[en] The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

[en] The TDD-C/4 program (FORTRAN-CERN, BESM-6 computer) for three-dimensional calculation of the IRT-M reactor fuel assemblies in the two-group diffusion approximation is described. The efficient neutron multiplication factor, neutron flux density, power density, power density nonuniformity factors by cells, coefficients of heat flux density nonuniformity on the surface of fuel elements are calculated. The system of mesh equations is solved using the method of successive upper relaxation with optimum relaxation factor

[en] The KSENIYa (FORTRAN-CERN, BESM-6 computer and FORTRAN-4, ES-1055 computer) and KSENIYa-1 (FORTRAN-CERN, BESM-6 computer) programs for calculation of xenon oscillations in the WWER type reactor and determining their optimum quenching are described. The neutron-physical problem for axial power distribution is solved in one-group approximation. Quenching of axial xenon oscillations is realized by variation of control rod position by height. Different optimum modes of the control rod displacement (relay and continuous) are defined

[en] The TUBE-1 program (FORTRAN, BESM-6 computer) for computerized simulation of crack propagation kinetics in pipelines under the effect of hydrogen embrittlement is described. The pipe is supposed to be endless. The crack-type defect begins to develope from an external surface and propagates along the forming pipe

[en] The PRAKTINEP (AR) program (ALGOL-GDR, BESM-6 computer) for a reactor plane cell one-group kinetic calculation by the surface pseudosource method in the G_N-approximation is described. The thermal neutron utilization factor, mean by zone neutron fluxes, fluxes, currents and second angular moments of neutron distribution function at zone boundaries and the quantity of absorbed neutrons in each zone of a cell are calculated

[en] The ATIKA program (ALGOL, BESM-6 computer) for two-dimensional multigroup calculation of spatial and energy distribution of the neutron flux density in a nuclear reactor biological shield is described. Diffusion-extraction method is a physical basis of the program. The diffusion component describes the neutron transport in the SP₁-approximation of the Boltzmann equation expansion by spherical harmonics. A system of algebraic equations is formed using finite difference representation and integration-interpolation method with the five-point approximation of the Laplacian operator. The solution is realized with the help of iterative method of varied directions. The program has a segment-type modular structure

[en] Algorithm and possibilities of the GBRCD geometrical module for simulation of a particle hystory in a system with complex configuration when calculating the reactor lattice by the Monte Carlo method are described. The module permits to carry out calculations for a medium consisting of separate regions of cylindrical and regular polyhedral configurations. The described module may be used in solving the dosimetry problems

[en] Results of experimental and calculational studies permitting to estimate the neutron flux spatial distribution in ionization chamber channels of the commercial WWER-1000 and WWER-440 reactors and also of the WWER-440 reactor with water biological shield are presented. The integral neutron flux density distribution along the channel cross section approximately at height of the core middle and the corresponding thermal and fast neutron flux density distributions are measured by the activation detectors. It is shown that the difference in fast neutron flux density exceeds that of thermal neutrons. The commercial WWER-1000 type reactor the fast neutron flux density is decreased by the factor of 1.7, and thermal neutron flux density - by the factor of 1.2, for the commercial WWER-440 reactor these values are 1.37 and 1.18, and for the WWER-440 one with water shield - 1.5 and 1.18

[en] The PRAKTINEK program (FORTRAN, BESM-6 computer) for one-group kinetic calculation of the LWGR type reactor cluster cell using the method of surface pseudosources in the G_N^P-approximation is described. The mean by zones neutron fluxes, neutron fluxes and currents at the cell boundaries, the probabilities of neutron absorption in each zone and the thermal neutron utilization factor are calculated

[en] The RNDT program (FORTRAN-CERN, BESM-6 computer) for calculation of deformed-stressed state of fuel elements with continuous or hollow pellet rods in the case of tight fuel-cladding bond is described. The fuel element is represented as two endless cylinders with tight bond in axial and radial directions. A fuel element being under the effect of nonstationary fields of radiation and temperature loaded by internal and external pressures as well as by an axial force is considered. The theories of thermoradiation plasticity and Birger type creep are used to solve the problem