VVER-1000 Reactor Research Articles

Thermohydraulic Calculation of VVER and SCWR Reactor Channels

Thermohydraulic Calculation of VVER and SCWR Reactor Channels

Neutronic Performance of Vver-1200 Reactor with Advanced Fuel and Cladding Materials

Neutronic Performance of Vver-1200 Reactor with Advanced Fuel and Cladding Materials

Investigation of Small Break Loca with Failure of the Emergency Core Cooling System Accident Management in Vver-1000 Reactor Type

Investigation of Small Break Loca with Failure of the Emergency Core Cooling System Accident Management in Vver-1000 Reactor Type

Факторы и механизмы удлинения твэлов ВВЭР-1000 при термических испытаниях, моделирующих режимы сухого хранения

To prove safety of dry storage conditions, thermal tests of the VVER-1000 fuel rods were performed in electrically heated furnaces in helium gas environment under stationary conditions (they were held at Т = 380 °С for 468 days) and thermal cycling ones (48 temperature cycles in a temperature range of 20 to 380 °С) with duration of cycles varying from 1 to 10 days and the total time of thermal testing of 427 days). The effect of thermal testing on elongation of fuel rods, which were under operation in the VVER-1000 reactors for one year (burnup was ~ 20 (MW⋅days)/kgU) and for six years (burnup was ~ 70 (MW⋅days)/kgU), was studied. It was revealed that high-burnup fuel rods elongated less compared to low burnup fuel rods. The highest elongation values were obtained for the low-burnup fuel rods: it made up 6.0 mm under the stationary operating conditions and 9.5 mm under thermal cycling. Hoop strain of fuel rods increased insignificantly. As there was gas pressure under the cladding, it induced tensile stresses in it at a temperature of 380 °С and thus, stimulated thermal creep. Lower burnup fuel rods could preserve their fuel-cladding gap throughout the entire height of their fuel columns. A model of gas-filled tubular specimen with a textured cladding in which the effective axial stress is enhanced by additional tensile load due to the pressure from fuel column on the spring plunger and a greater recovery of yield strength in the longitudinal direction compared to the transverse one, is applicable to these fuel rods. For high-burnup fuel rods, this model is applicable to for that part of the cladding where it is not in contact with the fuel, i.e. for a shorter length compared to low-burnup fuel rods. Thus, the idea was proved according to which the main mechanism of fuel rod elongation is the thermal creep of the cladding induced by enhanced (due to anisotropic recovery of the yield strength) effective axial stresses resulted from the fuel gas and fuel column pressure on the spring plunger. Thermal cycling leads to greater elongation of fuel rods due to a higher creep rate of the cladding at the unsteady-state stage during short cycles.

Analysis of computer code and method used in thermal-hydraulic safety justification of VVER reactor plants

This article presents the analysis of the calculation procedure and computer code KLAST used in calculations of the control rod dynamic characteristics during safety justification of water-cooled water-moderated power reactor plants. The code allows accounting for pressure differentials as a function of time occurred under the design conditions on the reactor core and on the drive extension shaft as well change of coolant density in the core. The code can be used to calculate dynamic characteristics of the control and protection system of control rod of VVER-1000 reactor types under the design accident conditions with rupture of the drive housing and to calculate the control and protection system of control rod dynamic characteristics during drop and damping in case of reactor damage during design accident conditions with pipeline break. In calculation, the control and protection system of control rod dynamic characteristics are determined versus time.

Neural Network Model of Changes in the Nuclide Composition of VVER Reactor Fuel

Neural Network Model of Changes in the Nuclide Composition of VVER Reactor Fuel

Обоснование теплотехнической надежности ВВЭР-1200 с установленными в ТВС перемешивающими решетками

One of the most important issues when it comes to the safety of a nuclear power plant is the need to ensure it not only during operation, but also at the design stage. To accomplish this task, it is important to correctly organize and accurately establish the conditions of heat transfer. This is necessary for the performance of a nuclear power plant. The work was focused on substantiating the thermal reliability of the VVER-1200 core with improved fuel assemblies. In this assembly, the design solution was the inclusion of mixing grates (MR) of the «vortex» and «running» types. As a result, it was found that this design feature makes it possible to increase the amount of electricity generation by increasing the critical power of fuel assemblies for the VVER reactor. The subject of the study were: the phenomenon of the crisis of heat transfer of the first kind, as well as thermalhydraulic processes in the core. In accordance with the intended purpose of the study, a solution technique was determined, for which the whole work was divided into two stages. During the first stage, a thermalhydraulic calculation was carried out, the temperature and heat flux distributions were obtained in the cross section and along the height of the fuel element, in which it is maximally loaded, taking into account the surface boiling zone. As a result of the second stage, it was found: the safety factor before the heat transfer crisis. Thus, this technique made it possible to obtain a result on the assessment of the heat engineering reliability of the reactor plant, based on the criteria found. As a final result, the following factors were determined: the reserve before the heat transfer crisis, pressure losses in the reactor plant in the near-wall boiling zone, flow parameters in a two-phase state, and uneven heat flows. Based on the results, it was proved and shown that the improvement of the FA design by adding four «vortex» and «running» gratings made it possible to increase the thermal reliability of the reactor core. This, in turn, makes it possible to operate the power unit at an increased power level, which will lead to an increase in the energy efficiency of the NPP. The significance of this work is: determination of indicators of thermal reliability of the AZ, responsible for the safety of the reactor plant.

