Predicting growth of deposits on fuel assemblies of VVER-440 reactors

We discuss issues of simulating the growth of deposits on the surface of the core of VVER-440 reactors, which result in a higher pressure difference of coolant across the reactor and in the need to reduce the reactor power. We also present results of model calculations carried out for one fuel campaign of Unit 4 at the Novovoronezh nuclear power station.     

An Analysis of the Behavior of Zinc Compounds in the Primary Coolant Circuit at a Nuclear Power Station with VVER Power Reactors

Technology of zinc metering into the coolant has already been used for more than 20 years at foreign nuclear power stations (NPS) with PWR reactors for reducing radiation fields on the equipment and suppressing corrosion cracking of nickel-rich alloys. The most likely forms in which zinc compounds can exist in the primary coolant circuit of water-cooled power reactors (such as VVER, PWR, or BWR) were assessed, and the data on their solubility were analyzed. It is demonstrated that zinc oxides and silicates feature the minimum solubility under the primary coolant circuit conditions. The conditions for crystallization of compounds on the surface of fuel rod cladding in the core in case of local boiling (i.e., subcooled liquid boiling) were analyzed. A review is presented of foreign publications with the assessment of the risk that zinc compounds are contained in zinc deposits on the surface of fuel rods when zinc is metered into the coolant of PWR power units with a thermally stressed core. The predictions are presented of the limit thickness of the deposits at which zinc oxides and silicates can precipitate into them in case of local subcooled boiling on the fuel rod surface. The predictions are presented for the effect of the concentration of boric acid, silica, and zinc in the coolant on the limit thickness of depositions at which zinc compounds can crystalize. It is demonstrated that, in the presence of boric acid, zinc can interact with borates with the formation of borate complexes, which reduce the risk of deposition of zinc silicates in the layer of deposits on the fuel rod surface. The calculated results confirm that increasing the thickness of oxide film on fuel rods increases the risk of crystallization of zinc compounds in the core.     

Some features of the effect the pH value and the physicochemical properties of boric acid have on mass transfer in a VVER reactor’s core

Certain features of the effect of boric acid in the reactor coolant of nuclear power installations equipped with a VVER-440 reactor on mass transfer in the reactor core are considered. It is determined that formation of boric acid polyborate complexes begins under field conditions at a temperature of 300°C when the boric acid concentration is equal to around 0.065 mol/L (4 g/L). Operations for decontaminating the reactor coolant system entail a growth of corrosion product concentration in the coolant, which gives rise to formation of iron borates in the zones where subcooled boiling of coolant takes place and to the effect of axial offset anomalies. A model for simulating variation of pressure drop in a VVER-440 reactor’s core that has invariable parameters during the entire fuel campaign is developed by additionally taking into account the concentrations of boric acid polyborate complexes and the quantity of corrosion products (Fe, Ni) represented by the ratio of their solubilities.     

Predicting the growth of deposits on the fuel rods in a VK-50 reactor

A newly developed procedure for predicting the growth of deposits on the fuel rods of a water-cooled water-moderated reactor under coolant boiling conditions is described. The results obtained from an experimental validation of the procedure carried out for fuel assemblies at different stages of their operation in a vessel-type boiling-water reactor are presented. It is shown on the basis of experimental data that the deposits forming on the fuel rods in a boiling-water reactor consist mainly of copper and iron. Copper exists in dissolved form and precipitates in pores between the particles of iron compounds, and the thickness of deposits is determined by the particles of iron corrosion products themselves. The corrosion products incipience and growth processes were investigated, and the effect of deposit formation from fine iron particles on fuel rod claddings operating under coolant boiling conditions was predicted theoretically and revealed experimentally. Relatively large particles moving along a fuel rod cannot penetrate into the laminar sublayer due to the effect of Magnus force on them. Based on the results of theoretical and experimental investigations, recommendations on decreasing the content of iron corrosion product particles in transient operating modes of boiling water reactors are worked out. The method of very fast decrease of pressure at low levels of reactor power worked out on a VK-50 reactor makes it possible to remove relatively large particles of shutdown corrosion products to the coolant purification system while keeping them from depositing on the fuel rods. With the use of this routine operation, matters concerned with radiation safety and durability of fuel assemblies in boiling light-water reactors are solved in a more efficient manner.     

Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.     

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.     

The influence of operational and water chemistry parameters on the deposits of corrosion products on fuel assemblies at nuclear power plants with VVER reactors

The phenomenon involving a growth of pressure drop in the reactor core and redistribution of deposits in the reactor core and primary coolant circuit of a nuclear power station equipped with VVER-440 reactors is considered. A model is developed, the physicochemical foundation of which is based on the dependence of corrosion product transfer on the temperature and pH _t value of coolant and on the correlation between the formation rate of corrosion products (Fe) (after subjecting the steam generators to decontamination) and rate with which they are removed from the circuit. The purpose of the simulation carried on the model is to predict the growth of pressure drop on the basis of field data obtained from nuclear power installations and correct the water chemistry (by adjusting the concentrations of KOH, H₂, and NH₃) so as to keep the pressure drop in the reactor at a stable level.     

The influence of the high-temperature coolant purification system at nuclear power stations equipped with VVER-1000 reactors on the formation of dose burdens

It is shown that the effectiveness of using high-temperature filters for purifying the coolant at nuclear power stations equipped with VVER-1000 reactors is mainly determined by the precipitation constant of activated corrosion products dispersed in the coolant.     

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.     

A balance procedure for calculating the model fuel assemblies reflooding during design basis accident and its verification on PARAMETER test facility

A balance procedure is proposed for estimating the main parameters characterizing the process of model fuel assemblies reflooding of a VVER reactor made on different scales under the conditions of a design basis accident by subjecting them to bottom reflooding¹. The proposed procedure satisfactorily describes the experimental data obtained on PARAMETER test facility in the temperature range up to 1200°C. The times of fuel assemblies quenching by bottom reflooding calculated using the proposed procedure are in satisfactory agreement with the experimental data obtained on model fuel assemblies of VVER- and PWR-type reactors and can be used in developing measures aimed at enhancing the safety of nuclear power stations.     

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.     

A comprehensive approach to selecting the water chemistry of the secondary coolant circuit in the projects of nuclear power stations equipped with VVER-1200 reactors

The paper presents the results obtained from studies on selecting the water chemistry of the secondary coolant circuit carried out for the project of a nuclear power station equipped with a new-generation VVER-1200 reactor on the basis of case calculations and an analysis of field experience gained at operating nuclear power stations.     

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.     

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.     

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.     

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.     

Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit’s purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.     

The capacities of a corrosion-fatigue monitoring system in managing the service life of steam generators operating at nuclear power stations equipped with a VVER-1000M reactor

We consider the possibilities of using a corrosion-fatigue monitoring system for operatively monitoring the corrosion of heat-transfer tubes in steam generators operating at nuclear power stations equipped with VVER reactors. It is proposed that this system be tested on a power unit built around a VVER-1000 reactor.     

Measures for ensuring reliable operation of the welded joint connecting the reactor coolant circuit’s header to the shell of a steam generator used at a VVER-1000 reactor-based nuclear power station

Problems that arose around the weld joint connecting the reactor coolant circuit’s header to the steam generator shell during operation of steam generators at nuclear power stations equipped with VVER-1000 reactors are considered. Works on studying the defects occurred in the header’s metal are described, and ways for preventing their development are determined.     

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.