Nuclear, Plasma and Radiological Engineering | 1999 Summary of Engineering Research
REACTOR PHYSICS AND REACTOR KINETICS
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.
back
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. Two new features-lacking in current implementations-will be added. 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.
back
Nuclear, Plasma and Radiological Engineering | 1999 Summary of Engineering Research