[en] In order to decrease the computer calculation time when solving the prob.lems of radiation transport by the Monte-Carlo method the methods for modelling the neutron flight direction after scattering are suggested. The methods have been verified in computing the cosine path for oxygen. The first method has been verified at the energy of 10 MeV when indicatrix of scattering has strong resonance and saving of computer time amounts to 9%. The second method is independent of the type of the scattering indicatrix and gives the computer time saving of about 8%

[en] Lack of neutron cross sections, particularly in the range of unresolved resonances, is the main difficulty faced in using the Monte-Carlo method for calculation of reactors. The algorithm of distribution function calculation for ²³⁹Pu with account of specificity of the nucleus has been proposed. The method to obtain the probability tables for simulation of neutron nuclei with collisions in the unresolved resonance range is described. By interpolating the tables one can readily and precisely to determine cross sections in this range. The calculation of probability tables according to the proposed method for ²³⁹Pu nucleus is demonstrated

[en] Three methods to obtain ksub(eff) of a nuclear reactor by the Monte Carlo method are described and compared: 1) the method of generation; 2) the devision matrix method; 3) combination of the method of generation with the method of surface multiplication. The calculation results show, that in the range of a statistical error the values of the effective multiplication factor coincide for all three methods. So, two other methods can be used for calculation of multiplying systems alongside with the most common method of generation. The division matrix method gives the possibility to obtain the distribution of neutron importance relative to the fission process. With the help of the division matrix method the conjugated kinetic equation can be solved without complex procesure eeprocedure of re-definition of nuclear physical constants. Thus, the division matrix method possesses largest advantages

[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] Method for solution of conjugated equation of neutron transport by the Monte Carlo method with detailed account of energy dependence of neutron reaction cross sections is presented. Fast neutron fluxes for three groups, coming from water barriers of different thickness have been calculated. The method enables to calculate neutron fluxes at any attenuation with practically available accuracy. Limits are imposed only by a word length of particular computer. Agreement of calculated data with those obtained by authors on the basis of group approach has been obtained

[en] By means of computerized simulation methods calculation of vacancies redistribution in nickel films, depending on the temperature according to the accepted cooling law taking account of the film structural state, is carried out. It is shown that at pulse high-energy methods of nickel film treatment characterized by nondimensional rates of cooling 10sup(-6)-10sup(-8) there exist a critical temperature below which vacancies don't attain external sinks (grain boundaries, film surface), and redistribution of vacancies along internal sinks as well as formation of vacancy complexes and clusters, i.e. quenching, take place. Above this temperature vacancies are significantly absorbed by film surface and grain boundaries. The most efficient is film thermal treatment by pulse methods with the grain sizes being d< h and cooling rate <10sup(-8), as in this case most of vacancies will be intensively absorbed by grain boundaries and film surface

[en] The Monte Carlo method is widely used for solving the problem of radiation-substance interaction. Special codes of modular type are being developed the main and laborious component of which is a geometrical module. The paper seeks to develop a universal geometrical module for simulating the range of neutral and charged particles in a complex geometry media. It also presents the algorithm and the modeule's potential for simulating a particle history when the complex geometry systems are calculated by the Monte Carlo method

[en] Two methods of neutron transverse cross-section representations in the region of resolved resonances have been compared, namely, the Breit-Wigner single-level formalism method and that of cubic splines. The correlation of two considered methods according to accuracy, computer time expenditures and the volume of the used computer memory has been ellustrated using ²³⁵U cross-section as an example. The performed calculations show that interpolation by cubic splines is offered to be used in a number of problems involving the computer time saving during cross-section computations as well as high accuracy of cross-section representations. It is preferable to specify cross-sections by the Breit-Wigner formalism in the problems where accurate cross-section specifications are not required

[en] An algorithm for the Monte Carlo calculation of local reactivity perturbations caused by insertion of small samples into critical assemblies, is suggested. The algorithm is based of the direct and adjoint transport equations from the perturbation area. The described algorithm is realized in a program. Central reactivity coefficients in the GODIVA critical assembly for different material samples were calculated in order to test the algorithm. The calculation result divergence with experimental data amounts to about 10%

[en] The dose of Compton and photoelectrons escaping from fuel cladding into the N₂O₄ coolant and absorbing by the gas is calculated by the Monte Carlo method using direct simulation of electron energy scattering. The dose value obtained for coolants of different density is compared with those calculated in terms of the Bragg theory for medium electron energies and using the Cormack-Johns approximation taking into account the resulting electron energy spectrum in the fuel cladding. It is shown, that the methods developed earlier are the good approximations for the absorbed dose calculation in the density range less than 0.258 g/cm³