DOI: 10.55176/2414-1038-2021-4-131-146

Authors & Affiliations

Zborovskii V.G.1,2, Khoruzhii O.V.1,2, Likhanskii V.V.1,2, Elkin N.N.1, Chernetskii M.G.1
1 National Research Center “Kurchatov Institute”, Moscow, Russia
2 Lebedev Physical Institute of the Russian Academy of Science, Moscow, Russia

Zborovskii V.G. – Leading Researcher, Cand. Sci. (Phys.-Math.), Lebedev Physical Institute of the Russian Academy of Science, Senior Researcher, National Research Center “Kurchatov Institute”. Contacts: 1, Akademika Kurchatova pl., Moscow, 123182. Tel.: +7 (926) 569-38-48; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Khoruzhii O.V. – Leading Researcher, Dr. Sci. (Phys.-Math.).
Likhanskii V.V. – Head of the Department, Dr. Sci. (Phys.-Math.).
Elkin N.N. – Leading Researcher, Dr. Sci. (Phys.-Math.).
Chernetskii M.G. — Engineer.


The paper describes the software module FRC-SCP intended for thermohydraulic simulation of a coolant under supercritical pressure (SCP) and a cooled fuel rod. Several designs of supercritical water reactors utilize the transition of the coolant from the pseudoliquid state to pseudogas while it is heated in the reactor core. SCP coolant under pseudophase transition exhibits specific behavior as its density changes significantly. Furthermore, coolant thermophysical properties (density, heat capacity, viscosity, thermal conductivity) can also vary across the coolant channel affecting heat transfer from the fuel rod to the coolant and consequently the fuel temperature. Existing feedback dependencies on coolant density and fuel temperature are important for the nuclear safety analysis of the reactor. The paper considers the present version of the FRC-SCP module. It implements the steady-state thermohydraulic channel solver to calculate coolant parameters: temperatures of the flow core and a heater, as well as coolant pressures, densities etc. User-specified correlations define the heat transfer law under the normal conditions. The module solves the thermal problem for the fuel rod consistently with the channel thermohydraulic problem. It is also possible to couple the FRC-SCP module with the neutron physical codes. Thermohydraulic module is tested against experiments on the heat transfer to SCP water in heated tubes. We discuss the behavior of the fuel cladding under conditions imitating the deteriorated heat transfer modes and the effect of the nuclear fuel thermal conductivity.

coolant, fuel rod, nuclear fuel, cladding, supercritical pressure, pseudophase transition, thermohydraulic simulation, heat transfer coefficient, temperature field, experiments with heated tubes, deteriorated heat transfer modes

Article Text (PDF, in Russian)


UDC 621.039.546:536.24

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2021, issue 4, 4:12