Improving the Accuracy of Neutron Flux Calculations in a CANDU Reactor by Combining the Diffusion Equation With in-Core Detector Readings

Abstract

The neutron flux in a CANDU nuclear reactor needs to be known both for refuelling decisions and for on-line flux mapping. Traditionally, different methods have been employed for the two purposes. While core-follow calculations have been performed off-line by solving the steady-state diffusion equation, flux mapping calculations have been performed on-line by fitting a linear combination of flux shapes (modes) to the detector readings from 102 vanadium detectors. Of the two, the first method is more accurate, while the second is faster. The present work introduces a new method, which combines the diffusion equation with detector readings to determine the neutron flux. Results for a one-energy-group one-dimensional implementation of the method are presented and shown to be more accurate than those of the diffusion-only method.