Journal
ANNALS OF NUCLEAR ENERGY
Volume 174, Issue -, Pages -Publisher
PERGAMON-ELSEVIER SCIENCE LTD
DOI: 10.1016/j.anucene.2022.109174
Keywords
Departure from nucleate boiling; Critical heat flux; Wall boiling model; Subcooled flow boiling; Critical heat flux modeling
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Funding
- National Natural Science Foundation of China [51876128, 11705035]
- Natural Science Foundation of Heilongjiang Province [YQ2020A003, LH2019A009]
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This paper investigates the capability of the open-source CFD code OpenFOAM to simulate the critical heat flux (CHF) in vertical pipe under extreme conditions, close to PWR operating conditions. The two-fluid approach combined with an extended wall boiling model was used to forecast CHF. The findings indicate that OpenFOAM promises to forecast boiling flow and CHF in the PWR reactor fuel assembly geometry.
This paper aims to see how well the open-source CFD code OpenFOAM can simulate the critical heat flux (CHF) in vertical pipe under extreme conditions, close to PWR operating conditions. The two-fluid approach combined with an extended wall boiling model was used to forecast CHF. The 2006 CHF look-up table was employed as benchmark data. The wall temperature jump was used as the criterion for predicting departure from nucleate boiling (DNB) type CHF. Good agreement was obtained between CHF from look-up table and calculated values. The effects of pressures, mass flow rates, and equilibrium qualities on CHF are investigated. CHF location and maximum volume fraction at corresponding CHF were also obtained. Most of the deviations are within 15% of error. The average absolute error of the simulated CHF was 7%. The findings indicate that OpenFOAM promises to forecast boiling flow and CHF in the PWR reactor fuel assembly geometry.(c) 2022 Published by Elsevier Ltd.
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