Analytical Evaluation of Unsteady Thermal Hydraulic Characteristics in Shell and Tube Heat Exchangers

2006 ◽  
Author(s):  
Shiro Takahashi ◽  
Yuichi Narumi ◽  
Kiyoshi Ishihama ◽  
Akihito Yokoyama ◽  
Toyohiko Tsuge ◽  
...  
Keyword(s):  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Temperature ◽  
Detached Eddy Simulation ◽  
Tube Heat Exchanger ◽  
Eddy Simulation ◽  
Safety And Reliability ◽  
Cfd Analyses

Many shell & tube heat exchangers are used in nuclear power plants. Unsteady thermal hydraulic phenomena have been studied in shell & tube heat exchangers to improve their safety and reliability and to extend their lifetime based on experience obtained from long periods of plant operation. We investigated unsteady flow in shell & tube heat exchangers by using computational fluid dynamics (CFD) analyses. The inlet flow on the shell side was separated and flow in several directions. A large part of the flow crossed over the tube bundle, and some parts of the flow took two circuitous roots (up and down) along the inner surface of the shell. Separated circuitous flows collided again where a baffle plate had been cut off. A pair of symmetric vortexes could be seen in that location. Some parts of the circuitous flow moved backwards into the tube bundle due to vortexes. These vortexes were unstable and changed their size and location. A pair of vortexes changed from symmetric to asymmetric. As a result, the direction of flow in the tube bundle near the vortexes changed continuously. Variations in vortexes simulated through CFD analyses could also be seen in tests on the actual size. Fluid temperature fluctuations around tubes were also evaluated through CFD analyses. Unsteady phenomena with changes from symmetric to asymmetric vortexes could be observed in the shell & tube heat exchanger and were simulated through CFD analyses with a detached eddy simulation (DES) turbulence model.

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.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

RevospECT® PRO Software for Automated Interpretation of Results and Automatic Tracking of Defects during Eddy Current Inspection of Heat Exchangers

NDT World ◽  
2021 ◽  
pp. 21-23
Author(s):  
Denis Shorikov ◽  
Aleksandra Melnikova
Keyword(s):  
Eddy Current ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Level Ii ◽  
Unique System ◽  
Self Study ◽  
Software Product ◽  
Eddy Current Inspection ◽  
Automated Interpretation

The eddy current NDT method has been successfully used at Russian nuclear power plants for more than 20 years, but there are still problems with assessing the reliability of the results. Software product of Zetec Inc. (USA) RevospECT® Pro allows you to automatically analyze and compare the monitoring results of the same object, obtained at different times, which allows you to track the development of defects. Thanks to a unique system for collecting and analyzing information, its ability to self-study, RevospECT® Pro is able to make decisions on its own, replacing the level II specialist in full.

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.

Research of Technology for Repair of Heat Exchangers of Nuclear Power Plants by Laser Welding

2021 ◽  
Vol 313 ◽  
pp. 94-105
Author(s):  
A. Bernatskyi ◽  
V. Sydorets ◽  
O.M. Berdnikova ◽  
I. Krivtsun ◽  
O. Kushnarova
Keyword(s):  
Service Life ◽  
Laser Welding ◽  
Welded Joints ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Test ◽  
Austenitic Steels ◽  
Depth Of Penetration

Extending the lifetime of energy facilities is extremely important today. This is especially true of nuclear power plants, the closure (or modernization) of which poses enormous financial and environmental problems. High-quality repair of reactors can significantly extend their service life. One of the critical parts is heat exchangers, the tubes of which quite often fail. Sealing, as a type of repair of heat exchanger tubes by the plugs, is promising provided that the joint quality is high. Practical experience in the use of welding to solve this problem has shown the need to search technological solutions associated with increasing the depth of penetration and reducing the area of thermal effect. The aim of the work was to develop a highly efficient technology for repair and extension of service life of heat exchangers of nuclear power plants based on the results of studying the technological features of laser welding of joints of dissimilar austenitic steels AISI 321 and AISI 316Ti in the vertical spatial position. Based on the results of the analysis of mechanical test data, visual and radiographic control, impermeability tests and metallographic studies of welded joints, the appropriate modes of laser welding of plugs have been determined. The principal causes of defects during laser welding of annular welded joints of dissimilar stainless steels are determined and techniques for their elimination and prevention of their formation are proposed. Based on the results of the research, technological recommendations for laser welding of plugs in the heat exchange tube of the collector are formulated, which significantly improves the technology of repair of steam generators of nuclear power plants and extends the service life of reactors.

Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve

2018 ◽  
Author(s):  
Shiro Takahashi ◽  
Eiji Ozaki ◽  
Atsuyuki Minenaga
Keyword(s):  
Pressure Loss ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Acoustic Noise ◽  
Acoustic Resonance ◽  
Stop Valve ◽  
Cfd Analyses ◽  
Main Steam Line ◽  
Main Steam

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

scholarly journals Determination of Transient Fluid Temperature and Thermal Stresses in Pressure Thick-Walled Elements Using a New Design Thermometer

Energies ◽  
2019 ◽  
Vol 12 (2) ◽  
pp. 222 ◽  
Author(s):  
Magdalena Jaremkiewicz ◽  
Dawid Taler ◽  
Piotr Dzierwa ◽  
Jan Taler
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Pressure ◽  
Thermal Stresses ◽  
Thermal Inertia ◽  
Internal Surface ◽  
Transient Temperature ◽  
Fluid Temperature ◽  
Start Up

In both conventional and nuclear power plants, the high thermal load of thick-walled elements occurs during start-up and shutdown. Therefore, thermal stresses should be determined on-line during plant start-up to avoid shortening the lifetime of critical pressure elements. It is necessary to know the fluid temperature and heat transfer coefficient on the internal surface of the elements, which vary over time to determine transient temperature distribution and thermal stresses in boilers critical pressure elements. For this reason, accurate measurement of transient fluid temperature is very significant, and the correct determination of transient thermal stresses depends to a large extent on it. However, thermometers used in power plants are not able to measure the transient fluid temperature with adequate accuracy due to their massive housing and high thermal inertia. The article aims to present a new technique of measuring transient superheated steam temperature and the results of its application on a real object.

Tackling FSI Simulation for FIV Problems in Tube Bundle Systems With POD Approach

2011 ◽  
Author(s):  
Marie Pomarede ◽  
Aziz Hamdouni ◽  
Erwan Liberge ◽  
Elisabeth Longatte ◽  
Jean-Franc¸ois Sigrist
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Structure Interaction ◽  
Flowing Fluid ◽  
Fluid Structure ◽  
Reduced Order ◽  
Structure Interaction ◽  
Flow Past ◽  
Steam Boilers

Tube bundles in steam boilers of nuclear power plants and nuclear on-board stokehold are known to be exposed to high levels of vibrations under flowing fluid. This coupled fluid-structure problem is still a challenge for engineers, first because of the difficulty to fully understand it, second because of the complexity for setting it up numerically. Although numerical techniques could help the understanding of such a mechanism, a complete simulation of a fluid past a whole elastically mounted tube bundle is currently out of reach for engineering purposes. To get round this problem, the use of a reduced-order model has been proposed with the introduction of the widely used Proper Orthogonal Decomposition (POD) method for a flow past a fixed structure [M. Pomare`de, E. Liberge, A. Hamdouni, E.Longatte, & J.F. Sigrist - Simulation of a fluid flow using a reduced-order modelling by POD approach applied to academic cases; PVP2010, July 18–22, Seattle]. Interesting results have been obtained for the reconstruction of the flow. Here a first step is to propose to consider the case of a flow past a fixed tube bundle configuration in order to check the good reconstruction of the flow. Then, an original approach proposed by Liberge (E. Liberge; POD-Galerking Reduction Models for Fluid-Structure Interaction Problems, PhD Thesis, Universite´ de La Rochelle, 2008) is applied to take into account the fluid-structure interaction characteristic; the so-called “multiphase” approach. This technique allows applying the POD method to a configuration of a flow past an elastically mounted structure. First results on a single circular cylinder and on a tube bundle configuration are encouraging and let us hope that parametric studies or prediction calculations could be set up with such an approach in a future work.

Sign in / Sign up

Export Citation Format

Share Document