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Автор: Grafiati
Опубліковано: 14 березня 2023

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Статті в журналах з теми "Thermal-hydraulic modeling"

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Li, Dong, Sujun Dong, Jun Wang, and Yunhua Li. "Temperature Dynamic Characteristics Analysis and Thermal Load Dissipation Assessment for Airliner Hydraulic System in a Full Flight Mission Profile." Machines 10, no. 4 (April 2, 2022): 258. http://dx.doi.org/10.3390/machines10040258.

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This paper deals with the modeling of the thermal load and the simulation of thermal dynamic characteristics for the hydraulic system of a large airliner in a full mission profile. Firstly, the formation mechanism of the thermal load in the hydraulic system is analyzed, and thermal dynamic modeling is conducted of the hydraulic components of an hydraulic system with an immersed heat exchanger employing the lumped parameter thermal node method and oil temperature and power loss of each key node within the hydraulic system within a full mission profile. Then, a thermal dynamic simulation model based on MATLAB/Simulink is established, and the temperatures at the nodes of different components and the absorptive capacity of the fuel heat sink in the thermal management module are calculated. The simulation results show that the thermal management scheme of the heat exchanger, located in the return oil pipeline of the hydraulic piston pump housing and immersed in the central fuel tank, can dissipate the thermal load of the system. This work is of important significance for temperature analysis and thermal load dissipation of the airliner hydraulic system.
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2

Oriolo, F., W. Ambrosini, G. Fruttuoso, F. Parozzi, and R. Fontana. "Thermal-Hydraulic Modeling and Severe Accident Radionuclide Transport." Nuclear Technology 112, no. 2 (November 1995): 238–49. http://dx.doi.org/10.13182/nt95-a35177.

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3

LI, Cheng-gong, and Zong-xia JIAO. "Thermal-hydraulic Modeling and Simulation of Piston Pump." Chinese Journal of Aeronautics 19, no. 4 (November 2006): 354–58. http://dx.doi.org/10.1016/s1000-9361(11)60340-3.

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4

Jiang, S. Y., X. X. Wu, Y. J. Zhang, and H. J. Jia. "Thermal hydraulic modeling of a natural circulation loop." Heat and Mass Transfer 37, no. 4-5 (July 1, 2001): 387–95. http://dx.doi.org/10.1007/s002310000136.

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5

Sunagatullin, Rustam Z., Rinat M. Karimov, Radmir R. Tashbulatov, and Boris N. Mastobaev. "Modeling the thermal-hydraulic effect of wax layer." SCIENCE & TECHNOLOGIES OIL AND OIL PRODUCTS PIPELINE TRANSPORTATION 9, no. 2 (April 30, 2019): 158–62. http://dx.doi.org/10.28999/2541-9595-2019-9-2-158-162.

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6

Hu, Jun-ping, and Ke-jun Li. "Thermal-hydraulic modeling and analysis of hydraulic system by pseudo-bond graph." Journal of Central South University 22, no. 7 (July 2015): 2578–85. http://dx.doi.org/10.1007/s11771-015-2787-0.

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7

Li, Dong, Sujun Dong, Jun Wang, and Yunhua Li. "Thermal dynamics and thermal management strategy for a civil aircraft hydraulic system." Thermal Science 24, no. 4 (2020): 2311–18. http://dx.doi.org/10.2298/tsci2004311l.

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Addressing the growing severe heat-generation and temperature-rise issues of the civil aircraft hydraulic system, this paper proposes a thermal dynamic model and thermal management strategies for the system within full mission profile. Firstly, a new thermal dynamic modeling towards general hydraulic components is conducted. Secondly, thermal dynamic governing equations are derived. Then a thermal management mechanism of the system is proposed. The conducted research is prerequisite to future numerical simulation of the thermal dynamic characteristics, evaluation and improvement of thermal management strategies for the system.
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8

Khater, H., T. Abu-El-Maty, and S. El-Din El-Morshdy. "Thermal-hydraulic modeling of reactivity accidents in MTR reactors." Kerntechnik 72, no. 1-2 (March 2007): 44–52. http://dx.doi.org/10.3139/124.100317.

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Khater, Hany, Talal Abu-El-Maty, and El-Din El-Morshdy. "Thermal-hydraulic modeling of reactivity accidents in MTR reactors." Nuclear Technology and Radiation Protection 21, no. 2 (2006): 21–32. http://dx.doi.org/10.2298/ntrp0602021k.

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This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.
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Bottura, L. "Thermal, Hydraulic, and Electromagnetic Modeling of Superconducting Magnet Systems." IEEE Transactions on Applied Superconductivity 26, no. 3 (April 2016): 1–7. http://dx.doi.org/10.1109/tasc.2016.2544253.

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Дисертації з теми "Thermal-hydraulic modeling"

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Pegonen, Reijo. "Development of an Improved Thermal-Hydraulic Modeling of the Jules Horowitz Reactor." Doctoral thesis, KTH, Reaktorteknologi, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-197712.

