Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.

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Boiling experiments on eutectic sodium-potassium alloy in the model of fast reactor subassembly under conditions of low-velocity circulation carried out at the IPPE call for further investigations into numerical modeling of the process. The paper presents analysis of pin bundle liquid metal boiling, stages of the process, its characteristics (wall temperature, coolant temperature, flow rate. pressure void fraction and others), that allowed the pattern map to be drawn. The problem of conversion of the data gained in Na-K mock-up experiments to in-pile sodium reactor operating conditions is analyzed here, as well as thermodynamic similarity of liquid metal coolants and eutectic Na-K alloy. Data on bundle boiling in Na-K are presented in comparison with those in different liquid metals. Analysis of data on liquid metal heat transfer in cases of pool boiling, boiling in tubes, in slots, and in pin bundles, as well as data on critical heat flux in tubes was performed and discussed in the paper. The relationship for calculation of critical heat flux in liquid metal derived by the authors is presented. Results of numerical modeling of liquid metal boiling heat transfer during accident cooling of reactor core applied to experimental conditions of going from forced to natural circulation are presented, too.

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Studies show that in principle it is possible by relying on the BN-600 design to secure the required parameters of a reactor setup with a fast reactor with high-temperature sodium coolant (BN-VT) for hydrogen production. The calculations for 600 MW(t) BN-VT for electricity and hydrogen production based on solid-oxide electrolysis of water taking account of a fundamentally different method of purificationremoval of hydrogen and tritium from the sodium by vacuum pumping through special membranes -have shown that the efficiency of such a system is ~40% and hydrogen production volume is 2.8·10 4 liters/sec. The purification of hydrogen by removing tritium to a concentration ensuring the maximum admissible hydrogen concentration in the hydrogen produced 3.26 Bq/liter, imposes additional requirements on the hydrogen removal system.The most important problem determining the future development of environmentally clean power generation is bringing hydrogen into the fuel cycle. Hydrogen is an attractive element for replacing fossil fuels (oil, gas), though it is not a source but rather a carrier of energy. As expected, the demand for hydrogen production is increasing rapidly. At the present time, the main method of hydrogen production is steam conversion of methane. However, from the standpoint of the longterm future of the large-scale production of hydrogen this method is not viable because it requires the use of nonrenewable resources and is accompanied by the release of greenhouse gases. For this reason, alternative methods of hydrogen production by splitting water using thermochemical or electrolytic heat sources are being developed.At present, the best developed high-temperature sources of heat could be Gen-IV nuclear reactors because the proposed heat carriers are gases, liquid metals (sodium, lead) and fused salts. The temperature at the exit from the core in these reactors can reach 900-950°C. They represent a new class of reactors intended for electricity production with high efficiency (50%) and supporting important technological processes in hydrogen production, coal gasification and liquefaction, deep reprocessing of oil, conversion of biomass into liquid fuel as well as in the chemical industry and metallurgy.The conceptual studies performed at the Leipunskii Institute for Physics and Power Engineering (FEI) on the development of a high-energy high-temperature fast sodium reactor (BN-VT) show that the development of such a reactor is technically feasible [1]. The technological problems associated with the high temperature in the reactor setup are paramount. In the present work, a great deal of attention is given to validating the development of a system that secures the quality of the sodium coolant at high temperature and the hydrogen concentration for a long service life, heat-and radiation-resistant hightemperature construction materials and their corrosion resistance with oxygen content in the sodium coolant at the level 0.1 ppm. Concrete estimates were made for 600 MW(t) BN-VT.

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