Determination of Void Fraction Profile in a Boiling Water Reactor Channel Using Neutron Noise Analysis
- 12 May 1978
- journal article
- research article
- Published by Taylor & Francis in Nuclear Science and Engineering
- Vol. 66 (2) , 264-268
- https://doi.org/10.13182/NSE78-A27209
Abstract
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.Keywords
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