One of the limiting conditions during operation of a Pressurized Water Reactor is cladding integrity in case of occurrence of any conditions I or II events. The decoupling criterion is the absence of Departure from Nucleate Boiling (DNB) during the full sequence of any of these transients. Heat transfer between the clad and the water is limited by the DNB phenomenon when local surface heat flux is greater than the so-called Critical Heat Flux (CHF). Heat production at the surface is higher than heat removal capacity by the coolant therefore a vapor blanket is formed around the clad; consequently the heat transfer will drastically drop resulting in a sudden significant increase of the local wall temperature and clad damage may appear if no corrective action is initiated. DNB can not be estimated with physical principles only. Experimental support is needed for evaluation. Occurrence of DNB is evaluated using the Departure from Nucleate Boiling Ratio (DNBR) which is a function of both core thermal hydraulic (T/H) parameters and design of the fuel assembly. Advanced fuel assemblies claim higher CHF values compared to previous designs. Along with increased DNB performances for advanced fuel assemblies, CHF correlation development and advanced methodologies enable to extend normal operating conditions of a nuclear plant. On the one hand, CHF performances really increased allow additional margin related to the loss of fuel cladding integrity whereas on the other hand optimized correlations and advanced methodologies reduce this margin. An accurate assessment of the CHF performance of the advanced fuel assemblies is therefore required. This paper will raise issues regarding the assessment of the CHF performance of new advanced fuel assemblies design. The issues will be focused on the reliability of the experimental assessment of the CHF values and the accuracy of the transposition of mock up geometries to plant core configuration (representativity of the experiments). The verification that the tests conditions (pressure, flowrate, quality, heat flux …) ensure a proper coverage of all core conditions encountered during any of the conditions I & II transients is closely linked to DNBR methods and will not be extensively covered in this paper. This paper suggests some thoughts about relevance of the demonstration carried out by vendors on these matters.

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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.

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