VVER-1000 Reactor Research Articles

Three-Dimensional Simulation of a VVER-1000 Reactor’s Pressure Chamber in the Modes with Asymmetrical Loop Operation Using a KORSAR/CFD Computation Code

The KORSAR/CFD code results from the development of the KORSAR/GP system code certified in 2009 by the Rostekhnadzor (Federal Service for Ecological, Technological, and Nuclear Supervision) as applied to the calculated justification of the safety for VVER reactors. One of the important aspects of development consists in the introduction of the CFD-module code into functional content for the simulation of spatial turbulent flows in the mixing chambers of reactors using a nested boundary method in the RANS-approximation. The CFD module is combined with a one-dimensional model according to a semi-implicit scheme as a standard code element. Calculation results using the KORSAR/CFD code are presented for the following three modes with the asymmetrical operation of a VVER-1000 reactor’s flow-circuit loops. They consist in breaking the steam pipeline in the steam generator, in connecting the main circulation pump while initially operating three pumps at the reactor power of 71% with respect to the nominal one, and in connecting a pump while initially operating two opposite pumps at the reactor power of 52% with respect to the nominal one. The calculations have been carried out based on the input data file for the NPP power unit with a VVER-1000 developed by the specialists in VVER design at OKB Gidropress, the Chief Designer in the field of VVER reactor units. A three-dimensional simulation of coupled neutron-physical and thermohydraulic processes in the reactor core has been performed. A thermohydraulic model of the reactor core has been used in a channel-by-channel approximation and a program block for the calculation of three-dimensional neutron kinetics. In the problems under consideration, the three-dimensional simulation domain for the CFD module includes four inlet manifolds and a part of the reactor pressure chamber before entering the holes in the elliptical bottom of the shaft. The holes in the elliptical bottom and the area beyond the shaft up to the outlet manifolds have been represented by the elements of a one-dimensional model. Based on the results of the calculations, the heat-carrier flow pattern in the reactor pressure chamber has been analyzed. A flow pattern effect exerted on the dynamics of the liquid temperature distribution at the entry into the fuel assemblies of the core and on the energy release power of the fuel assemblies in the simulated modes has been demonstrated.

Validation of the STEG code using PGV experiments on hydrodynamics of horizontal steam generator

The results of the validation of the STEG code using the experiments on the PGV test facility are presented. The PGV test facility (Electrogorsk, Russia) is a slice of the upper part of the horizontal steam generator for Russian type VVER reactors. The main aim of the tests is to investigate the influence of submerged perforated sheets (SPSs) with uniform and non-uniform perforation degrees on the hydrodynamics of two-phase water-steam flows. The brief description of the STEG code based on the 3D two-fluid model is presented. The crucial constitutive laws that allow an accurate simulation of the PGV tests are the correlation for SPS pressure drop and the interfacial drag force model. The SPS pressure drop is calculated with empirical correlation obtained at the PGV experimental data. It is demonstrated that three different models of interfacial drag force are necessary in different regions of the test facility (above the SPS, below the SPS and zone between the tube bundle and the SPS). The formulations of the models are presented. All experiments are analyzed using the STEG code. The peculiarities of the spatial two-phase flows have been established, and a quantitative comparison with experimental data (SPS pressure drops and void fractions) has been performed. It can be argued that the results generally indicate the relatively good predictive ability of the STEG code.

