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The paper presents results of experimental research, the purpose of which was to simulate the phenomenon of mixing of loop coolant flows inside the model of the pressure chamber of the water-cooled reactor. The study was conducted at the high-pressure aerodynamic test bench of NNSTU n.a. R.E. Alekseev. The scale model consisted of elements characteristic of water-cooled nuclear reactors (lowering annular channel, lower pressure chamber). Experimental studies were carried out in the range of Reynolds numbers from 20,000 to 50,000. The axial velocity field was studied at the entrance to the reactor core simulator, using a pneumometric research probe. The temperature field was simulated using the impurity diffusion method by introducing a contrast tracer into one of the loops of the model. Propane was used as a passive tracer. Tracer concentration at the inlet core simulator allowed to evaluate the degree of flow mixing and thereby studied the direction of coolant flow. As a result of the studies, the spatial distribution of the tracer concentration in the coolant flow in the lower annular channel, as well as in the lower pressure chamber, was determined. It was shown that in the lower pressure chamber, an axial central vortex influences the mixing intensity. The estimation of the parameters of mixing of the passive tracer in the model of reactor pressure was given.

A modernized experimental setup diagram is presented for studying the cooling of a high-temperature surface with a dispersed coolant flow. The results of an experimental study of unsteady heat removal from a target with a temperature exceeding the Leidenfrost temperature are presented. The cooling curves are obtained, the regions of effective heat transfer are determined. The primary experimental data were obtained during stationary heat transfer in the following range of operating parameters of the coolant: mass water flow rate 0.0038 ÷ 0.0145 kg/s; water pressure 0.4 ÷ 0.2 MPa, induction heating power 7 ÷ 20 kW.

A conversion of the TRIGA reactor Bandung has been planned by replacing fuel from rod type to plate type. The fuel replacement causes the fuel arrangement or reactor core configuration should be changed as well as the cooling system. Currently, the reactor cooling system takes place by natural convection with the flow direction from the bottom to the top of the core. The replacement of plate-type fuel with a small distance between plates causes the core cooling process to decrease, therefore to make more effective the forced convection cooling system with the direction of upward flow is required. It is expected that the cooling process will increase. The ability of the cooling system depends on flow distribution in the core at the time of reactor in operating, therefore it is necessary to analyze the distribution of coolant flow and the velocity distribution of the cooling system from the top reactor pool to the reactor core to determine that the coolant flow from the inlet pipe mostly goes to the core. The purpose of this research is to analyze the distribution of coolant flow and flow velocity to the reactor core. The analysis was carried out in a simulation using computational fluid dynamics FLUENT software by creating reactor core modeling. The result shows that the cooling water flows from above along the reactor pool, then the stored energy is reduced so that the flowing slowdown. In this case, resulting in evenly flow distribution to the reactor core.

This paper provides data on the performed analysis of hydrodynamics characteristics and calculation of crossflow distribution in a PWR mixed core. The analysis was realized using the subchannel thermal-hydraulic code VIPRE-01 and the CFD code STAR-CCM+. The paper sets out recommendations on choosing the computational model detailing and adjustment for the subchannel code. The analysis has shown that the mixed core model in VIPRE-01 takes into account the coolant flow redistribution correctly and provides the possibility of using this code for DNBR analysis in the mixed cores.

The present work is dedicated to verification of numerical model in standard solver of open-source CFD code OpenFOAM for two-phase flow simulation and to determination of so-called "baseline" model parameters. Investigation of heterogeneous coolant flow parameters, which leads to abnormal friction increase of channel in two-phase adiabatic "water-gas" flows with low void fractions, presented.

The results of experimental studies of pebble bed hydrodynamics are presented. For the first time experimental data on the pressure loss in a radial flow of fluid through the pebble bed was obtained. Experiments were carried out in the liquid flow rate ranging from 0.09 to 0.4 kg / s, fluid temperature is 20°C.