Preparations for shifting the power units of nuclear power stations equipped with RBMK-1000 reactors for operation with a 2 year interval between repairs

A general approach for carrying out works on justifying the shifting of power units used at nuclear power stations equipped with RBMK-1000 reactors for operation with an increased interval between repairs is formulated. The technical and organizational measures ensuring reliable operation of equipment and pipelines and acceptable safety of power units at nuclear power stations equipped with RBMK-1000 reactors in the new schedule of operation are described.     

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6–12 to 2.0–2.5 μg/dm³. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains рН₂₅ of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm³ in the steam generator feedwater.     

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World’s First nuclear power plant]

Successful commissioning in the 1954 of the World’s First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.     

Improving the transient stability of VVER-1000 reactor-based nuclear power plant units

Ensuring transient stability of nuclear power plant units in transient modes of their operation is one of the goals aimed at achieving enhanced safety and reliability of nuclear power plants. Field experience shows that for nuclear power plants equipped with VVER-1000 reactors, matters relating to transient stability in modes involving disconnection of the power unit main equipment, such as reactor coolant pumps, turbine-driven feedwater pumps, and turbine generator, are of most concern. Specialists of the Institute for Nuclear Power Plant Research perform comprehensive investigations of the technological processes aimed at working out measures for achieving better transient stability of nuclear power plant units. The article presents the results of joint activities carried out by specialists of the Institute for Nuclear Power Plant Research and the Novovoronezh nuclear power plant on enhancing the transient stability of Unit 5 at the Novovoronezh nuclear power plant.     

The alternative strategies of the development of the nuclear power industry in the 21st century

This paper emphasizes the urgency of scientific-and-technical and sociopolitical problems of the modern nuclear power industry without solving of which the transition from local nuclear power systems now in operation to a large-scale nuclear power industry would be impossible. The existing concepts of the longterm strategy of the development of the nuclear power industry have been analyzed. On the basis of the scenarios having been developed it was shown that the most promising alternative is the orientation towards the closed nuclear fuel cycle with fast neutron reactors (hereinafter referred to as fast reactors) that would meet the requirements on the acceptable safety. It was concluded that the main provisions of “The Strategy of the Development of the Nuclear Power Industry of Russia for the First Half of the 21st Century” approved by the Government of the Russian Federation in the year 2000 remain the same at present as well, although they require to be elaborated with due regard for new realities in the market for fossil fuels, the state of both the Russian and the world economy, as well as tightening of requirements related to safe operation of nuclear power stations (NPSs) (for example, after the severe accident at the Fukushima nuclear power station, Japan) and nonproliferation of nuclear weapons.     

Thermal and nuclear power plants: Competitiveness in the new economic conditions

In recent years, the conditions of development and functionality of power generating assets have notably changed. Considering the decline in the price of hydrocarbon fuel on the global market, the efficiency of combined-cycle gas-turbine plants in the European part of Russia is growing in comparison with nuclear power plants. Capital investments in the construction of nuclear power plants have also increased as a result of stiffening the safety requirements. In view of this, there has been an increasing interest in exploration of effective lines of development of generating assets in the European part of Russia, taking consideration of the conditions that may arise in the nearest long-term perspective. In particular, the assessment of comparative efficiency of developing combined-cycle gas-turbine plants (operating on natural gas) in the European part of Russia and nuclear power plants is of academic and practical interest. In this article, we analyze the trends of changes in the regional price of hydrocarbon fuel. Using the prognosis of net-weighted import prices of natural gas in Western European countries—prepared by the International Energy Agency (IEA) and the Energy Research Institute of the Russian Academy of Sciences (ERIRAS)—the prices of natural gas in the European part of Russia equilibrated with import prices of this heat carrier in Western Europe were determined. The methodology of determining the comparative efficiency of combined-cycle gas turbine plants (CCGT) and nuclear power plants (NPP) were described; based on this, the possible development of basic CCGTs and NPPs with regard to the European part of Russia for various scenarios in the prognosis of prices of gaseous fuel in a broad range of change of specific investments in the given generating sources were assessed, and the extents of their comparative efficiency were shown. It was proven that, at specific investments in the construction of new NPPs in the amount of 5000 dollars/kW, nuclear power plants in the European part of Russia become less efficient as compared to CCGTs operating on natural gas.     