Primary frequency regulation in the power system by nuclear power plants based on hydrogen-thermal storage

According to the Technical Requirements for Generating Equipment of Participants in the Wholesale Market of the Unified Energy System (UES) of Russia, from 2016 to participate in the general primary frequency regulation (PFR), the maneuverable characteristics of generating equipment of nuclear power plants with VVER reactors put into operation before 2009 should ensure frequency deviations guaranteed realization of the required primary power for loading up to 2% of the nominal electric power. For this, the current capacity of the reactor installation should be maintained at a level of not more than 98% of the nominal thermal power. The fulfillment of this requirement significantly reduces the installed capacity utilization factor (ICUF) of reactor plant.In addition, at present in the UES of the Russian Federation there is a tendency towards an increase in the deficit of peak and half-peak capacities. The majority of fossil fuel-fired thermal stations are switched to the half-peak mode, which negatively affects their efficiency and reliability. In addition, the rise in price of natural gas makes it more profitable to sell it abroad instead of burning at power plants. On the other hand, an increase in the share of nuclear power plants is observed in the UES, which exacerbates the problems associated with the passage of minima and maxima of the daily load in the power system, due to the economically and technically justified need to load NPPs with maximum CUF.The authors developed an approach to solving this problem by combining NPPs with an environmentally friendly energy source – an autonomous hydrogen power complex (AHPC), which includes heat accumulators and an additional multifunctional steam turbine unit. The developed energy complex will allow energy to be accumulated during hours of minima load in the power system due to the electrolysis of water to produce hydrogen and oxygen, as well as the accumulation of hot water in the storage tanks. The accumulated energy can be used to generate super-nominal electricity to cover the half-peak load zone in the power system. In addition, the presence of a low-power steam turbine installation will ensure uninterrupted power supply to consumers of their own needs at the NPP by using the energy of the residual heat from the reactor when the station is completely de-energized.Based on the proposed power complex, a method has been developed to ensure the participation of NPPs in the PFR in an energy system with a constant CUF. To assess the effectiveness of the proposed solution, a methodology for thermodynamic analysis of the power complex based on the combination of NPPs with AHPC was developed. The dependence of the required hydrogen fuel consumption and the efficiency of using off-peak electricity on the temperature of the feed water supplied to the hydrogen-oxygen steam generator from the hot water tanks is constructed.Based on the results obtained, the technical and economic efficiency of the developed power complex is considered. The accumulated net present value was determined depending on off-peak electricity tariffs with three variants of the forecast dynamics of the half-peak electricity tariff, taking into account natural gas savings, reduced investment in NPP safety systems and the economic effect of ensuring the participation of NPPs in the PFR with the plant load at 100%.

Effect of Boric Acid Solubility in Steam on the Process of Mass Transfer during Emergency Cooling of VVER-1200 Nuclear Reactor

The mechanisms of boric acid mass transfer in a VVER-1200 reactor core are studied in this work in the event of a major circulatory pipeline rupture and loss of all AC power. The VVER-1200's passive core cooling technology is made up of two levels of hydro accumulators. They use boric acid solution with a concentration of 16 g H3BO3/kg H2O to control the reactivity. Because of the long duration of the accident process, the coolant with high boron content starts boiling and steam with low concentration of boric acid departs the core. So, conditions could arise in the reactor for possible accumulation and subsequent crystallization of boric acid, causing the core heat removal process to deteriorate. Calculations were carried out to estimate the likelihood of H3BO3 build-up and subsequent crystallization in the core of the VVER reactor. According to the calculations, during emergency the boric acid concentration in the reactor core is 0.153 kg/ kg and 0.158 kg/kg in both the events of solubility of steam and without solubility of steam respectively and it does not exceed the solubility limit which is about 0.415 kg/kg at water saturation temperature. No precipitation of boric acid occurs within this time during the whole emergency process. Therefore, findings of the study can be used to verify whether the process of decay heat removal is affected or not.