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The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at CEA Cadarache research center in France. The reactor will support existing and future nuclear reactor technologies, with the first criticality expected at the end of this decade. The current/reference CEA methodology for simulating the thermalhydraulic behavior of the reactor gives reliable results. The CATHARE2 code simulates the full reactor circuit with a simplified approach for the core. The results of this model are used as boundary conditions in a three-dimensional FLICA4 core simulation. However this procedure needs further improvement and simplification to shorten the computational requirements and give more accurate core level data. The reactor’s high performance (e.g. high neutron fluxes, high power densities) and its design (e.g. narrow flow channels in the core) render the reactor modeling challenging compared to more conventional designs. It is possible via thermal-hydraulic or solely hydraulic Computational Fluid Dynamics (CFD) simulations to achieve a better insight of the flow and thermal aspects of the reactor’s performance. This approach is utilized to assess the initial modeling assumptions and to detect if more accurate modeling is necessary. There were no CFD thermal-hydraulic publications available on the JHR prior to the current PhD thesis project. The improvement process is split into five steps. In the first step, the state-of-the-art CEA methodology for thermal-hydraulic modeling of the reactor using the system code CATHARE2 and the core analysis code FLICA4 is described. In the second and third steps, a CFD thermal-hydraulic simulations of the reactor’s hot fuel element are undertaken with the code STAR-CCM+. Moreover, a conjugate heat transfer analysis is performed for the hot channel. The knowledge of the flow and temperature fields between different channels is important for performing safety analyses and for accurate modeling. In the fourth step, the flow field of the full reactor vessel is investigated by conducting CFD hydraulic simulations in order to identify the mass flow split between the 36 fuel elements and to describe the flow field in the upper and lower plenums. As a side study a thermal-hydraulic calculation, similar to those performed in previous steps is undertaken utilizing the outcome of the hydraulic calculation as an input. The final step culminates by producing an improved, more realistic, purely CATHARE2 based, JHR model, incorporating all the new knowledge acquired from the previous steps. The primary outcome of this four year PhD research project is the improved, more realistic, CATHARE2 model of the JHR with two approaches for the hot fuel element. Furthermore, the project has led to improved thermal-hydraulic knowledge of the complex reactor (including the hot fuel element), with the most prominent findings presented.

QC 20161208


DEMO-JHR
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Chen, Qiang. "Simulation of thermal plant optimization and hydraulic aspects of thermal distribution loops for large campuses." Texas A&M University, 2005. http://hdl.handle.net/1969.1/2451.

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Following an introduction, the author describes Texas A&M University and its utilities system. After that, the author presents how to construct simulation models for chilled water and heating hot water distribution systems. The simulation model was used in a $2.3 million Ross Street chilled water pipe replacement project at Texas A&M University. A second project conducted at the University of Texas at San Antonio was used as an example to demonstrate how to identify and design an optimal distribution system by using a simulation model. The author found that the minor losses of these closed loop thermal distribution systems are significantly higher than potable water distribution systems. In the second part of the report, the author presents the latest development of software called the Plant Optimization Program, which can simulate cogeneration plant operation, estimate its operation cost and provide optimized operation suggestions. The author also developed detailed simulation models for a gas turbine and heat recovery steam generator and identified significant potential savings. Finally, the author also used a steam turbine as an example to present a multi-regression method on constructing simulation models by using basic statistics and optimization algorithms. This report presents a survey of the author??s working experience at the Energy Systems Laboratory (ESL) at Texas A&M University during the period of January 2002 through March 2004. The purpose of the above work was to allow the author to become familiar with the practice of engineering. The result is that the author knows how to complete a project from start to finish and understands how both technical and nontechnical aspects of a project need to be considered in order to ensure a quality deliverable and bring a project to successful completion. This report concludes that the objectives of the internship were successfully accomplished and that the requirements for the degree of Degree of Engineering have been satisfied.
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Leem, Junghun. "Micromechanical fracture modeling on underground nuclear waste storage: Coupled mechanical, thermal, and hydraulic effects." Diss., The University of Arizona, 1999. http://hdl.handle.net/10150/284062.

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Coupling effects between thermal, hydraulic, chemical and mechanical (THCM) processes for rock materials are one of major issues in Geological engineering, Civil engineering, Hydrology, Petroleum engineering, and Environmental engineering. In all of these fields, at least two mechanisms of THCM coupling are considered. For an example, thermal, hydraulic, and mechanical coupling effects are important in Geological engineering and Civil engineering. The THM coupling produces effects on underground structures, since the underground structures are under influences of geothermal gradient, groundwater, gravitational stresses, and tectonic forces. In particular, underground repository of high-level nuclear waste involves all four of the THCM coupling processes. Thermo-hydro-mechanical coupling model for fractured rock media has been developed based on micromechanical fracture model [Kemeny 1991, Kemeny & Cook 1987]. The THM coupling model is able to simulate time- and rate-dependent fracture propagation on rock materials, and quantify characteristics of damage by extensile and shear fracture growth. The THM coupling model can also simulate coupled thermal effects on underground structures such as high-level nuclear waste repository. The results of thermo-mechanical coupling model are used in conducting a risk analysis on the structures. In addition, the THM coupling model is able to investigate variations of fluid flow and hydraulic characteristics on rock materials by measuring coupled anisotropic permeability. Later, effects of chemical coupling on rock materials are investigated and modified in the THM coupling model in order to develop a thermo-hydro-chemo-mechanical coupling model on fractured rocks. The THCM coupling model is compared with thermal, hydraulic, chemical, and mechanical coupling tests conducted at the University of Arizona. The comparison provides a reasonable prediction for the THCM coupling tests on various rock materials. Finally, the THCM coupling model for fractured rocks simulates the underground nuclear waste storage in Yucca Mountain, Nevada, and conducted performance and risk analysis on the repository.
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Han, Gee Yang. "A mathematical dynamic modeling and thermal hydraulic analysis of boiling water reactors using moving boundaries." Diss., The University of Arizona, 1993. http://hdl.handle.net/10150/186191.

