scholarly journals Analysis of coolant flow at the outlet of the physical model of the fuel assembly II. of the nuclear reactor VVER 440 / V213

2020 ◽  
Vol 328 ◽  
pp. 01010
Author(s):  
Peter Mlynár ◽  
František Világi ◽  
Zdenko Závodný ◽  
František Urban ◽  
František Ridzoň
Keyword(s):  
Velocity Profile ◽  
Fuel Assembly ◽  
Physical Model ◽  
Nuclear Reactor ◽  
Coolant Flow ◽  
Coolant Temperature ◽  
Cfd Simulations ◽  
Central Tube ◽  
Fuel Assemblies ◽  
The Mean

To safely and efficiently load the fuel assemblies of the VVER 440 / V 213 nuclear reactor, the relation between the temperature of the coolant at the outlet of the fuel assembly, measured by a thermocouple in the assembly’s axis, and the mean coolant temperature, present in the plane of the thermocouple, must be analysed. Based on the analysis of the coolant flow at the output of the physical model of the fuel assembly I. [1] and published CFD simulations [2,3,4] it was shown, that a special attention has to be paid to the influence of the water flow in the central tube on the temperature and velocity profile of the coolant at the thermocouple’s plane in the fuel assembly. For this reason, an experimental device with a physical model of the fuel assembly II. of the nuclear reactor VVER 440 / V 213 was designed, manufactured, and operated at the Faculty of Mechanical Engineering STU in Bratislava.

scholarly journals LATERAL COOLANT FLOW BETWEEN FUEL ASSEMBLIES IN MIXED CORES

2018 ◽  
Vol 14 ◽  
pp. 1
Author(s):  
Vojtech Caha ◽  
Jiří Čížek
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Coolant Flow ◽  
Coolant Temperature ◽  
Hydraulic Characteristics ◽  
Axial Coordinate ◽  
Comparison Of Results ◽  
Ansys Cfx ◽  
Fuel Assemblies ◽  
Fuel Reload

This paper presents the results of an analysis of lateral coolant flow between adjacent fuel assemblies with non-identical spacing grids in a mixed core consisting of TVSA-T mod.1 and TVSA-T mod.2 fuel assemblies. The calculation was carried out using modified subchannel code SUBCAL which allows to calculate 3D thermo-hydraulic characteristics of the coolant flow in the full three fuel assemblies model. This full three fuel assemblies model was created in two variants. The first variant consisted of three hydraulically identical fuel assemblies TVSA-T mod.1, whereas the second variant consisted of two fuel assemblies TVSA-T mod.1 and one fuel assembly TVSA-T mod.2 which mainly differ in types, number and axial coordinate of spacing grids and also in diameter of guide tubes. The influence of mixed core to lateral coolant flow and hence coolant temperature was obtained by comparing these two variants. The power distribution was taken from presumed mixed core fuel reload calculated by macro-code ANDREA. Finally there were also provided a comparison of results achieved by subchannel analysis approach with calculation of similar problem using CFD code ANSYS CFX by TVEL, the fuel supplier.

scholarly journals Introductory measurements on the physical model of the VVER 440 nuclear reactor II. fuel cell assembly

2018 ◽  
Vol 168 ◽  
pp. 06007
Author(s):  
Zdenko Závodný ◽  
Peter Mlynár ◽  
František Urban ◽  
Ján Hollý ◽  
Zoltán Fuszko ◽  
...  
Keyword(s):  
Fuel Cell ◽  
Physical Model ◽  
Nuclear Reactor ◽  
Coolant Temperature ◽  
Cell Assembly ◽  
Qualitative And Quantitative Analysis ◽  
Nuclear Reactor Fuel ◽  
The Mean ◽  
The Relationship ◽  
Coolant Velocity

Qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel cell assembly outlet and the mean coolant temperature profile in the thermocouple plane is required for safe and effective loading of nuclear fuel cells. Physical model of the VVER 440 nuclear reactor fuel cell assembly serves to analyze the influence of the coolant mass flow on the coolant velocity and temperature profiles at the plane of the thermocouple position in the fuel cell assembly.

