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.     

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.     

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.     

Layout and structural solutions for the reloading system of an advanced fast reactor

The particulars of fuel reloading in fast reactors are described. General data on a BN-1200 power-generating unit, the principles of the construction of its reloading system, the purpose and construction of individual types of reloading equipment, and required experimental work are presented. The comparative characteristics of the BN-800 and BN-1200 reloading equipment are presented.     

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.     

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.     

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.     

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.     

Plutonium use in nuclear power in Russia

Conclusions The following concept of plutonium utilization based on the evolutionary development of the traditional technology in our country arises: The main problem of any short-term program of dealing with plutonium must be solved — reliable and safe storage of separated energy plutonium and freed weapons plutonium before utilization in reactors. Plutonium (mainly energy plutonium) is utilized primarily in BN-800 fast reactors and the development of technology using weapons plutonium in BN-600 reactors starts. In the future attention should be focused on nuclear-power centers patterned after the Industrial Association “Mayak” (RT-1 plant, Complex 300, BN-800) with reliable nonproliferation of weapons plutonium. It is extremely important to speed up work on the completion of Complex 300: This work must be completed before BN-800 is ready. In the future efforts must be concentrated on the following: development and implementation, in BN-800, of an economically more efficient plutonium-burning core; the possibility of building light-water reactors with the required degree of safety for effective plutonium utilization must be justified (including a “cold” core based on cermet fuel); and, development and implementation of technology for a safe and an ecologically acceptable closed nuclear fuel cycle based on plutonium and²³³U with burnout of Am, Np, and Cm. Ministry of Atomic Energy of the Russian Federation. Institute of Physics and Power Engineering. A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Building. Special Design Office for Machines. Translated from Atomnaya énergiya, Vol. 76, No. 4, pp. 326–332, April, 1994.     

Behavior of vibrationally compacted mixed fuel in BN-800

The modified computer code MACROS was used to study the behavior of fuel elements with vibrationally compacted mixed fuel under irradiation in BN-800. Verification calculations were performed for fuel elements irradiated in BN-600 under close to nominal conditions. Attention was focused mainly on the factors playing an important role in their serviceability: swelling, creep, corrosion behavior of cladding, temperature distribution, gas-release, and degree of structural reformation of the fuel core.     

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.     

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.     

Control Principles for Reactors with Regulatable Coolant Circulation

It is shown that the design of the reactor control scheme used in BN-600-type power-producing systems does not provide for the operation of a nuclear power plant in the base and peak regimes. A design of the control scheme for a power system with regulatable circulation of the coolant in the first loop of the reactor that allows a nuclear power plant to operate in the base and peak regimes are presented and substantiated.     

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.     

基于示范快堆的ADS次临界快堆堆芯研究

选取中国示范快堆作为次临界快堆参考堆芯,研究次临界快堆作为嬗变PWR(U)乏燃料中次锕系元素的可行性。中国示范快堆堆芯设计是参考目前正在建设的俄罗斯示范快堆BN-800。次临界快堆堆芯在示范快堆堆芯基础上去掉中间7盒组件放置铅靶组件,控制棒组件用含贫铀和次锕系元素(MA)的组件代替,转换区组件用反射层组件代替。采用MCNPX和ORIGEN2程序作为计算软件。计算结果表明:次临界快堆中加入MA后能够保持一定的次临界度且具有较好的嬗变效果,因此,选取示范快堆堆芯作为ADS次临界快堆的参考堆芯研究是可行的。     

Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors

When designing new fast reactors, it is desirable to increase as much as possible the breeding occurring in the core in order to ensure the minimum excess reactivity for burnup on the one hand and a closed fuel cycle without replenishment with external plutonium and without separating plutonium from uranium during chemical reprocessing of irradiated fuel on the other. The latter requirement greatly decreases the risk of plutonium proliferation in such a fuel cycle. This requires a core breeding ratio 1.05–1.08. Such values can be achieved by using technologically perfected and tested oxide fuel with its volume fraction in the core increased to 55–60%. The results of computational-theoretical studies on the selection and optimization of cores with high fuel fractions for BN-1600 and BN-800 reactors are presented in this article. It is shown that such cores can be built in principle.