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A new development and practical application of a mathematical dynamic modeling for simulating normal and accidental transient analysis for the boiling water reactor system is presented in this dissertation. The mathematical dynamic modeling represents a new technology based on a moving boundary concept. The mathematical model developed for fluid flows is based on a set of the four equation mixture model, one-dimensional, single channel with a drift flux model in the two-phase flow regime. The four conservation equations used in the mathematical model formulation include the vapor phase mass equation, the liquid phase mass equation, the mixture energy equation, and the one-dimensional mixture momentum equation for the boiling channel. The formulation of the core thermal-hydraulic model utilizes a transient moving boundary technique which tracks the movements of the phase change and boiling transition boundaries. Such a moving boundary model has been developed to allow a smooth representation of the boiling boundary movement based on empirical heat transfer correlations and the local thermal-hydraulic conditions of the coolant flow along fuel pin channels. The mathematical models have been implemented to accommodate three-dimensional reactor kinetics, with detailed thermal conduction in fuel elements. Also, an accurate minimum departure from nucleate boiling ratio (MDNBR) boundary is predicted during transients. Several test calculations were performed to assess the accuracy and applicability of the moving boundary model. Comparison between the calculated results and the experimental data are favorable. Overall system studies show that some thermal margin is gained using the transient MDNBR approach vs the traditional quasi-static methodology. The model predicts accurate void fraction profiles for kinetic feedback and boiling stability analysis for the BWR. The moving boundary formulation and improved numerical solution scheme are an efficient and suitable tool which can be useful for realistic simulation of degraded nuclear power plant transients.
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Chen, Minghui. "DESIGN, FABRICATION, TESTING, AND MODELING OF A HIGH-TEMPERATURE PRINTED CIRCUIT HEAT EXCHANGER." The Ohio State University, 2015. http://rave.ohiolink.edu/etdc/view?acc_num=osu1431072434.

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Svensson, Oskar. "Electrohydraulic Power Steering Simulation : Dynamic, thermal and hydraulic modelling." Thesis, KTH, Skolan för elektroteknik och datavetenskap (EECS), 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-265674.

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There are several benets of electrohydraulic power steering systems, as compared to hydraulicpower steering systems where the pump is driven directly by the engine of the vehicle. Someof these benets are increased eciency and improved steering performance. The purpose ofthis project is to create a simulation model of the electrohydraulic power steering system inSimulink, excluding the hydraulic circuit. The model should thus consist of the electric motor,the drive electronics, the control system, the hydraulic pump as well as the communication andinterface to the master simulation system in which the model will be used.As a start a mathematical model of the motor is derived. Then the motor controller includingtwo current controllers and a speed controller is developed. The switching signals for the threephase bridge that drives the motor are calculated using space vector modulation. The motordrives a hydraulic pump, which is modeled using data sheet eciency curves. Finally a thermalmodel of the drive is developed. To fulll real time requirements, a lumped parameter approachis chosen. The nal model is exported as a Functional Mock-up Unit, which is a black-boxencapsulation of the complete simulation model.The simulation model is compared to measurement data to conrm its validity. Thesecomparisons shows that the dynamic response of the motor and its controller are close to themeasured values and that the thermal model adequately corresponds to the thermal tests. Thehydraulic pump model varied from measurements more than the other sub-modules. It was,however, seen as acceptable. Overall the system response was satisfactory, but naturally a lotof future improvements and new features could be made to improve the model.
Det finns flera fördelar med elektrohydraulisk servostyrning, där hydraulpumpen drivs av en el-motor, jämfört med hydraulisk servostyrning, där pumpen drivs direkt av fordonets förbränningsmotor. Några av dessa fördelar är ökad effektivitet och förbättrad styrprestanda. Syftet med detta projekt är att skapa en Simulink-modell av ett elektrohydraulisk system för servostyrning, exklusive hydraulkretsen. Modellen ska alltså bestå av delmodeller för elmotorn, drivelektroniken, styrsystemet, hydraulpumpen samt kommunikation med den övergripande simuleringsplattformen.Inledningsvis beskrivs en matematisk modell av elmotorn och efter det utvecklas motorstyrningen, bestående av två strömregulatorer samt en hastighetsregulator. Spänningen från strömregulatorerna uppnås genom space vector-modulation, som beräknar de pulskvoter som krävs för att uppnå denna spänning. Elmotorn driver en pump. Denna pump modelleras med hjälp av data från pumpens datablad. Slutligen modelleras drivelektronikens termiska egenskaper med ett termiskt nätverk. Den slutliga modellen omsluts av en Functional Mock-up Unit somintegreras i den övergripande simuleringsplattformen.
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Keshmiri, Amir. "Thermal-hydraulic analysis of gas-cooled reactor core flows." Thesis, University of Manchester, 2010. https://www.research.manchester.ac.uk/portal/en/theses/thermalhydraulic-analysis-of-gascooled-reactor-core-flows(29335acf-a397-4b8c-8217-fd2ee0d26967).html.

