CFD Analysis of Downcomer of Nuclear Reactor VVER 440

Kutiš, Vladimı́r; Jakubec, Jakub; Paulech, Juraj; Gálik, Gálik; Sedlár, Tibor

doi:10.1515/scjme-2016-0018

Cited by 10 publications

(8 citation statements)

References 4 publications

(2 reference statements)

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“…Coolant temperature measured by the thermocouple that is placed at the outlet of the fuel assembly in the reactor part, protective tube unit, could be slightly different from the average coolant temperature at the outlet. This could be caused by nuclear radiation heating of the thermocouple and of course by poor coolant mixing in the upper part of the fuel assembly [5]. This is the reason why the Kurchatov Institute built a test facility to examine the processes affecting mixing processes such as bypass and central tube flow.…”

Section: Cfd Simulations and Resultsmentioning

confidence: 99%

Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

Jakubec

Paulech

Kutiš

et al. 2017

Strojnícky Casopis – Journal of Mechanical Engineering

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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).

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Section: Cfd Simulations and Resultsmentioning

confidence: 99%

Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

Jakubec

Paulech

Kutiš

et al. 2017

Strojnícky Casopis – Journal of Mechanical Engineering

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“…The dry cask storage building represents mid to long term solution for the storage of depleted nuclear fuel assemblies. Individual Fuel assemblies [1] of reactor VVER440 [2] are bundled in hermetically sealed steel containers (casks). These casks are stored in a concrete air channel, where their decay heat is cooled by buoyancy driven air flow.…”

Section: Introductionmentioning

confidence: 99%

CFD Analysis of Dry Cask Nuclear Fuel Storage

Kutiš

Gálik

Paulech

et al. 2019

Strojnícky Časopis - Journal of Mechanical Engineering

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“…For this reason, the research of helium cooling loop systems is very topical. The research activities in the given area from several aspects are documented in the various publications, as for example [1,2,3,4,5].…”

Section: Introductionmentioning

confidence: 99%

Hydrodynamic Properties of High Temperature Natural Circulating Helium Cooling Loop

Dzianik

Gužela

2017

Strojnícky Casopis – Journal of Mechanical Engineering

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Abstract:The paper deals with the hydrodynamic properties, i.e. the consumption of mechanical energy expressed by pressure drops within a helium loop intended for the testing of decay heat removal (DHR) from the model of a gas-cooled fast reactor (GFR). The system is characterised by the natural circulation of helium, as a coolant, and assume steady operating conditions of circulation. The helium loop consists of four main components: model of gas-cooled fast reactor, model of the heat exchanger for decay heat removal, hot piping branch and cold piping branch. Using the process hydrodynamic calculations, the pressure drops of circulating helium within the main components of the helium loop were determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.

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CFD Analysis of Downcomer of Nuclear Reactor VVER 440

Cited by 10 publications

References 4 publications

Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

CFD Analysis of Dry Cask Nuclear Fuel Storage

Hydrodynamic Properties of High Temperature Natural Circulating Helium Cooling Loop

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