Developing and applying modern methods of leakage monitoring and state estimation of fuel at the Novovoronezh nuclear power plant

The results of developing and implementing the modernized fuel leakage monitoring methods at the shut-down and running reactor of the Novovoronezh nuclear power plant (NPP) are presented. An automated computerized expert system integrated with an in-core monitoring system (ICMS) and installed at the Novovoronezh NPP unit no. 5 is described. If leaky fuel elements appear in the core, the system allows one to perform on-line assessment of the parameters of leaky fuel assemblies (FAs). The computer expert system units designed for optimizing the operating regimes and enhancing the fuel usage efficiency at the Novovoronezh NPP unit no. 5 are now being developed.     

Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.     

液力驱动泵在汽水分离再热系统中的应用

田湾核电站一期工程是由俄罗斯设计、供货的2台NPP91/WWER-1000型机组。电站汽轮机汽水分离再热系统中的加热蒸汽凝结水疏水泵,采用1台液力驱动泵作为疏水泵。鉴于此类型的疏水泵在国内核电站二回路系统中首次采用,简要介绍了田湾核电站汽水分离再热系统流程设计特点,液力驱动泵的结构和运行原理,以及该泵在调试期间可能产生的问题和解决方法。     

The problem of optimizing the water chemistry used in the primary coolant circuit of a nuclear power station equipped with VVER reactors under the conditions of longer fuel cycle campaigns and increased capacity of power units

It is shown that the optimal water chemistry of the primary coolant circuit must be substantiated while introducing measures aimed at increasing the power output in operating power units and for the project called AES-2006/AES TOI (a typical optimized project of a nuclear power station with enhanced information support). The experience gained from operation of PWR reactors with an elongated fuel cycle at an increased level of power is analyzed. Conditions under which boron compounds are locally concentrated on the fuel rod surfaces (the hideout phenomenon) and axial offset anomaly occurs are enlisted, and the influence of lithium on the hideout in the pores of deposits on the surfaces of fuel assemblies is shown.     

Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.     

Achieving more reliable operation of turbine generators at nuclear power plants by improving the water chemistry of the generator stator cooling system

Ways of improving the water chemistry used in the turbine generator stator’s cooling systems at Russian nuclear power plants are considered. Data obtained from operational chemical monitoring of indicators characterizing the quality of cooling water in the turbine generator stator cooling systems of operating power units at nuclear power plants are presented.     

The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of ⁵⁸Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.     

The current state,main problems and directions in improving water chemistry at NPSs

An analysis of the current state of managing water-chemistry (WC) at Russian nuclear power plants with type-VVER and-RBMK reactors presently in operation is presented. The main directions for improvement of WC are shown.     

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium–plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.     

Experience of development and implementation of new electrohydraulic turbine control systems of the Leningrad Metal Plant for nuclear power plants

An experience of development and implementation of electrohydraulic control systems with the application of microprocessor systems is considered by the example of a K-1000-60/3000 turbine of plants nos. 3 and 4 of the Kalininskaya nuclear power plant (NPP). The long-term operation of the electrohydraulic control system on power generation unit no. 3 of the Kalininskaya NPP, which encloses the electronic part of the control system (EPCS), confirmed wide possibilities of the system. Modern requirements to the power control of power generating units were a basis for the development of the control system with the extended functions of the microprocessor part and with the individual control of control valves with the use of a hydroelectric drive. The system was developed and implemented on the K-1000-60/3000 turbine of plant no. 4. Test results for power generating unit no. 4 allowed one to make a decision of putting into commercial operation of the mode of the total primary control of frequency (TPCF). For the first time, the possibility of the operation of the power generating unit of the NPP with the water-moderated water-cooled power reactor (WWPR) was tested and proved in the mode of the normalized primary control of frequency (NPCF) with the power errors up to ±2% of nominal without the displacement of control rods of the reactor only due to the use of the self-control effect of the reactor plant.     

