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Information is presented on the BN-800 design, the second design following BN-600, power-generating unit with a fast reactor. The main stages of the development of the design begun in the 1980s, modified in the 1990s after the Chernobyl accident, and accepted for construction within the government program starting in 2000 are presented. The fundamental differences of BN-800 from BN-600 are characterized, and current R&D work is briefly described. Information is presented on the construction of BN-800 at the Beloyarskaya nuclear power plant, where the BN-600 has been operating since 1980.The development of the BN-800 reactor started in the USSR immediately after the completion of the BN-600 design, which reactor was started up in 1980 at the Beloyarskaya nuclear power plant. At that time, the BN-800 was regarded as an intermediate stage in the development of large-scale power generation using fast reactors. A small series of four reactors was planned for construction at the Beloyarskaya and Southern Urals nuclear power plants. Subsequently, a transition was planned to high-capacity reactors -up to 1600 MW. However, the construction of the first two BN-800 started at these sites was stopped because of the Chernobyl accident in 1986.Nonetheless, work on the BN-800 design continued. This was directed toward increasing safety and improving costeffectiveness. The research performed in these directions in 1990 was recognized as being successful. A license was obtained for resuming BN-800 construction at the Beloyarskaya nuclear power plant in 1997 and a license for construction at the Southern Urals nuclear power plant was obtained in 1998. These were the first licenses for building nuclear power plants after the Chernobyl accident.Together with accelerated construction of the high-capacity power-generating units with VVER-1000, a special federal program provides for the development of innovative technologies for nuclear power generation. This includes work on fast reactors to which the future transition to a closed fuel cycle is tied; such a fuel cycle will permit the most efficient use of uranium resources and solving the ecological problems of handling spent fuel and radioactive wastes. In the innovative part of the special program, a central role is given to the construction of a BN-800 sodium-cooled fast reactor, which should become an important stage in the development of fast breeder reactors and the formation of a closed fuel cycle in nuclear power [1,2].

Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), , VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.Analysis of world energy consumption with limited resources for conventional power generation shows that intensive economic development is impossible without the establishment of large-scale nuclear power capable of supplying the energy required for a substantial part of the growth in energy needs. At the present time, the largest amount of fuel-energy resources, including the most expensive and scarce -oil and gas, is used for producing heat with diverse potentials, approximately three times more than for electricity production [1, 2]. The expansion of nuclear power into high-temperature industrial production, where the scarcity of fossil fuel is even more acutely felt than in the production of electricity, can be attained by bringing high-temperature gas-cooled reactors (HTGR) which are capable of generating high-potential heat, into the nuclear power system. HTGR Development in the USA and Russia. HTGR use fuel with specific qualities as compared with other types of reactors; it contains spherical fuel pellets or fuel blocks containing fuel compacts. The foundation of the fuel composition is microfuel containing a spherical kernel with ceramic multilayer coatings, which are the main barriers for confining fission products. The first US facility with an experimental reactor, developed by the General Atomics Company, was introduced in 1967 at the Peach Bottom 115 MW(t) nuclear power plant. The objective was to demonstrate the possibilities of using HTGR for commercial production of electricity with high-temperature steam and to obtain the experience required for developing a more powerful nuclear power plant. The problem of developing a reactor with high fuel burnup (75 MW·days/kg) was posed at the same time. Thirty-three fuel modifications, including the fuel rods at the Fort St. Vrain nuclear power plant, were tested [3].The pilot nuclear power plant at Fort St. Vrain with a high-temperature gas-cooled reactor was put into operation in 1976. The main equipment in the first loop of the facility consisted of a 842 MW(t) reactor with a core comprised of prismatic fuel assemblies, steam-generators, and axial gas blowers with steam-turbine drive placed beneath the core in a vessel

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