Investigation of Rod Vibrations in Droplet Two-Phase Flows

2011 ◽  
Author(s):  
Yoshiteru Komuro ◽  
Zensaku Kawara ◽  
Tomoaki Kunugi
Keyword(s):  
Numerical Simulation ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Motion Tracking ◽  
High Speed Camera ◽  
Two Phase Flows ◽  
Two Phase ◽  
Wet Condition ◽  
Flow Induced Vibrations

Flow-induced vibrations are important problems in nuclear power plants from the view point of reactor safety. In the investigations of these vibrations especially those induced by two-phase flows, a numerical simulation plays a significant role, so it is necessary to obtain the experimental datasets that can validate the results of the numerical simulation. This paper deals with the experimental data of one-end-supported rod vibration, and focuses on the differences between the rod vibrations induced by single-phase air flows and those induced by droplet two-phase flows. In the experiments, the displacement of the non-supported end of the test rod was visualized by the high speed camera with high spatial and temporal resolutions, namely 9.5 μm and 500 μsec. Using an image analyzing software, the rod vibration displacements were measured by the motion tracking method. The curved surface of the rod was observed by another high speed camera and the relationship between the rod vibrations and the wet condition on the surface of the rod was investigated. In addition, the vibrations measured by the strain gages and those by the high speed camera were compared to discuss the differences in these two ways of the measurements.

Experimental and Numerical Simulation on Hydraulic Behavior of Molten Flow in the Lower Part in Reactor Core

2016 ◽  
Author(s):  
Kota Matsuura ◽  
Hideaki Monji ◽  
Susumu Yamashita ◽  
Hiroyuki Yoshida
Keyword(s):  
Numerical Simulation ◽  
Liquid Film ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Behavior ◽  
Reactor Core ◽  
Video Camera ◽  
Working Fluid ◽  
Simulation Code

In the decommissioning work of nuclear power plants, it is important to grasp the sedimentation place of molten materials. However, the technique to grasp exactly sedimentation place is not established now. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is developed. In the study, visualization experiment and numerical simulation were performed to validate the applicability of the JUPITER to the hydraulic relocation behavior in core internals. The test section used in this experiment simulated the structure of the core internals, such as a control rod and a fuel support piece, simply. The working fluid is water under the atmospheric pressure. The experiment uses a high-speed video camera to visualize the flow behavior. The behavior and the speed of the liquid film in a narrow flow channel is obtained. For the numerical analysis carried out prior to the experiment, the behavior of flow down liquid was shown. The typical behavior was also observed that the tip of a liquid film flowing down splits into.

An Overview of Numerical Method for Fluid-Structure Interaction

2014 ◽  
Author(s):  
J.-H. Jeong ◽  
M. Kim ◽  
P. Hughes
Keyword(s):  
Numerical Simulation ◽  
Numerical Methods ◽  
Nuclear Power ◽  
Power Plants ◽  
Fluid Structure Interaction ◽  
Phase Flow ◽  
Two Phase ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Deformable Structure

Fluid-structure interaction (FSI) is the interaction of some movable or deformable structure with an internal or surrounding fluid flow. Therefore, fluid-structure interaction problems are too complex to solve analytically and so they have to be analysed by means of experiments or numerical simulation. This paper provides an overview of numerical methods for fluid-structure interaction evaluation in an draft IAEA technical guideline: large eddy simulation (LES), direct numerical simulation (DNS), Lattice-Boltzmann method (LBM), finite element method (FEM) and computational fluid dynamics (CFD) method. In addition to providing general applications of numerical methods for fluid-structure interaction evaluation, the paper also describes some cases applied for problems associated with single-phase flow and two-phase flow in nuclear power plants.

