VVER-1000 Reactor Core Research Articles

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

The Russian fuel assembly TVS-2M is a newly proposed fuel configuration that contains gadolinium-oxide at different concentrations mixed with enriched UO2 at different U-235 enrichment values. The purpose of this study is to investigate the neutronic behavior of the Bushehr VVER-1000 reactor core by replacing the current standard Fuel Assemblies (FAs) called TVS with the new generation of TVS-2M FAs. In such manner, the reactor core was simulated under both modes of TVS-2M and TVS FAs arrangement using the WIMS, CITATION, MCNPX, and SuperMC codes in equal terms. Neutronic parameters of the reactor core such as effective multiplication factor, excess reactivity, neutron flux distribution, power peaking factor, and control rod worth are calculated and discussed within a comparative framework of both configurations of FAs; TVS and TVS-2M.

Investigation of VVER-1000 fuel assembly bowing effect on power distribution during cycle using neutron noise adiabatic approximation

Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to variations in boric acid concentration during the cycle. Neutron noise theory is employed to analyze these time-dependent neutron perturbations. In this research, it is investigated the bowing effect during the cycle in a VVER-1000 reactor using the neutron noise adiabatic approximation. It was initially considered a VVER-1000 reactor core to verify the calculations algorithm without bowed FAs. Results had acceptable accuracy in power, initial boric acid, and cycle length. Subsequently, it was explored various modes of FA bowing as neutron noise sources the results revealed that the power noise caused by bowing varies throughout the cycle, as expected. The noise amplitude during the cycle depends on changes in boric acid concentration, burnup, size, and the direction of bowing.

Loading pattern optimization of VVER-1000 reactor core based on the discrete golden eagle optimization algorithm

The main features of the loading pattern optimization (LPO) problem, such as high-dimensionality, multi-modality, and non-linearity, make it difficult to achieve a truly optimal configuration. In recent years, metaheuristic methods have been successfully used to solve this problem. In this research, a discrete golden eagle optimization (DGEO) algorithm has been developed to solve the LPO problem in the first cycle of VVER-1000 reactor core. To evaluate the proposed algorithm, a linear multi-purpose fitness function has been used to improve the safety parameters of the reactor core by obtaining a flatter power distribution during the first cycle, and also to enhance the economic parameters by increasing the cycle length and reducing the cost of fuel recycling. For this purpose, a FORTRAN program has been written to map the DGEO algorithm for the LPO problem using the PMAX and PARCS core calculation code to compute the fitness function in each iteration. To speed up the calculations, parallel computing has been applied in the written program. The results demonstrated that the loading pattern, which is suggested by the DGEO algorithm, enhances all the safety and economic parameters in the fitness function. Thus, the DGEO algorithm is highly reliable for the LPO problems in the VVER 1000 reactor core.

Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel

The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.

Evaluation of Neutronic Performance for the VVER-1000 Reactor Core with Regenerated Uranium-Plutonium Fuel

Evaluation of Neutronic Performance for the VVER-1000 Reactor Core with Regenerated Uranium-Plutonium Fuel

Loading Pattern Optimization of Vver-1000 Reactor Core Based on the Discrete Golden Eagle Optimization Algorithm

Loading Pattern Optimization of Vver-1000 Reactor Core Based on the Discrete Golden Eagle Optimization Algorithm

Preliminary design of soluble boron free small modular reactor core based on hexagonal lattice

This study made an effort to design a novel high-power density SMR core based on a hexagonal lattice for the soluble boron free environment. The VVER-1000 reactor core was converted to an SMR core with a “halved scale”. The core design process is divided into two parts; Part I: assembly-level analysis and Part II: whole-core analysis. Important design parameters such as core dimensions, number, and arrangement of fuel assemblies, fuel enrichment, appropriate burnable poison, and placement of burnable poison pins have been studied in Part I. Also, by designing an optimized burnable poison pattern, the excess reactivity and acceptable radial pin power distribution were treated without soluble boron. WIMS, and PARCS codes have been investigated for fuel assembly and whole core analysis. Finding the most appropriate fuel-loading pattern involved an iterative manual optimization between the core loading pattern and FA design to assess the acceptable neutronics and thermal-hydraulics core features and satisfy the safety margins.The design of high-power density core requires accurate prediction of the thermal–hydraulic parameters to ensure safe operation. In Part II coupled 3-D neutronic/thermal–hydraulic analysis has been done to identify the main safety challenges of the developed core. To access the proper discharge burn-up, an initial enrichment of 6.6% U235 was determined for the developed core. The developed core maximum discharge burn-up is equal to ∼49 Mwd/KgU at full power operation. Radial-zoning burnable poison is investigated in the seek of a proper pattern to limit power peaking.

A Possibility for Large-Scale Production of 238Pu in Light-Water Reactor VVER-1000

This paper considers the possibility for large-scale production of plutonium isotope 238Pu in the light-water nuclear power reactor VVER-1000. 238Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with 237NpO2 as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/237Np) by optimizing the pitch of NpO2-rod lattice. The computations showed that the scale of 238Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope 208Pb) around the NpO2 fuel assembly (FA) made it possible to obtain high-purity 238Pu with the content of 236Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years).

