STP286

    Neutron Dosimetry for Materials Irradiation Studies

    Published: Jan 1960


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    Abstract

    The correct interpretation of radiation effects upon materials depends upon accurate knowledge of neutron exposures. An analysis of the problems associated with neutron dosimetry for materials irradiation experiments in research reactors is presented along with a discussion of neutron flux data as a factor in the experimental environment. Some of the problems presented include: choosing the best monitors, interpreting preliminary neutron flux surveys, measuring and interpreting flux levels under changing reactor conditions, and using flux data in the analysis of radiation effects. These problems are discussed, citing examples from practical experience in the Argonne CP-5 Reactor, the Brookhaven Graphite Reactor, the Oak Ridge Low-Intensity Test Reactor (LITR), and the Materials Testing Reactor (MTR).


    Author Information:

    Steele, L. E.
    U. S. Naval Research Laboratory, Washington, D. C.,

    Hawthorne, J. R.
    U. S. Naval Research Laboratory, Washington, D. C.,


    Paper ID: STP46371S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP46371S


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