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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 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.

In this contribution modeling and simulation of surface acoustic waves (SAW) sensor using finite element method will be presented. SAW sensor is made from piezoelectric GaN layer and SiC substrate. Two different analysis types are investigated-modal and transient. Both analyses are only 2D. The goal of modal analysis, is to determine the eigenfrequency of SAW, which is used in following transient analysis. In transient analysis, wave propagation in SAW sensor is investigated. Both analyses were performed using FEM code ANSYS.