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This paper is devoted to the application of the method of numerical solution of the Boltzmann equation for the solution of the problem of modeling the behavior of radionuclides in the cavity of the interelectric gap of a multielement electrogenerating channel. The analysis of gas-kinetic processes of thermionic converters is important for proving the design of the power-generating channel. The paper reviews two constructive schemes of the channel: with one- and two-way withdrawal of gaseous fission products into a vacuum-cesium system. The analysis uses a two-dimensional transport equation of the second-order accuracy for the solution of the left-hand side and the projection method for solving the right-hand side — the collision integral. In the course of the work, a software package was implemented that makes it possible to calculate on the cluster architecture by using the algorithm of parallelizing the left-hand side of the equation; the paper contains the results of the analysis of the dependence of the calculation efficiency on the number of parallel nodes. The paper contains calculations of data on the distribution of pressures of gaseous fission products in the gap cavity, calculations use various sets of initial pressures and flows; the dependency of the radionuclide pressure in the collector region was determined as a function of cesium pressures at the ends of the gap. The tests in the loop channel of a nuclear reactor confirm the obtained results.

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 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.

Represented study considers numerical simulation of corium cooling driven by natural convection within a horizontal hemicylindrical cavity, boundaries of which are assumed isothermal. Corium is a melt of ceramic fuel of a nuclear reactor and oxides of construction materials.

Corium cooling is a process occurring during severe accident associated with core melt. According to invessel retention conception, the accident may be restrained and localized, if the corium is contained within the vessel, only if it is cooled externally. This conception has a clear advantage over the melt trap, it can be implemented at already operating nuclear power plants. Thereby proper numerical analysis of the corium cooling has become such a relevant area of studies.

In the research, we assume the corium is contained within a horizontal semitube. The corium initially has temperature of the walls. In spite of reactor shutdown, the corium still generates heat owing to radioactive decays, and the amount of heat released decreases with time accordingly to Way–Wigner formula. The system of equations in Boussinesq approximation including momentum equation, continuity equation and energy equation, describes the natural convection within the cavity. Convective flows are taken to be laminar and two-dimensional.

The boundary-value problem of mathematical physics is formulated using the non-dimensional nonprimitive variables «stream function – vorticity». The obtained differential equations are solved numerically using the finite difference method and locally one-dimensional Samarskii scheme for the equations of parabolic type.

As a result of the present research, we have obtained the time behavior of mean Nusselt number at top and bottom walls for Rayleigh number ranged from 10³ to 10⁶. These mentioned dependences have been analyzed for various dimensionless operation periods before the accident. Investigations have been performed using streamlines and isotherms as well as time dependences for convective flow and heat transfer rates.