Monte Carlo current-based diffusion coefficients: Application to few-group constants generation in Serpent

E. Dorval, J. Leppänen

    Research output: Contribution to journalArticleScientificpeer-review

    8 Citations (Scopus)

    Abstract

    Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core were studied by combined diffusion theory and Monte Carlo methods. Few-group XS data sets generated by the Monte Carlo code Serpent for both normal and sodium-voided cells were used in the multi-group diffusion code TRIZ. Two different approaches were adopted in order to account for neutron leakage at fuel assembly level: a radially-reflected and axially-heterogeneous model with vacuum boundary conditions at the bottom and the top; and a more typical infinite cell calculation, followed by criticality spectrum corrections. In addition to the standard diffusion coefficients calculated by Serpent, a novel method for the calculation of directional diffusion coefficients was implemented and tested, yielding satisfactory results for normal and sodium-voided conditions, using Monte Carlo results as a reference. The feasibility of using the B1B1 criticality spectrum as a weighting function for these diffusion coefficients was also tested, and slightly better estimates of k-eff were obtained when compared to the direct use of diffusion coefficients arising from the solution of the B1B1 equations.
    Original languageEnglish
    Pages (from-to)104-116
    JournalAnnals of Nuclear Energy
    Volume78
    DOIs
    Publication statusPublished - 2015
    MoE publication typeA1 Journal article-refereed

    Fingerprint

    Sodium
    Reactor cores
    Fast reactors
    Neutrons
    Monte Carlo methods
    Boundary conditions
    Vacuum

    Keywords

    • Serpent
    • MOnte Carlo
    • directional diffusion coefficient
    • TRIZ
    • SFR
    • critically spectrum

    Cite this

    @article{39f472b1453a4cefae5c599713373d13,
    title = "Monte Carlo current-based diffusion coefficients: Application to few-group constants generation in Serpent",
    abstract = "Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core were studied by combined diffusion theory and Monte Carlo methods. Few-group XS data sets generated by the Monte Carlo code Serpent for both normal and sodium-voided cells were used in the multi-group diffusion code TRIZ. Two different approaches were adopted in order to account for neutron leakage at fuel assembly level: a radially-reflected and axially-heterogeneous model with vacuum boundary conditions at the bottom and the top; and a more typical infinite cell calculation, followed by criticality spectrum corrections. In addition to the standard diffusion coefficients calculated by Serpent, a novel method for the calculation of directional diffusion coefficients was implemented and tested, yielding satisfactory results for normal and sodium-voided conditions, using Monte Carlo results as a reference. The feasibility of using the B1B1 criticality spectrum as a weighting function for these diffusion coefficients was also tested, and slightly better estimates of k-eff were obtained when compared to the direct use of diffusion coefficients arising from the solution of the B1B1 equations.",
    keywords = "Serpent, MOnte Carlo, directional diffusion coefficient, TRIZ, SFR, critically spectrum",
    author = "E. Dorval and J. Lepp{\"a}nen",
    year = "2015",
    doi = "10.1016/j.anucene.2014.12.011",
    language = "English",
    volume = "78",
    pages = "104--116",
    journal = "Annals of Nuclear Energy",
    issn = "0306-4549",
    publisher = "Elsevier",

    }

    Monte Carlo current-based diffusion coefficients : Application to few-group constants generation in Serpent. / Dorval, E.; Leppänen, J.

    In: Annals of Nuclear Energy, Vol. 78, 2015, p. 104-116.

    Research output: Contribution to journalArticleScientificpeer-review

    TY - JOUR

    T1 - Monte Carlo current-based diffusion coefficients

    T2 - Application to few-group constants generation in Serpent

    AU - Dorval, E.

    AU - Leppänen, J.

    PY - 2015

    Y1 - 2015

    N2 - Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core were studied by combined diffusion theory and Monte Carlo methods. Few-group XS data sets generated by the Monte Carlo code Serpent for both normal and sodium-voided cells were used in the multi-group diffusion code TRIZ. Two different approaches were adopted in order to account for neutron leakage at fuel assembly level: a radially-reflected and axially-heterogeneous model with vacuum boundary conditions at the bottom and the top; and a more typical infinite cell calculation, followed by criticality spectrum corrections. In addition to the standard diffusion coefficients calculated by Serpent, a novel method for the calculation of directional diffusion coefficients was implemented and tested, yielding satisfactory results for normal and sodium-voided conditions, using Monte Carlo results as a reference. The feasibility of using the B1B1 criticality spectrum as a weighting function for these diffusion coefficients was also tested, and slightly better estimates of k-eff were obtained when compared to the direct use of diffusion coefficients arising from the solution of the B1B1 equations.

    AB - Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core were studied by combined diffusion theory and Monte Carlo methods. Few-group XS data sets generated by the Monte Carlo code Serpent for both normal and sodium-voided cells were used in the multi-group diffusion code TRIZ. Two different approaches were adopted in order to account for neutron leakage at fuel assembly level: a radially-reflected and axially-heterogeneous model with vacuum boundary conditions at the bottom and the top; and a more typical infinite cell calculation, followed by criticality spectrum corrections. In addition to the standard diffusion coefficients calculated by Serpent, a novel method for the calculation of directional diffusion coefficients was implemented and tested, yielding satisfactory results for normal and sodium-voided conditions, using Monte Carlo results as a reference. The feasibility of using the B1B1 criticality spectrum as a weighting function for these diffusion coefficients was also tested, and slightly better estimates of k-eff were obtained when compared to the direct use of diffusion coefficients arising from the solution of the B1B1 equations.

    KW - Serpent

    KW - MOnte Carlo

    KW - directional diffusion coefficient

    KW - TRIZ

    KW - SFR

    KW - critically spectrum

    U2 - 10.1016/j.anucene.2014.12.011

    DO - 10.1016/j.anucene.2014.12.011

    M3 - Article

    VL - 78

    SP - 104

    EP - 116

    JO - Annals of Nuclear Energy

    JF - Annals of Nuclear Energy

    SN - 0306-4549

    ER -