CFD Validation of Void Distribution in a Rod Bundle With Spacer

2014 ◽  
Author(s):  
Kevin Goodheart ◽  
Arto Ylönen ◽  
Victor De Cacqueray ◽  
Horst-Michael Prasser
Keyword(s):  
Nuclear Industry ◽  
Wire Mesh ◽  
Cfd Validation ◽  
Cfd Models ◽  
Rod Bundle ◽  
Void Distribution ◽  
Test Loop ◽  
Cfd Method ◽  
Stability And Accuracy ◽  
Paul Scherrer

As CFD (Computational Fluid Dynamics) continues to grow in the nuclear industry the need for validation is essential to this growth. The focus of this paper is to highlight the validation of AREVA’s 2-Phase CFD models in a rod bundle with spacer. The experimental work is based on void distribution measurements using a 64×64 wire mesh sensor at the Paul Scherrer Institute adiabatic test loop fuel rod bundle (SUBFLOW). The 2-Phase models are based on the Eulerian multiphase framework where the interaction models are further developed using user routines in a commercial CFD tool to achieve stability and accuracy in a rod bundle configuration with a spacer. The 2-Phase CFD models coincide with the SUBFLOW experiments in showing the effect of void collecting in the center of the sub-channels slightly downstream of the spacer and further downstream the smaller bubbles migrate toward the rod surfaces whereas the larger bubbles stay in the sub-channels. Good agreement is achieved between CFD and void distribution experiments. The multi-phase CFD method was used by AREVA to improve the performance of the new products GAIA and ATRIUM™ 11.

Transient Boiling and Cross Flow in 5×5 Rod Bundle With Rapid Heating

2018 ◽  
Author(s):  
Hiroki Takiguchi ◽  
Masahiro Furuya ◽  
Takahiro Arai ◽  
Kenetsu Shirakawa
Keyword(s):  
Fuel Assembly ◽  
Heat Generation ◽  
Cross Flow ◽  
Nuclear Safety ◽  
Wire Mesh ◽  
Coolant Temperature ◽  
Heat Output ◽  
Rod Bundle ◽  
Test Loop ◽  
The Cross

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped. This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 × 5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40°C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

scholarly journals Computational Fluid Dynamics Model of Wells Turbine for Oscillating Water Column System: A Review

2021 ◽  
Vol 2053 (1) ◽  
pp. 012013
Author(s):  
N. Abdul Settar ◽  
S. Sarip ◽  
H.M. Kaidi
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Water Column ◽  
Cfd Simulation ◽  
Oscillating Water Column ◽  
Wells Turbine ◽  
Flow Problems ◽  
Cfd Models ◽  
Column System ◽  
Cfd Method

Abstract Wells turbine is an important component in the oscillating water column (OWC) system. Thus, many researchers tend to improve the performance via experiment or computational fluid dynamics (CFD) simulation, which is cheaper. As the CFD method becomes more popular, the lack of evidence to support the parameters used during the CFD simulation becomes a big issue. This paper aims to review the CFD models applied to the Wells turbine for the OWC system. Journal papers from the past ten years were summarized in brief critique. As a summary, the FLUENT and CFX software are mostly used to simulate the Wells turbine flow problems while SST k-ω turbulence model is the widely used model. A grid independence test is essential when doing CFD simulation. In conclusion, this review paper can show the research gap for CFD simulation and can reduce the time in selecting suitable parameters when involving simulation in the Wells turbine.

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.

Structural and Hydraulic Design of Mechanical Pump Using in LBE Test Loop

2012 ◽  
Vol 588-589 ◽  
pp. 190-193 ◽  
Author(s):  
Bo Gao ◽  
Min Guan Yang ◽  
Xin Kai Sun ◽  
Ning Zhang
Keyword(s):  
Design Process ◽  
Hydraulic Model ◽  
Future Design ◽  
Model Based ◽  
Hydraulic Design ◽  
Test Loop ◽  
The Future ◽  
Mechanical Pump ◽  
Cfd Method ◽  
Design Ideas

In order to design a mechanical pump that can satisfy the special requirements of the LBE test loop, structural and hydraulic design ideas were discussed in this paper. A new vertical centrifugal submerged pump was proposed, including installation and hydraulic model. Based on the provided parameter, hydraulic design of the pump has been done by CFD method. Velocity caused erosion problem was considered primarily in the design process. It is helpful for the future design of pumps in various loops and ADS.

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.

Post-Test CFD Analysis of Non-Uniformly Heated 19-Pin Fuel Bundle Cooled by HLM

2018 ◽  
Author(s):  
R. Marinari ◽  
I. Di Piazza ◽  
M. Angelucci ◽  
D. Martelli
Keyword(s):  
Mixed Convection ◽  
Large Scale ◽  
Fluid Dynamic ◽  
Cfd Validation ◽  
Cfd Models ◽  
Fuel Bundle ◽  
Post Test ◽  
Fuel Pin ◽  
Experimental Facilities ◽  
Gen Iv

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the Front-End Engineering Design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments. The MYRRHA-19 like Fuel Pin Bundle Simulator installed in the NACIE-UP facility allows to make non-uniform and dissymmetric tests with only a few pins heated. This technical feature of the FPS is very interesting for CFD validation and this kind of data tests in HLM fuel bundles are not so common in the literature. In the present paper, a post-test validation is made by a detailed CFD model of the test section. Experimental data, statistically treated by the error propagation theory, are briefly presented and a preliminary comparison with CFD results using different models/turbulent Prandtl numbers are shown. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.

