A Noniterative Method of Computing Neutron Flux Distributions and Coolant Density Profile in a Light-Water Moderated Reactor

A one-dimensional noniterative method for calculating the fast- and thermal-neutron flux distribution, effective neutron multiplication factor, power density, enthalpy profile, water density distribution, and steam void map of a light-water moderated reactor core is presented and programmed as a computer code — ANDREA. In this method, the spatial dependence of the neutron spectrum is accounted for explicitly. The method outlined in this paper can be used as one of the design tools for pressurized water reactor (PWR) cores as well as for boiling water reactors (BWR). The novelty of this method lies in its noniterative mathematical formulation which takes into account the nuclear-thermal interaction in a reactor channel. Fission density is directly related to heat generation and heat generation causes density changes in the coolant with subsequent formation of steam voids. The method described here is based on the fact that the above relationships are interdependent. As a result of this noniterative formulation, a significant amount of computer time is saved. Finally, it is to be noted that the method presented in this paper is primarily intended for the analysis of large power reactors.