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The paper considers the influence of the spatial grid type on the result of solving the equation of neutron transport in the nuclear power plant (NPP) shielding. Neutron fields were calculated in a realistic model of a liquid metal cooled fast neutron tank reactor with an integral equipment layout. Structured cubic and unstructured hexahedral grids (pmsnsys and FRIGATE codes) and unstructured tetrahedral and prismatic grids (RADUGA T code) are used. Limiting values of the group fluxes averaged over the material zones for refined grids have been obtained. It has been shown that the calculation results depend on the type of approximation for the curvilinear inner boundaries between the material zones rather than on the grid cell type (cube, hexahedron, tetrahedron, prism). Using “toothed” approximations for curvilinear boundaries leads to an increase in the area of the boundaries, as well as to the neutron flux refraction condition arising on them. These effects lead to an upward bias in the transport equation solution, and for all energy groups. Conclusion. When solving an equation of neutron transport in the NPP shielding by a grid technique, it is necessary to use grids other than leading to “toothed” approximations of the inner boundaries. Tetrahedral or prismatic grids, or grids of arbitrary hexahedrons can be recommended, as well as composite grids in which cubic cells are located inside the material zone, and hexahedron cells are located near the zone boundary.

Purpose. The paper considers the influence of the spatial grid type on the result of solving the equation of neutron transport in the nuclear power plant (NPP) shielding.Method. Neutron fields have been calculated in a realistic model of a liquid metal cooled fast neutron tank reactor with an integral equipment layout. Structured cubic and unstructured hexahedral grids (PMSNSYS and FRIGATE codes) and unstructured tetrahedral and prismatic grids (RADUGA T code) are used. Limiting values of the group fluxes averaged

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.