Determination of void fraction profile in a boiling water reactor channel using neutron noise analysis
Journal Article
·
· Nucl. Sci. Eng.; (United States)
OSTI ID:6722721
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.
- Research Organization:
- Oak Ridge National Lab., TN
- OSTI ID:
- 6722721
- Journal Information:
- Nucl. Sci. Eng.; (United States), Journal Name: Nucl. Sci. Eng.; (United States) Vol. 66:2; ISSN NSENA
- Country of Publication:
- United States
- Language:
- English
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