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Abstract This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al2O3 nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

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.

Burnup and neutronic analysis for UOX and MOX fuel assemblies in VVER reactor

The present work examined the feasibility of utilizing Mixed Uranium–Plutonium Oxide (MOX) fuel in the VVER-1000 reactor core. In this study, about 30% of Uranium Oxide (UOX) fuel assemblies of the original core are replaced by MOX fuel assemblies. The effect of MOX fuel on the safety parameters is studied by using the MCNP6 code and the nuclear data library (ENDF/B)-VII.1. The safety parameters examined included the effective multiplication factor, temperature coefficients, void coefficient, boron coefficient, shutdown margin and the kinetic parameters. The excess reactivity, neutron flux and power density distributions and power peaking factor are also examined. All estimated values of reactivity coefficients analyzed for the core contains 30% MOX fuel confirmed to be negative, which guarantees the reactor safety and power stability in a normal state. Also, the effective delayed neutron fraction was lower than that of the full UOX core as well as the neutron lifetime was shorter, this shows that the MOX core responds faster than the UOX core. The maximum power factor for the MOX core was equal to the design limit. The shutdown margin satisfied the design requirement. The results confirmed that the VVER-1000 is able to meet the safety requirement for core designs with MOX loadings of up to 30%.

Analysis of the Kalinin-3 coolant transient benchmark by SKETCH-N/SKAZKA code system

The switching off of the one of four main circulation pumps results in the Kalinin-3 Coolant Transient Benchmark. The benchmark contains four exercises, but this article shows the results of the exercise two of the VVER-1000 reactor core calculation. Reactor core analysis code SKETCH-N/SKAZKA is used to simulate the steady-state and transient conditions of the VVER reactor. The reactor core boundary conditions for the problem are provided by ATHLET/BIPR-VVER code.In the initial steady-state condition and at the end of the transient the power distribution is compared with the nuclear power plant (NPP) data. The average error in 3-D power distribution is 3.6%. The average difference in fuel assembly (FA) power is less than 2%. In the transient calculation, we compared with the experimental data the reactor power and axial offset. There is close agreement between the numerical results and NPP data.

A coupled thermohydraulic/neutronic investigation of noise propagation in VVER-1000 reactor using[formula omitted] nanofluid coolant

Thermohydraulic and neutronic coupling for the VVER-1000 reactor carried out by applying Al2O3 nanofluid as coolant in this paper. The simulation took place in two sections; with the neutronic section, the discretization of the nodal (average current) method for dissociation of two-group diffusion equations accomplished and with the thermohydraulic section, the finite volume method for decomposition of conservation equations (mass, momentum and energy) in static and transient states was used.In this study, neutronic and thermohydraulic noises (Fourier transform) in the frequency domain were analyzed by applying propagation noise sources in the VVER-1000 reactor core. With the neutronic section, macroscopic cross sections were generated making use of Al2O3 nanofluid as coolant with different volume fractions by means of WIMS-D5 computation code.The ultimate goal of the paper is to perform neutronic and thermohydraulic simulations of VVER-1000 reactor using Al2O3 nanofluid as coolant in the frequency domain. Two separate noise scenarios for the inlet nanofluid coolant (coolant temperature and velocity) with different volume fractions of Al2O3 nanoparticles (0.001%, 0.1% and 1.0%) were considered in the analysis. Final results showed that the core power peaking factors will be less influenced by inlet fluctuations when applying more nanoparticles in the reactor coolant.

"Full-Core" VVER-1000 calculation benchmark

Abstract This work deals with the \Full-Core" VVER-1000 calculation benchmark which was proposed on the 26th Symposium of AER [1]. Recently, the calculation benchmarks \Full-Core" VVER-440 [2] and its extension [3] have been introduced in the AER community with positive response [4, 5]. Therefore we have decided to prepare a similar benchmark for VVER- 1000. This benchmark is also a 2D calculation benchmark based on the VVER-1000 reactor core cold state geometry, explicitly taking into account the geometry of the radial reflector. The loading pattern for this core is very similar to the fresh fuel loading of cycle 9 at Unit 1 of the Temelin NPP (Czech Republic). This core is filled with six types of fuel assemblies with enrichment from 1.3%w 235U to 4.0%w 235U with up to 9 fuel pins with Gd burnable absorber per FA. The main task of this benchmark is to test the pin-by-pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron- physics calculations especially for VVER reactors. In this contribution we present the overview of available macro-codes results.

A coupled neutronic/thermal–hydraulic module for the transient analysis of VVER-1000 reactor during reactivity insertion accidents

For the safety analysis of nuclear power plants, it is essential to provide a simple and fast model to investigate the time-dependent behavior of reactor cores. Most research papers in this field have generally been limited to use coupling techniques between existing codes. In this paper, an integrated module called CMS (nuclear reactor Core Modular Simulator) is proposed to simulate the transient thermo-neutronic behavior of a VVER-1000 reactor core during reactivity insertion accidents. CMS contains three modules: a neutronic module based on the point kinetic equations, a thermal-hydraulic module based on the time-dependent single heated channel approximation and a reactivity calculation one. A two-stage fourth order implicit Runge-Kutta method was used for solving the point kinetics equations and the homogeneous two-phase flow model was considered for the thermal-hydraulic part. To verify the proposed module, two reactivity insertion accidents of the Bushehr VVER-1000 reactor, a design basis accident and an anticipated operational occurrence, have been simulated and the obtained results have been compared with its final safety analysis report showing a good agreement.

