Tesis sobre el tema "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.
Texto completoQC 20161208
DEMO-JHR
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.
Texto completoLeem, 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.
Texto completoHan, 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.
Texto completoChen, 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.
Texto completoSvensson, 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.
Texto completoDet 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.
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.
Texto completoBladh, 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.
Texto completoThere 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.
Minav, Tatiana, Luca Papini y 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.
Texto completoLin, Fangcheng y 林芳正. "Investigations of Control system and Thermal-Hydraulic modeling in PCTRAN". Thesis, 2003. http://ndltd.ncl.edu.tw/handle/56601177843256355459.
Texto completo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.
Xu, Kang. "Thermal and hydraulic modeling and control of a district heating system". Thesis, 2012. http://spectrum.library.concordia.ca/975015/1/Xu_MASc_S2013.pdf.
Texto completo邱茗秀. "Investigations of Thermal-Hydraulic modeling and Setup of Offsite Dose Calculation Capabilities in PCTRAN". Thesis, 2003. http://ndltd.ncl.edu.tw/handle/41493362633215289001.
Texto completo國立清華大學
工程與系統科學系
91
The thesis investigates thermal-hydraulic modeling of PCTRAN-PWR and establishes offsite Dose Calculation Capabilities in PCTRAN-PWR.PCTRAN-PWR is referred to a PCTRAN version for Westinghouse nuclear power plant. The thesis has generalized major thermal-hydraulic theory about core, pressurizer and steam generator and major portion of calculation function form PCTRAN-PWR The thesis has also established XOQ calculation modeling in PCTRAN-PWR. XOQ is the rate of concentration of the effluent over its flow rate. Using the XOQ calculation modeling of this thesis and dose calculation modeling of PCTRAN-PWR, we can calculate and display the whole body dose rate and thyroid dose rate and their integrated dose in the range of offsite radius 5 km according to weather condition and wind velocity and direction at real time. It will be useful for Emergency plan maneuvers.
Rulff, David. "Modeling Satellite District Heating and Cooling Networks". Thesis, 2011. http://hdl.handle.net/1807/31418.
Texto completoChiang, Keng-Yen y 江庚晏. "Phenomenal Investigations of the Thermal-hydraulic Responses of Multi-Dimensional simulation and Modeling in Core and Downcomer during the L2-5 Test of LOFT Using RELAP5-3D/K". Thesis, 2009. http://ndltd.ncl.edu.tw/handle/07358771141141666665.
Texto completo國立清華大學
工程與系統科學系
97
In this study, RELAP5-3D/K code is used to simulate the L2-5 test of Loss of Fluid Test (LOFT) facility. RELAP5-3D is a multi-dimensional reactor system thermal-hydraulic analysis code. In the present simulation, the core and downcomer are modeled as inconnected three dimensional components. The results of the simulation are compared with the results of the RELAP5-3D/K one-dimensional analysis. The purpose of this study is to qualify the margin of the safety analysis relatd to the design criteria of loss of coolant accident. The results show that the peak cladding temperatures (PCT) as predicted by the 3-D model of core and downcomer is about 180℉ lower than that of the 1-D model of the corresponding components. The results show that the predicted rise of cladding temperature in the blowdown phase of accident is higher in the case that core is simulated three-dimensionally. It is also demonstrated that the chimney effect in the 3-D core simulation is stronger than the case of 1-D simulation of core, which tends to lower the PCT in LOCA analysis. Chimney effect is referred as the coolant flow rate in a hot channel will increase by sucking in the coolant from nearby colder region. Modeling the downocmer three-dimensionally has a tendency to reduce the predicted crossflow in the annular region surrounded the core barrel. It implies that large amount of injected emergency cooling water will flow downward into the lower plenum. The initiation of core reflooding will be earlier for the core with three-dimensional simulation of downcomer.
Oh, Myung-Do. "Thermal-hydraulic modelling and analysis for large-scale vapor explosions". 1985. http://catalog.hathitrust.org/api/volumes/oclc/13190715.html.
Texto completoTypescript. Vita. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaves 392-401).