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Recent assessments of the strategic prospects of the nuclear energy industry tend to show a somewhat condescending attitude towards fusion, which, regretfully, has grounds given the current state of affairs in nuclear science. However, the analysis of the problems and the potentialities of the fusion of light nuclei and fission of heavy nuclei suggests that a full-scale development of each of the two nuclear technologies inevitably involves the need to overcome some formidable technological, materials science, environmental, and economic problems, some of which may put the feasibility of pursuing these nuclear energy programs at question. At the same time, the physical features of the fission and fusion processes objectively suggest that it may be reasonable to combine the two nuclear technologies within a joint nuclear energy system to enable a greater synergetic effect. This apparently would markedly lessen the problems attendant upon the application of each of the discussed technologies in the course of the full-scale development of nuclear energy.

Analysis of the problems and potential of two nuclear technologies based on fusion reactions of light nuclei and fission of heavy nuclei shows the following. Independent large-scale development of each one will make it necessary to overcome still unsolved technological, materials science, environmental, and economic problems, raising the question of the expediency of further development of these energy industries. At the same time, the physical features of the fission and fusion processes objectively indicate the expediency of combining them into a unified nuclear energy system, which will have a large synergistic effect. This will greatly mitigate the adverse aspects of each technology that will appear if the two systems develop independently. Calculations of neutron multiplication in the blanket of a hybrid thermonuclear reactor that confirm the physical feasibility and reliability of picking the direction of development in the form of a unified nuclear energy system are presented.

Nuclear energy system development in the current century is considered in the article. The importance of this issue both for the future and for the choice of ways to solve current energy problems is high, which is largely due to the increasing environmental restrictions. Nuclear energy using promises negative impact of the energy system on enviroment minimizing. However, the danger of radiative pollution of the environment is a significant deterrent to the large scale industry development. The article discusses the problem of the nuclear power system development in the variant of the coordinated use of fission reactors for energy production and hybrid thermonuclear reactors for production of artificial fuel from thorium raw materials for fission reactors. Estimates are presented to substantiate the preferences of the structure of the nuclear power system under consideration.

The DAREUS software package designed for modeling dynamic processes in the cores of experimental solution reactors is described. The KIR program based on the Monte Carlo method is used in the package to compute the necessary kinetic parameters. The results of the calculations of some test cases are given.

The Argus research solution reactor and the main principles of experimental validation of its safety are described. The described validation is computationally confirmed in accordance with the requirements of the normative documents using the DAREUS software package designed for modeling dynamic processes in the cores of experimental solution reactors.

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.

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.

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.

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.

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.

The growth prospects for nuclear power are examined: impeding and driving forces. The main problem facing nuclear power today is the management of spent fuel and wastes. Eventually, it will become economically inexpedient to store spent fuel; the natural solution is to reprocess spent fuel, extract the fi ssile isotopes, and recycle it in thermal or fast reactors. Since the fast-reactor fraction in the current system is trivial and the situation will remain the same in the near- and medium-term futures, a variant of plutonium recycling in thermal reactors with a different fuel arrangement is examined. It is shown that a heterogeneous fuel arrangement can give a 10-fold reduction of the radiation load in reprocessing enterprises. Different setups are compared in terms of the potential danger presented by spent-fuel reprocessing. The advantages of the thorium fuel cycle are examined from the standpoint of radiation and environmental safety.

The concentration of neutron radiation far from a source is discussed. A new method is proposed for solving this problem by choosing from among the scattered neutrons the ones that have acquired the required direction of flight and energy. The material along which such a neutron will travel must possess a small neutron cross section at the given energy. Materials consisting of isotopes with a large resonance cross section, for example, 56Fe, meet the requirements. The method discussed for concentrating high-energy neutrons makes it possible to increase their concentration more than 1000-fold in the detection area far from the source compared with isotropic propagation away from the source in all directions with equal probability.

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.

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.

The management of accumulated spent fuel is now an urgent problem. Questions concerning the need for reprocessing and subsequent isolation of high-level wastes are elucidated. It is shown that the BN-1200 and BREST fast reactors now under consideration for closing the fuel cycle do not permit efficient use of plutonium, because for the existing spent-fuel reprocessing technologies these reactors could be very expensive or a long cool-down time may be required, which will reduce the positive effect from closing the fuel cycle. The requirements of fast reactors that make fast reprocessing of spent fuel possible with the minimal environmental load on the fuel cycle are formulated.

