[en] An algorithm for calculation of prompt fission neutron lifetimes in a nuclear reactor by the Monte Carlo method is described. Evaluation of the importance function is carried out with solution of the neutron transport equation without solving the adjoint equation. The results of the prompt neutron lifetime calculations performed within some critical experiments are presented and compared with the experimental results.

[en] In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

[en] An algorithm for calculation of the neutron importance function and the delayed neutron effective fraction by the Monte Carlo method implemented in the KIR program is presented. The results of calculation of the delayed neutron effective fraction in some critical experiments are given in comparison with the experimental results.

[en] The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

[en] The DAREUS software package designed for modeling dynamic processes in the cores of experimental solution reactors is described. The KIR program based on the Monte Carlo method is used in the package to compute the necessary kinetic parameters. The results of the calculations of some test cases are given.

[en] The Argus research solution reactor and the main principles of experimental validation of its safety are described. The described validation is computationally confirmed in accordance with the requirements of the normative documents using the DAREUS software package designed for modeling dynamic processes in the cores of experimental solution reactors.

[en] The design of the ASTRA facility and critical assemblies that simulate physics features of modular high-temperature reactors (HTHR-Ms) with a graphite moderator and reflectors loaded with fuel particles having multilayer ceramic coatings is described in detail. Geometrical dimensions of the main elements and regions of the critical assemblies, composition of the materials used, and experimental results for various configurations of the critical assemblies are presented. A detailed computational benchmark model allowing for the structural and compositional features of the critical assembly configurations in question is developed on the basis of all the above data. The results are to be used for verification of the neutronics codes used for calculations of high-temperature helium-cooled reactors.

[en] The software developed within the framework of the MCU project which implements the Monte Carlo method for solving the transport equations for various kinds of radiation is widely used for calculating nuclear reactors of different types. The final version of MCU-5 is described. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the codes from the MCU family are presented.

[en] The methods and algorithms used in the PERST-5 package are described. This package is part of the MCU-5 code and is intended for neutron-physical calculation of the cells and parts of nuclear reactors using a generalized method of first collision probabilities.

[en] The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark-results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (AP0LL02, TRIP0LI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section. (authors)