[en] For the calculation justification and optimization of the fuel cycle on the basis of thermonuclear operating units and nuclear reactors - fuel consumers and as part of the development of the UNK software package, the GAMSOUR software module was developed for calculating the specific and group spectrum of delayed gamma rays in irradiated material and the effective radiation doses of personnel when working with it. The results of verification of the UNK software package for calculating the dose rate of materials containing U-233 with U-232 impurities are presented. The results of the comparison with the dose rate calculations obtained by other programs showed that the differences in the assessment of the dose rate depending on the exposure time of uranium-233 do not exceed several cents over the entire time interval up to 15 years. The results of the research have shown that in order to correctly account for gamma rays sources, it is necessary to take into account the Tl-209 nuclide. The exclusion of this nuclide from the U-233 decay chain leads to a systematic underestimation of the gamma rays source and, consequently, the effective radiation dose depending on the exposure time of U-233. The GAMSOUR software module is implemented as an independent software tool and can be used in software complexes for solving the gamma rays transfer equation.

[en] Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem

[en] An algorithm for calculation of prompt fission neutron lifetimes in a nuclear reactor by the Monte Carlo method is described. Evaluation of the importance function is carried out with solution of the neutron transport equation without solving the adjoint equation. The results of the prompt neutron lifetime calculations performed within some critical experiments are presented and compared with the experimental results.

[en] Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed

[en] Verification of the SUHAM-U code has been carried out by the calculation of two-dimensional benchmark-experiment on critical light-water facility VENUS-2. Comparisons with experimental data and calculations by Monte-Carlo code UNK with the same nuclear data library B645 for basic isotopes have been fulfilled. Calculations of two-dimensional facility were carried out with using experimentally measured buckling values. Possibility of SUHAM code application for computations of PWR reactor with uranium and MOX fuel has been demonstrated. (authors)

[en] Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

[en] Integral equations for the shape functions in the adiabatic, quasi-static, and improved quasi-static approximations are presented. The approach to solving these equations by the Monte Carlo method is described.

[en] A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

[en] An algorithm for calculation of the neutron importance function and the delayed neutron effective fraction by the Monte Carlo method implemented in the KIR program is presented. The results of calculation of the delayed neutron effective fraction in some critical experiments are given in comparison with the experimental results.

[en] The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.