Water chemistry in power units at nuclear power stations equipped with RBMK-1000 reactors

The modified standards for the water chemistry at nuclear power stations equipped with RBMK-1000 reactors for the entire service life of power units are presented. Stages through which the information-analytical system Center for Providing Chemical Support to Nuclear Power Stations was established and developed and its current structure are described. An example of analyzing the water chemistry of coolant used in the multiple forced circulation loop and operational data obtained through the communication channels with the above-mentioned system is given, and the main objectives pursued by the Center for Providing Chemical Support to nuclear power stations equipped with RBMK reactors are described in the part of conducting their water chemistry and using means and systems for maintaining water chemistry parameters.     

Erosion-corrosion wear of the components of pipeline systems at nuclear power stations equipped with RBMK-1000 reactors

The erosion-corrosion wear of components of pipeline systems at nuclear power stations equipped with RBMK-1000 reactors is analyzed. It is shown that the mechanism of erosion-corrosion wear involves two parallel processes: thinning of pipeline walls in some sections and sedimentation of corrosion products in the other sections.     

Increasing the rated capacity of power units installed at Russian nuclear power stations equipped with VVER-1000 reactors

Safe operation of the Balakovo nuclear power station’s Unit 2 built around a VVER-1000 reactor at a thermal power output of 3120 MW with meeting of the safety criteria and compliance with the requirements of existing regulatory documents is substantiated. Results from measurements of process parameters at a power output equal to 104% of its nominal value are presented.     

Preventing inadmissible erosion-corrosion thinning from occurring in the diffuser segments of feedwater supply control systems of power units at nuclear power stations equipped with RBMK-1000 reactors

Results obtained from investigations of erosion-corrosion processes that occur during operation of the feedwater supply control systems used in power units of nuclear power stations equipped with RBMK-1000 reactors and the sensitivity of these processes to variations in the chemical composition of metal and in the flow path geometry are presented. It is found that local erosion-corrosion thinning of the walls in the diffuser segments of feedwater supply control systems occur mainly due to intense mass transfer in the near-wall region taken in combination with a low content of chromium. Hydrodynamic simulation was carried out, and it was shown based on its results that local erosion-corrosion thinning of the walls of pipeline segments downstream of the valves controlling the supply of feedwater to power units of nuclear power stations equipped with RBMK-1000 reactors can be prevented by subjecting them to appropriate modernization. It is found that the above mentioned diffuser parts can be made more resistant to erosion-corrosion wear by keeping the content of chromium in the main metal and weld joints at a level of no less than 0.25% and concurrently reducing the hydrodynamic effect in the zones of weld connections.     

Prospects for development of nuclear power stations equipped with VVER reactors

The prospects for further development in Russia of nuclear stations equipped with water-cooled water-moderated reactors are considered.     

Assessing the influence of uncertain factors in analyzing emergency processes at nuclear power stations equipped with VVER-1000 reactors

We present the results of taking into account, by means of a newly developed procedure, uncertainty factors in a simulation of the emergency process for a VVER-1000 reactor installation during the accident involving a small leak and failure of the pumps of the high-pressure emergency core cooling system.     

An investigation of the effect of hydrodynamics of flow on erosion-corrosion resistance of components of a feedwater control assembly used in power units of nuclear power plants equipped with RBMK-1000 reactors

The results of computational hydrodynamic simulation of the flow of a working medium in the flow path of the feedwater control assembly used in power units of nuclear power plants (NPPs) equipped with the reactors of the RBMK-1000 type are presented It was established that the rate of control valve opening has an influence on the position of the areas of the intense local erosion-corrosion thinning of metal of the diffuser section downstream of the valve.     

Lines of improving the economic efficiency of nuclear power stations equipped with VVER reactors

The economic efficiency of measures on optimizing technical solutions for the equipment of nuclear power stations—specifically, on improving turbine units—is shown.     

Investigation of thermohydraulic processes in steam generators for nuclear power stations equipped with VVER reactors

The results obtained from experimental investigations and mathematical simulation of horizontal steam generators are considered. Recommendations for continuing these works are given.     

Problems connected with polishing turbine condensate of power units at nuclear power stations equipped with VVER reactors and ways of improving its quality

Experience gained from operation of the turbine condensate polishing system used at existing nuclear power stations equipped with VVER reactors is analyzed. The optimal scheme of polishing turbine condensate intended for use at newly designed nuclear power stations is substantiated.     

Specific features of corrosion damage to heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors

Specific features of corrosion damage occurring to the heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors are considered. The results obtained from metallographic studies of flaws found in samples cut out from steam-generator tubes are analyzed. Regularities with which flaws of steam-generator tubes are distributed over the tube bundle volume are discussed. Approaches for assessing the technical state and remaining service life of steam-generator tubes are presented.     

Introduction of water chemistry conditions of the secondary coolant circuit with metering organic amines at nuclear power stations equipped with VVER-1000 reactors

Results from introduction of new water chemistry conditions involving metering of organic amines (morpholine and ethanolamine) at nuclear power stations equipped with VVER-1000 reactors are presented.     

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.     

Developing the concept of maintenance and repairs in projects of power units for new-generation nuclear power stations

Results from conceptual elaboration of individual requirements for the system of maintenance and repairs that must be implemented in the projects of new-generation nuclear power stations are presented taking as an example the power unit project for a nuclear power station equipped with a standard optimized VVER reactor with enhanced information support (the so-called VVER TOI reactor). Implementation of these concepts will help to achieve competitiveness of such nuclear power stations in the domestic and international markets.     

Field examination of the erosion-corrosion wear of pipelines at Russian nuclear power stations equipped with VVER reactors

Distributions of wall thicknesses differing from their nominal values are analyzed, and factors facilitating the formation of thicknesses equal to and exceeding their nominal values are considered. It is shown that the nominal and mean thickness values calculated using the full volume of field examination data have to be used in calculating the rates of erosion-corrosion wear and predicting the service life of pipelines.     

Thermal-hydraulics, physical chemistry, and technology at nuclear power stations equipped with fast-neutron sodium-cooled reactors

Main results of investigations aimed at developing a verified system of computer codes that take into account the interrelation among nuclear-physical, thermal-hydraulic, physicochemical, thermal-mechanical, mass-transfer, and technological processes in nuclear power installations and at substantiating the models used as the core of these codes are presented together with the results of tests carried out to obtain data for verifying the codes. Original Russian Text ￠ V.V. Alekseev, A.D. Efanov, F.A. Kozlov, A.P. Sorokin, 2007, published in Teploenergetika.     

Consideration of field experience in developing new projects of steam generators for nuclear power stations equipped with VVER reactors

The main problems encountered during the operation of horizontal steam generators are considered. Design features of the new PGV-1000MK and PGV-1500 steam generators are analyzed.     

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.     

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.     

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.