Analysis of coolant flow in central tube of VVER-440 fuel assemblies

Kerntechnik ◽  
2012 ◽  
Vol 77 (4) ◽  
pp. 258-264 ◽  
Author(s):  
G. Zsíros ◽  
S. Tóth ◽  
A. Aszódi
Keyword(s):  
Coolant Flow ◽  
Central Tube ◽  
Fuel Assemblies

scholarly journals CFD Analysis of Downcomer of Nuclear Reactor VVER 440

2016 ◽  
Vol 66 (2) ◽  
pp. 55-62
Author(s):  
Vladimír Kutiš ◽  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Gálik Gálik ◽  
Tibor Sedlár
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Flow ◽  
Orifice Diameter ◽  
Cfd Analysis ◽  
Flow Conditions ◽  
Fuel Assemblies ◽  
Flow Through ◽  
Cfd Analyses

Abstract 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.

PWR Fuel Assembly Modal Testing and Analysis

2003 ◽  
Author(s):  
Bruno Collard ◽  
Ste´phane Pisapia ◽  
Sergio Bellizzi ◽  
Fre´de´ric Witters
Keyword(s):  
Fuel Assembly ◽  
Flow Rate ◽  
Nuclear Reactor ◽  
Experimental Tests ◽  
Mechanical Tests ◽  
Full Scale ◽  
Modal Testing ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Non Linear

Pressurized Water Reactor (PWR) seismic or Lost Of Coolant Accident (LOCA) loads could result in impacts between nuclear fuel assemblies or between fuel assemblies and the core baffles. Forces generated during these shocks are often the basis for the determination of the maximum loads and of the spacer grid and fuel rod design. The knowledge of the fuel assembly kinematics is essential to compute these maximum loads, and this requires experimental tests. Our study aims at characterizing the behavior of a full-scale fuel assembly subjected to various excitations. The effect of the assembly environment (air, still water and water under flow) is studied. The French Nuclear Reactor Directorate experimental facility HERMES T allows hydraulic and mechanical testing of full-scale fuel assemblies. It is designed for flow rate up to 1200 m3/h and temperature up to 170°C. Specific excitation devices allow mechanical tests with amplitudes of motion up to 20 mm. Laser vibrometry, displacement transducers and tracking camera apparatus measure the fuel assembly displacement. To identify this Multi Degree Of Freedom (MDOF) system (assembly or assembly + fluid), two dependent problems have to be addressed: the linear or non-linear model selection, and the estimation of the corresponding parameters. Under different environments and excitation types, it is shown that the mechanical system is strongly non-linear. The damping term, essentially fluid, increases with flow rate and with motion amplitude, while the stiffness decreases with amplitude. The main results, the measuring and identification methods and the extrapolation to the reactor thermohydraulic conditions are presented and discussed.

scholarly journals Validation of SCALE Model of VVR-M Reactor

2020 ◽  
Vol 18 ◽  
pp. 42-47
Author(s):  
V. I. Borysenko ◽  
◽  
V. V. Goranchuk ◽  
Keyword(s):  
Physical Model ◽  
Nuclear Reactor ◽  
Physical Characteristics ◽  
Multiplication Factor ◽  
The Core ◽  
Fuel Assemblies ◽  
Neutron Multiplication ◽  
Research Nuclear Reactor ◽  
Effective Neutron ◽  
Calculation Code

The peculiarities of development of neutron-physical model of the VVR-M research nuclear reactor in the SCALE calculation code are considered in the article. Models of separate core elements, which influence neutron-physical characteristics of VVR-M, have been developed. Simulation was performed using the CSAS6 control module. Validation of the VVR-M neutron-physical model, built in the SCALE calculation code, has been carried out by comparing the calculated value of the effective neutron multiplication factor with the critical reactor state at the beginning of seven fuel loads with the number of fuel assemblies in the core from 72 to 129. The model is developmed to determine the effective neutron multiplication factor in the reactor, as well as other neutron-physical characteristics, such as neutron spectrum, neutron flux density in various cells of the reactor. Thus, it is possible to conduct numerical experiments to determine the most optimal locations of research channels in the core of the VVR-M, to conduct physical experiments on the irradiation of the research samples, detectors, structural materials, etc. In the article, the simplifications accepted at construction of neutron-physical model of research nuclear reactor VVR-M in SCALE calculation code are presented. The main elements of the model are described: fuel assemblies, beryllium displacer, control rods.