ANSYS-CFX simulation of the SRBTL test loop core with nanofluid coolant

AbstractIn the present study, CFD calculations are presented for the three types of water-based nanofluids Al2O3/water, CuO/water and TiO2/water with 0.1% volume fraction. These calculations are done with ANSYS-CFX and as geometry the SRBTL test loop as scaled down test loop for a VVER-1000 reactor core design is used. The goal of this study is to evaluate the CFD program against the SRBTL test loop core as a scaled core for applying water-based nanofluids as coolant. ANSYS-CFX simulation data are validated against the RELAP5/MOD3.2 simulation data for pure water. This comparison shows a good agreement. The simulation results for the nanofluids and water including Re number, temperature, viscosity, pressure drop and heat transfer coefficient through the SRBTL test loop core are compared. The results of the comparisons show that the SRBTL test loop core is suitable to extract experimental data of water-based nanofluids for using them as coolant in the VVER-1000 reactor.

Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm

Abstract The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Influence of the heat transfer coefficient on the level of residual stress after heat treatment of the VVER-1000 reactor baffle

Background. Improvement of the methodology for the computational analysis of residual stresses in the structural elements of the reactor is an integral part of the work when extending the service life of NPP power units.
 Objective. Determine the value of residual technological stress arising in the baffle of a VVER-1000 reactor during welding and postweld heat treatment according to the austenitizing mode. To evaluate the effect of considering the dependence of the heat transfer coefficient on the temperature of the baffle surface at cooling in air during heat treatment.
 Methods. Numerical modeling of the stress-strain state of the baffle during welding and postweld heat treatment was carried out using the finite element method.
 Results. It was determined that in the process of heat treatment according to the austenitizing mode, the residual welding stress in the baffle are almost completely relaxed. Due to the high temperature gradient during rapid cooling in air after heating in the process of austenitization, new rather high residual stresses are formed in the zones of the baffle with the greatest metal thickness.
 Conclusions. Based on the results of the investigation, a high level of residual technological stress was determined, which should be considered when calculating the justification for extending the service life of the VVER-1000 reactor baffle.

Heat Supply for the Power-Biological Complex from TPP and NPP

This work is devoted to the problem of utilization of waste heat from condensers of thermal power plants and nuclear power plants. The waste heat of the condensers of TPPs and NPPs, together with the circulating water, enters the environment, causing its thermal pollution. The use of this heat in an energy-biological complex, for example, in fisheries, increases their efficiency and solves an environmental problem. Compared to ordinary ponds, this energy complex has an almost year-round increase in biomass and accelerated maturation of producers. The article presents a developed methodology that makes it possible to assess the effectiveness of such a fishery. Calculations using this method were carried out for a fish farm raising sturgeons on the basis of the waste heat of a nuclear power plant with a VVER-1200 reactor and a K-1200-6.8/50 turbine

Verification of a Reactor’s Digital Acoustic Model in the Startup and Nominal Operation Modes of NPPs Equipped with VVER Reactors

Verification of a Reactor’s Digital Acoustic Model in the Startup and Nominal Operation Modes of NPPs Equipped with VVER Reactors

Development and verification of methods for predicting the frequency of self-oscillations in swirling coolant flows of nuclear power plants

Methods and algorithms for calculating the frequency of self-oscillations in swirling coolant flows of nuclear power plants with VVER reactors have been developed. The frequency of self-oscillations occurring in the acoustic sections of the primary circuit and their connections in the starting modes of the power unit and when operating at the rated power level is predicted. It is established that the self-oscillation frequencies can fall into the frequency bandwidth of mechanical vibrations and vibrations of the reactor plant’s internal devices. It is shown that in order to prevent the resonance of self-oscillations of the coolant with the vibration frequencies of internal devices, it is necessary and sufficient to take the frequency of self-oscillations outside the bandwidth of the vibration frequency of structures. The results of verification of the results of forecasting the frequency of self-oscillations in swirling coolant flows at a nuclear power plant unit with VVER are presented. The application of the developed technique shows that the pendulum oscillations of the VVER - 1200 body are caused by an increase in the parameters of the coolant and the geometric dimensions of the VVER - 1200 reactor compared to VVER-1000.

