scholarly journals Numerical Investigation on the Flow Characteristics in a 17 × 17 Full-Scale Fuel Assembly

Energies ◽  
2020 ◽  
Vol 13 (2) ◽  
pp. 397 ◽  
Author(s):  
Zihao Tian ◽  
Lixin Yang ◽  
Shuang Han ◽  
Xiaofei Yuan ◽  
Hongyan Lu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Flow Characteristics ◽  
Full Scale ◽  
Pressurized Water Reactor ◽  
Lateral Velocity ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Real Size ◽  
Cfd Method

In a previous study, several computational fluid dynamics (CFD) simulations of fuel assembly thermal-hydraulic problems were presented that contained fewer fuel rods, such as 3 × 3 and 5 × 5, due to limited computer capacity. However, a typical AFA-3G fuel assembly consists of 17 × 17 rods. The pressure drop levels and flow details in the whole fuel assembly, and even in the pressurized water reactor (PWR), are not available. Hence, an appropriate CFD method for a full-scale 17 × 17 fuel assembly was the focus of this study. The spacer grids with mixing vanes, springs, and dimples were considered. The polyhedral and extruded mesh was generated using Star-CCM+ software and the total mesh number was about 200 million. The axial and lateral velocity distribution in the sub-channels was investigated. The pressure distribution downstream of different spacer grids were also obtained. As a result, an appropriate method for full-scale rod bundle simulations was obtained. The CFD analysis of thermal-hydraulic problems in a reactor coolant system can be widely conducted by using real-size fuel assembly models.

An Estimation of the Dynamic Buckling Load for the Spacer Grid of Pressurized Water Reactor Fuel Assembly

2006 ◽  
Vol 326-328 ◽  
pp. 1603-1606 ◽  
Author(s):  
Sang Youn Jeon ◽  
Young Shin Lee
Keyword(s):  
Fuel Assembly ◽  
Dynamic Buckling ◽  
Buckling Load ◽  
Estimation Methods ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Water Reactor ◽  
Spacer Grids ◽  
The Impact

This study contains an estimation of the dynamic buckling load for the spacer grid of fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the dynamic buckling loads of spacer grid. The dynamic impact tests and analyses were performed to evaluate the impact characteristics of the spacer grids and to predict the dynamic buckling load of the full size spacer grid. The estimation results were compared with the test results for the verification of the estimation methods.

scholarly journals Numerical and experimental investigation of the water flow through PWR spacer grids

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Higor Fabiano Pereira de Castro ◽  
Guilherme Augusto Moura Vidal ◽  
Tiago Augusto Santiago Vieira ◽  
Vitor Vasconcelos Araújo Silva ◽  
Daniel De Almeida Magalhães Campolina ◽  
...  
Keyword(s):  
Water Flow ◽  
Technology Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Ansys Cfx ◽  
Flow Through ◽  
Belo Horizonte ◽  
Main Components

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.

Confinement Dependency on Added Stiffness Effect on a Fuel Assembly Under Axial Flow

2017 ◽  
Author(s):  
G. Ricciardi
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Axial Flow ◽  
Dynamical Behaviour ◽  
Seismic Excitation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Spacer Grids ◽  
The Core ◽  
Fuel Assemblies

Safety measures are required to insure the drop of control rods and that the core is cooled when the fuel assemblies of a Pressurized Water Reactor (PWR) are subjected to a seismic excitation. A way to insure these two criteria is to prevent the spacer grids from buckling. The reactor core made of fuel assemblies is subjected to an axial water flow to cool the reactor. The flow modifies the dynamical behaviour of the fuel assemblies. Tests made on a real fuel assembly highlighted an added stiffness effect under axial flow. In previous studies simulations were compared to experiment involving by-passes significantly larger than the distance between two fuel assemblies in a PWR core. Thus, one could wonder if the observations made on a fuel assembly with large by-passes are representative of core geometry. Simulations using a fluid-structure model of the core to a seismic excitation have been proposed. A parametric study has been conducted to observe the effect of confinement on the added stiffness effect for several confinements and bulk velocities. Simulations showed that the added stiffness reaches a maximum for a confinement around 20 mm, and that the added stiffness should be negligible in a real core configuration.

