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.

Relationship Between Flow Pattern and Pressure Distribution in Venturi Scrubber

2014 ◽  
Author(s):  
Naoki Horiguchi ◽  
Hiroyuki Yoshida ◽  
Akiko Kaneko ◽  
Yutaka Abe
Keyword(s):  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
The Self ◽  
Radioactive Materials ◽  
Venturi Scrubber ◽  
Nuclear Disaster ◽  
Severe Accidents ◽  
Main Components

As revealed by Fukushima Daiichi nuclear disaster, countermeasures against severe accidents in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing diffusion of the radioactive materials, it is most important to install filtered venting devices to release high pressure contaminated gas to the atmosphere with elimination radioactive materials in the gas. A Multi Venturi Scrubber System (MVSS) is one of the filtered venting devices, and used in European reactors [1, 2]. One of the main components of the MVSS is a Venturi Scrubber (VS). It is considered that a dispersed or dispersed annular flow is formed in the VS by a self-priming phenomena. In the self-priming phenomena, the liquid was suctioned from a surrounding region of the VS to the inside of the VS. And a part of the radioactive materials are eliminated through the gas-liquid interface of the dispersed or annular dispersed flow. Therefore, to consider the MVSS operation characteristics, it is important whether to occur the self-priming or not and the liquid flow rate of the self-priming of the VS. The objective of this paper is to understand the self-priming phenomena of the VS for the filtered venting. And theoretical analysis and experiment were conducted. By comparing these results, we discussed about the mechanism of the self-priming phenomena. As results, the self-priming phenomena in the VS was confirmed and, at a high gas flow rate, the suspension of the self-priming is confirmed experimentally and theoretically.

Development of High Performance Moisture Separator Reheater

2009 ◽  
Author(s):  
Issaku Fujita ◽  
Kotaro Machii ◽  
Teruaki Sakata
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
High Performance ◽  
Power Plants ◽  
Tube Bundle ◽  
Separation Performance ◽  
Heating Steam ◽  
Severe Erosion ◽  
Moisture Separator ◽  
And Performance

Moisture Separator Reheaters (MSRs) of Nuclear power plants, especially 1st generation type (commercial operation started from between 1970 and 1982), has been suffered from various problems like severe erosion, moisture separation performance deterioration, drain sub cooling. To solve these problems and performance improvement, improved MSR was developed. At the new MSR, high performance SS439 stainless steel round type tube bundle was applied, where heating steam distribution is optimized by orifice plate in order to minimize the drain sub cooling. Based on the CFD approach, cycle steam distribution was optimized and FAC resistant material application for the internal parts of MSRs was determined. As a result, pressure drop was reduced by 0.6% against the HP turbine exhaust pressure. Performance of moisture separation was improved by the latest chevron type separator. Where, the reverse pressure is locally caused at the drainage area of the separator because remarkable longitudinal pressure distribution is formed by the high-speed steam flow in the manifold. Then, a new moisture separation structure was developed in consideration of the influence that this reverse pressure gave to the separator performance.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

Film Fraction and Droplet Size inside of a Rectangular Venturi Scrubber

2010 ◽  
Vol 660-661 ◽  
pp. 549-554
Author(s):  
Vádila Giovana Guerra ◽  
M.A.F. Daher ◽  
José Antônio Silveira Gonçalves ◽  
José Renato Coury
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Gas Flow ◽  
Liquid Flow Rate ◽  
Dimensionless Number ◽  
Gas Velocity ◽  
Flow Rates ◽  
Venturi Scrubber ◽  
Liquid Injection ◽  
Gas Drag

The Venturi scrubber, equipment frequently used in the removal of particles from gases, is constituted basically by a duct with a convergent section followed by a constriction, or throat, and a divergent section. A liquid, usually injected in the throat, is atomized by the flowing air at high speed. The formed droplets act as collectors of particles from the gas. The process of droplet formation from an injected liquid can be described as follows: the liquid enters the gas stream in the form of a jet, perpendicular to the gas flow. As the jet penetrates the gas stream, it is bent by the gas drag. After a given penetration distance, a burst occurs, and the remaining jet is disintegrated as a droplet cloud. Depending on the liquid and gas flow rates, the penetration on the jet into the gas stream may reach the walls of the equipment, and a fraction of liquid deposits in the form of a film. This film contributes little for the removal of particles from the dust laden gas. Few studies have analyzed the formation of film at the scrubber walls and its influence in the droplet size inside the Venturi scrubber. For this reason, the present study is focused on the experimental measurement of the deposition of the liquid film on the walls of a rectangular Venturi scrubber and, simultaneously, the estimation of the droplet size measured in the Venturi throat. The experiments were carried out varying the liquid flow rate, the gas velocity and the number of orifices of liquid injection. A correlation, using a dimensionless number, was proposed to quantify the influence of each experimental condition. The results indicate that film fraction has a significant influence in the droplet size measured inside of Venturi scrubber.

The Fluid Elastic Instability of Concentric Arrays of Tube Bundles Subjected on Cross Flow

2018 ◽  
Author(s):  
Liyan Liu ◽  
Wei Xu ◽  
Kai Guo ◽  
Zhanbin Jia ◽  
Yang Wang ◽  
...  
Keyword(s):  
Heat Exchangers ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Experimental Studies ◽  
Cross Flow ◽  
High Speed Photography ◽  
Elastic Instability ◽  
Tube Bundles

Concentric arrays of tube bundles are applied extensively in heat exchangers at nuclear power plants. Flow induced vibration is one of the main causes of heat exchanger failures. However, there is no corresponding standard and basic parameters in the design code of different countries for concentric arrays of tube bundles. The fluid elastic instability of this type of heat exchangers cannot be calculated, and the design criteria is lacked. In this paper, a circulating water tunnel experimental facility were set up to test the vibration characteristic of concentric arrays subjected to cross flow. A non-contact measurement method based on high-speed photography imaging technology were adopted, which improved the accuracy of the test. Three kinds of tube bundles (0-degree angle, 15-degree angle and 30-degree angle arrangement, radial/circumferential pitch being 33.6/36.4 mm) were studied. The vibration frequency, amplitude and critical velocity of the tube bundle were investigated by changing the flow velocity. Computational fluid dynamics and fluid-structure interaction method were applied to simulate the fluid elastic instability of tube bundles, that were further verified by the experiments. Meanwhile, the numerical simulation supplements the contents of the experimental studies, which is utilizable to investigate and research the fluid elastic instability. The results of this work could provide references for the design of concentric array heat exchangers.

Buckling Behaviors of Metallic Columns Under Compressive Load at Extremely High Temperatures

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Temperatures ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Compressive Load ◽  
Buckling Load ◽  
Severe Accident ◽  
Lateral Deflection ◽  
Wide Range ◽  
Creep Buckling

This study aims to investigate buckling behaviors of a slender stainless steel column under compressive loads in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling load, defined a load which generates a failure of the column (plastic collapse) was experimentally measured in a wide range of temperatures from 25 °C up to 1200 °C. The buckling load values measured were compared to numerical estimations for both an ideal column and for a column initially bent. Secondly, creep buckling tests were also performed for extremely high temperatures (800 °C, 900 °C, and 1000 °C). Creep buckling was found to occur very quickly compared to general creep times under tensile stresses. Time to creep buckling was exponentially increased with decrease of loads applied. Lateral deflection of a test column was estimated using captured images by a high speed camera. It was suggested to represent creep buckling behaviors as a time-lateral deflection curve. Moreover, an empirical correlation was developed to predict creep buckling time, based on the Larson-Miller model with experimental results obtained in present study.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

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