Analysis of critical pipe break sizes leading to reactor pressure vessel liquid level collapse and core uncovery with APROS

Collapse of liquid level in reactor pressure vessel (RPV) and uncovery of core in typical pipeline rupture accidents, such as steam generator tube rupture (SGTR), main steam line break (MSLB), and loss of coolant accident (LOCA) for pressurized water reactor (PWR), may occur. This study used APROS to establish a full-scale simulation model of VVER440 reactor and evaluated different break sizes that can lead to RPV liquid level collapse and core uncovery during these accident processes. It is demonstrated that in SGTR accident, the liquid level in RPV collapsed under the condition of four or more heat transfer tube ruptures without the risk of core uncovery in all analyzed cases. In the MSLB accident, RPV liquid level collapse and core uncovery were absent due to the integrity of the primary circuit. In the LOCA accident, the RPV liquid level collapsed under a break size of 0.5% of pipe area (PA) of cold and hot leg. Notably, core uncovery occurred when the break area was 1.8% of PA in the cold leg and 65% of PA in the hot leg. The time required until core uncovery was earlier in the cold leg than that in the hot leg.

Comparative study of a stationary two-component nuclear energy system with light water and fast breeder reactors versus one-component one with self-sufficient non-breeding fast neutron reactors

Russia is currently in its initial phase of transitioning to a two-component nuclear power structure, consisting of nuclear power plants with thermal and fast reactors operating in a closed nuclear fuel cycle. Alternative approaches to closing the nuclear fuel cycle often have different assumptions regarding the role of fast reactors in this new structure. They can either imply that fast reactors are self-sufficient and are economically viable in such a way, that they can completely replace existing thermal reactor technology, or that they are used for breeding fuel for the whole nuclear energy system and are essentially complementary to VVER reactors in a permanent two component symbiotic structure. The results of this study show that nuclear power scenarios favoring the development of a large-scale fast reactor fleet with a breeding ratio ∼1 without any additional VVER type component in the nuclear power structure is optimal from an economic standpoint.

Calculation researches for the formation of high-level and long-life medium-level radioactive waste of activation origin for the WWER-1200 reactor of Belarusian NPP

The study of the issue of constructing a disposal facility for high-activity radioactive waste in the Republic of Belarus is associated with an assessment of the volume of radioactive waste to be deeply disposed of, generated as a result of the activation of structural materials of the VVER-1200 reactor of the Belarusian NPP. This paper presents the results of computational studies of the formation of solid high-level radioactive waste (HLW) and long-lived intermediate level radioactive waste (ILW-LL) during neutron activation of materials adjacent to the core of the VVER-1200 reactor structures of the Belarusian NPP. The assessment of the volumes of HLW and ILW-LL of activation origin, formed over 60 years of operation of the VVER-1200 reactor, was carried out on the basis of computational studies of the induced activity of structural and shielding materials using reactor and Monte Carlo program codes (SERPENT 2, TVS-M, DYN3D, MCU- PD). As a result of the research, it was found that when neutrons activate materials of the sections of the VVER-1200 reactor structures of the Belarusian NPP adjacent to the core (in the steel structures of the baffle, shaft, surfacing, part of the reactor vessel, structures of the protective tube unit (PTU), the space under the core, in heat insulation materials, rods of absorbing elements (Dy2TiO5)) are formed by HLW and ILW-LL with a total weight of 272,5 tons and a volume of 43 m3 . Calculated studies of the activation of dry protection materials, building concrete, support truss and biological protection showed that these structural elements will not belong to either HLW or ILW-LL.

Application of metaheuristics optimization in fuel rod design: A case study for helium charging pressure

In this paper, it is suggested that the design parameters of the fuel rod can be optimized by using a metaheuristics algorithm so that the fuel rod is entirely safe in its normal operation and has a more efficient performance than usual. We applied Artificial Bee Colony (ABC) in fuel rod design. The fuel burnup, irradiation history, and operation data of the heavily burned fuel rod are obtained for typical VVER-1000 NPP by modeling core cycle design data from the first to the fourth cycle in the KASKAD package. This package intended for core calculation of VVER reactors consists of several primary modules such as BIPR-7A for assembly-wise core calculation, PERMAK-A for pin-by-pin core calculation, and TVS-M for macroscopic cross-section generation.The conventional initial gas pressure in the VVER reactor is 2 MPa. In this paper, considering the effects of gap gas pressure on the heat distribution of the fuel rod, the objective function has been selected to reduce the maximum temperature and the average temperature of the fuel pellet while maintaining safety standards. The optimal pressure obtained in this paper for this particular reactor with a specific power history is 0.467 MPa which is entirely safe in terms of safety throughout the reactor operation. It reduces the average fuel temperature during the process. Furthermore, the core calculations show that using this optimal pressure increases the quantity of energy produced. Fuel burnup increases at optimum pressure with an overall decrease in fuel temperature.