STP956

    Comparison of Calculated Integral Values Using Measured and Calculated Neutron Spectra for Fusion Neutronics Analyses

    Published: Jan 1987


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    Abstract

    The kerma heat production density, tritium production density, and dose in a lithiumfluoride pile with a deuterium-tritium neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both of the UFO and MORSE-CV values agreed within the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source.

    Keywords:

    nuclear fusion, integral values, kerma heat production, tritium production, dose, neutron spectrum, deuterium-tritium neutron source, fusion blanket, lithium-fluoride, neutron spectrometer, data processing code, unfolding, Monte Carlo code, group cross section set, covariance


    Author Information:

    Sekimoto, H
    Associate professor, Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo,


    Paper ID: STP25691S

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP25691S


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