抄録
In the framework of the investigation on the thermal-fluid dynamics phenomena in helical pipes of the innovative nuclear power reactor steam generator, at the Department of Nuclear Engineering at the University of Palermo, various research activities were performed relating to validations works of the models implemented in RELAP5/MOD3.2b thermal-hydraulic advanced code in order to simulate two phase flow phenomena taking place in these systems, though the one-dimensional nature of it. In this paper it is shown that the results obtained by the analyses of the experiments performed in different international laboratories and related to various helical pipes geometries, allows to prove the good performance of the so modified RELAP5 code.
