CHAPTER II

       APPLICATION OF DIFFUSION THEORY TO NEUTRONS
             IN A FISSIONING URANIUM PLASMA


       A General Boltzmann Equation for Neutrons
            in a Fissioning Uranium Plasma


        A fundamental macroscopic description of the

neutron population in a solid fuel reactor is given by the Boltzmann neutron transport equation. This integrodifferential equation specifies the neutron distribution in space, velocity, time and direction of motion. Using notation similar to that of Weinberg and Wigner (18), the transport equation is


1 a  4 .(,                ,  t
    (r,v   t) +   •             + Et(r'v) (r'v''t)



    x(v) f  dv, f  df ' cv'Zfc ,v') ( ,' , ' T 1



  + f dv' f d     sE rv'-v, '   w'r'v"'t)           (2.1)



External neutron sources and delayed neutrons are not included in (2.1).

        The symbols are defined in the usual manner:


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