The capability of SN methodology to accurately predict the neutronics behavior of a compact, light water‐moderated reactor is examined. This includes examining the effects of cross‐section modeling and the choice of spatial and angular representation. The isothermal temperature coefficient in the range of 293 K to 355 K is analyzed, as well as the radial fission density profile across the central fuel element. Measured data from a series of critical experiments are used for these analyses.


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    Title :

    SN transport analyses of critical reactor experiments for the SNTP program


    Contributors:

    Conference:

    10th SymposiumùSpace Nuclear Power and Propulsion ; 1993 ; Albuquerque, NM, USA


    Published in:

    AIP Conference Proceedings ; 271 , 2 ; 811-815


    Publication date :

    1993-01-15


    Size :

    5 pages





    Type of media :

    Conference paper


    Type of material :

    Electronic Resource


    Language :

    English




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