Improvement of the equipment in fast-neutron reactor facilities

Information is presented on the development of the main equipment of the BN-1200 advanced reactor facility: first- and second-loop main circulation pumps, intermediate heat exchanger, actuation mechanism of the cold filter-trap control and protection system, autonomous and air heat exchangers, steam generator. The approach to the development of the equipment is based on maximum use of the experience gained in operating BN-350 and -600 as well as experience in developing the BN-800 design, which gives a basis for ensuring reliable operation of the BN-1200 equipment. New solutions for improving the technical-economic indicators and increasing the safety of a power-generating unit, whose validation required R&D work, are examined at the same time.     

Core and fuel cycle for an advanced sodium-cooled fast reactor

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in BN-350, -600, 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.     

BN-800 design validation and construction status

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.     

Concept of an advanced power-generating unit with a BN-1200 sodium-cooled fast reactor

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.     

Fuel for advanced sodium-cooled fast reactors: current status and plans

Different fuel compositions have been irradiated in the BR-10, BOR-60, BN-350, and BN-600 reactors and investigated: PuO₂, UC, UN, UPuN, UO₂, UPuO₂, metallic doped and undoped alloys, and fuel compositions with inert matrices. Studies of fuel elements with UPuN as well as with fuel compositions based in MgO and ZrN irradiated in the BOR-60 reactor were completed in 2009. The main results of the investigations of different fuel compositions are presented and the problems of ensuring the serviceability of BN-800 and -1200 fuel elements are discussed.     

Safety properties of bn-1200 with mixed oxide and nitride fuel: a comparison

Along with mixed oxide fuel, the possibility of using in BN-1200 dense nitride fuel, making it possible to attain higher technical-economic performance, is also studied. However, safety analysis will determine the choice of fuel type. In this connection, it is important to perform a comparative analysis of the inherent safety properties for variants of the BN-1200 core with mixed uranium-plutonium and nitride fuel for the most serious unanticipated loss-of-power accident with failure of all emergency protection organs of the reactor simultaneously. A two-dimensional version of the COREMELT computer code was used in the calculations. The computational analysis showed that the inherent safety of BN-1200 is much greater with nitride than with mixed uranium-plutonium fuel.     

Experience in Operating the BN-600 Unit at the Belyi Yar Nuclear Power Plant

Data on the experience in operating the BN-600 reactor with sodium coolant in the first and second loops are presented. It is shown that the equipment and the systems in the sodium loops, which have operated for more than 23 years, are highly realiable. The average installed capacity utilization factor for this period reached about 74%. The losses due to rupture of the heat-transfer pipes in the steam-generator modules and the sodium leaks from the loops were about 0.3%. Disruptions of normal operation are detected reliably and contained by the safety systems present in the unit. Unique experience in performing maintenance and repair work on the sodium equipment has been gained on the BN-600 reactor.     

Optimization of the Technical and Economic Performance Indicators of Nuclear Power Plants with Fast Reactors

The results of design analyses for improving nuclear plants with fast reactors, specifically, by using cartridge-vessel generators instead of sectional-modular generators, are presented. It is shown for a nuclear power plant with a BN-800 reactor that the cartridge-vessel steam generators designed by the Special Machine Design Office substantially decrease the metal content, dimensions, mass, amount of construction work, and construction costs of the main vessel of the nuclear power plant.In the BN-800 design, a cartridge-vessel steam generator decreases the specific capital costs for constructing a power-generating unit of a nuclear power plant by approximately 8%, which substantially closes the gap between these costs for nuclear power plnats with BN-800 and VVER-1000 reactors.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 403–412, June 2005.     

Investigation of the influence of the sodium void effect of reactivity on the technical-economic performance and safety of an advanced fast reactor

Measures to decrease the sodium void effect of reactivity and the influence of this effect on the technical-economic 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.     

Successive development phases for sodium-cooled fast reactors

The development and operation of sodium-cooled fast reactors and the prospects for developing the next generation of such reactors are reviewed. The main phases, the problems of each phase, and the results obtained by solving these problems are shown. The main results obtained by adopting innovative technical design solutions, making it possible to consider the problem of developing a competitive power-generating unit with a BN-1200 reactor, are examined and described.     

