[en] Main equations describing neutron flux and values distribution in the one-dimensional geometry, main algorithmic peculiarities of the programs for macroscopic constants preparation, methods of solution and main parameters of programs for solving main and conjugate reactor equations in P₁,P₂ and Ssub(n) approximations are given. The program for calculation of main reactor functionals is described. All programs mentioned are prepared for computers M-220 and M-222 and are combined in two programs complexes, M-26 and DIOR1. (author)

[en] To obtain the library of group constants derived from the estimated nuclear data the programs system FEDGROUP was developed. This system is flexible enough to obtain any group constants library derived from any source of estimated data. The FEDGROUP system is open one which means that it can be connected to any system of constants maintenance and any algorithm of group constants calculation can be added. Teh system is written on FORTRAN and intended for BESM-6 and CDC-3300 computers. (I.T.)

[en] The one-dimensional diffuse theory is used for reactor physics calculations of fast reactors. Computer program based on the one-dimensional diffuse theory is speedy and not memory consuming. The algorithm is described for the three-zone fast reactor criticality computation in one-dimensional diffusion approximation. This algorithm is realised on IBM 370/135 computer. (I.T.)

[en] The physical algorithm of mathematical program for fast reactor computation by the Monte-Carlo method is described (program ''MAK''). Continuous by energy (but not the group) presentation of the neutron cross sections, fission neutrons spectra and other probability quantities are used. The programs are developed for Minsk-22 and Minsk-32 computers. (author)

[en] The experience of small group methods application for computation of fast reactors in two-dimensional hexagonal geometry is discussed. Basic information on algorithms and computer programs used is provided. Influence of number and limits of the energy groups on the accuracy of main physical characteristics computation is discussed. It is shown that two group approximation has acceptable precision. Comparison of calculated and experimental magnitudes of the modelling assembly of the reactor BN-600 is given. (author)

[en] Computer program for the neutron flux calculation based on the multigroup diffuse theory used for some years in the Institute is described. This program can be applied not only for solution of standard problems but for solving some of special ones. The calculation programs system RHEIN contains programs for creation of the multigroup parameters tables, for one and two-dimensional calculations of the neutron flux and for the integral reactor parameters calculations. (I.T.)

[en] The task of optimization of physical parameters of fast power reactor is formulated as a task of minimization of some non-linear multiparameter function in presence of limits. With given type of limitations and on the base of characteristics of the function optimized, development of the method of minimization was done. This method is a combination of the method of local linear approximation and iterative method of penal functions developed by the authors. Convergence and accuracy of determination of extremum of the method proposed are investigated. Efficiency of the search algorithm applied for the method is illustrated by the example of minimizing of test functions. (author)

[en] Application of the non-linear programming methods on optimization of nuclear materials distribution in fast reactor is discussed. The programming task composition is made on the basis of the reactor calculation dependent on the fuel distribution strategy. As an illustration of this method application the solution of simple example is given. Solution of the non-linear program is done on the basis of the numerical method SUMT. (I.T.)

[en] Variation method for solution of the single velocity kinetic equations with anisotropic scattering in periodic cells is described. The problems of convergence and stability of this method are discussed. Estimations of the linear functionals resulted from approximate solution of the problem and estimation of the approximate solution itself are given. (author)

[en] The problems of practical applications of amplitude-subgroup method of space-energy neutrons distribution is discussed. The method is applied taking into account fine structure of spectrum and scattering anisotropy in the fast heterogeneous critical assemblies and reactors. (author)