Determination of Void Fraction Profile in a Boiling Water Reactor Channel Using Neutron Noise Analysis

Fluctuations in the neutron flux caused by steam bubbles were analyzed to infer the average void fraction in the four fuel bundles that surround an in-core detector string in a boiling water reactor. The velocity of steam bubbles was inferred from the phase lag between axially displaced in-core fission detectors. This velocity, together with the measured power distribution and mass flow rate, was used to obtain the void fraction as a function of axial position. The results are in agreement with the predictions based on the Zuber et al. model, except near the top of the fuel channel.