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In this thesis a numerical study has been undertaken to investigate turbulent flow and heat transfer in a number of flow problems, representing the gas-cooled reactor core flows. The first part of the research consisted of a meticulous assessment of various advanced RANS models of fluid turbulence against experimental and numerical data for buoyancy-modified mixed convection flows, such flows being representative of low-flow-rate flows in the cores of nuclear reactors, both presently-operating Advanced Gas-cooled Reactors (AGRs) and proposed ‘Generation IV’ designs. For this part of the project, an in-house code (‘CONVERT’), a commercial CFD package (‘STAR-CD’) and an industrial code (‘Code_Saturne’) were used to generate results. Wide variations in turbulence model performance were identified. Comparison with the DNS data showed that the Launder-Sharma model best captures the phenomenon of heat transfer impairment that occurs in the ascending flow case; v^2-f formulations also performed well. The k-omega-SST model was found to be in the poorest agreement with the data. Cross-code comparison was also carried out and satisfactory agreement was found between the results.The research described above concerned flow in smooth passages; a second distinct contribution made in this thesis concerned the thermal-hydraulic performance of rib-roughened surfaces, these being representative of the fuel elements employed in the UK fleet of AGRs. All computations in this part of the study were undertaken using STAR-CD. This part of the research took four continuous and four discrete design factors into consideration including the effects of rib profile, rib height-to-channel height ratio, rib width-to-height ratio, rib pitch-to-height ratio, and Reynolds number. For each design factor, the optimum configuration was identified using the ‘efficiency index’. Through comparison with experimental data, the performance of different RANS turbulence models was also assessed. Of the four models, the v^2-f was found to be in the best agreement with the experimental data as, to a somewhat lesser degree were the results of the k-omega-SST model. The k-epsilon and Suga models, however, performed poorly. Structured and unstructured meshes were also compared, where some discrepancies were found, especially in the heat transfer results. The final stage of the study involved a simulation of a simplified 3-dimensional representation of an AGR fuel element using a 30 degree sector configuration. The v^2-f model was employed and comparison was made against the results of a 2D rib-roughened channel in order to assess the validity and relevance of the precursor 2D simulations of rib-roughened channels. It was shown that although a 2D approach is extremely useful and economical for ‘parametric studies’, it does not provide an accurate representation of a 3D fuel element configuration, especially for the velocity and pressure coefficient distributions, where large discrepancies were found between the results of the 2D channel and azimuthal planes of the 3D configuration.
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Bladh, Lisa. "Thermal-hydraulic modelling of Forsmark 1 NPP in TRACE : Validation versus the 25th of July, 2006 plant transient." Thesis, Uppsala University, Department of Physics and Astronomy, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-125297.

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There is a widespread use of thermal hydraulic codes in nuclear industry. The codesare used to analyse the transient and steady-state behavior of the nuclear powerplants. The US Nuclear Regulatory Commission that has long experience of developing such codes are now incorporating the capabilities of their earlier codes into one modern simulation tool, called TRACE. The code is under development and validation work is required especially in the field of BWR applications. Eventually the code is expected to replace similar codes such as TRAC and Relap5.

With this in mind, a TRACE model of Forsmark 1 has been set up to investigate how well it can simulate a plant transient. On the 25th of July, 2006 there was a disturbance at Forsmark 1 that caused the RPV water level and pressure to decrease.In this project, plant data acquired during the event are used to validate the model of Forsmark 1. The validation work is focused on comparing measured and calculated water and pressure levels in the RPC during the transient.

The results show qualitatively good agreement with the validation data, however during a period of the simulations there are large discrepancies concerning the pressure and water level in the RPV. In total, 13 simulations are performed, studying the influences of parameters such as simulation time-step size, the feed water flow boundary conditions and the steam line isolation valve characteristics. Based on the results of the simulations, a number of recommendations are made regarding suggestions for further work.

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Minav, Tatiana, Luca Papini, and Matti Pietola. "A Thermal Analysis of Direct Driven Hydraulics." Saechsische Landesbibliothek- Staats- und Universitaetsbibliothek Dresden, 2016. http://nbn-resolving.de/urn:nbn:de:bsz:14-qucosa-200125.

