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

    Published: Jan 1987

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    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.


    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,

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP25691S

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