Prediction of Void Fraction in PWR Subchannel by CATHARE2 Code

The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, in drawing attention to their weak points, in identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. 3 field codes, 2 phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are also used for the OECD/NRC PSBT benchmark. The paper presents the validation activity performed by CATHARE2 v2.5_1 (six equation, two field) code on the basis of the sub-channel experiments available in the database and performed in different test sections. Four sub-channel test sections are addressed in different thermal-hydraulic conditions (i.e. pressure, coolant temperature, mass flow and power). Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests.