ENERGY-DEPENDENT NEUTRON-TRANSPORT PROBLEM WITH ARBITRARY GEOMETRY

The stationary neutron transport problem is considered which describes the transport of energy-dependent anisotropic scattered neutrons in a one, two, or three-dimensional inhomogeneous medium of arbitrary convex shape containing distributed source and incident neutrons. The approach to the problem is by successive approximation which leads to the existence of a unique solution as well as a recursion formula for the determination of the solution and error estimates for the approximations. In addition to the energy-dependent problem, a discussion is given to the constant-energy neutron transport and radiative transfer problems, including a slab problem. Some recursion formulas for calculating the solutions are obtained. All the formulas obtained can be used for computer calculations.