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This paper focuses on thermal analysis of a direct driven hydraulic setup (DDH). DDH combines the benefits of electric with hydraulic technology in compact package with high power density, high performance and good controllability. DDH enables for reduction of parasitic losses for better fuel efficiency and lower operating costs. This one-piece housing design delivers system simplicity and lowers both installation and maintenance costs. Advantages of the presented architecture are the reduced hydraulic tubing and the amount of potential leakage points. The prediction of the thermal behavior and its management represents an open challenge for the system as temperature is a determinant parameter in terms of performance, lifespan and safety. Therefore, the electro-hydraulic model of a DDH involving a variable motor speed, fixed-displacement internal gear pump/motors was developed at system level for thermal analysis. In addition, a generic model was proposed for the electric machine, energy losses dependent on velocity, torque and temperature was validated by measurements under various operative conditions. Results of model investigation predict ricing of temperature during lifting cycle, and flattened during lowering in pimp/motor. Conclusions are drawn concerning the DDH thermal behavior.
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Lin, Fangcheng, and 林芳正. "Investigations of Control system and Thermal-Hydraulic modeling in PCTRAN." Thesis, 2003. http://ndltd.ncl.edu.tw/handle/56601177843256355459.

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碩士
國立清華大學
工程與系統科學系
91
ABSTRACTS PCTRAN is a reactor transient and accident simulation software program that operates on a personal computer. It was developed by Taiwan Power Company and Micro-Simulation Technology (MST). PCTRAN have high resolution color display and interactive control capability enable versatile, high speed simulation, yet low cost transient simulation. We can use it to simulate various transients and events in order to assess the safety of nuclear power plants. In the present thesis, we will descriptive all of the PCTRAN model structure that it is include source code, VB interface and the data base structure correlation. We also detail investigations into PCTRAN system control blocks. Due to the fact that PCTRAN can not include all of the plant systems and transient initiation events, the operator should be familiar with plant basics in order to complete a reasonable and logical PCTRAN simulation run with its built-in existing functions. Under current basic PCTRAN structures, we can add or modify necessary VB objects and source codes to develop a proper tool for transient analysis in a nuclear power plant.
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Книги з теми "Thermal-hydraulic modeling"

1

Rimkevicius, S. Modelling of thermal hydraulic transient processes in nuclear power plants: Ignalina compartments. Redding, NY: Begell House, 2007.

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2

Ušpuras, Eugenijus, and Algirdas Kaliatka. Basis of Modeling of Thermal Hydraulic Processes in Nuclear Reactors. KTU leidykla „Technologija“, 2013. http://dx.doi.org/10.5755/e01.9786090209356.

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Частини книг з теми "Thermal-hydraulic modeling"

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Genc, Derya, Jeramy C. Ashlock, Bora Cetin, Kristen Cetin, Masrur Mahedi, Robert Horton, and Halil Ceylan. "Monitoring and Modeling of Soil Thermal and Hydraulic Behavior Beneath a Granular-Surfaced Roadway." In Lecture Notes in Civil Engineering, 877–88. Cham: Springer International Publishing, 2021. http://dx.doi.org/10.1007/978-3-030-77234-5_72.

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2

Sidi-Ali, Kamel, Djaber Ailem, El Moundir Medouri, and Toufik Belmrabet. "Thermal Hydraulic Modeling of a Nuclear Reactor Core Channel Using CFD; Application for an EPR." In Lecture Notes in Mechanical Engineering, 9–16. Cham: Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-11827-3_2.

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Essen, D., G. Küpers, H. Mes, and B. V. Neratoom. "Thermal Hydraulic Modelling Studies on Heat Exchanging Components." In Research in Numerical Fluid mechanics, 30–44. Wiesbaden: Vieweg+Teubner Verlag, 1987. http://dx.doi.org/10.1007/978-3-322-89729-9_3.

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Huang, Hai, Paul Meakin, and Jing Zhou. "Quasistatic Discrete Element Modeling of Hydraulic and Thermal Fracturing Processes in Shale and Low-Permeability Crystalline Rocks." In Hydraulic Fracture Modeling, 75–109. Elsevier, 2018. http://dx.doi.org/10.1016/b978-0-12-812998-2.00004-7.

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Dixit, Uday Shanker, and Rajkumar Shufen. "Finite element method modeling of hydraulic and thermal autofrettage processes." In Mechanics of Materials in Modern Manufacturing Methods and Processing Techniques, 31–69. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-818232-1.00002-3.

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Li, Hongzhi, and Yifan Zhang. "Heat Transfer and Fluid Flow Modeling for Supercritical Fluids in Advanced Energy Systems." In Handbook of Research on Advancements in Supercritical Fluids Applications for Sustainable Energy Systems, 388–422. IGI Global, 2021. http://dx.doi.org/10.4018/978-1-7998-5796-9.ch011.

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This chapter aims to model the supercritical fluids thermal hydraulics behaviors including heat transfer, pressure drops, and flow instabilities for the purpose of accurate design and efficient safe operation of advanced energy systems. At first, the convection heat transfer models considering the effect of nonlinear properties and the effect of buoyancy and acceleration have been provided and discussed. Secondly, the hydraulic resistance models for supercritical fluids have been selected and suggested for different conditions. Thirdly, the published models for supercritical flow instabilities based on four different regional partitions are summarized and clarified. At last, two typical case studies have been provided to further intuitively elaborate the thermal hydraulics of supercritical fluids within the advanced energy systems.
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Chalaev, Djamalutdin, and Nina Silnyagina. "DEVELOPMENT OF HIGH EFFICIENT SHELL-AND-TUBE HEAT EXCHANGERS BASED ON PROFILED TUBES." In Integration of traditional and innovation processes of development of modern science. Publishing House “Baltija Publishing”, 2020. http://dx.doi.org/10.30525/978-9934-26-021-6-42.

