A fission matrix approach to calculate pin-wise 3-D fission density distribution

William J. Walters, Nathan Roskoff, Alireza Haghighat

    Research output: Chapter in Book/Report/Conference proceedingConference contribution

    5 Citations (Scopus)

    Abstract

    This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

    Original languageEnglish (US)
    Title of host publicationMathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
    PublisherAmerican Nuclear Society
    Pages1351-1361
    Number of pages11
    ISBN (Electronic)9781510808041
    StatePublished - Jan 1 2015
    EventMathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 - Nashville, United States
    Duration: Apr 19 2015Apr 23 2015

    Publication series

    NameMathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
    Volume2

    Other

    OtherMathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
    CountryUnited States
    CityNashville
    Period4/19/154/23/15

    Fingerprint

    3D
    fission
    density distribution
    Calculate
    matrices
    spent fuels
    Arrangement
    Coefficient
    Criticality
    coefficients
    Cooling
    Multiplication
    multiplication
    methodology
    Methodology
    Dependent
    cooling

    All Science Journal Classification (ASJC) codes

    • Mathematics(all)
    • Nuclear and High Energy Physics

    Cite this

    Walters, W. J., Roskoff, N., & Haghighat, A. (2015). A fission matrix approach to calculate pin-wise 3-D fission density distribution. In Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 (pp. 1351-1361). (Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015; Vol. 2). American Nuclear Society.
    Walters, William J. ; Roskoff, Nathan ; Haghighat, Alireza. / A fission matrix approach to calculate pin-wise 3-D fission density distribution. Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015. American Nuclear Society, 2015. pp. 1351-1361 (Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015).
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    abstract = "This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.",
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    Walters, WJ, Roskoff, N & Haghighat, A 2015, A fission matrix approach to calculate pin-wise 3-D fission density distribution. in Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015. Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015, vol. 2, American Nuclear Society, pp. 1351-1361, Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015, Nashville, United States, 4/19/15.

    A fission matrix approach to calculate pin-wise 3-D fission density distribution. / Walters, William J.; Roskoff, Nathan; Haghighat, Alireza.

    Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015. American Nuclear Society, 2015. p. 1351-1361 (Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015; Vol. 2).

    Research output: Chapter in Book/Report/Conference proceedingConference contribution

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    AU - Roskoff, Nathan

    AU - Haghighat, Alireza

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    N2 - This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

    AB - This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

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    M3 - Conference contribution

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    Walters WJ, Roskoff N, Haghighat A. A fission matrix approach to calculate pin-wise 3-D fission density distribution. In Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015. American Nuclear Society. 2015. p. 1351-1361. (Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015).