Large Eddy Simulations of a Coolant Flow in Spacer Grid Fuel Assemblies With a Spectral Element Solver

2018 ◽  
Author(s):  
Haomin Yuan ◽  
Vakhtang Makarashvili ◽  
Elia Merzari ◽  
Aleksandr Obabko ◽  
Yiqi Yu
Keyword(s):  
Fuel Assembly ◽  
Spectral Element ◽  
Coolant Flow ◽  
Scale Model ◽  
Tetrahedral Mesh ◽  
Spacer Grid ◽  
Fuel Rod ◽  
Hexahedral Elements ◽  
Large Eddy ◽  
Fuel Assemblies

In this study we used Nek5000, an open-source, high-order spectral element CFD code developed at Argonne National Laboratory (ANL), to model the coolant flow in spacer grids. Two fuel assembly configurations were studied: 2 × 2 and 5 × 5 fuel rod arrangements. The simulations for the 2 × 2 case were based on previous studies, simulating one span of the 2 × 2 fuel rod configuration including a surrogate spacer grid and mixing vane design with typical features of spacers for energy production. Dual periodic boundary conditions were applied in the spanwise direction to take the crossflow into consideration. The study of the 5 × 5 fuel assembly was performed as part of the ANL–Framatome collaboration for advancing computational fluid dynamics (CFD) tools. An advanced numerical model was developed to simulate the experimental setup provided by Framatome. For the 5 × 5 fuel assembly study, two cases of flow geometry were simulated with Nek5000: balanced and unbalanced configurations. In the balanced flow the coolant was entering the fuel rod assembly through 121 uniformly spaced inlet holes arranged in an 11 × 11 matrix. The unbalanced case, on the other hand, featured 14 larger holes placed on only one side of the horizontal plane. Nek5000 accepts only hexahedral meshes, which bring a great challenge to the meshing process for a spacer grid fuel assembly. A tet-to-hex meshing strategy was applied to handle the complex geometric features. A tetrahedral mesh was created first, and then each tetrahedral element was converted into four hexahedral elements. Boundary layers were extruded to fit to the exact geometry. In order to account for transient flow characteristics, the large eddy simulation approach was applied in this study. The employed subgrid-scale model relies on explicit filtering, which has been proven valid for many engineering-scale simulations. We present here the simulation results obtained for both the 2 × 2 and 5 × 5 fuel assemblies.

Modeling the Flow Redistribution Upstream From the Spacer Grid of a PWR

2020 ◽  
Author(s):  
Stanislas de Lambert ◽  
Jérome Cardolaccia ◽  
Vincent Faucher
Keyword(s):  
Fuel Assembly ◽  
Cfd Simulations ◽  
Spacer Grid ◽  
Flow Rates ◽  
Spacer Grids ◽  
Lateral Forces ◽  
Fuel Assemblies ◽  
Hydraulic Network ◽  
Set Up ◽  
Management Issues

Abstract Fuel assemblies’ deformation is an industrial issue that has been first reported in the 90’s. This phenomenon has originally been pointed out for being the explanation of IRI (incomplete rod cluster insertion). Recently, fuel assembly bowing has regained attention for its impact over several core’s management issues, including core neutronics. When deformation occurs, it tends to alter bypasses geometry around the affected fuel assembly. The water gaps’ thicknesses along the assembly’s height does not match the nominal value anymore. As a result, spacer grids can get closer of farther to the surrounding ones. The redistribution between the bypasses and the grid is then involved, depending on the bypasses’ thicknesses and the grid geometry. This unfolding effect entails differences in pressure laterally along a grid, which thus brings about a lateral hydraulic force exerting on the grid. The following paper presents a method to esteem this redistribution thanks to an hydraulic network. Hydraulic resistances can be set up according to the bypass thickness. As a result, both pressure and volumetric flow rates can be calculated to further estimate lateral forces. The approach has been validated with both CFD simulations and an experimental mock-up.

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

2017 ◽  
Vol 67 (1) ◽  
pp. 69-76
Author(s):  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Vladimír Kutiš ◽  
Gabriel Gálik
Keyword(s):  
Steady State ◽  
Temperature Distribution ◽  
Pressure Drop ◽  
Fuel Assembly ◽  
Nuclear Reactor ◽  
Software Tool ◽  
Operating Conditions ◽  
Coolant Flow ◽  
Cfd Modelling ◽  
Ansys Cfx

AbstractThe 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.

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