STP538: Theoretical Analysis of Sodium Removal from Fast Flux Test Facility Fuel Subassemblies

    Borisch, R. R.
    Development engineer, Westinghouse Hanford Company, Richland, Wash.

    Pages: 10    Published: Jan 1937


    Abstract

    The Fast Flux Test Facility (FFTF) is a 400 MWt, sodium cooled, nuclear fuels and materials test reactor being developed at the USAEC's Hanford Engineering Development Laboratory. The reactor fuel is clad with Type 316 stainless steel. In order to examine fuel, the sodium coolant must be removed in a manner which will not degrade the fuel cladding. Several possible processes are reviewed and a moderately detailed sizing analysis is given for the chosen process which is argon-moist argon-water rinsing.

    Keywords:

    cleaning, stainless steels, sodium, subassemblies


    Paper ID: STP41384S

    Committee/Subcommittee: A01.06

    DOI: 10.1520/STP41384S


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