Data Processing Technology of Interference Fringe for Particle Decontamination Measurement

2019 ◽  
Author(s):  
K. Fujiwara ◽  
Y. Nakamura ◽  
A. Kaneko ◽  
Y. Abe
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Bubbly Flow ◽  
Wave Length ◽  
Fission Products ◽  
Gaussian Mixture ◽  
Single Bubble ◽  
High Speed Camera ◽  
Phase Information

Abstract In severe accidents (SAs) of nuclear power plants, release of gas containing fission products (FPs) from the reactor vessel is thought to be a major issue. As to reduce the leakage of FPs into the environment, gas containing FPs are generally discharged though the wet well, and decontaminated by the transfer effect of FPs from the gas-phase to the liquid phase. This effect is called pool scrubbing. In SA analysis codes such as MELCOR, it is predicted in the model that FP particle motion inside a single bubble, created by the bubbly flow inside the wet well as a major factor in decontamination. However, there are almost no experimental data to investigate the decontamination behavior. Therefore, in our experiment, we used an advanced M/Z interferometer in order to visualize the particle decontamination behavior by adopting Maki prism and installing a high-speed camera. However, since the interferometer experiments are not specialized in non-stable phenomena and, there were several problems to be solved. The first problem was the phase extraction method in the FFT measurement. Since the FFT information of interference is complicated, the existing extract the phase information by hand from the overall amplitude. However, since the high-speed camera visualization provide a large amount of information, this is not a realistic solution in our experiment. Therefore, in order to obtain the threshold between phase information from the overall amplitude quantitatively, we applied the Gaussian mixture model (GMM) as to cluster the data. From the measurement results, we succeed in obtaining a threshold from the fitting results of GMM. The next problem was to obtain a fine image of the bubble interface in order to obtain the decontamination behavior in the bubble interface. However, the interference image contains a stripe on the background which makes it difficult to obtain the interface information. Therefore, in our experiment, we added a LED backlight coaxial to the laser of interferometer in order to obtain the backlight image on the bubble. The interference and backlight image are divided by wave length with a dichroic mirror. We have done a synchronous visualization of interference and backlight image. A fine mask to extract the interface of bubble is obtained from the calibration and comparison of two images. Using the visualization of continuous image of particle decontamination behavior from a single bubble, the decontamination behavior of particle from a single bubble was clearly obtained. Although the existing model predict the decontamination behavior as stable, the non-stationary decontamination of particle from the bubble has been measured.

Solution of Governing Equations for Six-Field System Code

2017 ◽  
Author(s):  
Glenn A. Roth ◽  
George L. Mesina ◽  
Fatih Aydogan
Keyword(s):  
Gas Phase ◽  
Nuclear Power ◽  
Power Plants ◽  
Governing Equations ◽  
Field Equations ◽  
Two Phase Flows ◽  
Two Phase ◽  
Drift Flux ◽  
Continuous Fields ◽  
Analysis System

Modeling of two phase flows in nuclear power plants is very important for design, licensing, and operator training and therefore must be performed accurately. As requirements have increased, the form and accuracy of the models and computer codes have improved along with them. Early formulations for the field equations include: single phase liquid with algebraic drift flux for the gas phase, modeled with mass, momentum and energy governing equations; and two separate fields, liquid phase and gas phase, typically modeled with six governing equations. These lump bubbles and droplets into the gas and liquid phases respectively and use flow regime maps based upon available experimental data. However, the experiments do not cover the entire spectrum of reactor conditions, so that transitions and extrapolations, which are inherently inaccurate, must be employed. Further, some reactor scenarios, such as boiling and condensation, can be more accurately resolved by modeling bubbles or droplets separately from the continuous fields. Introduction of an additional field, droplet or bubble, apart from the continuous liquid and gas fields, generally uses nine governing equations. Despite the successful development of the above-mentioned methods for modeling reactor coolant flow in modern software, such as RELAP, TRAC, TRACE, CATHARE and many others, there remain reactor scenarios that require greater resolution to model. This is particularly true of conditions during reflood, where emergency spray flows dominate the cooling profile within the core. Existing system codes use a lumped approach for two phase flows that groups the fields by their phase, thereby losing track of the physical interactions between the discrete fluid fields. The accuracy of these accident analysis system codes can be improved by characterizing the interactions between additional coolant fields. To capture the effect of the various field interactions, governing equations involving six-fields have been developed. The six fields are 1) continuous liquid, 2) continuous vapor, 3) large droplets, 4) small droplets, 5) large bubbles and 6) small bubbles. The additional fields and the related governing equations introduce additional variables and source terms that require new closure relationships and primary variables. This article presents the equations and variables and develops the discrete set of 18 equations that must be solved to model the system.