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The use of advanced heat transfer surfaces (corrugated tubes of various modifications) is an effective way to intensify the heat transfer and improve the hydraulic characteristics of tubular heat exchangers. The methods for evaluating the use of such surfaces as working elements in tubular heat exchangers have not been developed so far. The thermal and hydrodynamic processes occurring in the tubes with the developed surfaces were studied to evaluate the efficiency of heat exchange therein. Thin-walled corrugated flexible stainless steel tubes of various modifications were used in experimental studies. The researches were carried out on a laboratory stand, which was designed as a heat exchanger type "tube in tube" with a corrugated inner tube. The stand was equipped with sensors to measure the thermal hydraulic flow conditions. The comparative analysis of operation modes of the heat exchanger with a corrugated inner tube of various modifications and the heat exchanger with a smooth inner tube was performed according to the obtained data. Materials and methods. A convective component of the heat transfer coefficient of corrugated tube increased significantly at identical flow conditions comparing with a smooth tube. Increasing the heat transfer coefficient was in the range of 2.0 to 2.6, and increased with increasing Reynolds number. The increase in heat transfer of specified range outstripped the gain of hydraulic resistance caused by increase of the flow. Results and discussion. CFD model in the software ANSYS CFX 14.5 was adapted to estimate the effect of the tube geometry on the intensity of the heat transfer process. A two-dimensional axially symmetric computer model was used for the calculation. The model is based on Reynolds equation (Navier-Stokes equations for turbulent flow), the continuity equation and the energy equation supplemented by the conditions of uniqueness. SST-turbulence model was used for the solution of the equations. The problem was solved in the conjugate formulation, which allowed assessing the efficiency of heat exchange, depending on various parameters (coolant temperature, coolant velocity, pressure). The criteria dependences were obtained Nu = f (Re, Pr). Conclusions. The use a corrugated tube as a working element in tubular heat exchangers can improve the heat transfer coefficient of 2.0 - 2.6 times, with an increase in hydraulic resistance in the heat exchanger of 2 times (compared with the use of smooth tubes). The criteria dependences obtained on the basis of experimental studies and mathematical modeling allow developing a methodology for engineering calculations for the design of new efficient heat exchangers with corrugated tubes.
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Demazière, Christophe. "Neutronic/thermal-hydraulic coupling." In Modelling of Nuclear Reactor Multi-physics, 311–36. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-815069-6.00006-4.

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ur Rehman, Obaid, Nor Erniza Mohammad Rozali, and Marappa Gounder Ramasamy. "Fouling and Mechanism." In Heat Transfer [Working Title]. IntechOpen, 2022. http://dx.doi.org/10.5772/intechopen.105878.

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Fouling is the deposition of material on the heat transfer surface which reduces the film heat transfer coefficient. The impact of fouling on the heat exchanger is manifested as the reduction of thermal and hydraulic performance, in which the latter has a minor effect. This factor needs to be considered when calculating the effectiveness of the heat exchanger. During the design of heat exchangers, the fouling factor increases the required heat transfer area, which adds extra manufacturing costs. With less efficient heat exchangers, the economic cost of fouling is related to excess fuel consumption, loss of production, and maintenance or cleaning. The extra fuel consumption also damages the environment by increasing greenhouse gas production. Although much of the research work has been done on modeling and predicting fouling, it is still a poorly understood phenomenon representing the complexity of its mechanism. The common fouling mitigation action after the onset of fouling is to optimize the operating condition, e.g., increase the bulk flow velocity or decrease surface temperature. However, many quantitative and semi-empirical models have been developed to predict the fouling rate for preventive actions and optimizing cleaning schedules.
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Abootalebi, P., and G. Siemens. "Thermal-hydraulic modelling a Canadian deep geological repository." In Energy Geotechnics, 265–70. CRC Press, 2016. http://dx.doi.org/10.1201/b21938-43.

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Тези доповідей конференцій з теми "Thermal-hydraulic modeling"

1

Bruyere, Vincent, Nicolas Authier, and Patrick Namy. "Thermal-hydraulic modeling and acoustic correlation (Conference Presentation)." In Laser Applications in Microelectronic and Optoelectronic Manufacturing (LAMOM) XXV, edited by Gediminas Račiukaitis, Carlos Molpeceres, Aiko Narazaki, and Jie X. Qiao. SPIE, 2020. http://dx.doi.org/10.1117/12.2544115.

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2

Tarrad, Ali H., Rafea A. Al-Baldawi, and Ahmad A. Al-Issa. "Implementation of Expert System Modeling to Thermal-Hydraulic Design of Hydraulic Systems." In ASME 2014 Power Conference. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/power2014-32038.

