Heat -- Transmission


The term Critical Heat Flux (CHF) is used in boiling heat transfer to describe the local value of the heat flux at which a characteristic reduction in heat transfer coefficient first occurs. In today’s technology, the CHF is a phenomenon related to the design and safety of various important devices.

A major limitation on the thermal design of a light-water reactor (LWR) is the necessity to maintain an adequate safety margin between the CHF and the local heat flux. Extended operations at local power levels in excess of the CHF can lead to high temperature oxidation and embrittlement or melting of the zircaloy cladding, thus jeopardizing the fuel rod’s integrity. In the nuclear industry, there have been many empirical CHF correlations developed over the years. These correlations are mostly based on steady-state (or quasi-steady) data obtained from various experiments covering different ranges of CHF parameters. Therefore, the application of such correlations is not only restricted by their parametric ranges, but is also limited to steady (or quasi-steady) operating conditions. In nuclear reactors, however, the CHF level is more likely to be reached during abnormal (transient) operating conditions, rather than during normal (steady) operations. Depending upon the type of accident, a wide range of thermal-hydraulic conditions may arise. For accurate nuclear reactor modeling, the accurate prediction of CHF as a function of time-dependent, thermal-hydraulic conditions is essential. This was the motivation of the subject study.

This research was a two-fold study. In the first part, the quasi-steady approach in predicting the CHF is defined and analyzed. In this part, data from blowdown experiments are compared to commonly used steady-state correlations on a local-instantaneous basis. This is done as a continuation of the previous studies conducted at Argonne National Laboratory. In all these studies, including the present study, a simple computer program Coolant Dynamic Analysis (CODA), is used. The results are discussed in terms of the limitations of the quasi-steady approach.

In the second part of this study, faster transients, where the quasi-steady approach is unable to predict the CHF, are considered. A new theory is developed to predict the CHF in power transients, which are typical of Reactivity Initiated Accidents (RIA) in LWRs. The proposed theory is purely analytical. It incorporates the effect of the hydrodynamic instability on the surface dryout in addition to evaporation, which was studied as the unique mechanism by previous researchers. The presented theory compares quite favorable with the available data from electrical heaters.

Two other types of transients which are of interest to the nuclear industry are rapid flow reduction. These are typical of Loss-of-Coolant Accidents (LOCA) and Loss-of-Flow Accidents (LOFA). The effects of these transients on CHF are also discussed from first principles. Finally, the important parameters of a generalized transient CHF correlation are identified and are grouped into dimensionless numbers. The physical significance of each group is discussed. Based on experimental and theoretical observations, a general mathematical model is developed to correlate transient CHF.


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Gunnerson, Fred S.


Doctor of Philosophy (Ph.D.)


College of Engineering


Mechanical Engineering and Aerospace Sciences




368 p.




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Doctoral Dissertation (Open Access)



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