ОЦЕНКА ВЛИЯНИЯ ИЗМЕНЕНИЯ ПЛОТНОСТИ ТЕПЛОНОСИТЕЛЯ НА ТОЧНОСТЬ РАСЧЕТА ПОТВЭЛЬНОГО ЭНЕРГОВЫДЕЛЕНИЯ ВВЭР

This paper presents the verification results of SAPFIR_95&RC_micro program that is intended for fine-grid three-dimensional pin-by-pin calculation of power in VVER reactor core at change of coolant density. In this paper, fine-grid three-dimensional pin-by-pin power calculation is typically seen as calculation where pin-by-pin power is computed on the finite-difference grid that matches the fuel rod spac-ing in fuel assembly. MCNP-based calculation was used as a benchmark; this calculation was prepared for one of the first fuel loadings of the NPP-2006 reactor. A series of two-dimensional calculations has been performed: in each test the same coolant density over the reactor core (variation of water density from 0.3 to 0.8 g/cm3) is set, coolant temperature is considered as invariable and is 315 °C. Three-dimensional calculation has been performed: different density (range of 0.46-0.73 g/cm3) in ten layers along the core axis is set, coolant temperature is varying from 299 to 348 °C. The influence of number of energy groups on the accuracy of power calculation for fuel rods and gadolinium fuel rods is analyzed. Two-, six- and twelve-group diffusion approximation are calculated. The calculation model is prepared using SAPFIR_95.1 cell calculation program and SAPFIR_95&RC_micro pin-by-pin power calculation program.

СРАВНЕНИЕ КОЭФФИЦИЕНТОВ ВЛИЯНИЯ ДЛЯ РАСЧЕТОВ ТОКОВ ИОНИЗАЦИОННЫХ КАМЕР ДЛЯ РЕАКТОРА РУ В-392М (ВВЭР-1200), РАССЧИТАННЫХ РАЗЛИЧНЫМИ СПОСОБАМИ

The article represents the results of calculations of the influence factors of fuel assemblies on the current of ionizing chambers (IC) in the simulation of a reaktimetr of reactor VVER. Two independent approaches and two programs were used to solve the neutron transport equation. The influence factors of the extreme series of fuel assemblies nearest to the reflection shield and IC on the IC readings are calculated. According to the known software "MCNP-4C" factors are obtained from the solution of forward and adjoint problems. Software "TDMCC", developed in RFNC-VNIIEF, the influence factors obtained from the solution of the forward problem. Both of programs implement the Monte Carlo method to solve the transport equation. The obtained coefficients were used in the calculations of measurements that were carried out at the physical start-up of unit 1 of Novovoronezh-2 NPP, including the simulation of measurement of scram-system efficiency. The article represents the dependence of the relative current on time in the logarithmic scale, as well as the dependence of reactivity on time, built on the results of numerical simulation using three sets of the influence factors in the process of resetting the scram-system. As an illustration, the dependencies are not different from each other.

Prospects of creation of the two-component nuclear energy system

Possible options of organization of two-component energy system with closed nuclear fuel cycle (CNFC) and new business potential associated with provision of CFC services to foreign customers are examined. Dominating role in the development of nuclear power generation is assigned to VVER reactors with gradually increasing fraction of sodium-cooled fast breeder reactors (FR) incorporated within the joint nuclear fuel cycle operated on MOX-fuel. Components of such energy system perform the following functions: 1. Fast reactors: Generate electric power in base-load mode (possibility of fine tuning of reactor power within limited range (100 – 75 – 100%) is assumed); Utilize waste and/or regenerated uranium for re-fueling power reactors, produce plutonium applicable to the maximum extent for manufacturing MOX-fuel for VVER reactors; Incinerate long-lived highly radioactive wastes – minor actinides separated during reprocessing spent nuclear fuel of FR and VVER reactors. 2. VVER reactors: Generate electricity in compliance with manoeuvrability requirements imposed by the utility company operating the energy system; Utilize MOX-fuel instead of UO2 fuel; Are offered for export with the option of returning SNF back to Russia; Plutonium extracted from VVER spent fuel is used for manufacturing MOX-fuel for SFR. 3. Nuclear fuel cycle facilities: Provide reprocessing of SNF from VVER and SFR with extraction of nuclear materials for recycling; Use depleted or reprocessed uranium and plutonium extracted from spent nuclear fuel for manufacturing MOX-fuel; Provide partitioning of RAW for subsequent utilization of minor actinides and reduction of risks of proliferation of nuclear materials, conditioning and disposal of RAW. Russia possesses capacities for establishing the two-component system with CNFC, as well as the new business approach to rendering CNFC services to foreign customers.