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Thermal and hydrodynamic concepts are of a vital importance; therefore their assessments are unavoidable for the purpose of hydraulic systems. The present study implements the practical updated knowledge of the expertise for both of the hydraulic and thermal fields in an expert system model. This is implemented in order to improve the performance of hydraulic system by considering the thermal effect on the hydraulic system operation. Accordingly, a computer program (Hydraulic System Calculations), designated as (HSC) implements a Visual Basic language in the Microsoft Visual Studio 2010 software has been built. Regardless of the design requirements, the code is capable to deal with (18) possible connection types of the actuators, in series or parallel, arrangements. The suggested code provides the designer with a number of choices, different kind of connections, to resolve the problem of hydraulic oil overheating which may arise during the continuous operation of the hydraulic unit. As a result, the (HSC) is able to minimize the human errors, effort, time and cost of hydraulic machine design.
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Teixeira, Jose C. F., Antonio C. Oliveira, and Senhorinha F. C. F. Teixeira. "Thermal Hydraulic Modeling of Shell and Tube Heat Exchangers." In The 15th International Heat Transfer Conference. Connecticut: Begellhouse, 2014. http://dx.doi.org/10.1615/ihtc15.cpm.009525.

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4

Yonglin Li, Xinbing Su, Haojun Xu, and Dawei Li. "Thermal-hydraulic modeling and simulation of high power hydro-motor." In 2008 Asia Simulation Conference - 7th International Conference on System Simulation and Scientific Computing (ICSC). IEEE, 2008. http://dx.doi.org/10.1109/asc-icsc.2008.4675479.

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5

Yang, Changjiang. "RELAP5 Core Modeling Study for Level 1 PRA Thermal-Hydraulic Analyses." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16125.

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Thermal-hydraulic (T/H) analyses are used to support the level 1 Probabilistic Risk Assessment (PRA) success criteria and the manual operation time. To address the multiple-failure accident scenarios that are considered in the PRA, usually numerous T/H analyses were performed. So it is meaningful to develop a relative simple T/H model with acceptable accuracy for level 1 PRA T/H analyses. To achieve this object, the core modeling effects on the core damage progression were studied according to ASME/ANS RA-Sa-2009. Two types of core modeling methods were studied, including single channel core modeling and multi-channel core modeling. For the single channel core modeling, the study was focused on the axial nodes number effect. For the multi-channel core modeling, the cross-flow effects were studied. Several cases were calculated on a 3-loop PWR medium size break LOCA core damage scenario with Relap5/MOD3.2. Some key parameters related to the core state, such as peak cladding temperature (PCT), core water level and coolant inventory, were compared and analyzed. A kind of core modeling for level 1 PRA T/H analyses was suggested at the end.
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6

Kozlowski, T., and T. Mui. "Preliminary Fuel Performance and Thermal Hydraulic Modeling of the MPCMIV Benchmark." In 2020 ANS Virtual Winter Meeting. AMNS, 2020. http://dx.doi.org/10.13182/t123-33402.

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7

Delfino, Claudio, and Birol Aktas. "Modeling of Safety/Relief Valves With Thermal-Hydraulic System Computer Codes." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49336.

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A generic guideline for thermal-hydraulic (T-H) simulation of multiple bank safety relief valves (SRVs) was developed. To test the guideline, the Full Integral Simulation Test (FIST) 6PMC2 was simulated with the consolidated T-H code of USNRC, TRACE. The FIST 6PMC2 experiment simulates the response of a generic BWR/6 plant to a Main Steam-line Isolation Valve (MSIV) closure without reactor scram. During the test, the HPCS is unavailable and not used. The only inventory make-up system available is the Reactor Core Isolation Cooling (RCIC). This experiment can also be considered proto-typical of a BWR ATWS-like scenario. The simulation is largely dominated by the transient behavior of the SRVs. In this study, the experimental data was analyzed and used to check the modeling guideline for SRVs. The guideline relies on only the data available prior to an experiment or any other analysis, e.g. valve flow coefficients, inlet hydraulic diameters, etc. The study also revealed deficiencies in the “then” current valve model of the TRACE code which were subsequently corrected. The study demonstrates that the T-H models can simulate the operational behavior of SRVs very accurately while rather simple mistakes can be very damaging at the same time.
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8

Sun, Peiwei, and Jin Jiang. "Thermal-Hydraulic Modeling of CANDU-SCWR and Linear Dynamic Model Development." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29780.

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CANDU-SCWR is one Generation IV reactor being developed in Canada. Significant amount of efforts has been made to develop CANDU-SCWR. Little work has been done on the dynamic analysis and control design. To observe the dynamic behaviours of CANDU-SCWR, the detailed CANDU-SCWR thermal-hydraulic model is developed. The movable boundary method is adopted for CANDU-SCWR thermal-hydraulic modeling. The benefits of adopting movable boundary are derived from the comparisons with the fixed boundary method. The steady-state results agree well with the design data. The responses of CANDU-SCWR reactor to different disturbances are simulated and analyzed and the results are reasonable in theory. Linear dynamic models are derived from simulation data of CANDU-SCWR thermal-hydraulic model around the design operating point using a system identification technique to facilitate the control system design. The linear dynamic models are validated and it is shown that they can describe the dynamic characteristics of CANDU-SCWR around the design point accurately.
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Heckmann, Klaus, Jürgen Sievers, and Fabian Weyermann. "Leak Rate Computation: Flow Resistance vs. Thermal-Hydraulic Aspect." In ASME 2018 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/pvp2018-84534.

