Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation

Jaakko Leppänen (Corresponding Author)

    Research output: Contribution to journalArticleScientificpeer-review

    60 Citations (Scopus)

    Abstract

    A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achieved by using the same unionized energy grid for all point-wise reaction cross sections. This speed-up results from the fact that time-consuming grid iteration is reduced to minimum, and if the unionized grid is constructed by combining the grids of all nuclides, there is no loss of data or accuracy in the calculation. The drawback of this approach is that computer memory is wasted for storing a large number of redundant data points. Memory usage may become a problem, especially in burnup calculation, in which the irradiated materials consist of several hundred actinide and fission product nuclides. The grid size easily increases to over 1 million points, requiring tens of gigabytes of memory for storing the cross section data. This paper presents two practical methods for reducing the memory demand, while trying preserve the accuracy of the original data. The calculation routines are included in the PSG2/Serpent Monte Carlo reactor physics burnup calculation code and the methods are tested in a BWR assembly burnup calculation.
    Original languageEnglish
    Pages (from-to)878-885
    Number of pages8
    JournalAnnals of Nuclear Energy
    Volume36
    Issue number7
    DOIs
    Publication statusPublished - 2009
    MoE publication typeA1 Journal article-refereed

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    Neutrons
    Data storage equipment
    Isotopes
    Fission products
    Actinides
    Physics

    Keywords

    • neutron transport theory
    • neutrons

    Cite this

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    title = "Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation",
    abstract = "A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achieved by using the same unionized energy grid for all point-wise reaction cross sections. This speed-up results from the fact that time-consuming grid iteration is reduced to minimum, and if the unionized grid is constructed by combining the grids of all nuclides, there is no loss of data or accuracy in the calculation. The drawback of this approach is that computer memory is wasted for storing a large number of redundant data points. Memory usage may become a problem, especially in burnup calculation, in which the irradiated materials consist of several hundred actinide and fission product nuclides. The grid size easily increases to over 1 million points, requiring tens of gigabytes of memory for storing the cross section data. This paper presents two practical methods for reducing the memory demand, while trying preserve the accuracy of the original data. The calculation routines are included in the PSG2/Serpent Monte Carlo reactor physics burnup calculation code and the methods are tested in a BWR assembly burnup calculation.",
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    Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation. / Leppänen, Jaakko (Corresponding Author).

    In: Annals of Nuclear Energy, Vol. 36, No. 7, 2009, p. 878-885.

    Research output: Contribution to journalArticleScientificpeer-review

    TY - JOUR

    T1 - Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation

    AU - Leppänen, Jaakko

    PY - 2009

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    AB - A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achieved by using the same unionized energy grid for all point-wise reaction cross sections. This speed-up results from the fact that time-consuming grid iteration is reduced to minimum, and if the unionized grid is constructed by combining the grids of all nuclides, there is no loss of data or accuracy in the calculation. The drawback of this approach is that computer memory is wasted for storing a large number of redundant data points. Memory usage may become a problem, especially in burnup calculation, in which the irradiated materials consist of several hundred actinide and fission product nuclides. The grid size easily increases to over 1 million points, requiring tens of gigabytes of memory for storing the cross section data. This paper presents two practical methods for reducing the memory demand, while trying preserve the accuracy of the original data. The calculation routines are included in the PSG2/Serpent Monte Carlo reactor physics burnup calculation code and the methods are tested in a BWR assembly burnup calculation.

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    KW - neutrons

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