[en] This paper presents the impact of different types of burnable poisons and filling methods on burnup in the VVER-1000 reactor. In this study, SuperMC was used to visually model the fuel assembly of the VVER-1000 reactor, perform burnup calculations and compare the results. Five groups (Homogeneous Gd2O3, Heterogeneous Gd2O3 1and 2, Homogeneous AmO2, Pure UO2) of different burnable poisons are modeled with fuel assembly as a unit, and the results will be analyzed by comparing the difference of keff. (paper)

[en] At present there is a variety of experimental and calculation nuclear data which are rather entirely presented in the following evaluated nuclear data libraries: ENDF (USA), JEFF (Europe), JENDL (Japan), TENDL (Russian Federation), ROSFOND (Russian Federation). Libraries of nuclear data, used for neutron-physics calculations in programs: Scale (Origen-Arp), MCNP, WIMS, MCU, and others. Nevertheless all existing nuclear data bases, including evaluated ones, contain practically no information about threshold neutron reactions on ²³²Th nuclei; available values of outputs and cross-sections significantly differ by orders. The work shows necessity of nuclear constants corrections which are used in the calculations of grids and thorium storage systems. The results of numerical experiments lattices and storage systems with thorium. (paper)

[en] The results of multiplying properties design research of the unit cell and LWR fuel assembly with the high temperature gas-cooled thorium reactor fuel pellet are presented in the work. The calculation evaluation showed the possibility of using thorium in LWR effectively. In this case the amount of fissile isotope is 2.45 times smaller in comparison with the standard loading of LWR. The research and numerical experiments were carried out using the verified accounting code of the program MCU5, modern libraries of evaluated nuclear data and multigroup approximations. (paper)

[en] Within the framework of international initiatives on the nuclear weapon non-proliferation in the Republic of Kazakhstan, conversion of research reactors to low-enriched fuel is implemented. This work aims at determining the spatial power distribution of the IVG.1M reactor with low enrichment fuel. The focused attention to this parameter is explained by the fact that energy release (or power density) determines the power, composition of technological systems, and efficiency of a nuclear reactor. In order to assess the change in the parameters of the reactor after conversion and to check the reliability of the simulation, the method of experimental study of the distribution of power density at the stage of physical start-up of the IVG.1M reactor is considered. The results of MCNP6 calculation of the power distribution in fuel elements, fuel assemblies and in technological channels are presented. (author)

[en] For the past ten years, the IVG.1M research reactor has been converted from HEU fuel to LEU fuel. Currently, the final stages of conversion (physical and energy start-ups of reactor) are being carried out. One of the tasks of physical start-up was to conduct physical studies of power distribution by radius and height of water-cooled technological channels. Studies were carried out using the activation gamma spectrometric method based on correspondence between energy release and measured gamma-radiation activity of fission-products or activation products. Study object was activation of indicators from copper wire and fuel element fragments placed in physical mock-up of technological channels with LEU fuel. ENREDI program, based on polynomial interpolation and approximation methods, was used to determine the detailed (in each fuel elements) relative power distribution by radial cross section of fuel assembly. During these experimental studies, relative power distribution over the height and radial cross section of technological channels of each rows, power peaking factors and value of absolute power were obtained. These studies will make it possible to evaluate changes in power profile in the IVG.1M reactor fuel after reducing of the enrichment, and also to check the accuracy of neutronic simulation. (author)

[en] The hybrid use of Nedis-2m and Serpent 2.1.30 codes to predict the radiation characteristics (i.e., neutron yield and energy spectrum) of an Am-Be source with a fine-grained mixture of americium dioxide (AmO₂) and beryllium (Be) core was studied with a focus on the grain size influence on the simulation results. The study showed that the fine-grained structure of the source core would decrease the number of alpha particles participating in the nuclear reactions with ¹⁷,¹⁸O and ⁹Be nuclei, which softened the neutron energy spectrum and reduced the neutron yield. The simulations also confirmed that the source core made of the stable crystals of AmBe₁₃ intermetallic alloy would improve the neutron yield to maximum 50% compared to the core made of AmO₂. Moreover, a source with a variable neutron yield was proposed with a heterogeneous core of AmO₂ rods embedded in Be. The neutron energy spectrum of heterogeneous source resembled the energy spectrum of Deuterium-Tritium (D T) neutrons which were generated in a long magnetic trap with high-temperature plasma. The subcritical irradiation facility assembled from the n^th number of heterogeneous Am-Be source can be used to study the properties of materials and the equipment operating in the epithermal and fast neutron spectra. The use of a heterogeneous ²⁴¹Am-Be assembly, as a basic element of an irradiation installation, simplifies the handling and operation procedures because it is easily disabled by removing the Be layer, or by inserting a sheet of the appropriate size and material between the Be and Am rod. (Author)