Сортування половинок зразків-свідків відносно їхнього положення при опроміненні в реакторі ВВЕР-1000

Сортування половинок зразків-свідків відносно їхнього положення при опроміненні в реакторі ВВЕР-1000

Development of the model to determine the fuel temperature field in a two-dimensional problem statement

Water-cooled water-moderated reactors (VVER) are widely used at Russian nuclear power plants. The VVER reactor core is formed by fuel assemblies consisting of fuel rods. The fuel in fuel rods is uranium dioxide. The safety of the reactor operation is ensured through stringent requirements for the maximum nuclear fuel temperature. Calculation of temperature fields within the reactor core requires associated problems to be solved to determine the internal energy release in fuel based on neutronic characteristics. Dedicated software for such calculations is not available to a broad range of users. At the present time, there are numerical thermophysical modeling packages available for training or noncommercial applications which are used extensively, including Elcut, Flow Vision, Ansys Fluent, and Comsol Multiphysics. Verification of the obtained results is becoming an important issue in building models using these calculation packages. An analytical solution was obtained as part of the study for the fuel temperature field determination. A program was developed in MathCAD based on this solution. A model was developed in Comsol Multiphysics to determine the fuel temperature field with constant thermophysical properties in a two-dimensional problem statement. The numerical model was verified using the analytical solution. The influence of the number of the grid nodes on the solution accuracy was established. The analytical solution can be used to determine the fuel temperature field at any radial coordinate of the reactor. The temperature field determination model developed in MathCAD can be used to verify numerical models of the fuel temperature field determination developed in dedicated packages.

Some aspects of the VVER-440 reactor plant life re-extension: a case study of the Novovoronezh NPP Unit 4

The re-modernization of Unit 4 at the Novovoronezh NPP (Novovoronezh-4) made it possible to take a new approach to the problem of extending the VVER-440 reactor plant life and operation. The authors analyze the existing problems of the VVER-440/179 power unit, showing possible solutions to the identified shortcomings and the final state of the updated power unit. Modernization works significantly expanded the range of design-basis accidents from the primary coolant leak from an opening (DN = 100 mm) to the maximum possible, associated with a rupture of the main circulation pipelines (MCP) (DN = 500 mm). A unique experience was gained in using the safety systems of Unit 3, which was finally shutdown for decommissioning, to increase reliability and provide additional redundancy for the safety systems of Unit 4. The results of the performed works showed the correctness of the adopted concept of re-extending the service life of Unit 4 and ensured its compliance with the modern safety requirements in nuclear power engineering, including as it relates to the safety impact of the first-level probabilistic safety analysis model (PSA-1) for internal initiating events.

Radioecological monitoring in the vicinity of Rostov NPP. The analysis of results of long-term investigations

The purpose of this work was to assess the influence of the Rostov NPP (consisting of 4 units with VVER1000 reactors) on the radioecological situation in the region of the station location within 18 years from the start of the first power reactor. Based on the results of observations at the radioecological monitoring network established in 2001, which included seven control sites and five control points, the content of natural and artificial radionuclides in environmental objects, agricultural products and foodstuffs was analyzed. It was shown that over the entire period under consideration, the average content of 90Sr in the soils of agroecosystems of the 15-km zone of influence of the Rostov NPP varied within 1.7-7.4 Bq/kg, and 137Cs – 7.5-14.9 Bq/kg. It was not detected any trends to increase in the concentrations of artificial radionuclides in the soil throughout the considered territory. The variation range of the average content of natural radionuclides in soils was 561-634 Bq/kg for 40K, 23.4-27.5 Bq/kg for 226Ra and 32.7-35.9 Bq/kg for 232Th. The average concentration of 90Sr in the grain was in the range of 0.1-0.68 Bq/kg, and 137Cs – 0.23-0.54 Bq/kg. Even the maximum specific activity values of artificial radionuclides in food grain were below the SanPiN standards 55 times for 90Sr and 65 times for 137Cs. The maximum levels of 137Cs in vegetables and potatoes and vine crops were 80 times lower than the SanPiN standards. The maximum levels of specific activity of 90Sr in milk were more than 400 times lower than the SanPiN standard (25 Bq/kg), and for 137Cs (standard 100 Bq/kg) this difference was 600 times. Vegetable crops were characterized by minimal transfer factors (TFs) of radionuclides. TFs of 90Sr in vegetables, depending on the product type, was in the range of 0.04-0.17, and for 137Cs it was in the range of 0.008-0.2 (Bq/kg)/(kBq/m2 ). Maximum TFs of radionuclides were noted in grasses. TF of 90Sr in natural and perennial grass was 0.75-2.2, and for 137Cs it was 0.28-0.86 (Bq/kg)/(kBq/m2 ). The differences in transition factors between vegetables and grass reached 50 times, and on average, they were 10-20 times. It was noted that all types of crops accumulated 90Sr on average 2-5 times more in comparison to 137Cs ac cumulation. It was shown that the content of artificial radionuclides in the water of the Tsymlyansk reservoir is 163 and 183 times lower than the permissible level for 90Sr and 137Cs, respectively. The analysis of 18-year results of observations of the radioecological situation in the vicinity of the Rostov NPP allows to conclude that the operation of this NPP in normal mode and the commissioning of new power-generating units did not lead to a recorded increase in the content of artificial radionuclides in agricultural products, foodstuffs and environmental objects.