A variant of a fuel load for burning power-grade plutonium in VVER is examined. Using power-grade plutonium in fuel elements without the raw isotope makes it possible to increase its power potential almost 1.5-fold compared with the conventional mixed fuel and eliminates the need for repeated processing of the spent fuel obtained when plutonium is burned.

A concept is proposed for a fast reactor which has a potential for overcoming the safety and fuel-cycle related fears of large-scale development of nuclear power. This proposal is based on ideas advanced in 1958 at the Second Geneva Conference on World Use of Atomic Energy and is aimed at the implementation of the innovation potential of nuclear power for solving the most complex safety and fuel supply problems of reactor technology.

An algorithm for direct Monte Carlo modeling of the isotopic kinetic is described. The nuclear reactions, including those that change the isotopic composition (n, α), (n, 2n), and others, are recorded for each neutron or group of neutrons when performing a Monte Carlo calculation of neutron transport. At the same time as the neutron transport calculations, the change of the concentration of the nuclides and the irradiation time are calculated on the basis of the number of recorded interactions and energy released. The computational results presented show that it is in principle possible to apply this approach to isotopic kinetic problems.

In the latest assessments of the strategic prospects for the development of nuclear energy, we can note the tendency of a condescendingly arrogant attitude towards thermonuclear fusion, which, unfortunately, largely corresponds to the real state of affairs. At the same time, an analysis of the problems and potential of two nuclear technologies based on reactions of fusion of light nuclei and fission of heavy ones shows the following. Independent large-scale development of each of these areas will inevitably lead to the need to overcome the still unresolved technological, material science, environmental and economic problems, some of which raise the question of the appropriateness of further development of these energy sectors. At the same time, the physical features of fission processes and fusion objectively indicate the expediency of combining them within the framework of a single nuclear energy system. Such a combination will provide a great synergistic effect, as a result of which the negative manifestations of the complex problems of each technology will significantly weaken on the path of large-scale development of the industry.

A method for calculating, using a matrix neutron first-collisions probabilities which is reconstructed at each step, the burnup of cells and fuel assemblies is presented. A method for reconstruction and correction of the first-collisions probabilities using average chords to the first collision of a neutron, which are calculated using the geometric module for reconstructing the stochastic trajectories of neutrons, is described. The results of a calculation of the multiplication coefficient of elementary cells with different material composition relative to the reference cell are presented. Computational results are presented for the burnup of a VVER fuel-assembly fragment with a consumable absorber are presented.

The results of a search for the optimal water-uranium ratio for VVER with no changes to the fuel elements and assemblies are presented. It is shown that decreasing the volume of a fuel pellet while maintaining the same 235 U load as a result of increasing the enrichment will decrease the specific consumption of uranium and increase the run time. The proposed modifications to the fuel can be implemented in operating reactors and advanced VVER designs.

The DESAE computer program has been developed, within the framework of the INPRO project implemented under IAEA aegis, as a program complex for making predictive calculations of nuclear power growth. The main purpose of these calculations is to determine the most likely trends in the structural development and scales of the industry and the expected time frames when the new technologies will be needed. Their main benefit lies in the possibility of constructing a balanced picture, as close to reality as possible, of the future structure and developing on this basis a well-substantiated program for strategic scientific-technical policy. A brief presentation of a DESAE calculation of a scenario for nuclear power growth in the country up to 2100 is given. This scenario corresponds to the above-mentioned requirement of balance with respect to the consumption of natural uranium as well as the amount of electricity and heat that will be required.

The main results of computational and experimental investigations of the neutron-physical characteristics of the GT-MHR high-temperature gas-cooled reactor are presented: analysis of the possible reasons for the uncertainties in the calculations of the neutron-physical characteristics of a core with plutonium fuel using different computer programs, computational and experimental investigations taking account of the experiments on the Astra critical assembly simulating the ring-shaped GT-MHR core, and the effect of using different nuclear data on the temperature coefficient of reactivity computed with high-accuracy computational codes such as MONTEBURNS-MCNP 5-ORIGEN 2.1.

The results of comparative calculations of material balances are presented for two concepts for the growth of nuclear power in the 21st century. The concepts of growth on the basis of fast reactors without expanded breeding of fuel and the conventional concept with a two-component structure of nuclear power growth are examined. The advantages of the conventional concept, capable of producing substantial savings of natural uranium and/or making it possible to increase substantially the installed capacities of nuclear power plants, are demonstrated.

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark–results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (APOLLO2, TRIPOLI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section.Copyright © 2008 by ASME