Uncertainty evaluation of the Chexal-Lellouche correlation for void fraction in rod bundles

The prediction of the peak cladding temperature is key for the evaluation of the safety characteristics of nuclear reactor designs. One of the most important parameters in determining the peak cladding temperature is the two-phase mixture level in the core, which in turn depends on the void fraction in the core. In the two-fluid model, which is used in best-estimate thermal-hydraulic analysis codes such as RELAP and TRACE, the void fraction is largely determined by the interfacial drag. In the codes, the interfacial drag is determined based on drift-flux type correlations for the distribution parameter and void-weighted area-averaged drift velocity. It is essential that the model for the prediction of void fraction in the reactor core be accurate to ensure safe shutdown of the reactor. In light of this, a thorough review of the Chexal–Lellouche drift-flux type correlation currently used in RELAP5/MOD3.2.2 has been initiated to evaluate the scalability and uncertainty of the correlation. Analysis of the physical dependencies indicates some serious concerns regarding the presence of compensating errors in the models for the distribution parameter and void-weighted area-average drift velocity.