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 MIXED CORE SUBCHANNEL MODEL FOR SUBCAL AND COUPLING WITH ANDREA

2018 ◽  
Vol 19 ◽  
pp. 1
Author(s):  
Vojtěch Caha ◽  
Jiří Čížek
Keyword(s):  
Flow Distribution ◽  
Coolant Flow ◽  
The Other ◽  
Hydraulic Characteristics ◽  
Axial Coordinate ◽  
Fuel Assemblies ◽  
Homogeneous Core ◽  
Full Core

This paper presents the results of an analysis of flow distribution in VVER-1000 mixed core consisting of fuel assemblies with non-identical spacing grids. The calculation was carried out using the modified subchannel code SUBCAL-AZ which allows to calculate 3D thermal-hydraulic characteristics of the coolant flow in the full core subchannel model coupled with the neutron-physical code ANDREA. This full core subchannel model was created in three variants depending on the ANDREA calculations. The first variant (homogeneous core) consisted of 163 hydraulically identical fuel assemblies TVSA-T mod.2, whereas the other variants (mixed cores) consisted of fuel assemblies TVSA-T mod.0, mod.1 and mod.2. These fuel assemblies mainly differ in types, number and axial coordinate of spacing grids and also in diameter of guide tubes. The influence of mixed core to flow distribution was obtained by comparing these variants.

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 Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Jingwen Yan ◽  
Yuxiang Zhang ◽  
Baowen Yang ◽  
Weicai Li ◽  
Yuemin Zhou
Keyword(s):  
Fuel Assembly ◽  
Cross Flow ◽  
Flow Distribution ◽  
Hydraulic Performance ◽  
Computational Domain ◽  
Spacer Grid ◽  
Typical Structure ◽  
Mixing Performance ◽  
Ansys Cfx ◽  
Fuel Assemblies

The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD) code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly.

Development of a Multiphysics Approach With ANSYS CFX and PARCS for Local Analysis of Neutron Oscillation

2017 ◽  
Author(s):  
Zhuoqi Du ◽  
Marcus Seidl ◽  
Rafael Macián-Juan
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Noise Analysis ◽  
Flow Distribution ◽  
Local Analysis ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Ansys Cfx ◽  
Neutron Noise

Neutron noise analysis has been done over the decades to predict fuel assembly vibrations and to evaluate safety related issues. Neutron noise occurs due to several reasons: the vibration of the fuel rods, flow obstacles such as rod bending and crud deposition, the moderator temperature and time dependent changes caused by varying flow distributions within a fuel assembly, etc. In order to have a better insight of the neutron noise, a fluid mechanics, structural and neutronics coupled code is developed to perform detailed multiphysics simulations at the level of the fuel rods inside a fuel assembly. In this paper the coupling routine of both steady state and transient calculation is described and the outcome is discussed under several scenarios to understand the influence of rod vibration, moderator temperature and flow distribution on the neutronic field. This paper presents the methodology to couple the multiphysics Computational Fluid Dynamics (CFD) code ANSYS-CFX 16.0 with the 3D neutron diffusion code PARCS v3.0. The model for a 16×16 Pressurized Water Reactor (PWR) fuel assembly is set up for ANSYS-CFX. A sensitivity analysis is carried out to obtain the optimal mesh parameters which results in a good accuracy, as well as a small need for computation capability. Transient cases are studied on a quarter fuel assembly applying oscillating moderator inlet boundary conditions in which the inlet moderator temperature and the inlet moderator velocity are varying over time. In order to simulate the vibration of the fuel rod, the fuel rod part is implemented as immersed solid in ANSYS-CFX. Different vibration modes are applied to both cases: individual single rods of the fuel assembly, and all rods of the fuel assembly. The results of each case are shown in this paper giving a better understanding of how axial power distribution develops with varying flow conditions and vibrating fuel rods.

CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

2006 ◽  
Author(s):  
Sa´ndor To´th ◽  
Ga´bor Le´gra´di ◽  
Attila Aszo´di
Keyword(s):  
Flow Field ◽  
Fuel Assembly ◽  
Nuclear Reactors ◽  
Turbulence Models ◽  
Sensitivity Study ◽  
Cfd Analysis ◽  
Type Reactor ◽  
Ansys Cfx ◽  
Fuel Assemblies ◽  
Fuel Pin

From the aspect of planning the power upgrading of nuclear reactors — including the VVER-440 type reactor — it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960 mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper.

Investigation of Mixing Flow Process Using PIV and PLIF Techniques

2013 ◽  
Vol 816-817 ◽  
pp. 1054-1058
Author(s):  
Ezddin Hutli ◽  
Dániel Tar ◽  
Valer Gottlasz ◽  
Gyorgy Ezsol
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Cross Section ◽  
Temperature Difference ◽  
Power Distribution ◽  
Coolant Temperature ◽  
Mixing Process ◽  
Flow Process ◽  
Flow Rates ◽  
Rod Bundle

A coolant mixing investigation in a head of a half-size model of VVER-440 fuel assembly (simulator) has been performed at KFKI. The PIV and PLIF measurements have been done under a selected list of power distribution options, flow rates and powers. The experiments were focused on obtaining a data for investigating the trends in temperature difference between the value registered by a thermocouple and that obtained using PLIF technique. The coolant temperature distribution has been measured in many positions along the coolant trajectory and where coolant flow leaves the rod bundle and in the cross section location of thermocouple, thus the dynamics of effect of mixing process is also declared. PIV and LPIF results show their ability to verify the primary results of CFD calculations.

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.

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.

scholarly journals MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE

2021 ◽  
Vol 247 ◽  
pp. 02028
Author(s):  
Wojciech Rydlewicz ◽  
Emil Fridman ◽  
Eugene Shwageraus
Keyword(s):  
Monte Carlo ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fast Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Complex Core ◽  
Nodal Diffusion ◽  
Full Core

This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.

scholarly journals KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR

2021 ◽  
Vol 247 ◽  
pp. 10020
Author(s):  
Dongyong Wang ◽  
Yingrui Yu ◽  
Xingjie Peng ◽  
Chenlin Wang ◽  
Kun Liu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Transport Equation ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Calculation Ability

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.

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