The design features and construction status of the BN-800 reactor being built at the Beloyarskaya nuclear power plant and the main scientific and technical problems which will be solved with the construction of this reactor are examined. The most important ones are mastering a closed fuel cycle, checking new technical solutions, and testing improved fuel and construction materials. The directions for improving the technical-economic performance characteristics of fast reactors that the BN-800 and -1800 designs can provide are described. It is shown that economic performance indicators which are at least as good as those of VVÉR reactor with close power levels can be achieved.

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in , and the BN-800 design as well as new technical solutions that increase facility cost-effectiveness and safety. The BN-1200 design must permit the reactor to operate with different cores, including with denser fuel. The main fuel variant considered is oxide fuel and for the nearest term nitride fuel, for which the production technology involves the same steps as the oxide technology. The main approaches for choosing the parameters of the BN-1200 core as well as the results of computational studies are presented.An advanced sodium-cooled fast reactor satisfying the requirements for nuclear power in the 21st century is now being developed in our country. Initially, 1800 MW(e) capacity was being considered for this reactor [1]. However, for many reasons, specifically, to ensure transportability of the main equipment by rail, its capacity was lowered to 1200 MW. The development of BN-1200 is based on the maximum use of the solutions implemented in BN-350 and -600 and in the BN-800 design as well as new technical solutions which increase cost-effectiveness and safety.BN-1200 Core. The main variant considered is oxide fuel and for the nearest term nitride fuel, whose production technology is largely identical to that of oxide fuel. Variants with heterogeneous introduction of regions with depleted metallic uranium are also considered. Metallic fuel does not satisfy the BN-1200 thermodynamic indicators which have been adopted. For fuel elements to work reliably with such fuel, the maximum coolant temperature must be lowered by approximately 100°C.The high burnup adopted for the fuel (16-18% h.a.) is attained by using in the fuel-element cladding EP-450 type ferrite-martensite steel; for this, the maximum cladding temperature in the design is limited by 670°C. For such fuel-element cladding temperature, the coolant heating in the reactor is lowered (140°C). This makes it possible to lower the maximum sodium temperature at the entrance to the fuel assemblies; this temperature is associated with the nonuniformity of the coolant heating in the core and the fuel assemblies. In the future, a transition is planned to low-swelling ferrite-class high-temperature steel, dispersion hardened by oxides of rare-earth elements.An important characteristic of the core is its volume heat density, since it actually determines the core size and critical load. Initially, a high heat density (~500 MW/m 3 ) was chosen for fast reactors. This choice was made to minimize the

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].

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Methodological program techniques for performing neutron-physical calculations of fast-reactor cores with a description of the main codes TRIGEX, JARFR, GEFEST, MMKKENO, and ModExSys are described. The results of verification and assessment of the methodological errors on the basis of a test model of BN-600 are presented. Transport effects and mesh errors in the reaction rate distributions in the core, lateral screen, and in-reactor storage area are evaluated. An integral assessment of the computational accuracy is performed by comparing with the experimental data obtained by γ-scans of BN-600 fuel assemblies. It is shown that calculations describe the experimental data over the core with an error no worse than 5%. In the lateral screen and the in-reactor storage area, the discrepancy between the calculations and experiment does not exceed 20-30%.Computational modeling of the neutron field is the main method of determining the energy release in a fast-reactor core, where, as a rule, there is no apparatus for tracking the power density. Correspondingly, computational accuracy is one of the key elements of the operational reliability of fuel elements and assemblies. Another aspect of the problem is substantiating the conservatism of fast-reactor core design -design limits or margins affecting the technical-economic characteristics. Experience in operating BN-600 makes it possible to evaluate the accuracy of the design and operational programs under real conditions of a commercial medium-power fast reactor and to take it into account in subsequent development work.Such an analysis is based on the measurement of the power density in BN-600 by means of γ-scanning. Eleven such experiments were completed in the time period from physical startup of the reactor in 1980 and up to 2008. The present article analyzes the last experiments performed in 2003-2006 in the course of transitioning from a 01M1 core to a 01M2 core with maximum burnup 11.1% [1]. This upgrade was an important next step in making use of the serviceability margin of reactor fuel assemblies, which required greater attention to monitoring the characteristics of the core [2].Program and Constants System for Neutron-Physical Calculations of Fast Reactors. A consistent system for performing neutron-physical calculations of fast reactors is now available. It includes the program systems TRIGEX [3], JARFR [4], GEFEST [5], and MMKKENO [6] as well as the BNAB constants library [7,8]. It should be noted that the base