Performances Analyses of Crud of EPR Fuel Assembly

2017 ◽  
Author(s):  
Xuming Wang ◽  
Cenxi Yuan ◽  
Chen Ye
Keyword(s):  
Risk Assessment ◽  
Fuel Assembly ◽  
Low Risk ◽  
Third Generation ◽  
Pressurized Water Reactor ◽  
Worst Case ◽  
Pressurized Water ◽  
Water Reactor ◽  
Advanced Fuel ◽  
New Type

Taishan European Pressurized Water Reactor (EPR) is a third generation advanced pressurized water reactor (PWR), which adopts the third generation advanced fuel assembly (AFA-3G-LE) from AREVA for the first time. As suggested by American Electric Power Research Institute (EPRI), an EPRI level III crud risk assessment is necessary for new type of plants. Because crud induced power offset (CIPS) and crud induced local corrosion (CILC) can lead to axial offset anomaly (AOA) and fuel cladding failure, respectively. A EPRI level III CIPS/CILC risk assessment for Taishan EPR is performed with a new framework of simulation by using sub-channel code FLICA, crud code BOA, and Monte Carlo transport code Tripoli-4. Such framework enables a self-consistent calculation, including a detailed description on neutronics contributed by boron. The validation of present work is confirmed because of the good agreement with the experienced data of EPRI. The results show that AFA-3G-LE has a good performance on crud risk assessment. Even in the worst case, the boron-10 deposition (2.6 g) and the maximum thickness of crud (59 μm) are lower than the low risk threshold, 31.33 g and 75 μm, respectively. Hence, It is expected that Taishan EPR has a very low risk on CIPS and CILC.

Numerical Prediction of Void Distribution in Two-Phase Turbulent Flow in a Sub-Channel

2009 ◽  
Author(s):  
Tsutomu Ikeno ◽  
Tatsuya Sasakawa ◽  
Isao Kataoka
Keyword(s):  
Turbulent Flow ◽  
Transport Equation ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Simulation Code ◽  
Pressurized Water ◽  
Void Fractions ◽  
Rod Bundle ◽  
Void Distribution

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.

Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

Advanced Alternatives of VVER-440 Fuel Assembly

2014 ◽  
Author(s):  
Juraj Tomaškovič ◽  
Petr Dařílek ◽  
Radoslav Zajac ◽  
Vladimír Nečas
Keyword(s):  
Fuel Assembly ◽  
Economic Efficiency ◽  
Fuel Cycle ◽  
Gradual Increase ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Burn ◽  
Fuel Pin ◽  
Basic Characteristics

The main goals of fuel development for pressurized water reactor are effectiveness and economic efficiency. Both requirements can be achieved by gradual increase of discharged fuel burn-up and prolongation of fuel cycle. The mentioned effects can be reached by optimisation of fuel assembly profiling, fuel enrichment raise, and by parasitic absorption reduction. These methods were used in VVER-440 fuel assembly optimisation, described in this paper. Fuel pin configurations with enrichment limit 5 % and also enlarged one up to 5.95 % U235 were designed. Reduction of parasitic absorption was limited by carcass frame of the assembly. Basic characteristics of the best assembly proposals are presented and effects on equilibrium fuel cycle of VVER-440 reactor are characterized.

Improvement of the CATHARE 3D Code Prediction on PIERO Transient

2008 ◽  
Author(s):  
Tony Glantz ◽  
Roberto Freitas
Keyword(s):  
Direct Application ◽  
Full Scale ◽  
The Other ◽  
Second Step ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Lower Head ◽  
Scale Modelling

The PIERO experiment has been carried out to study phase’s separation in the lower plenum and the downcomer of a Pressurized Water Reactor (PWR) during the end of the depressurization of a large break loss of coolant accident (LB-LOCA). This experiment has been used for the validation and assessment of the 3D module of CATHARE code [1] but the results are not good because of an overestimation of the liquid entrainment in the lower plenum in one hand and the use of a coarsed meshing for modelling the PIERO experiment in the other hand. Two ways of improvement are possible: the first one and the most complicated is to introduce a stratification model in the 3D module of CATHARE. The other one is a possibility to use a refining meshing in order to simulate PIERO experiment. This second way has been performed and the computations results are greatly improve. Nevertheless, PIERO experiment is not on a reactor scale and a direct application of the meshing recommendations made on PIERO is impossible to translate directly on the reactor case. So, the strategy of validation applied to the reactor case consisted in reproducing a PIERO transient with a full scale lower plenum in a first step. In a second step, a converged meshing for the full scale modelling has been determined. In a last step, results obtained with this kind of modelling have been validated via two correlations developed by Wallis and al., that define boundaries conditions between which the water level remaining in the lower head is allowed to vary. This strategy of validation led to model the reactor’s lower plenum with the more axial meshes in order to have good results.

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