Dynamic Interaction between the Coolant Flow and Fuel Assembly

Vestnik MEI ◽  
2019 ◽  
Vol 5 ◽  
pp. 11-23
Author(s):  
Konstantin N. Proskuryakov ◽  
Keyword(s):  
Fuel Assembly ◽  
Dynamic Interaction ◽  
Coolant Flow

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 REGULARITIES OF FORMATION OF FLOW OF COOLANT BEHIND THE TVS-KVADRAT MIXING SPACING GRID OF THE PWR-TYPE REACTOR

2018 ◽  
Vol 61 (3) ◽  
pp. 258-268
Author(s):  
S. M. Dmitriev ◽  
D. V. Doronkov ◽  
M. A. Legchanov ◽  
V. D. Sorokin ◽  
A. E. Khrobostov
Keyword(s):  
Fuel Assembly ◽  
Experimental Model ◽  
Experimental Studies ◽  
Coolant Flow ◽  
Hydrodynamic Characteristics ◽  
Experimental Investigations ◽  
Thermal Reliability ◽  
Flow Rates ◽  
Type Reactor ◽  
Active Zones

Tgratinghis paper presents the results of experimental investigations of the influence of mixing spacer gratings with different types of deflectors on the coolant flow in the TVSKvadrat fuel assembly of the PWR-type reactor. Experimental model of the TVS-Kvadrat of the PWR reactor was made in complete geometric similarity with the full-scale cassettes. Studies were carried out by modeling the flow of coolant in the core with the use of an experimental stand; the latter was an aerodynamic open loop through which air is pumped. To measure the local hydrodynamic characteristics of the coolant flow, special pneumatic sensors were used that were able to measure the full velocity vector at the point by its three components. During the studies of the local fluid dynamics of the coolant, the transverse flow rates were measured; also, the coolant flow rates were measured by cells of the TVS-Kvadrat experimental model. The analysis of the spatial distribution of the projections of the absolute flow velocity made it possible to detail the pattern of the coolant flow behind the mixing spacing gratings with different variants of the deflector design, as well as to choose the deflector of the optimal design. Accumulated data base on the flow of the coolant in the TVS-Kvadrat fuel assembly formed the basis of the engineering justification of the structures of the active zones of PWR reactors. Guidelines for choosing optimal designs mixing spacing grids have been considered by designers of the “Afrikantov OKBM” JSC when they created implementations of the latest TVS-Kvadrat assemblies. The results of experimental studies are used to verify CFD-codes of both foreign and domestic origin, as well as the programs for detailed cell-by-cell calculation of active zones in order to reduce conservatism in the justification of thermal reliability.

Benchmarking the NEM Real-Time Core Model for VVER-1000 Simulator Application: Asymmetric Core

2016 ◽  
Author(s):  
Emiliya Georgieva ◽  
Yavor Dinkov ◽  
Kostadin Ivanov ◽  
Robert Stieglitz
Keyword(s):  
Fuel Assembly ◽  
Real Time ◽  
Monitoring System ◽  
Pressure Loss ◽  
Reactor Core ◽  
Coolant Flow ◽  
Core Model ◽  
Secondary Sources ◽  
Main Circulation ◽  
Simulation Results

A real-time version of the Nodal Expansion Method (NEM) code is developed and implemented into Kozloduy 6 full-scope replica control room simulator. Combined with an enhanced thermal-hydraulics and I&C models the whole package is a high-fidelity simulation tool for operator training and various other applications. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. The transient of ‘Switching-off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power’ as described in OECD/NEA Kalinin 3 Coolant Transient Benchmark is an example of an asymmetric core scenario with a range of parameter changes. Simulation results concerning fuel assembly power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system. Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superfluous description of the reference unit. In such a case, an approach based on a ‘generic’ V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion. One example is core coolant flow and pressure loss during the transient. Pump head and pressure loss across reactor vessel are measured and recorded and in-core monitoring system provides estimation of core coolant flow, which is quite high in comparison with some other V320 units (e.g. by about 5 % larger). Without more detailed pressure loss data across the main circulation loop and specific pump characteristics, however, one can only guess how much simulation is off the mark. Another detail of the same problem is coolant flow through a specific fuel assembly. The presence of a fuel assembly of different design (TVS-M type) surrounded by TVSA type fuel assemblies shall be thoroughly considered, because secondary sources indicate significant differences in fuel assembly pressure loss coefficients between the two types. Coolant flow affects coolant (and fuel) temperature profile and thus neutron cross-sections. Yet another example, even more strongly affecting our ability to interpret simulation results is core power reconstruction provided by the in-core monitoring system of the unit. The SPND (Self-Powered Neutron Detector) current readings are subject of conversion by an algorithm based upon simulated spatial neutron flux distribution across the reactor core. While error estimation of the parameters in stationary conditions is available from secondary sources, there is no reliable estimation of error magnitude during the transient.

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.

Dynamic characteristics of the mockup of the fuel assembly in coolant flow

2012 ◽  
Vol 41 (2) ◽  
pp. 168-171 ◽  
Author(s):  
Yu. N. Drozdov ◽  
V. V. Makarov ◽  
A. V. Afanas’ev ◽  
I. V. Matvienko ◽  
E. P. Osipova ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Dynamic Characteristics ◽  
Coolant Flow

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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