École Nationale Supérieure de
Techniques Avancées

Abstract

In this project, a numerical model is developed to assess the thermal-hydraulic system code RELAP5-3D correlations’ used for two-phase flow in MYRRHA secondary circuit conditions. The well-known drift-flux model is applicable to a wide range of thermal-hydraulic problems in nuclear reactors. The selected correlations for the model, based on the 4 equations drift-flux model, show better performance at lower pressures due to its approach to interfacial interaction and it is suitable for a wide range of thermal-hydraulic problems in nuclear reactors. All field and constitutive equations were properly selected and derived to get as acceptable with respect to the simplified approach. For the solution methodology, an iterative scheme was chosen with convergence criteria on exit pressure. The results show great compliance with experimental results and for some conditions are even better than RELAP5-3D results.