The paper is focused on CFD analyses of the coolant flow in the nuclear reactor VVER 440. The goal of the analyses is to investigate the influence of the orifice diameter on the mass flow through individual fuel assemblies in the reactor core. The diameter of orifice can be changed during the operation of a nuclear power plant. Considered boundary conditions in the investigated region of the coolant are based on nominal coolant flow conditions in the nuclear reactor VVER 440.

Abstract:The paper presents the numerical simulation of thermo-hydraulic behaviour of coolant in the VVER-440 nuclear reactor under standard outage conditions. Heating-up and flow of coolant between the reactor pressure vessel and spent fuel storage pool are discussed.KEYWORDS: CFD analysis, ANSYS CFX, Nuclear Reactor, VVER-440, refuelling IntroductionThermo-hydraulic conditions in nuclear reactors are important not only in the operation mode but also under refuelling and outage conditions. During outages, several components of the nuclear reactor system (NR) are flooded by coolant [1]. These interconnected components are the reactor vessel (RV), reactor pool (RP) and spent fuel storage pool (SFSP). Values of pressure and temperature are significantly lower than in the operating mode (above 12 MPa), because the system is operated under atmospheric pressure conditions. For the thermal behaviour, only residual heat of fuel assemblies is considered. This residual heat is caused by decay of the secondary fission products and it is necessary to ensure that the cooling process during outages of the reactor continues. The thermo-hydraulic conditions are important also for outage conditions, because:• reactor vessel is interconnected with reactor pool and spent fuel storage pool (flooded by coolant) • fuel assemblies with their residual heat are situated in the reactor vessel and in the spent fuel storage pool • thermal and hydraulic influence between reactor vessel and pools occurs • transfer of impurities may occur The paper presents thermo-hydraulic conditions calculated using Computational Fluid Dynamics -CFD code [2, 3] ANSYS CFX in nuclear reactor during outage conditions where the above mentioned phenomena are discussed. CFD AnalysisThe CFD analysis was performed considering the geometric model in Fig. 1, that represents the volume of the coolant in the system during outage conditions.

The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.KEYWORDS: CFD analysis, ANSYS CFX, Fuel assembly, VVER 440, thermal-hydraulics IntroductionNuclear reactor safety, thermohydraulics is a very important subject [1]. Thermohydraulics as a multiphysical domain influences not only the thermal conditions of nuclear fuel, but also the distribution of neutron flux within the reactor core, thermal and pressure loading of reactor pressure vessel and dictates the critical value of heat flux, which can flow form the fuel rod to coolant. For many years, thermohydraulics of nuclear reactors has been investigated only by specialized system codes, like RELAP and ATHLET. In the last decade, computational fluid dynamics -CFD [2] emerged as a very useful alternative tool to analyse thermohydraulics, where real 3D geometry can be considered. The paper presents the application of CFD for the investigation of fuel assembly bypass coolant mass flow and its influence on the coolant temperature distribution within the fuel assembly head. Geometric model and discretizationTo perform thermo-hydraulic analysis of the fuel assembly in the reactor VVER440, it is necessary to create an equivalent 3D geometric model of the coolant in the fuel assembly (FA). Creating the geometric model of coolant is divided into three steps (Fig.1).In the first step, an accurate geometric model of the fuel assembly with all details is created. This model includes parts of the protective tubing known as the fixator, where the thermocouple housing is placed. This 3D geometric model represents real geometry of FA, which also can be used for structural analysis. Fig.1 shows fully detailed 3D CAD model of fuel assembly. In the Fig.1 there is bypass outlet from fuel assembly in the bottom and bypass inlet in top, marked with blue circle.Second step, detailed geometric model of fuel assembly is simplified because of the future mesh generation and computational hardware limitations. Simplifications are performed on input and also on output parts of fuel assembly. Those modifications won't have significant influence on the coolant flow (Fig.1).

Abstract:The article deals with modelling of coolant flow within the spent fuel storage pool of a VVER 440 reactor. The spent fuel storage pool is modelled in a state of standard reactor operation. The coolant heating from the remaining thermal power of stored spent fuel assemblies was also modelled. KEYWORDS:CFD, Thermo-Hydraulics, BSVP IntroductionThe spent fuel storage pool is used for long term storage and cooling of spent fuel assemblies of the VVER 440 reactor. The intention of this paper is to model the steady state coolant flow within the storage pool. Computational Fluid Dynamics solutions were calculated using ANSYS CFX. Geometric modelThe spent fuel storage pool consists of the primary pool, two inlet pipelines located at the bottom of the storage pool, two outlet pipe lines located at the top, including the submerged reserve and storage grids and their support structures. The storage grid contains hexagonal absorbers around every fuel assembly storage location. The model contains spent fuel assemblies stored in all storage locations, 290 in total. The heat transfer through the solid walls of the absorbers and fuel assembly shroud was not modelled.Although, the physical models describing the processes within the storage pool are relatively simple, the structures submerged within the pool are of highly complex design and geometry. Significant simplifications were necessary to save computing power and to speed up solution times. The reserve grid was removed from the model together with its supporting structure, as it is not present during standard operation. A number of details were simplified on the remaining structures that were deemed to have insignificant impact on flow profiles. Simplified fuel assemblies were used, where individual fuel rods in fuel assemblies were represented by an equivalent assembly model.

Abstract. New smart materials have been developed in material science, that are suitable for mechatronic applications. Modern mechatronic systems are focusing on minimizing size, active control and low energy consumption. All this attributes can be incorporated into term Micro Electro Mechanical Systems (MEMS). To improve performance of MEMS system, new materials and technologies are developed -one of them, which found broad application usage is Functionally graded material (FGM). MEMS application usually contains multilayer structure and in some application class MEMS systems contain piezoelectric layers. Piezoelectric structures offer facilities to make motions. Piezoelectric layers can be also used to damp vibrations as an active damping or as an active sensor. For better understanding these multiphysical problems new mathematical models are developed. The paper deals with finite beam element with piezoelectric layers and functionally graded material of core. In the paper homogenization of FGM material properties and homogenization of core and piezoelectric layers is presented. In the process of homogenization direct integration method and multilayer method is used. There is also presented the derivation of individual submatrices of local stiffness and mass matrix, where concept of transfer constants is used. Functionality of new FGM finite beam with piezoelectric layers is presented by numerical experiments.