STP1490: Mesh Tally Radiation Damage Calculations and Application to the SNS Target System

    Ferguson, P. D.
    Oak Ridge National Laboratory, Oak Ridge, TN

    Gallmeier, F. X.
    Oak Ridge National Laboratory, Oak Ridge, TN

    Mansur, L. K.
    Oak Ridge National Laboratory, Oak Ridge, TN

    Wechsler, M. S.
    North Carolina State University, Raleigh, NC

    Pages: 6    Published: Jan 2007


    Abstract

    A new method for the calculation of radiation damage parameters in large geometries encompassing multiple materials using the MCNPX mesh tally has been developed. The method has been tested against previously published calculations of the displacement rate for protons and neutrons at the center of the SNS 316LN stainless steel target vessel nose. Displacement rates for neutrons, protons, and the total using the mesh tally method are shown to agree with previous work. The mesh tally method is also applied to the SNS aluminum moderator vessels and to the SNS inner reflector plug composed of aluminum, beryllium, and stainless steel. Results are given for displacement, helium, and silicon production rates.

    Keywords:

    displacements, helium production, MCNPX, mesh tally method, neutron irradiation, proton irradiation, radiation damage, silicon production, SNS


    Paper ID: STP45442S

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP45442S


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