Helical coil thermal-hydraulic model for supercritical lead cooled fast reactor steam generators

This paper deals with the presentation of the developments of an in-house code created for the steady state thermal-hydraulic analysis of helical coil steam generators. These components are of particular interest in the nuclear field both for normal and emergency operation because they are able to remove thermal power with a very high degree of compactness. This quality is particularly attractive for small modular reactors where constraints related to the volume of components are a critical issue. The model is able to perform calculations to validate the design of these components, run fast sensitivity analyses and show the spatial behaviour of the most important parameters of the flow such as temperature, pressure, heat transfer coefficients and heat flux. In this paper the additional features of the model such as the possibility to simulate supercritical water and liquid metals are presented. The choices made for the selection of semi-empirical correlations available in literature for the estimation of the heat transfer coefficients in these conditions are described and justified. Finally, the predictive capability of the updated model is validated against the nominal data of the BREST reactor, which is a lead fast reactor that make use of supercritical water.