Radiation transport capabilities in the Serpent 2 Monte Carlo code

    Research output: Contribution to conferenceConference articleScientificpeer-review

    Abstract

    The Serpent Monte Carlo code was originally developed as a computational tool for various neutron transport problems encountered in reactor physics applications, but in recent years the scope has been broadened to new fields, including radiation transport and fusion neutronics. The development work has been focused on three topics: 1) Photon transport calculations involving source terms comprised activated materials; 2) Advanced CAD-based geometry types and 3) Weight-window based variance reduction methods with a built-in importance solver.

    The photon physics model was originally implemented for the purpose of accurate heat deposition calculations in reactor analysis, but the capability is also applicable to general radiation transport applications. The radioactive decay source mode combines material compositions obtained from a separate burnup or activation calculation with ENDF decay spectra, and forms the source terms automatically without additional user input.

    The CAD-based geometry type is based on the STL data format, which is widely used for 3D printing and therefore supported by virtually every CAD tool. Geometries comprised of STL solids are handled as separate universes, which can be combined with conventional CSG and other available geometry types. Apart from exporting the native CAD format into STL, the procedure involves no format conversions.

    Variance reduction in Serpent is based on a conventional mesh-based weight-window technique. The weight-window mesh can be read from MCNP WWINP format files, or produced using an built-in solver based on the response matrix method. The routine is capable of calculating importances with respect to one or multiple responses, and an iterative global variance reduction scheme can be used to gradually expand the mesh throughout the whole geometry. The supported mesh types include conventional Cartesian, cylindrical and hexagonal mesh, but also unevenly-spaced and adaptive octree mesh types.

    The purpose of this paper is to provide a general overview and practical examples of the methodology used in Serpent for radiation transport calculations, including its advantages, major limitations and future prospects.
    Original languageEnglish
    Publication statusPublished - 2019
    MoE publication typeNot Eligible
    Event28th International Conference Nuclear Energy for New Europe, NENE 2019 - Grand Hotel Bernardin, Portorož, Slovenia
    Duration: 9 Sep 201912 Sep 2019
    Conference number: 28
    http://www.nss.si/nene2019/

    Conference

    Conference28th International Conference Nuclear Energy for New Europe, NENE 2019
    Abbreviated titleNENE 2019
    CountrySlovenia
    CityPortorož
    Period9/09/1912/09/19
    Internet address

    Fingerprint

    radiation transport
    mesh
    computer aided design
    format
    geometry
    reactor physics
    general overviews
    radioactive decay
    photons
    files
    printing
    matrix methods
    universe
    fusion
    reactors
    activation
    methodology
    neutrons
    heat
    physics

    Cite this

    Leppänen, J. (2019). Radiation transport capabilities in the Serpent 2 Monte Carlo code. Paper presented at 28th International Conference Nuclear Energy for New Europe, NENE 2019, Portorož, Slovenia.
    Leppänen, Jaakko. / Radiation transport capabilities in the Serpent 2 Monte Carlo code. Paper presented at 28th International Conference Nuclear Energy for New Europe, NENE 2019, Portorož, Slovenia.
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    abstract = "The Serpent Monte Carlo code was originally developed as a computational tool for various neutron transport problems encountered in reactor physics applications, but in recent years the scope has been broadened to new fields, including radiation transport and fusion neutronics. The development work has been focused on three topics: 1) Photon transport calculations involving source terms comprised activated materials; 2) Advanced CAD-based geometry types and 3) Weight-window based variance reduction methods with a built-in importance solver.The photon physics model was originally implemented for the purpose of accurate heat deposition calculations in reactor analysis, but the capability is also applicable to general radiation transport applications. The radioactive decay source mode combines material compositions obtained from a separate burnup or activation calculation with ENDF decay spectra, and forms the source terms automatically without additional user input.The CAD-based geometry type is based on the STL data format, which is widely used for 3D printing and therefore supported by virtually every CAD tool. Geometries comprised of STL solids are handled as separate universes, which can be combined with conventional CSG and other available geometry types. Apart from exporting the native CAD format into STL, the procedure involves no format conversions.Variance reduction in Serpent is based on a conventional mesh-based weight-window technique. The weight-window mesh can be read from MCNP WWINP format files, or produced using an built-in solver based on the response matrix method. The routine is capable of calculating importances with respect to one or multiple responses, and an iterative global variance reduction scheme can be used to gradually expand the mesh throughout the whole geometry. The supported mesh types include conventional Cartesian, cylindrical and hexagonal mesh, but also unevenly-spaced and adaptive octree mesh types.The purpose of this paper is to provide a general overview and practical examples of the methodology used in Serpent for radiation transport calculations, including its advantages, major limitations and future prospects.",
    author = "Jaakko Lepp{\"a}nen",
    year = "2019",
    language = "English",
    note = "28th International Conference Nuclear Energy for New Europe, NENE 2019, NENE 2019 ; Conference date: 09-09-2019 Through 12-09-2019",
    url = "http://www.nss.si/nene2019/",

    }

    Leppänen, J 2019, 'Radiation transport capabilities in the Serpent 2 Monte Carlo code', Paper presented at 28th International Conference Nuclear Energy for New Europe, NENE 2019, Portorož, Slovenia, 9/09/19 - 12/09/19.