Расчет теплогидравлических характеристик исследовательского реактора на основе ТВС реактора ВВЭР-440

Расчет теплогидравлических характеристик исследовательского реактора на основе ТВС реактора ВВЭР-440

Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320

Abstract The paper gives a description of conservative analysis of initiating event associated with uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín NPP. This event is included in the group of beyond design basis accidents. The aim of analysis is to determine also the time interval which is necessary for interventions leading to the deterrence of fuel damage. The failure of operator intervention to isolate dilution routes at intervals shorter than 30 min is assumed. Since the response of the whole NPP system influences the course of safety important parameters of the reactor core, the calculations were made by an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic system code ATHLET. It is shown that, in addition to exceeding the DNBR limit of more than 99 min from the start of the transient, the remaining safety acceptance criteria will not be violated until the end of the calculation.

Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018

Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018

Fuel cycles with PK-3+ FAs for VVER-440 reactors

Abstract In order to increase the efficiency of fuel utilization at Dukovany NPP, the design of FA was changed by shroud removal and replacement with a structure called “Karkas”. Optimization of PK-3+ type FAs with different average enrichments was performed in order to find out those enrichment profiles with minimized non-uniform energy generation in FA (during burn-up). In addition, it was assumed that such a radial enrichment profile in FA could be achieved by making a change in the location of the fuel pin with a Gd2O3 burnable absorber – from the 2nd row to the 3rd row of pins from the edge of the fuel assembly on the fuel assembly diagonal. The aim of this study was to achieve a full quadruplicate cycle, every 15 months (approx. 450 days) at 1475 MWt nominal power. Preliminary results indicate that combination of PK-3+ and Gd-2M+ fuel assemblies does not show any unusual phenomena from the point of view of reactor physics. The proposed strategy is based on B1C33 cycle implemented at Dukovany NPP that is designed to be 395 FPDs. Already in the first “transient” cycle (34th) loaded with 60 fresh PK-3+ FAs and 12 Gd-2M++ CAs, the reached length at EOR is 424 FPDs, which means stretch-out 26 effective days. Averaged, the transition cycle stretch-out length is 23.5 effective days. For steady cycles, this average value is 19.2 effective days.

Serpent2-SUBCHANFLOW pin-by-pin modelling capabilities for VVER geometries

In the framework of the EU Horizon 2020 McSAFE project, a Serpent2-SUBCHANFLOW coupling has been developed with the aim at performing large-scale pin-by-pin depletion and transient calculations in Light Water Reactors. While this tool is not tied to a specific type of geometry, a set of capabilities have been developed to generate the pin-level hexagonal models corresponding to VVER reactors. The handling of VVER geometries is based on the use of nested hexagonal regular meshes in Serpent2, pin-level subchannel models in SUBCHANFLOW and unstructured meshes for feedback exchange and interpolation. In this work, these features are explained in detail and illustrated in a VVER-1000 fuel assembly. For this test case, an in-depth analysis is made regarding the modelling considerations in SUBCHANFLOW and their impact on the neutronic solution. In particular, the use of coolant- and fuel-centered subchannels and the explicit modelling of stiffener plates is discussed.

