This study presents the results of testing nuclear data libraries by analyzing statistical criteria obtained from comparing experimental and calculated rates for (n,2n), (n,p), (n,pn), (n,n'gamma) (n,alpha) and (n,gamma) reactions measured on samples Ni-nat, Zr-nat, Nb-nat, Cd-nat, Ti-nat, Co-nat,Cu-63(96%),Cu- 65(99.70%), Zn-64(99.70%), In-nat, Al-nat, Mg-nat, Fe-nat, Au-nat and Th-nat, which were placed in the experimental channels of micromodels of the fusion blanket. The "fast" (the cylinder & Oslash; 230 mm and 520 mm length was filled with similar to 67 kg of molten salt 0.52NaF + 0.48ZrF4) and the "thermal" blanket (the same cylinder was placed in a dry channel inside a cubic container filled with water with dimensions of 52.0 x 52.0 x 52.0 cm were investigated. The reaction rates were measured using the activation method. Modeling with transport codes MCNP5, KIR, PHITS-3.31, SuperMC3.4.0 was performed using the ENDF/B-VII.0 library for neutron transport as well as seven neutron data libraries for reaction rates simulation, including: JEFF-3.3, JENDL-4.0, ENDF/B-VIII.0, ROSFOND-2010, FENDL-3.0, TENDL - 2019 and IRDFF-II.

The article considers the possibilities of a fusion neutron source with a molten-salt blanket to provide fuel for thermal fission reactors with uranium-thorium nuclear fuel cycle, and it also formulates recommendations for closing fuel nuclide balances in a nuclear power system with fission and fusion reactors.

The KIR2 software complex for simulating neutronic stationary and nonstationary processes in reactors based on the Monte Carlo method [1, 2] is discussed. The paper presents a brief description of the software complex being developed, its key capabilities, and the features of the architecture.

The paper considers the KIR-TG software package, which is designed for the interconnected neutron-physical and thermohydraulic test calculations of the infinite grid of VVER-SKD fuel elements. As part of the work, a software program written in the Python and Lua dynamic programming languages for linking between the module for calculating neutron physics (KIR computer program) and the TK-SKD thermohydraulic module was created and applied. The use of precision calculation performed by the Monte Carlo method in the neutronphysical part avoids the error of deterministic approximation methods in the calculation results. The problem statement, as well as the nature and results of the work, revealed the advantages of this approach to solving problems related to thermal-hydraulic calculations: parallel neutronic and sequential thermal-hydraulic calculations do not affect the overall speed of the associated calculation, the calculation codes are developed in parallel to save time and resources, and errors are tracked separately.

The article discusses the problem of fuel supply for promising large-scale nuclear power engineering. Because of the limited resource of natural uranium, the fuel for the future should be artificial fissionable isotopes. This task involves closing the fuel cycle and processing the entire volume of spent nuclear fuel (SNF) in order to extract new fissile isotopes from it. However, as shown in the article, the irretrievable losses inherent in the SNF processing, when accumulated, create a radiation load on the environment, which is hundreds of times higher than the existing level. This makes it difficult to extract fissionable isotopes from SNF of a fission reactor and motivates the use of other neutron sources to generate the required amount of fissionable isotopes while maintaining a low level of radiation load. Such a source of neutrons can be hybrid fusion reactors with 233 U accumulated in the blanket from thorium raw materials.

This paper presents the results of the experimental determination and computational simulation of the ambient dose equivalent rate for a metallic thorium cylindrical miniblock and the (n,2n), (n,f), and (n,gamma) reaction rates in a thin Th-232 metal foil irradiated with neutrons of the NG-24M generator spectrum. The ambient dose equivalent rate was determined by dosimeters-radiometers. The reaction rates were determined by the activation method using Ge spectrometers without destroying the irradiated samples. Computational simulations of ambient dose equivalent and reaction rates were performed, respectively, using the radiation transport codes PHITS, MCNP5, and KIR2, which use various nuclear data libraries: JEFF-3.2 and -3.3; JENDL4.0; ENDF/B-VII.0, -VII.1, and -VIII.0; ROSFOND; FENDL; and TENDL. The authors give an estimate of the U-232/U-233 relative accumulation upon natural thorium irradiation in a fusion facility blanket with defined neutron spectrum. The nonirradiated and irradiated thorium nuclide composition change simulation and visualization were performed using analytical solutions of an ordinary system of homogeneous linear differential equations describing nuclide transmutations.

