STP484

    Neutron Fluence Limit Determinations for Some Fast Flux Test Facility Components

    Published: Jan 1970


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    Abstract

    This paper describes the analytical procedures that have been developed at Battelle-Northwest Laboratory (BNWL) and Westinghouse Advanced Reactors Division (ARD) to determine neutron fluence limits for certain components of the Fast Flux Test Facility (FFTF), namely the core barrel, core support structure, and the reactor vessel.

    Keywords:

    irradiation, neutron flux, radiation effects, fast reactors (nuclear), degradation, mechanical properties, crystal lattices, helium, grain boundaries, deformation, ductility, nuclear reactor materials, stainless steels, temperature, tests


    Author Information:

    Moen, RA
    Research engineer and manager, Pacific Northwest Laboratory, Battelle Memorial Institute, Richland, Wash.

    Tobin, JC
    Research engineer and manager, Pacific Northwest Laboratory, Battelle Memorial Institute, Richland, Wash.

    Thomas, KC
    Manager, Westinghouse Electric Corp., Madison, Pa.


    Paper ID: STP26640S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP26640S


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