РЕЗУЛЬТАТЫ МОДЕЛИРОВАНИЯ ТЕСТОВОЙ ЗАДАЧИ «BENCHMARK ROSTOV 2» С ИСПОЛЬЗОВАНИЕМ КОМПЛЕКСА ПРОГРАММ САПФИР_95&RC_ВВЭР И КОРСАР

This paper presents the results of “Benchmark Rostov 2” test task simulation using SAPFIR_95&RC_VVER program package; the test task is based on a full-scale experiment performed at the Rostov NPP, where the transition mode with boric acid dilution and a working group insertion to compensate the reactivity were implemented. The purpose of this task is to develop the methods of non-stationary computation of pin-by-pin power energy distribution in calculation programs for neutronic and thermohydraulic characteristics of VVER-type reactors. The SAPFIR_95&RC_VVER program package uses two calculation methods of pin-by-pin power energy distribution: micro- and macro-flux superposition method and combined fine-grid method, where the nodes of radial computational grid in the reactor core coincide with the centers of fuel elements in fuel rod assembly. The combined fine-grid method for calculating pin-by-pin power energy distribution is carried out with the known values of burn-up distributions, fuel temperature, and coolant density obtained from superposition method. Both approaches are used for solving stationary and non-stationary tasks. When simulating the non-stationary processes, the test task can be solved in conjunction with KORSAR thermohydraulic code. This paper describes neutronic models of VVER-1000 reactor core used in “Benchmark Rostov 2” test task for superposition method and combined method. Calculation results for power energy distribution fields are given.

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.

Reactivity worth calculation for control rods groups of the VVER-1000 nuclear reactor core considering shadowing effect using Monte Carlo method

Global core calculations use the diffusion equation to predict theoretically the nuclear reactor behaviour. However, this equation is not valid in strong absorbing media where the neutron spectrum is a rapidly varying function of the position, such as control rods or burnable poisons. In this paper, to overcome this misleading, the Monte Carlo simulation has been performed and the VVER-1000 reactor core in the MCL (Reactor Minimally Controlled Power Level) condition is modelled using the MCNPX code to calculate the reactivity worth of the control rod groups. The calculations in this model are divided into four steps. At first, the integral and differential worth are calculated for control groups 8, 9 and 10 with 50% overlapping and shadowing effect is considered. And in three other steps the integral and differential reactivity worths of control groups 8, 9 and 10 are calculated separately (without overlapping). In each step, the core is maintained critical by variation of the boron concentration. In these processes, the boric acid coefficient is achieved while the core is critical. The results are compared with experimental values and are in good agreement with them.

Reactivity worth calculation for control rods groups of the VVER-1000 nuclear reactor core considering shadowing effect using Monte Carlo method

Global core calculations use the diffusion equation to predict theoretically the nuclear reactor behaviour. However, this equation is not valid in strong absorbing media where the neutron spectrum is a rapidly varying function of the position, such as control rods or burnable poisons. In this paper, to overcome this misleading, the Monte Carlo simulation has been performed and the VVER-1000 reactor core in the MCL (Reactor Minimally Controlled Power Level) condition is modelled using the MCNPX code to calculate the reactivity worth of the control rod groups. The calculations in this model are divided into four steps. At first, the integral and differential worth are calculated for control groups 8, 9 and 10 with 50% overlapping and shadowing effect is considered. And in three other steps the integral and differential reactivity worths of control groups 8, 9 and 10 are calculated separately (without overlapping). In each step, the core is maintained critical by variation of the boron concentration. In these processes, the boric acid coefficient is achieved while the core is critical. The results are compared with experimental values and are in good agreement with them.

A novel multi-objective optimization method, imperialist competitive algorithm, for fuel loading pattern of nuclear reactors

Imperialist Competitive Algorithm (ICA) has been successfully applied to the various optimization problems and has demonstrated great results in both the global optimal achievement and the convergence rate. In this work, the ICA has been applied to achieve the best fuel arrangement of the VVER-1000 reactor core during the cycle. Bushehr Nuclear Power Plant (BNPP) core optimal arrangement has been searched by considering the minimization of the power peaking factor, flattening of the radial neutron flux distribution and maximizing the effective multiplication factor (Keff) of the core. Cross section and power distribution calculations have been performed by DRAGON-4 and CITATION codes, respectively. The algorithm implementation has been carried out in MATLAB framework for optimization of one-sixth (1/6) and one-twelfth (1/12) symmetry of the core. The results show that this method can be used as an efficient computational method for finding an optimized multi-objective loading pattern in a reactor core.