The article discusses the issue of tritium production in VVER reactors in order to use it as fuel material in fusion reactors. The production of one tritium nucleus from the raw isotope 6 Li requires the expenditure of one neutron. This is equivalent to neutron losses in the fuel cycle of a fission reactor. The absorption of neutrons by 6 Li is partially compensated by the fuel burnup reactivity reserve in the fission reactor. However, owing to the insufficiently rapid burnup of 6 Li, it acts as an absorber throughout the entire campaign. This leads to its reduction.

The paper presents the results of numerical studies of nonstationary processes without feedback in a research reactor. The obtained results show a large degree of uncertainty of the calculated functionals depending on the parameters of the time, spatial, and energy grids. A brief description of the parallel version of the CTART4 code is given.

This paper presents the results of an experiment determining (n,2n), (n,p), (n,pn), (n,alpha), (n,n'gamma), and (n,gamma) reaction rates in 15 test samples of both natural and high-enriched composition: Mg-nat, Al-27, Ti-nat, Fe-nat, Co-59, Ni-nat, Cu-63 (99.5%), Cu-65 (99.7%), Zn-64 (99.4%), Zr-nat, Nb-93, Cd-nat, In-nat, Tm-169, and Au-197. Computer simulations in the NG-24M neutron generator spectrum were carried out using the MCNP5 and KIR2 radiation transport codes with different nuclear data libraries (JEFF-3.2, JEFF-3.3, JENDL-4.0, ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, ROSFOND-2010, FENDL-3.0, TENDL-2019, and IRDFF-II). The elaborated full-scale model for neutron transport analysis included the geometry and composition of the neutron generator, experimental samples, and laboratory room. The mean square deviation factor was used to compare the experimental and the simulated results. The best predictive results for both the MCNP5 code and the KIR2 code were obtained with the FENDL-3.0 and ENDF/B-VIII.0 libraries.

Recent assessments of the strategic prospects of the nuclear energy industry tend to show a somewhat condescending attitude towards fusion, which, regretfully, has grounds given the current state of affairs in nuclear science. However, the analysis of the problems and the potentialities of the fusion of light nuclei and fission of heavy nuclei suggests that a full-scale development of each of the two nuclear technologies inevitably involves the need to overcome some formidable technological, materials science, environmental, and economic problems, some of which may put the feasibility of pursuing these nuclear energy programs at question. At the same time, the physical features of the fission and fusion processes objectively suggest that it may be reasonable to combine the two nuclear technologies within a joint nuclear energy system to enable a greater synergetic effect. This apparently would markedly lessen the problems attendant upon the application of each of the discussed technologies in the course of the full-scale development of nuclear energy.

The work is devoted to the problem of initial neutron source distribution for modeling kinetic transient processes by the direct analogue Monte Carlo method. An algorithm is proposed for modeling such a source. The source algorithm implemented in the KIR2 program and designed to solve the problem of taking the system out of the critical state by introducing a perturbation (reactivity). Only prompt neutrons are considered. For the neutron source distribution, a phase coordinates of the neutron are used. The phase coordinates are determined in the preliminary calculation of the critical state. The results of testing the algorithm on non-stationary test problems (an infinite environment and RP1GS) are presented. In each task, the geometric area is filled with material presented in the form of single-group macroparameters. The results of calculating the integral neutron flux density were compared with the solution of the point kinetics equations. Good agreement was obtained - the deviation during all the calculated processes does not exceed 0.4 % in the integrated flux density, but it increases at the end of the processes with the introduction of negative reactivity up to 2.5 % with a small number of neutrons in the system.

The article presents the results of testing the KIR code designed to solve the neutron transfer equation by the Monte Carlo method, based on available experimental data from the international Bank ICSBEP on solution benchmarks. The KIR program is part of the DAREUS program complex designed for modeling dynamic processes in solution reactors. Computer models of 13 critical assemblies were created. The total number of calculated configurations was 137. The results were analyzed and the calculation errors were determined both in comparison with the experiment and with other programs. In most cases, the calculated keff multiplication factor falls within the experimental error. The obtained calculation results show a mean square deviation of the multiplication coefficient of 0.7 % for all calculated configurations of experimental assemblies. However, for a number of assemblies, there are significant differences in the assessment. Similar deviations in the calculated values of keff were observed in a variety of previously performed calculations for other codes. The article presents the main factors affecting the results of calculations. Mainly they are related to the insufficiently complete description of experiments submitted to the ICSBEP International Bank.

Analysis of the problems and potential of two nuclear technologies based on fusion reactions of light nuclei and fission of heavy nuclei shows the following. Independent large-scale development of each one will make it necessary to overcome still unsolved technological, materials science, environmental, and economic problems, raising the question of the expediency of further development of these energy industries. At the same time, the physical features of the fission and fusion processes objectively indicate the expediency of combining them into a unified nuclear energy system, which will have a large synergistic effect. This will greatly mitigate the adverse aspects of each technology that will appear if the two systems develop independently. Calculations of neutron multiplication in the blanket of a hybrid thermonuclear reactor that confirm the physical feasibility and reliability of picking the direction of development in the form of a unified nuclear energy system are presented.