Evaluation of loads on the main joint of reactor pressure vessel under hypothetical severe accident at VVER-1200 NPP

This paper presents the results of calculations of the loadings that reactor main joint may experience after energetic fuel-coolant interaction in a hypothetical severe accident at VVER-1200 reactor. Estimates of the loadings are performed using a semi-empirical procedure with conservative assumptions. The initial data for estimates are deduced from the modeling of representative severe accident with an integral code SOCRAT/V1. The loadings can be used for further thermomechanical calculations of the integrity of elements that are fixing the upper head of VVER-1200 RPV.

Quantification and propagation of neutronics uncertainties of the Kozloduy-6 VVER-1000 fuel assembly using SCALE 6.2.1 within the NEA/OECD benchmark for uncertainty analysis in modelling of LWRs

This work is based on the benchmark for uncertainty analysis in modelling of light water reactors compiled by the Nuclear Energy Agency within the Organisation for Economic Cooperation and Development (OECD/NEA). The objective of the benchmark is to determine and verify uncertainty bounds for results of calculations of LWRs based on operating data using best-estimate codes. The main contribution of this paper is the quantification of uncertainties in the Kozloduy-6 VVER-1000 fuel assembly using SCALE-6.2.1 methodology. The benchmark consists of three phases, each with three exercises. Three reactor systems are also studied, viz. the PWR, VVER and BWR reactors. In this study, Phase I of the benchmark was considered for the uncertainty quantification. The sources of uncertainties are classified into three groups, namely uncertainties due to nuclear data, manufacturing tolerances and numerical uncertainties due to methods’ implementations. The calculations are carried out using KENO-VI to perform the neutronics calculations and TSUNAMI-2D/3D and SAMPLER to perform the sensitivity and uncertainty analysis. Nuclear data uncertainty has been identified to be the highest contributor of uncertainty of the VVER-1000 fuel assembly. Although this is true, the uncertainty due to other parameters must always be considered together with nuclear data, since some of them could be significant.

Short-term strength properties and features of high-temperature deformation of VVER reactor pressure vessel steel 15Kh2NMFA-A within the temperature range 20–1200 °C

The problem discussed in this work concerns the estimate of integrity of the VVER reactor pressure vessel in the severe accident. In particular, the main goal of discussed investigation is to determine short-term mechanical properties of ferritic-pearlitic vessel steel 15Kh2NMFA-A (2% Cr, 1% Ni, 0.5% Mn, 0.5% Mo) within the temperature range 20–1200 °C. This steel type is one of basic structure steel types used in fabrication of reactor pressure vessels of Russian design, particularly, the VVER vessels. To perform short-term tests of this steel, several lots of steel specimens 8 and 10 mm in diameter with gauge part of length 40–50 mm, correspondingly, were made. The test specimens were made of the fragment of unirradiated standard VVER-1000 reactor pressure vessel of Russian design fabricated by OMZ Group “Izhorskiye zavody” (Russia). The short-term tensile tests of steel specimens were performed both in air within the temperature range 20–1100 °C and in vacuum within the temperature range 900–1200 °C. Basing on the results of the tests, strength-and plastic/strain parameters of this steel (ultimate tensile strength, yield stress, tensile Young’s modulus, steel elongation at fracture, the coefficient of linear thermal expansion) were determined. The analysis of obtained results made it possible to reveal a series of singularities of 15Kh2NMFA-A steel strain. In particular, it was shown that in the temperature range 770–820 °C (Ac1-Ac3), the investigated steel undergoes inner phase/structure transformations, which have an essential influence on strength- and strain properties of investigated steel. The analysis of temperature effect on strength properties (ultimate tensile strength and yield stress) and Young’s modulus of this steel made it possible to reveal three temperature ranges in which these parameters have characteristic features of change. In the first of these ranges (the temperature values from 20 to ~650 °C), a moderate degradation of strength- and elastic properties of steel is observed. In the second range (from ~650 to 850 °C), a sharp decrease of values of steel strength and elasticity occurs, and at test temperature of more than 900 °C, a monotonous decrease of these mechanical properties is observed. The comparison of the results obtained in the present work with the earlier results of tests at temperature of more than 900 °C showed their satisfactory agreement.