Finite Element Analyses of Residual Stresses in Typical Welded Joints Used in Nuclear Power Plants and Comparisons With Experiments

2010 ◽  
Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito
Keyword(s):  
Numerical Simulation ◽  
Residual Stress ◽  
Residual Stresses ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Butt Joint ◽  
Welding Residual Stress ◽  
Dissimilar Metal ◽  
Water Reactors

Recent discoveries of stress corrosion cracking (SCC) at nickel-based metals in pressurized water reactors (PWRs) and boiling water reactors (BWRs) have raised concerns about safety and integrity of plant components. It has been recognized that welding residual stress is an important factor causing the issue of SCC in a weldment. In this study, both numerical simulation technology and experimental method were employed to investigate the characteristics of welding residual stress distribution in several typical welded joints, which are used in nuclear power plants. These joints include a thick plate butt-welded Alloy 600 joint, a dissimilar metal J-groove set-in joint and a dissimilar metal girth-butt joint. First of all, numerical simulation technology was used to predict welding residual stresses in these three joints, and the influence of heat source model on welding residual stress was examined. Meanwhile, the influence of other thermal processes such as cladding, buttering and heat treatment on the final residual stresses in the dissimilar metal girth-butt joint was also clarified. Secondly, we also measured the residual stresses in three corresponding mock-ups. Finally, the comparisons of the simulation results and the measured data have shed light on how to effectively simulate welding residual stress in these typical joints.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

Measurement of Flow-Induced Forces: A Single Flexible Tube in a Rigid Triangular-Pitch Bundle Subjected to Two-Phase Cross-Flow

2015 ◽  
Author(s):  
Enrico Deri ◽  
Joël Nibas ◽  
Olivier Ries ◽  
André Adobes
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Dynamic ◽  
Cross Flow ◽  
Maintenance Policy ◽  
Excitation Spectra ◽  
Two Phase ◽  
Force Coefficients ◽  
Out Of Plane

Flow-induced vibrations of Steam Generator tube bundles are a major concern for the operators of nuclear power plants. In order to predict damages due to such vibrations, EDF has developed the numerical tool GeViBus, which allows one to asses risk and thereafter to optimize the SG maintenance policy. The software is based on a semi analytical model of fluid-dynamic forces and dimensionless fluid force coefficients which need to be assessed by experiment. The database of dimensionless coefficients is updated in order to cover all existing tube bundle configurations. Within this framework, a new test rig was presented in a previous conference with the aim of assessing parallel triangular tube arrangement submitted to a two-phase cross-flow. This paper presents the result of the first phase of the associated experiments in terms of force coefficients and two-phase flow excitation spectra for both in-plane and out-of-plane vibration.

Crack Growth Prediction and Leakage Potential of Steam Generator Tubes Subjected to Flow Induced Vibrations

2014 ◽  
Author(s):  
Salim El Bouzidi ◽  
Marwan Hassan ◽  
Jovica Riznic
Keyword(s):  
Crack Growth ◽  
Steam Generator ◽  
Power Plants ◽  
Mechanical Energy ◽  
Steam Generators ◽  
Two Phase ◽  
Element Analysis ◽  
Wide Range ◽  
Flow Induced Vibrations ◽  
Crack Growth Model

Nuclear steam generators are critical components of nuclear power plants. Flow-Induced Vibrations (FIV) are a major threat to the operation of nuclear steam generators. The two main manifestations of FIV in heat exchangers are turbulence and fluidelastic instability, which would add mechanical energy to the system resulting in great levels of vibrations. The consequences on the operation of steam generators are premature wear of the tubes, as well as development of cracks that may leak radioactive heavy water. This paper investigates the effect of tube support clearance on crack propagation. A crack growth model is used to simulate the growth of Surface Flaws and Through-Wall Cracks of various initial sizes due to a wide range of support clearances. Leakage rates are predicted using a two-phase flow leakage model. Non-linear finite element analysis is used to simulate a full U-bend subjected to fluidelastic and turbulence forces. Monte Carlo Simulations are then used to conduct a probabilistic assessment of steam generator life due to crack development.

Sign in / Sign up

Export Citation Format

Share Document