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.     

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.     

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.     

Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.     

Numerical simulation of fuel assemblies containing pebble fuel elements

A numerical calculation of the new fuel assembly for a VVER-1000 reactor containing fuel on the basis of pebble fuel elements is carried out. The reactor core pressure drop coefficients in the axial and radial directions are studied.     

Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.     

Combined numerical and experimental investigations of local hydrodynamics and coolant flow mass transfer in Kvadrat-type fuel assemblies of PWR reactors with mixing grids

Results of research works on studying local hydrodynamics and mass transfer for coolant flow in the characteristic zones of PWR reactor fuel assemblies in case of using belts of mixing spacer grids are presented. The investigations were carried out on an aerodynamic rig using the admixture diffusion method (the tracer-gas method). Certain specific features pertinent to coolant flow in the fuel rod bundles of Kvadrat-type fuel assemblies were revealed during the experiments. The obtained study results were included in the database for verifying computation fluid dynamics computer codes and detailed cell-wise calculations of reactor cores with Kvadrat-type fuel assemblies. The obtained results can also be used for more exact determination of local coolant flow hydrodynamic and mass transfer characteristics in assessing thermal reliability of PWR reactor cores.     

Calculations of heat flowrates to the VVER-440 reactor vessel during interaction of corium melt with the reactor vessel

Quasi-steady-state thermal conditions of the corium bath found on a reactor vessel bottom is considered. Using approximate heat transfer models proposed earlier, a conjugate problem is solved with taking into account radiation heat transfer in the space above the bath surface. The results of calculations of heat flow-rates from the corium bath to the VVER-440 reactor vessel are presented. It is shown that water delivery to the bath surface leads to an essential decrease in the heat flowrate towards the wall in the most dangerous area, where a layer of melted steel contacts the reactor vessel, and retains a possibility of keeping the melt inside the reactor vessel under external cooling.     

The concept of extending the service life of the VVER-440-based power units at the Novovoronezh nuclear power plant

Basic statements of the Concept of Extending the Service Life of the VVER-440-Based Power Units at the Novovoronezh NPP beyond 45 years are considered. This topic is raised in connection with the fact that that in December 2016 and in December 2017 the extended service lives of Units 3 and 4 at this NPP will expire. The adopted concept of repeatedly extending the service life of the Novovoronezh NPP Unit 4 implies fitting the power unit with additional reactor core cooling systems with a view to extend the (ultimate) design-basis accidents (which have hitherto been adopted to be a loss of coolant accident involving a leak of reactor coolant through a break with a nominal diameter of 100 mm) to a reactor coolant leak equivalent to rupture of the main reactor coolant pipeline. The modified Unit 4 will also use the safety systems of Unit 3 that is going to be decommissioned. Preliminary calculated assessments of the new design-basis accident scenario involving rupture of the reactor coolant pipeline in Unit 4 fitted with a new configuration of safety systems confirmed the correctness of the adopted concept of repeatedly extending the service life of Unit 4.     

Some methods for achieving more efficient performance of fuel assemblies

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.     

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.     

Analysis of operation of filters for post-accident decontamination of pressurized rooms of a nuclear power plants with a type VVER-440 reactor

Operation of filters of postaccident decontamination of pressurized rooms of a nuclear power plant with a type-VVER-440 reactor is analyzed. The distribution of radioactive nuclides over filter stages, the time variation of the thermal state of filter, and the characteristic features of the processes of sorption in the section of fine cleaning 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.     

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.     

A combined thermal and strength analysis of the behavior of a VVER-440 reactor’s vessel during an accident involving core melting

Results from parametric calculations of the thermal and strength behavior of a VVER-440 reactor’s vessel during a beyond-design-basis accident are presented. The ranges of gage pressure inside the vessel and the conditions of externally cooling the reactor with water are presented under which the corium that appears in the reactor vessel during a severe accident can be held there for a long time.     

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.     

Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGВ that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the “hot” header on the water level the “cold” end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.     

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.     

Thermohydraulic investigations of increasing the burnup of nuclear fuel in fast-neutron reactors

We present the results from experimental and calculated investigations of the effect the deformation of fuel assembly casings, fuel rod bundle, and single fuel rods in the cores of fast-neutron reactors has on their temperature conditions. It is shown that the distortion of a fuel assembly (FA) is determined to a considerable extent by temperature nonuniformities in it, which in the final analysis affects the burnup of nuclear fuel.     

Analysis of phase composition of corrosion-products deposits on the surfaces of a PG-440 steam-generator tube bundle by Mossbauer spectroscopy method

Possibilities of using Mossbauer spectroscopy for the analysis of the phase composition of corrosion deposits at the outer and inner surfaces of the tube bundle of a PG-440 steam generator of one of the Kola NPP power units are shown.