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.     

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.     

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.     

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.     

PGV-1500 steam generator. New problems and solutions

The basic questions concerning the development of a steam generator for a nuclear power plant with a VVé R-1500 reactor are presented. The basic design requirements which follow for steam generators from experience in operating analogs at nuclear power plants and taking account of the requirements for a reactor system are presented. The special features inherent to horizontal-type steam generators, which have been mastered and are used in nuclear power plants in our country, are noted. The domestic and world operating experience is taken into account in the development of the design. It is concluded that the design of the PGV-1500 steam generator satisfies the requirements for the concept of a VVéR reactor facility for a 1500 MW(e) unit of a nuclear power plant and is competitive on the world market for power-generating equipment for nuclear power plants. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 416–425, December, 2005. An erratum to this article is availabel at .     

VBÉR-300 reactor system and power-generating units for regional nuclear power production

The basic design solutions and characteristics of the VBéR-300 reactor system for the power-generating units of 150–300 MW(e) nuclear power plants and regional nuclear heat-and-electricity plants are described. The reactor system implemented as a unit is based on the technologies and solutions used for marine nuclear power systems, which have been corroborated by experience in operating nuclear-powered icebreakers. The technical-economic advantages of floating power-generating units are substantiated. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 35–39, January, 2007.     

Experience in operating and upgrading the No. 5 unit of the Novovoronezh nuclear power plant – practical base for developing a reliable source of nuclear energy

The No. 5 unit of the Novovoronezh nuclear power plant, starting commercial operations on September 26, 1980, is the first power-generating unit with a 1000 MW VVER in our country. The assimilation of its power gave invaluable experience to designers, builders, and equipment manufacturers; this experience was taken into account in the design solutions for next-generation power-generating units. A large volume of work on increasing the efficiency, reliability, and safety was performed over a 30-year service life. At present, the power-generating unit has been shut down for a major overhaul for upgrading according a program for extending the service life by 25–30 years.     

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.     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。     

Activities for 4S USNRC licensing

4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper.     

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study.     

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.     

Radiation characteristics of reactor structures after the final shutdown of a nuclear power plant with VVER

The results of computational studies of the radiation characteristics and mass of the structures in the VVER-1200 reactor facility in the AES-2006 design with respect to the activity group as a function of the holding period after final shutdown of the power-generating unit are presented. The three-dimensional program KATRIN was used to obtain the radiation characteristics. It is shown that after final shutdown all metal structures in VVER-1200 will remain radioactive wastes for up to 150 years and the concrete structures after a holding period of more than 60 years will remain in the radioactive materials category. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 56–59, January, 2009.     

Direct-Flow Channel Reactor with Supercritical Coolant Pressure

The concept of a direct-flow channel reactor with supercritical-pressure water (CR-SCP) is presented. Neutron-physics, thermohydraulic, and strength calculations are used to substantiate the fundamental core design with a heavy-metal moderator which at supercritical pressure is competitive with other modern reactor designs with respect to fuel-cycle indicators. Two types of fuel-element and fuel-channel structures are examined. It is shown that fuel elements based on micropellets and a metal matrix are highly reliable and have higher operating characteristics (burnup, service life, geometric stability, and so on) than fuel elements with uranium-dioxide fuel. A CR-SCP design and the technological scheme of a power-generating unit are presented, and the systems which are required to ensure normal operation and safety are determined. The main technical-economic indicators of a power-generating unit with installed electric power 850 MW are estimated.Journal variant of a report presented at the International Scientific and Technical Conference on Channel Reactors: Problems and Solutions, October 2004, N. A. Dollezhal’ Scientific-Research and Design Institute of Power Engineering, Kursk Nuclear Power Plant.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 243–53, April, 2005.     

Composition and structure of simulation models for evaluating decommissioning costs for nuclear power plant units

The methodological and practical approaches to realizing an iterative multiple-criterion analysis for evaluating the decommissioning costs of the power-generating units of nuclear power plants and the optimal structure of the analysis using information technologies are examined. The objective prerequisites which have been established and facilitate the development and use of decommissioning simulation models in practical work are analyzed.     

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.     

Economic effect of the development and operation of serially produced propulsion nuclear power systems

The possibility of decreasing the capital cost of building a nuclear power plant by unifying the equipment and technological processes is examined for the example of the experience in building and operating propulsion nuclear power-generating units. It is shown that it is desirable to adopt for nuclear power plants the most effective solutions and organizational-technical, and technological approaches which have been implemented in the development of propulsion nuclear power systems. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, 39–43, January, 2007.     

Fast Reactors. Status and Prospects

The stages of the development of fast reactors in the world are analyzed. It is shown that substantial progress has been made in the development and operation of sodium-cooled fast reactors and accident-free operation of the main liquid-metal equipment, equal to the performance of general industrial equipment. Ways to make nuclear-power-plant units of this type competitive are discussed.The status of the work on fast reactors with other coolants – gas, steam, and heavy metals – is briefly reviewed. The main problems which must be solved to implement these directions of the development of fast reactors are indicated.