    Radiation transport capabilities in the Serpent 2 Monte Carlo code. / Leppänen, Jaakko.

    2019. Paper presented at 28th International Conference Nuclear Energy for New Europe, NENE 2019, Portorož, Slovenia.

    Research output: Contribution to conferenceConference articleScientificpeer-review

    TY - CONF

    T1 - Radiation transport capabilities in the Serpent 2 Monte Carlo code

    AU - Leppänen, Jaakko

    PY - 2019

    Y1 - 2019

    N2 - The Serpent Monte Carlo code was originally developed as a computational tool for various neutron transport problems encountered in reactor physics applications, but in recent years the scope has been broadened to new fields, including radiation transport and fusion neutronics. The development work has been focused on three topics: 1) Photon transport calculations involving source terms comprised activated materials; 2) Advanced CAD-based geometry types and 3) Weight-window based variance reduction methods with a built-in importance solver.The photon physics model was originally implemented for the purpose of accurate heat deposition calculations in reactor analysis, but the capability is also applicable to general radiation transport applications. The radioactive decay source mode combines material compositions obtained from a separate burnup or activation calculation with ENDF decay spectra, and forms the source terms automatically without additional user input.The CAD-based geometry type is based on the STL data format, which is widely used for 3D printing and therefore supported by virtually every CAD tool. Geometries comprised of STL solids are handled as separate universes, which can be combined with conventional CSG and other available geometry types. Apart from exporting the native CAD format into STL, the procedure involves no format conversions.Variance reduction in Serpent is based on a conventional mesh-based weight-window technique. The weight-window mesh can be read from MCNP WWINP format files, or produced using an built-in solver based on the response matrix method. The routine is capable of calculating importances with respect to one or multiple responses, and an iterative global variance reduction scheme can be used to gradually expand the mesh throughout the whole geometry. The supported mesh types include conventional Cartesian, cylindrical and hexagonal mesh, but also unevenly-spaced and adaptive octree mesh types.The purpose of this paper is to provide a general overview and practical examples of the methodology used in Serpent for radiation transport calculations, including its advantages, major limitations and future prospects.

    AB - The Serpent Monte Carlo code was originally developed as a computational tool for various neutron transport problems encountered in reactor physics applications, but in recent years the scope has been broadened to new fields, including radiation transport and fusion neutronics. The development work has been focused on three topics: 1) Photon transport calculations involving source terms comprised activated materials; 2) Advanced CAD-based geometry types and 3) Weight-window based variance reduction methods with a built-in importance solver.The photon physics model was originally implemented for the purpose of accurate heat deposition calculations in reactor analysis, but the capability is also applicable to general radiation transport applications. The radioactive decay source mode combines material compositions obtained from a separate burnup or activation calculation with ENDF decay spectra, and forms the source terms automatically without additional user input.The CAD-based geometry type is based on the STL data format, which is widely used for 3D printing and therefore supported by virtually every CAD tool. Geometries comprised of STL solids are handled as separate universes, which can be combined with conventional CSG and other available geometry types. Apart from exporting the native CAD format into STL, the procedure involves no format conversions.Variance reduction in Serpent is based on a conventional mesh-based weight-window technique. The weight-window mesh can be read from MCNP WWINP format files, or produced using an built-in solver based on the response matrix method. The routine is capable of calculating importances with respect to one or multiple responses, and an iterative global variance reduction scheme can be used to gradually expand the mesh throughout the whole geometry. The supported mesh types include conventional Cartesian, cylindrical and hexagonal mesh, but also unevenly-spaced and adaptive octree mesh types.The purpose of this paper is to provide a general overview and practical examples of the methodology used in Serpent for radiation transport calculations, including its advantages, major limitations and future prospects.

    M3 - Conference article

    ER -

    Leppänen J. Radiation transport capabilities in the Serpent 2 Monte Carlo code. 2019. Paper presented at 28th International Conference Nuclear Energy for New Europe, NENE 2019, Portorož, Slovenia.