The status of work on the development of a 1200 MW sodium-cooled reactor facility for serial construction is presented. The general characteristics of the facility and the power-generating unit as well as the objectives which must be attained as a result of the design are presented. The design of the power-generating group is based on solutions some of which have been checked during the operation of sodium-cooled reactors in Russia and some have been validated by the appropriate research and development work performed for BN-800. At the same time, new solutions are used which are aimed at improving the technical-economic indicators and increasing the level of safety. Additional R&D work will be needed to validate them.The nuclear future of Russia and most other countries developing nuclear power involves fast reactors and a closed fuel cycle. Thus far only sodium-cooled fast reactors are ready for commercial adoption on a wide scale. In our country, we have more than 50 years of experience in developing and successfully operating such reactors: BR-5 (1959) → BR-10 (1973-2002) → BOR-60 (1969 -in operation) → BN-350 (1973-1998) → BN-600 (1980 -in operation) → BN-800 (under construction). A design of an advanced power-generating unit with a sodium-cooled BN-1200 reactor as the foundation for the initial step in the serial construction of reactors of this type is under development.Target Indicators of Reactor Facility and Power-Generating Unit Design. The development of a power-generating unit must meet the following requirements set for new-generation reactors and nuclear power plants:1) competitiveness with advanced power-generating unit based on reactors of a different type and power-generating units operating on fossil fuel;2) safety increased to a level that makes it unnecessary to take measures to protect the general public outside the borders of the nuclear power plant site for any possible accidents;3) attainment of breeding ratios from 1.2 (stage 1) to 1.3-1.35 (stage 2) using mixed oxide fuel and 1.45 with nitride fuel; 4) reducing the construction time for serially built power-generating units to 48 months; and 5) possibility of introducing a series of reactors in 2-3 years after the startup of the main power-generating unit.
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].
Measures to decrease the sodium void effect of reactivity and the influence of this effect on the technicaleconomic performance and the safety of BN-1200 are analyzed. Three variants of the core structure differing by the structural implementation and dimensions are examined. It is shown that a flattened core with a sodium cavity, replacing the top end screen, gives self-protection of the reactor with respect to unanticipated accidents. The elimination of the sodium cavity and an increase of the core height result in degradation of the self-protection properties but at the same time improve the technical-economic properties of the reactor. The possibilities for optimizing the construction of the reactor from the standpoint of reaching a compromise between safety properties and the technical-economic characteristics are discussed.In the Soviet Union, after the Chernobyl accident, a requirement of zero integral void effect of reactivity was adopted for BN-800, which design was completed at the beginning of the 1990s. A new concept was developed to satisfy this requirement: a sodium cavity was to be organized above the core. A decrease of the density or removal of sodium from this cavity causes negative reactivity. It turned out that dimensions of the cavity that would ensure a negative or zero (integral) void effect of reactivity when sodium boils out of all fuel assemblies in the core can be found. Studies of accidents in BN-800 with the new core showed that the solution adopted is effective for increasing safety.A core with zero sodium void effect of reactivity is also proposed for advanced high-capacity sodium-cooled fast reactors which are now being designed. However, this makes it necessary to decrease the height of the core to 85 cm and, correspondingly, increase the radius, which creates difficulties in optimizing a high-power core, since the dimensions of the header plate and the rotating plug increase and the effectiveness of the control rods decreases. The sodium void effect of reactivity is important for unanticipated accidents with possible removal of a large quantity of sodium from the core.In this connection, discussions periodically arise about the desirability of a core with zero or close to zero sodium void effect of reactivity in advanced designs, especially since the normative documents on safety do not make the satisfaction of this condition mandatory. Rejection of this concept would make it possible to improve appreciably certain technicaleconomic characteristics of a reactor.Three models of the BN-1200 core, differing by the height of the core (85 and 100 cm), the diameter of the core, and composition of the top end screen (sodium cavity or breeding screen) and, correspondingly, the sodium void effect of
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