MODE-A模式核电机组参与调峰适应性分析及建议

就秦山第二核电厂机组MODE-A固有特性,不适当的调峰方式会不同程度引起放射性污水排放超标、核燃料包壳应力增加、堆芯寿期末硼浓度难以控制、局部反应堆功率瞬态、核燃料废弃及经济能损严重等症状,削弱了该系列反应堆运行安全裕度,极易诱发人因事件。针对这些非安全因素,结合区域电网电源分布结构,参照国外核电机组调峰现状及经验反馈,提供可参考的建议。     

再论山东海阳核电厂厂区标高和厂区护堤设计标准的确定

山东海阳核电厂规划建设容量为6×1000MWe级核电机组,并留有再扩建两台的可能性。厂区一次规划,分期建设。一期工程建设规模为2×1000MWe级核电机组。核电厂厂区标高和厂区护堤设计标准确定的合理与否,不仅可以节省大量资金,而且将直接影响到整个核电厂的工期,因此应该引起各方面重视。     

Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

One of the main objectives of severe accident management at a nuclear power plant is to protect the integrity of the containment, for which the most serious threat is possible ignition of the generated hydrogen. There should be a monitoring system providing information support of NPP personnel, ensuring data on the current state of a containment gaseous environment and trends in its composition changes. Monitoring systems’ requisite characteristics definition issues are considered by the example of a particular power unit. Major characteristics important for proper information support are discussed. Some features of progression of severe accident scenarios at considered power unit are described and a possible influence of the hydrogen concentration monitoring system performance on the information support reliability in a severe accident is analyzed. The analysis results show that the following technical characteristics of the combustible gas monitoring systems are important for the proper information support of NPP personnel in the event of a severe accident at a nuclear power plant: measured parameters, measuring ranges and errors, update rate, minimum detectable concentration of combustible gas, monitoring reference points, environmental qualification parameters of the system components. For NPP power units with WWER-440/270 (230) type reactors, which have a relatively small containment volume, the update period for measurement results is a critical characteristic of the containment combustible gas monitoring system, and the choice of monitoring reference points should be focused not so much on the definition of places of possible hydrogen pockets but rather on the definition of places of a possible combustible mixture formation. It may be necessary for the above-mentioned power units to include in the emergency operating procedures measures aimed at a timely heat removal reduction from the containment environment if there are signs of a severe accident phase approaching to prevent a combustible mixture formation in the containment.     

Substantiation of the feasibility of using fuel rods with tungsten spraying in power fast reactors of new generation

The contemporary development of nuclear power technologies in Russia made it possible to create projects of economic and safe fast reactors of new generation. These reactors will be a basis of the large-scale nuclear power engineering in the middle and end of the 21st century. Fast reactors of inherent safety (BREST), in which heavy emergencies are deterministically excluded [1,2], are of most interest among such projects. However, the limits of domestic power engineering implemented in BREST projects are not completely reached; there are reserves for further bettering of the safety and, probably, efficiency of new generation reactors. There are 1 to 2 decades left for improving technologies of fast reactors of inherent safety. One of the BREST concept reserves—the usage of fuel-rod shells with tungsten spraying—is the idea unattractive and unrealistic at first sight due to the tungsten’s high cost and the large cross section of fast-neutron absorption. However, the performed analysis and the calculation studies make it possible to draw a conclusion on the potential possibility of using the tungsten coatings of fuel rods for further improvement of reliability and safety of BREST-type reactors without deterioration (and probably with improvement) of economical characteristics of nuclear power plants with these reactors.