Radiation transport capabilities in the Serpent 2 Monte Carlo code

    Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review


    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
    Title of host publicationProceedings
    Subtitle of host publication28th International Conference Nuclear Energy for New Europe, NENE 2019
    ISBN (Electronic)978-961-6207-47-8
    Publication statusPublished - 2019
    MoE publication typeA4 Article in a conference publication
    Event28th International Conference Nuclear Energy for New Europe, NENE 2019 - Grand Hotel Bernardin, Portorož, Slovenia
    Duration: 9 Sept 201912 Sept 2019
    Conference number: 28


    Conference28th International Conference Nuclear Energy for New Europe, NENE 2019
    Abbreviated titleNENE 2019
    Internet address


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