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Nowadays, active development of 4th generation nuclear reactors with liquid metal coolants takes place. Therefore, simulation of their elements and units in 3D modelling software are relevant. The thermal-hydraulic analysis of reactor units with liquid metal coolant is recognized as one of the most important directions of the complex of interconnected tasks on reactor unit parameters justification. The complexity of getting necessary information about operating conditions of reactor equipment with liquid-metal coolant on the base of experimental investigations requires the involvement of numerical simulation. The domestic CFD code FlowVision has been used as a research tool. FlowVision software has a certificate of the Scientific and Engineering Centre for Nuclear and Radiation Safety for the nuclear reactor safety simulations. Previously it has been proved that this simulation code had been successfully used for modelling processes in nuclear reactors with sodium coolant. Since at the moment the nuclear industry considers plants with lead-bismuth coolant as promising reactors, it is necessary to justify the FlowVision code suitability also for modeling the flow of such coolant, which is the goal of this work. The paper presents the results of lead-bismuth eutectic flow numerical simulation in the heat exchange tube bundle of NPP steam generator. The convergence studies on a grid and step have been carried out, turbulence model has been selected, hydraulic resistance coefficients of lattices have been determined and simulations with and without $k_\theta^{}$-$e_\theta^{}$ model are compared within the framework of fluid dynamics and heat exchange modeling in the heat-exchange tube bundle. According to the results of the study, it was found that the results of the calculation using the $k_\theta^{}$-$e_\theta^{}$ turbulence model are more precisely consistent with the correlations. A cross-verification with STAR-CCM+ software has been performed as an additional verification on the accuracy of the results, the results obtained are within the error limits of the correlations used for comparison.

The paper presents the results of three-dimensional numerical simulation of thermal-hydraulic processes in the Intermediate Heat Exchanger of the advanced Sodium-Cooled Fast-Neutron (BN) Reactor considering a developed methodological approach.

The Intermediate Heat Exchanger (IHX) is located in the reactor vessel and intended to transfer heat from the primary sodium circulating on the shell side to the secondary sodium circulating on the tube side. In case of an integral layout of the primary equipment in the BN reactor, upstream the IHX inlet windows there is a temperature stratification of the coolant due to incomplete mixing of different temperature flows at the core outlet. Inside the IHX, in the area of the input and output windows, a complex longitudinal and transverse flow of the coolant also takes place resulting in an uneven distribution of the coolant flow rate on the tube side and, as a consequence, in an uneven temperature distribution and heat transfer efficiency along the height and radius of the tube bundle.

In order to confirm the thermal-hydraulic parameters of the IHX of the advanced BN reactor applied in the design, a methodological approach for three-dimensional numerical simulation of the heat exchanger located in the reactor vessel was developed, taking into account the three-dimensional sodium flow pattern at the IHX inlet and inside the IHX, as well as justifying the recommendations for simplifying the geometry of the computational model of the IHX.

Numerical simulation of thermal-hydraulic processes in the IHX of the advanced BN reactor was carried out using the FlowVision software package with the standard $k-\varepsilon$ turbulence model and the LMS turbulent heat transfer model.

To increase the representativeness of numerical simulation of the IHX tube bundle, verification calculations of singletube and multi-tube sodium-sodium heat exchangers were performed with the geometric characteristics corresponding to the IHX design.

To determine the input boundary conditions in the IHX model, an additional three-dimensional calculation was performed taking into account the uneven flow pattern in the upper mixing chamber of the reactor.

The IHX computational model was optimized by simplifying spacer belts and selecting a sector model.

As a result of numerical simulation of the IHX, the distributions of the primary sodium velocity and primary and secondary sodium temperature were obtained. Satisfactory agreement of the calculation results with the design data on integral parameters confirmed the adopted design thermal-hydraulic characteristics of the IHX of the advanced BN reactor.

The development of a high-temperature gas-cooled reactor (HTGR) constituting a part of nuclear power-and-process station and intended for large-scale hydrogen production is now in progress in the Russian Federation. One of the key objectives in development of the high-temperature gas-cooled reactor is the computational justification of the accepted design.

The article gives the procedure for the computational analysis of thermal and physical characteristics of the high-temperature gas-cooled reactor. The procedure is based on the use of the state-of-the-art codes for personal computer (PC).

The objective of thermal and physical analysis of the reactor as a whole and of the core in particular was achieved in three stages. The idea of the first stage is to justify the neutron physical characteristics of the block-type core during burn-up with the use of the MCU-HTR code based on the Monte Carlo method. The second and the third stages are intended to study the coolant flow and the temperature condition of the reactor and the core in 3D with the required degree of detailing using the FlowVision and the ANSYS codes.

For the purpose of carrying out the analytical studies the computational models of the reactor flow path and the fuel assembly column were developed.

As per the results of the computational modeling the design of the support columns and the neutron physical characteristics of the fuel assembly were optimized. This results in the reduction of the total hydraulic resistance of the reactor and decrease of the maximum temperature of the fuel elements.

The dependency of the maximum fuel temperature on the value of the power peaking factors determined by the arrangement of the absorber rods and of the compacts of burnable absorber in the fuel assembly is demonstrated.

An approach to numerical analysis of thermo-hydraulic processes proceeding in a fast-neutron reactor is described in the given article. The description covers physical models, numerical schemes and geometry simplifications accepted in the computational model. Steady-state and dynamic regimes of reactor operation are considered. The steady-state regimes simulate the reactor operation at nominal power. The dynamic regimes simulate the shutdown reactor cooling by means of the heat-removal system.

Simulation of thermo-hydraulic processes is carried out in the FlowVision CFD software. A mathematical model describing the coolant flow in the first loop of the fast-neutron reactor was developed on the basis of the available geometrical model. The flow of the working fluid in the reactor simulator is calculated under the assumption that the fluid density does not depend on pressure, with use a $k–\varepsilon$ turbulence model, with use of a model of dispersed medium, and with account of conjugate heat exchange. The model of dispersed medium implemented in the FlowVision software allowed taking into account heat exchange between the heat-exchanger lops. Due to geometric complexity of the core region, the zones occupied by the two heat exchangers were modeled by hydraulic resistances and heat sources.

Numerical simulation of the coolant flow in the FlowVision software enabled obtaining the distributions of temperature, velocity and pressure in the entire computational domain. Using the model of dispersed medium allowed calculation of the temperature distributions in the second loops of the heat exchangers. Besides that, the variation of the coolant temperature along the two thermal probes is determined. The probes were located in the cool and hot chambers of the fast-neutron reactor simulator. Comparative analysis of the numerical and experimental data has shown that the developed mathematical model is correct and, therefore, it can be used for simulation of thermo-hydraulic processes proceeding in fast-neutron reactors with sodium coolant.