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First Attempt to Determine a Critical Heat Flux Correlation for Thermalhydraulic System Codes

[+] Author Affiliations
Christophe Herer, Antoine Lejosne

Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-Roses Cedex, France

Paper No. ICONE26-81589, pp. V06AT08A057; 7 pages
doi:10.1115/ICONE26-81589
From:
  • 2018 26th International Conference on Nuclear Engineering
  • Volume 6A: Thermal-Hydraulics and Safety Analyses
  • London, England, July 22–26, 2018
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5148-7
  • Copyright © 2018 by ASME

abstract

Boiling crisis or departure from nucleate boiling is a key phenomenon in heat transfer processes. It appears when a vapor blanket is created at the heated wall and impedes its cooling. Due to either excessive heat flux (Critical Heat Flux - CHF) or high local void (dryout), a significant temperature rise is observed and the clad might be damaged. Thermalhydraulic accident analysis includes therefore the determination of appearance of boiling crisis in order to compute the evolution of the clad temperature during the transient. The timing of departure from nucleate boiling (DNB) provides the corresponding value of residual power and therefore the clad temperature. Boiling crisis cannot be predicted from first principles. Numerous experiments form a huge database from which hundreds of correlations have been derived. As the primary use of boiling crisis correlation is associated with subchannel analysis codes, the developed correlations of interest are constructed with two-fluid mixture local thermalhydraulic variables. This approach imposes that, when used in two-fluid six equation codes, the liquid and vapor variables are combined to compute the two-phase mixture mass velocity and thermodynamic quality to get the CHF. Therefore, it is logical to consider the possibility to set up a correlation directly connecting CHF and two-fluid (liquid and vapor) thermalhydraulic variables (velocities and enthalpies along with pressure and void fraction). Two additional features were considered. The first one was to establish a correlation using data obtained in rod cluster configuration and not in tubes as the ones currently implemented in thermalhydraulic system codes. For this, the measurements carried out at Columbia University on different rod bundle configurations were considered, as being now released by EPRI in the public domain. The second one was to attach to the coefficients of the correlation their uncertainties to allow best estimate plus uncertainties calculations. First calculations showed promising results such as obtaining a correlation with similar qualities as the already existing ones, based on two-phase mixture parameters. However, a lot still remains to be done in order to obtain a reliable correlation to be implemented in thermalhydraulic system codes.

Copyright © 2018 by ASME

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