neutron transport equation in the programs of MCU, MCNP and Scale type. In the work the neutrons intensity and spectrum (the neutron distribution function at zero iteration) for the system of multigroup neutron transport equations were obtained using approximation of a wide range of calculated and experimental data (with the error of not more than 5%) with high accuracy. Spectral and integral neutron-physical characteristics of the system were obtained as the result of the multigroup equations system solution, and the resulting calculation data were verified. The applied approach is considered economical from the point of view of computational cost (as the value of neutron flux density fractions agree at the 3rd iteration) and expenditures connected with nuclear data bank storage. This approach can be used in tasks of nuclear and radiation safety. The computations were performed using the system of group constants BNAB-78,-93 and other available libraries of evaluated nuclear data: ROSFOND, BROND, BNAB, EXFOR and ENDSF. High accuracy results agree with the results obtained in ANISN, MCNP (JENDL-3.2.) and SCALE-4.3(KENO-V.a, ENDF/B-V).

Original languageEnglish
Pages (from-to)38-49
Number of pages12
JournalIzvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika
Volume2017
Issue number4
DOIs
Publication statusPublished - 2017

Keywords

  • Diffusion equation
  • Multigroup approximation
  • Neutron transport
  • Subcritical systems

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

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