Two-Group Macroscopic Cross-Section Data Base for the Prototype Miniature Neutron Source Reactor

Abstract

A two-group macroscopic cross-section data base has been generated using a simplified model for the Miniature Neutron Source Reactor (MNSR). The correction terms to the basic macroscopic cross-sections due to varying effects of temperature along the axis of the fuel and burnup were expressed in the form of polynomials. The error analysis shows that the data base is accurate enough to be used for neutronic calculations for the reactor.