Neutron Detection-Based Void Monitoring in Boiling Water Reactors

The ratio between the thermal- and fast-neutron fluxes in a boiling water reactor depends on the void fraction. The density of the steam-water mixture present in the core determines the efficiency of the moderation of fast neutrons born in fission; therefore, the void fraction could be determined by means of a simultaneous measurement of the thermal- and fast-neutron fluxes. Such measurement could also be used to investigate channel bow of the nuclear fuel bundles surrounding the detector because of sensitivity of the thermal flux to geometry changes.

Calculations have been performed with both lattice and nodal codes to study the behavior of the void fraction correlation to the ratio of the thermal- and fast-neutron fluxes. The results prove the correlation to be nearly linear and robust. The rate of change of the correlation is insensitive to standard reactor operating parameters such as control rods and burnable absorbers; the sensitivity of the ratio to void fraction changes primarily depends on the geometry of the fuel bundles. A linear prediction model was used to represent the nodal code results. The absolute void fraction at over 792 positions in the core could be predicted with an absolute uncertainty of ±1.5%.