VVER Reactor Core Research Articles

Transmutation Investigation of a Typical VVER-1000 Reactor Burnup Products to less Toxicity Isotopes in a Fusion-Fission Hybrid Reactor

Transmutation of burnup products for a typical VVER-1000 reactor in a fusion-fission hybrid reactor has been investigated using MCNPX 2.6.0 calculation code. Burnup calculation of a fission reactor was performed for a 1000 MW practical power of a VVER reactor core using UO2 fuel and light water as its coolant and moderator, as well. The burnup products have been classified and separated according to their half-lives and their usages, and the remaining wastes were used as fuels for a fusion-fission hybrid reactor. The burnup calculation was also implemented for the fission wastes of the mentioned hybrid reactor, and the remaining products were separated and classified. The results showed that fission products can be transmuted into useful elements or to elements with less toxicity (with the decrement of activity) and less mass, and according to the IAEA toxicity limit of nuclear wastes, the levels of radioactivity and toxicity decreased. The results confirmed appropriately “the waste transmutation capability” of a fusion-fission hybrid reactor.

Experimental investigation of the coolant flow in the VVER reactor core with TVSA fuel assemblies

The paper presents the results of an experimental study to investigate the coolant interaction in adjoining fuel assemblies in the VVER reactor core composed of TVSA-T and upgraded TVSA FAs. The processes of the in-core coolant flow were simulated in a test wind tunnel. The experiments were conducted using models representing different portions of the VVER reactor core fuel bundle and consisted in measuring the radial and axial airflow velocities in representative areas within the FAs and in the interassembly space. The results of the experiments can be translated to the full-scale conditions of the coolant flow with the use of the fluid dynamics simulation theory. The measurements were performed using a five-channel pressure-tube probe. The coolant flow pattern in different portions of the fuel bundle is represented by distribution diagrams and distribution maps for the radial and axial velocity vector components in the representative areas of the models. An analysis for the spatial distribution of the radial and axial velocity vector components has made it possible to obtain a detailed pattern of the coolant flow about the FA spacer, mixing and combined spacer grids of different designs. The accumulated database for the coolant flow in FAs of different designs forms the basis for the engineering justification of the VVER reactor core reliability and serviceability. The investigation results for the coolant interaction in adjoining TVSA FAs of different designs have been adopted for the practical use at JSC Afrikantov OKBM to estimate the heat-engineering reliability of the VVER reactor cores and have been included in the database for verification of computational fluid dynamics (CFD) codes and detailed by-channel calculation codes.

ГИДРОДИНАМИКА ТЕПЛОНОСИТЕЛЯ В АКТИВНОЙ ЗОНЕ РЕАКТОРА ВВЭР С ТВСА РАЗЛИЧНЫХ КОНСТРУКЦИЙ

The results of experimental studies of the interassembly interaction of coolant in VVER reactor core which consists of TVSA-T and upgraded TVSA are presented. The modeling of coolant flow in the fuel assembly (FA) was carried out on an aerodynamic stand. The studies were carried out on a fragment model of VVER reactor core and consisted in measuring the velocity vector modulus in the characteristic zones of both TVSA and interassembly space of VVER reactor core. The measurements were carried out by a five-channel pneumometric probe. An analysis of the spatial distribution of projections of the absolute flow velocity allowed to detail the pattern of flow-round by the coolant flow of spacer, mixing and combined spacer grids TVSA. The results of study of interassembly interaction of coolant between neighboring TVSA-T and upgraded TVSA were adopted for practical use by JSC “Afrikantov OKBM” in assessing heat engineering reliability of VVER reactor cores and are included in the data-base for verification of computational fluid dynamics programs (CFD codes) and detailed cellular calculation of VVER reactor core.

Hydrodynamics of coolant in VVER reactor core with fuel assemblies of various designs

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and to determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of interassembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at JSC “Afrikantov OKBM” in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of VVER reactors’ core.

Use of erbium as a burnable absorber for the VVER reactor core life extension

The paper presents the results of a computational and theoretical analysis concerned with the use of erbium as a burnable absorber in VVER-type reactors. Partial refueling options for the reactor life extension to 18 and 24 months is considered, the refueling ratio being equal to three for the 18-month life and to two for the 24-month life. Erbium is expected to be present in all fuel elements in the FA with the same weight content. The influence of the erbium weight content on such neutronic characteristics of the reactor and fuel as burn-up, reactivity coefficients, residual volume of “liquid" control, and amounts of the liquid radioactive waste (LRW) formed was assessed. The calculations were performed using a simplified model of refueling without FA reshuffling. An infinite array of polycells consisting of FAs with different in-core times was considered. The escape of neutrons from the core was taken into account by selecting the critical value K∞ at the end of life. Erbium does not burn up in full for the lifetime which affects the fuel burn-up as compared with the liquid excessive reactivity compensation system. The reduction is 0.7% per 0.1% of the erbium weight load in the fuel elements. This, however, also reduces the maximum content of the boron absorber in the coolant and the LRW accumulation in the ratio of 5% per 0.1% of the erbium weight load. Erbium influences the spectral component of the coolant temperature reactivity coefficient which turns out to be negative even with its minor weight fraction in fuel elements, and a reduction in the boron absorber fraction leads to a positive value of the density reactivity coefficient. As a result, the overall coolant temperature reactivity coefficient has a negative value throughout the lifetime.