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The computation of mass flow rates through crack-like defects in piping systems of light water reactors requires typically the description of two-phase flow conditions. The computed discharge rate depends on the crack opening area, the thermal-hydraulic modeling of the flow, and the flow resistance of the crack. Several models have been proposed to characterize the critical flow through crack-like defects. An evaluation of advantages and shortcomings of the different models with regard to the interaction of the three different parts (crack opening area, thermal-hydraulic modeling, flow resistance) has been performed. In this paper, the flow resistance modeling from several approaches is discussed, and compared with a database from eight different testing programs. Five different flow models are applied to analyze a database of more than 800 leak rate measurements for subcooled water from twelve different experimental programs. It is shown that the correct modeling of the flow resistance is crucial for a best estimate reproduction of the measured data. It turns out that generally, equilibrium models are about as good as non-equilibrium models. The data were processed with the GRS software WinLeck which includes different analytical approaches for the calculation of crack sizes and leak rates in piping components. The most reliable results within the model selection are produced by the CDR model (Critical Discharge Rate) of the ATHLET code (Analysis of Thermal-hydraulics of Leaks and Transients) developed by GRS. As a conclusion, the accurate modeling of form losses and frictional pressure losses for critical discharge flow rates through crack-like leaks are essential for a reliable prediction of flow rates. Uncertainties in leak rate computations results arise due to the lack of information about the flow geometry and its associated drag. The assumed flow resistance of a through-wall crack influences the computed leak rate as significant as the phase-change- and flow-models. The manifest difference between equilibrium models (Pana, Estorf) and non-equilibrium models (Henry, ATHLET-CDR) seems to be less significant than the pressure loss issue. One can conjecture that, for crack-like through-wall defects, friction effects play a more important role than non-equilibrium effects.
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10

Wang, Cong, Danmei Xie, Peng Zhang, Xinggang Yu, and Xiuqun Hou. "Investigation on Modeling Thermal-Hydraulic System of CPR1000 NPP Based on RELAP5." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-31096.

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Based on the best-estimate program RELAP5/MOD4.0, a full-scope thermal-hydraulic model with reference to CPR1000 nuclear power plant is established in this paper, which includes the thermal-hydraulic systems of conventional island as well as the primary nuclear island which has already been researched in traditional safety analysis. Therefore, this paper mainly details the numerical model of the turbine and other parts of the conventional island thermodynamic system. A comparison between the calculated results in steady-state and the actual data of reactor demonstrates a fine consistency, thus verifying the accuracy and reliability of the model. In addition, the steam parameter changes are numerically simulated during the steam turbine’s off-design operating condition such as back pressure variation and the variation trends are the same as the actual situation of nuclear power plants.
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Звіти організацій з теми "Thermal-hydraulic modeling"

1

Keefer, R. H., and L. W. Keeton. Review of computational thermal-hydraulic modeling. Office of Scientific and Technical Information (OSTI), December 1995. http://dx.doi.org/10.2172/291150.

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Hamm, L. L., and M. A. Jr Shadday. Subchannel thermal-hydraulic modeling of an APT tungsten target rod bundle. Office of Scientific and Technical Information (OSTI), September 1997. http://dx.doi.org/10.2172/578631.

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3

Xia, Yidong, Joshua Hansel, Ray A. Berry, David Andrs, and Richard C. Martineau. Preliminary Study on the Suitability of a Second-Order Reconstructed Discontinuous Galerkin Method for RELAP-7 Thermal-Hydraulic Modeling. Office of Scientific and Technical Information (OSTI), September 2017. http://dx.doi.org/10.2172/1468483.

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4

Roberts, J. S., S. L. Woosley, D. L. Lessor, and C. Strachan. Preliminary investigation of the potential for transient vapor release events during in situ vitrification based on thermal- hydraulic modeling. Office of Scientific and Technical Information (OSTI), July 1992. http://dx.doi.org/10.2172/10166606.

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5

Roberts, J. S., S. L. Woosley, D. L. Lessor, and C. Strachan. Preliminary investigation of the potential for transient vapor release events during in situ vitrification based on thermal- hydraulic modeling. Office of Scientific and Technical Information (OSTI), July 1992. http://dx.doi.org/10.2172/7310002.

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6

McGraw, D., and P. Oberlander. Groundwater Flow and Thermal Modeling to Support a Preferred Conceptual Model for the Large Hydraulic Gradient North of Yucca Mountain. Office of Scientific and Technical Information (OSTI), December 2007. http://dx.doi.org/10.2172/921093.

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7

Corradin, Michael, M. Anderson, M. Muci, Yassin Hassan, A. Dominguez, Akira Tokuhiro, and K. Hamman. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling. Office of Scientific and Technical Information (OSTI), October 2014. http://dx.doi.org/10.2172/1183658.

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