Results of operation and current safety performance of nuclear facilities located in the Russian Federation

After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104–109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation “Rosatom”, should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn’t happen.The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP units is analyzed.     

New Technological Platform for the National Nuclear Energy Strategy Development

The paper considers the need to update the development strategy of Russia’s nuclear power industry and various approaches to the large-scale nuclear power development. Problems of making decisions on fast neutron reactors and closed nuclear fuel cycle (NFC) arrangement are discussed. The current state of the development of fast neutron reactors and closed NFC technologies in Russia is considered and major problems are highlighted.     

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.     

Prospects for development of VVER-type pressurized light-water reactor installations

The problems of development of the Russian nuclear power industry on the basis of advancement of VVER-type pressurized light-water reactors for power engineering are discussed. Characteristics of the VVER-1200 reactor installation for the NPS-2006 project are presented, and areas of further development of VVER-type reactors are considered.     

Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification of numerical codes, all examined configurations of the MHD flow are also investigated numerically.     

Improvement of the projects of reactor installations

Experience gained in the development of reactor installations (RIs) with VVER reactors and their operation as part of a nuclear power station is analyzed. It is shown that the newly designed projects of RIs that comply with modern requirements for safety and economic indicators are evolutionary in nature.     

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.     

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.     

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.     

Use of liquid metals in nuclear and thermonuclear engineering,and in other innovative technologies

By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.     

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.     

The SVBR-75/100 reactor installation. An assessment of the possibility of using it in projects for refurbishing power units of nuclear power stations equipped with VVER reactors

It is shown that use in the nuclear power industry of Russia of a nuclear technology employing modular multipurpose fast-neutron small-capacity reactors with lead-bismuth coolant will make it possible to considerably speed up a comprehensive solution to the problems faced by the industry, including those related to decommissioning the power units of nuclear power stations with VVER reactors after the reactor installation has worked through its service life.     

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.     

Coupling heat storage to nuclear reactors for variable electricity output with baseload reactor operation

Nuclear reactors produce heat and thus can couple to heat storage systems to provide dispacthable electricity while the reactor operates at full power. Six classes of heat storage technologies couple to light-water reactors with steam cycles. Firebrick Resistance-Heated Energy Storage (FIRES) converts low-price electricity into high-temperature stored heat for industry or power. FIRES and brick recuperators coupled to nuclear brayton power cycles may enable high-temperature reactors to buy electricity when prices are low and sell electricity at higher price.     

Decreasing the influence of arc feed currents and transient recovery voltages in single-phase conditions in EHV lines

One of the most efficient ways to solve the problem of reactive power compensation connected with restriction of unacceptable levels of voltage on extrahigh voltage (EHV) power-transmission lines is installation of additional shunting reactors (SR) in the EHV networks. Moreover, installation of shunting reactors leads to the occurrence of resonance overvoltages under single-phase operation conditions of power transmission (for example within the one-phase automatic reclosing time), as well as to the occurrence of deenergized current phase of the arc feed in the arc channel, which prevents fast deionization of another channel and increasing transient recovery voltage. In this paper, measures for decreasing the influence of arc feed currents and transient recovery voltages in single-phase conditions of EHV lines are considered.     

The results of the development of technology for using sodium as coolant for fast-neutron reactors

The main results obtained from research activities on the technology of sodium coolant as applied to nuclear power installations built around fast-neutron reactors are briefly described. Abnormal situations involving considerable contamination of sodium coolant that were observed in fast-neutron reactors are considered. Further investigations in this field are aimed at extending the service life of the BN-600 nuclear power installation and developing advanced nuclear power plants on the basis of fast-neutron reactors with increased parameters. Problems that have to be urgently solved for further development of the sodium usage technology are formulated.     

Comparison of the technical and economic parameters of different variants of the nuclear fuel cycle reactors of the nuclear power plants

The article presents the comparative economic analysis of the variants of implementation of the fuel cycles for fast and thermal reactors. Calculations are carried out for the formed reference conditions by the cost of processing stages of the fuel cycles for the foreign and expert Russian information sources. The comparative data on the resource supply, absolute and specific costs of the fuel cycle variants for the whole life cycle of the thermal power plant projects with the fast and thermal reactors are considered. The conclusions of the efficiency of the fuel cycle variants for the assumed reference conditions of the calculation are made.     

Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors

Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.     

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.