Экспериментальное исследование течения потока теплоносителя в активной зоне реактора ВВЭР с ТВСА

Экспериментальное исследование течения потока теплоносителя в активной зоне реактора ВВЭР с ТВСА

Использование эрбия в качестве выгорающего поглотителя в реакторах типа ВВЭР при работе на удлиненных кампаниях

Использование эрбия в качестве выгорающего поглотителя в реакторах типа ВВЭР при работе на удлиненных кампаниях

Determination of the physicochemical properties of boric acid for VVER emergency modes

The results of experimental research of physicochemical properties of water solutions of H3BO3 are presented in this article. The literature review about the influence of hydrogen ion exponent of H3BO3 solutions on its heat and mass transfer, and corrosiveness of an acid in the VVER reactor core in is given. It is found that the existing data does not cover the entire range of parameters (acid concentration, temperature and pressure) characteristic of a possible Loss-of-Coolant Accident (LOCA) at nuclear power plant with a VVER. The problem of the accumulation and crystallization of boric acid in case of LOCA and the operation of passive safety systems (such as hydro accumulators of the first, second and third stages, and a passive heat removal system) is formulated. Test equipment and techniques of research are described. The generalizing equations for the pH and acidity constant of hydrous solutions of H3BO3 in range of the concentration (5-100 gkg−1 H2O) and temperature (25-50 °C) has been obtained.

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.

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.

Diagnostics of the critical heat flux state of a VVER reactor based on a channel steaming model *

The purpose of the study is to develop a model for predicting the process of a critical heat flux state with the VVER reactor core channel steaming. The model describes the dynamics of the nuclear reactor behavior in conditions of uncertainty, which are typical of abnormal situations, based on information on the process of heat exchange in the core process channels. The use of the proposed model leads to an increase in the speed of response due to a simplified procedure to calculate the parameters of the heat exchange process in the reactor core. The quality of the reactor state assessment is improved through the prediction of the heat exchange process parameters and determination of the critical heat flux parameters in the core prior to the onset of surface boiling the potentiality of which is not predicted in modern VVER in-core monitoring systems. A modification of the mathematical model has been proposed which offers the simplest possible way of using the advantages of neural networks in diagnostics. The model can be used to develop systems for diagnostics of in-core anomalies and systems for adaptive control of the VVER-type reactor thermal power.

Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

A physical model of the radial and axial reflector of VVER-1200-like reactor core has been developed. Five types of radial reflector with different material composition exist for the VVER reactor core and 1D and 2D models were developed for all of them. Axial top and bottom reflectors are described by the 1D model. A two-group XS library for diffusion code DYN3D has been generated for all types of reflectors by using Serpent 2 Monte Carlo code. Power distribution in the reactor core calculated in DYN3D is flattened in the core central region to more extent in the 2D model of the radial reflector than in its 1D model.

Analysis of the uncertainties in the physical calculations of water-moderated power reactors of the VVER type by the parameters of models of preparing few-group constants

The article covers the uncertainty analysis of the physical calculations of the VVER reactor core for different meshes of the reference values of the feedback parameters (FBP). Various numbers of nodes of the parametric axes of FBPs and different ranges between them are investigated. The uncertainties of the dynamic calculations are analyzed using RTS RCCA ejection as an example within the framework of the model with the boundary conditions at the core inlet and outlet.

A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

The paper covers the results of VVER core reflooding studies in fuel assembly (FA) mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 °C and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation.

Optimization of designing the core fuel loading pattern in a VVER-1000 nuclear power reactor using the genetic algorithm

In this investigation, there are two prime objectives. These objectives are: firstly, enhancing the core effective multiplication factor (Keff), secondly, keeping the Pq lower than a predetermined value to maintain the fuel integrity, and consequently attaining maximum energy so that these two objectives satisfy optimizing to design the core fuel loading pattern better. In this research, on one hand, a new method has been developed to determine optimized status of the core fuel loading pattern based on the genetic algorithm in a typical VVER-1000 reactor.On the other hand, a simple discrete mathematical model in the first stage has been defined in order to clarify the performance of this complex problem, and the optimization of the given model by simulation, and through computer programming, that is, based on genetic algorithm method. The program of analyzing the core fuel pattern optimization of VVER-1000 reactor is as follows.First, one-sixth of VVER reactor core is simulated by MCNP4C code based on its symmetry, through information of FSAR. Using the given simulation obtained from the MCNP4C code, the neutron flux and Keff of the mentioned reactor are resulted. Then the input files of WIMS-D4 code are created according to the initial data of one-sixth of the symmetry of VVER reactor core fuel assemblies, based on its material characteristics. Then the output results (cross sections) of cell calculation obtained from WIMS-D4 are transferred to the CITATION code as input. Then, the input finally returns to the extended program cycle by assessing the genetic responses. According to the boundary conditions and analyzing method aforementioned, the acceptable response, and the best status of interchanging the fuel assemblies for acceptable responses are determined.