Artavazd M. Sujyan, Aleksey M. Osipov

National Research Center «Kurchatov Institute»

Computational studies of thermal-hydraulic processes in the core of nuclear reactors with supercritical water coolant are of high importance, especially at the stage of concept development of such reactors. At the same time, the complexity of determining the water properties at supercritical parameters leads to the need for further development of thermal-hydraulic models and forecasting tools, for which one has to turn to modern computer science and information systems. Such studies are necessary in justifying the safety of nuclear reactor concepts with supercritical coolant parameters. As a part of the safety justification for pressurized water reactors, the RELAP5 program code has been widely used. When performing calculations, the correlations used in the program code to describe the heat transfer process between the wall and the liquid are extremely important. In order to assess the correctness of the data on heat transfer in pipes and bundles with supercritical water used in RELAP5 code, a reproduction of an experiment on heat transfer in the vertical pipe was performed. The series of calculations for the wall and the coolant temperatures at the pressures of 23 and 25 MPa and various heat fluxes in the range from 600 kW/m2 to 1100 kW/m2 have been performed. The results of the experiment are compared with the results of its computational reproduction. The calculated and experimental results are compared with the theoretical characterization of heat transfer mechanism in supercritical pressure water.

nuclear reactor, computer science, information systems, supercritical water coolant, heat transfer, RELAP5