The article discusses the issue of the concept of “new generation code”, which has been actively used recently to characterize computer programs designed to solve problems of the transfer of neutrons and gamma quanta in nuclear facilities. As an example of a new generation code developed for solving the multigroup transport equation by the grid (deterministic) method, the first version of the new software package RADUGA-TV is considered, including, in particular, the UNK complex for calculating burnup. The article lists the main features of the RADUGA-TV code: the problems to be solved, the types of constants used, the methods for specifying the geometry of the calculation area, the methods for constructing an unstructured spatial mesh. The possibilities of the postprocessor for processing the obtained solution are presented. The article presents progressive algorithms included in the RADUGA-TV code, including grid schemes and methods for parallelizing computations. The advantages of using unstructured grids, including those consisting of cells of various types, are discussed. Methods for parallelizing computations on hybrid computing systems are considered. The question of the spatial grid decomposition when parallelizing computations on distributed memory systems is considered, as well as the question of organizing parallel computation on such systems. Comparison of the characteristics and capabilities of the RADUGA-TV code and other similar in purpose codes, foreign (ATTILA, AETIUS, ARES, THOR) and domestic ODETTA is performed. It is shown that the RADUGA-TV code is significantly advanced methodically and practically has no analogues. The article was written based on the materials of the report at the conference “Neutronika-19” and contains more detailed information on the issues discussed in the report.

For one dimensional benchmark BSS-6 calculation by SUHAMD-TD and SUHAM-TD codes (nonsta-tionary solution in diffusion and surface harmonic method models with using delayed neutrons), by KIR (time dependent Monte Carlo method) and LUCKY-A code (discrete ordinate method for time dependent transport equation solution with using of parallel calculations). Transport cross sections for codes KIR, LUCKY-A and SUHAM-TD were obtained from diffusion coefficients for neutron groups to make kinetic calculation. Right and left boundary conditions for plate were zero for diffusion model and vacuum (not angular fluxes into plate) for kinetic model.

In the latest assessments of the strategic prospects for the development of nuclear energy, we can note the tendency of a condescendingly arrogant attitude towards thermonuclear fusion, which, unfortunately, largely corresponds to the real state of affairs. At the same time, an analysis of the problems and potential of two nuclear technologies based on reactions of fusion of light nuclei and fission of heavy ones shows the following. Independent large-scale development of each of these areas will inevitably lead to the need to overcome the still unresolved technological, material science, environmental and economic problems, some of which raise the question of the appropriateness of further development of these energy sectors. At the same time, the physical features of fission processes and fusion objectively indicate the expediency of combining them within the framework of a single nuclear energy system. Such a combination will provide a great synergistic effect, as a result of which the negative manifestations of the complex problems of each technology will significantly weaken on the path of large-scale development of the industry.

Calculations for the fusion reactor were carried. The purpose this one is assessment of an opportunity to use blanket of these units for fission isotopes reproduction for fission reactors. This paper contains: - full scale 3D model, where 1/8 space part was used (symmetry), biological shielding and toroidal coil winding were included in model, model size about 5,8 m for every X,Y,Z geometry axis; - results by MCU and LUCKY-A codes of neutron fluxes, reaction rates; - reproduction coefficient for thorium-232 as relation to intensity of neutron source; - analysis of results. MCU and LUCKY-A computer codes use parallel calculation technologies to get solution transport equation.

В последних оценках стратегических перспектив развития ядерной энергетики можно отметить тенденцию снисходительно высокомерного отношения к термоядерному синтезу, которая, к сожалению, в значительной части корреспондируется с реальным состоянием дел. В то же время анализ проблем и потенциала двух ядерных технологий, основанных на реакциях синтеза лёгких ядер и деления тяжёлых, показывает, что независимый масштабный рост каждого из указанных направлений неизбежно приведёт к необходимости преодоления ещё не решённых проблем технологического, материаловедческого, экологического и экономического характера, некоторые из которых поставят вопрос о целесообразности дальнейшего развития этих отраслей энергетики. Вместе с тем физические особенности процессов деления и синтеза объективно указывают на целесообразность их объединения в рамках единой ядерной энергетической системы. Такое объединение обеспечит большой синергетический эффект, в результате которого существенно ослабнут негативные проявления сложных проблем каждой из технологий на пути крупномасштабного развития отрасли.

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.

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.