Assessment of the nano-fluid effects on the thermal reactivity feedback coefficients in the VVER-1000 nuclear reactor with nano-fluid as a coolant using thermal hydraulic and neutronics analysis

Increasing efficiency and improving energy consumption in the nuclear power plants have always been of interest for researchers. So, they try to improve the heat transfer in the applied systems. Hence, extensive research has been performed to apply an alternative fluid which has more suitable thermal properties instead of the conventional fluids such as water. One of these efforts is application of nanoparticles in a coolant fluid. The most important advantage of the nanoparticles is increase of the thermal conductivity and heat transfer coefficient. Considering the importance of the nanofluids effect as a coolant in the nuclear reactors on the reactor dynamic parameters, in this paper, for the first time, fuel and coolant temperature reactivity coefficients which have important contribution in the dynamic analysis and safety requirements of the nuclear reactors, are calculated in a VVER-1000 Nuclear Reactor with nanofluid as a coolant. In this study, using different volumetric percentages and sizes of Al2O3 (Alumina) nanoparticle, the important and fundamental parameters of the VVER-1000 reactor, including dynamic reactor parameters such as temperature reactivity coefficients are calculated. For this purpose, at the first, the equivalent cell of the fuel rod and the surrounding coolant nanofluid are simulated in the hexagonal fuel cell of the VVER-1000 reactor. Then, the thermal hydraulic calculations are carried out at different concentrations and sizes of the nanoparticle and their effects on the heat transfer parameters such as the heat transfer coefficient, temperature of coolant and fuel are assessed. Also, using the neutron calculating codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated and analyzed. Through the thermal hydraulic and neutronics calculations, the fuel and coolant temperature reactivity coefficients are calculated and analyzes versus variation of the concentration and size.

Neutronic and thermal-hydraulic aspects of loading pattern optimization during the first cycle of VVER-1000 reactor using Polar Bear Optimization method

Loading Pattern Optimization (LPO) is an extremely important issue for the safe and economic operation at nuclear power plants because the core configuration affects many of the neutronic and thermal-hydraulic parameters (NTPs) of the reactor. Most researches were done in this area limited to employing different heuristic optimizing algorithm and the neutronic aspects at the beginning of the cycle (BOC). In this research, a comprehensive code package has been developed using coupling the PARCS code and the COBRA-EN code with the new Polar Bear Optimization (PBO) algorithm, which tries to find an optimal loading pattern with maximized cycle length, maximized departure from the nucleate boiling ratio (DNBR), and flatter power distribution. Other core safety parameters such as maximum boric acid concentration at the BOC, maximum power peaking factor (PPF), the maximum temperature of the fuel center and the maximum clad surface temperature are considered as penalties for accepting or rejecting optimal configuration alongside the fitness function. This code Package which named PBLPO (Polar Bear Loading Pattern Optimization) code, was implemented for the core of VVER-1000 reactor in Bushehr Nuclear Power Plant (BNPP). The optimized pattern that found with PBLPO improved safety parameters comparing to loading pattern that the proposed in the Final Safety Analysis Report (FSAR) and shows that the polar bear algorithm has a good performance in LPO.

ЭКСПОРТНЫЙ ПОТЕНЦИАЛ РОССИЙСКИХ БЫСТРЫХ РЕАКТОРОВ И ТЕХНОЛОГИЙ ЗАМКНУТОГО ЯДЕРНОГО ТОПЛИВНОГО ЦИКЛА

The key competitive advantage of the Rosatom State Corporation on the world nuclear energy market is an integrated offer of a range of services, including construction, operation and maintenance of NPPs abroad for the entire life cycle up to 60 years, adopted for new projects with VVER-1200 reactors. In the long term, a significant expansion of the scale of international business is planned, which is reflected in the target indicators of Rosatom State Corporation and its organizations. Russia is a recognized world leader in the development of fast neutron reactors. To date, the technology of sodium fast reactors BN successfully demonstrated and entered the stage of commercialization. Russia is the only country in the world that has in its nuclear power structure operating BN 600 and BN 800 fast reactors. Technologies of fast reactors and closed nuclear fuel cycle, subject to their commercialization, may become one of the main export products of the Russian nuclear energy industry. Therefore, the leadership of the Rosatom State Corporation has set the task of developing the conditions for entering Russian NPPs with fast reactors the world nuclear energy markets.