A Nodal Scheme for Multigroup Neutron Diffusion Equations with Axially Varying Node Sizes
R. Uddin,* Q. Deng
University of Illinois

A nodal integral method is being developed and implemented to solve the multigroup neutron diffusion equations. Several new features-lacking in current implementations-are being explored. First, the nodal method will allow different node sizes in the axial direction to accurately accommodate physically distinct regions in the axial direction. Second, the iterative solution of the final set of equations will be carried out using the multigrid algorithm. In addition, improved approximations, based on actual local analytical solution for flux, are being tested for the leakage and source terms.

Stochastic Neutronics
R. A. Axford,* J. L. Hill
Los Alamos National Laboratory

The time evolution of the neutron population in a fissile assembly containing an insufficiently large enough number of neutrons to be able to define a probable number of neutrons per unit volume of phase space must be analyzed on the basis of stochastic formulations. New methods for obtaining the generating functions from which dynamics parameters and neutron fluctuations can be found have been developed. Methods for multidimensional assemblies are being studied.