Some results of the work of the COMECON Scientific-Technical Coordinating Council on fast reactors

Conclusions The experience of 6 years of work by the Council of Scientific-Technical Cooperation as well as the meetings of specialists have made it possible to determine the subject-matter for joint research on fast reactors with due account for the capabilities and interests of each country. Thanks to the contact between delegations of various countries at the CSTC sessions and at meetings it has been possible to effectively coordinate research and to find optimal solutions for particular problems. The work of the specialists has become more goal-oriented. The results of the work on individual problems are discussed at meeting of specialists from the COMECON member-countries.The regular information prepared by the USSR on the operation of the BOR-60 and BN-350 reactors and on the development of the BN-600 is of great importance.At its twelfth meeting, the CSTC established the subject on which efforts should be concentrated: the choice and substantiation of the principal technical solutions and parameters for the BN-1600, and the development of individual systems and equipment for this reactor, development and construction of reliable sodium-water steam generators, as well as instruments for monitoring the technological parameters of fast-neutron reactor plant.Translated from Atomnaya Énergiya, Vol. 44, No. 6, pp. 533–534, June, 1978.     

Prospects for the BN-1800 Sodium-Cooled Fast Reactor Satisfying 21st Century Nuclear Power Requirements

The BN-1800 power-generating unit is designed to meet the requirements of the strategy for developing atomic energy in Russia in the first half of the 21st century. The development time is the next 15 years and construction could start after 2020. The design is innovative and includes the development of key new technical solutions as compared with the BN-800 reactor which is now under construction.The new technical solutions are based on the substantial positive experience in operating fast reactors in Russia (~125 reactor·years), specifically the BN-600 reactor. The innovations make it possible not only to solve strategic problems, such as increasing safety, improving ecology (including by burning actinides), and nonproliferation but also to make large improvements in economic performance.     

Problems of BN-800 construction and the possibilities of developing advanced fast reactors

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. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 21–26, January, 2007.     

Construction,operation, and operating lifetime extension of the steam generators of a BN-600 power-generating unit

The construction of the PGN-200M steam generator of a BN-600 power-generating unit at the Beloyarskaya nuclear power plant is described. Data from 25 years of operation are presented and the basic questions which were solved during the startup-adjustment work, which increased the operational reliability of the steam generator, are elucidated. The advantages of steam generators with different designs are compared, and it is concluded that it is desirable to develop a high-power sodium-cooled vessel steam generator for future facilities. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 481–488, December, 2005.     

30 Years of experience in operating the BN-600 sodium-cooled fast reactor

Experience in operating the BN-600 sodium-cooled fast reactor during its nominal service life as well as its service life extension period, an additional 15 years, is described. Information is presented on the performance indicators which were achieved and deviations from the normal operating regime which occurred when the reactor was first started up. The degree to which they affect the safety and technical-economic performance of the facility is evaluated. It is concluded on the basis of an analysis of the BN-600 operating experience that sodium-cooled fast reactors have now been mastered commercially and that their prospects for further development are good.     

秦山二期长循环燃料管理方案研究

秦山二期目前采用的是使用36盒富集度为3.7%的换料组件的OUT-IN年换料燃料管理策略,从2号机组第11燃料循环开始第1个长循环过渡,采用的是使用44盒富集度为4.45%的换料组件的IN-OUT长循环燃料管理策略。为解决目前长循环燃料管理策略循环长度偏短、预留停堆维修时间较长、电站可利用率较低等问题,对使用48盒富集度为4.45%的换料组件的IN-OUT长循环燃料管理策略进行了研究。通过对上述3种燃料管理策略的主要计算结果进行分析和比较,推荐秦山二期长循环燃料管理采用48盒富集度为4.45%的燃料管理策略。     

Operation experience of the BN-600 fast reactor

This paper presents the main operating experience of the BN-600 Power Unit, which has been under operation since 1980 as a part of the Beloyarsk Nuclear Power Plant situated in Zarechny town, Sverdlovsk Region, Russia.Theperformance of the core, the reactor, the fuel handling systems, the steam generator unit, the sodium and steam circuits, and the electrical equipment is highlighted. The technological procedures and the events which occurred, including sodium leaks, are also considered.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.