An Alternate Approach to Neutronics Analysis of CANDU Reactors

Abstract

An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Uptil now CANDU reactors have been analysed by the methods developed at AECL and CGE using semi-empirical techniques. Relying on multi-group transport codes GAM and GATHER in combination with diffusion code CITATION a package of codes is established with . the object of using it for survey as well as production purposes. A typical 600 MW(e) CANDU reactor has been analysed and the results so obtained are quite reasonable.