The Applicability of NFI-1 DNB Correlation and Fluid-to-Fluid Similarities to Freon DNB Test

2008 ◽  
Author(s):  
Shumpei Kakinoki ◽  
Keizo Matsuura ◽  
Kenichi Kitagawa ◽  
Isao Kataoka
Keyword(s):  
Heat Flux ◽  
Equivalent Condition ◽  
High Heat Flux ◽  
Diffusivity Coefficient ◽  
Test Results ◽  
Hydraulic Test ◽  
Rod Bundle ◽  
Test Loop ◽  
R 134A ◽  
Subchannel Analysis

Freon thermal hydraulic test is expected to be one of the workable methods to develop high thermal hydraulic performance PWR fuel. That is, high pressure water and high heat flux condition in PWR core can be substituted with lower pressure Freon and lower heat flux by applying appropriate fluid-to-fluid similarity and modeling parameters. Freon DNB tests and mixing tests were carried out against a 4×4 rod bundle configuration where R-134A flowed vertically upwardly. The tests were carried out at Freon thermal hydraulic test loop in Korea Atomic Energy Research Institute (KAERI). The spacer grid used in these tests was modeled on that of conventional PWR fuel, that is, square lattice grid with split type mixing vanes. Diameter of heater rod simulating PWR fuel rod is about 10.7mm and heating length is about 2000 mm. Freon mixing tests were carried out to estimate Turbulence Diffusivity Coefficient (TDC), which was normally used in conventional thermal hydraulic design of nuclear reactor. Freon CHF test results showed that parametric trends agreed with those of existing CHF data. To predict CHF of 4×4 rod bundle, subchannel analysis code Modified COBRA-3C and NFI-1 DNB correlation were applied. TDC value used in subchannel analysis was determined by fitting Freon mixing test data. NFI-1 DNB correlation was developed for predicting DNB heat flux in rod bundle configuration by using water CHF test results at HTRF test loop at Columbia University. The design of spacer grids used in KAERI Freon DNB test was similar to that used in water CHF test at HTRF. Water equivalent flow condition of this R-134A test was estimated using fluid-to-fluid similarities. NFI-1 DNB correlation was applied to this water equivalent condition to estimate water equivalent DNB heat flux. Then R-134A equivalent DNB heat flux was estimated reversely, and compared to Freon DNB test result. The test results were predicted well and applicability of NFI-1 DNB correlation and fluid-to-fluid similarities in 4×4 rod bundle is discussed.

Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution

2006 ◽  
Author(s):  
Fatih Aydogan ◽  
Lawrence E. Hochreiter ◽  
Kostadin Ivanov
Keyword(s):  
Void Fraction ◽  
Channel Model ◽  
Test Facility ◽  
Cooling Power ◽  
Distribution Data ◽  
Ct Scanner ◽  
Benchmark Data ◽  
Fine Mesh ◽  
Rod Bundle ◽  
Void Distribution

Good quality experimental data is needed to refine the thermal hydraulic models for the prediction of rod bundle void distribution and critical heat flux (CHF) or dry-out. The Nuclear Power Engineering Corporation (NUPEC) has provided a valuable database to evaluate the thermal hydraulic codes [1]. Part of this database was selected for the NUPEC BWR Full-size Fine-Mesh Bundle Tests (BFBT) benchmark sponsored by US NRC, METI-Japan, NEA/OECD and Nuclear Engineering Program of the Pennsylvania State University (PSU). Twenty-five organizations from ten countries have confirmed their intention to participate and will provide code predictions to be compared to the measured data for a series of defined exercises within the framework of the BFBT benchmark. This benchmark data includes both the fine-mesh high quality sub-channel void fraction and critical power data. Using a full BWR rod bundle test facility, the void distribution was measured at mesh sizes smaller than the sub-channel by using a state-of-the-art computer tomography (CT) technology [1]. Experiments were performed for different pressures, flow rates, exit qualities, inlet sub-cooling, power distributions, spacer types and assembly designs. There are microscopic and sub-channel averaged void fraction data from the CT scanner at the bundle exit as well as X-ray densitometer void distribution data at different elevation levels in the rod bundle. Each sub-channel’s loss coefficient was calculated with using the Rehme method [2,3], and a COBRA-TF sub-channel model was developed for the NUPEC facility. The BWR assembly that was modeled with COBRA-TF includes two water rods at the center. The predicted sub-channel void fraction values from COBRA-TF are compared with the bundle exit void fraction values measured using the CT-scanner void fraction from the BFBT benchmark data. Different plots are used to examine the code prediction of the void distribution at a sub-channel level for the different sub-channels within the bundle.

CFD analysis of a void distribution benchmark in a rod bundle

Kerntechnik ◽  
2013 ◽  
Vol 78 (1) ◽  
pp. 38-42
Author(s):  
E. Krepper ◽  
R. Rzehak
Keyword(s):  
Cfd Analysis ◽  
Rod Bundle ◽  
Void Distribution
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