Дисертації з теми "Thermal-hydraulic modeling"

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

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.

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.

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.

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.

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.

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.

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.

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.

A district heating system is a centralized heating system widely used for space heating. They offer economic benefits and are acknowledged to be more energy efficient. Their energy efficiency can be further improved by optimally controlling and operating the overall system. With this as the motivation a thermal and hydraulic model of a district heating is developed in this thesis. The developed model consists of a boiler, six buildings, hot water distribution network, circulating pump, balancing and control valves and terminal heaters. Both dynamic and steady state hydraulic simulations were made to study supply, return water and zone temperature response; Pressure distribution in the piping network under different load conditions. A relationship between balance valve settings and outdoor temperature (heating load) was determined. It was shown that proper setting of balance valves as a function of heating load improves energy efficiency. PI controllers were designed for the boiler and zone temperature control. Closed loop simulations are presented to show the control performance. By using steady state optimization technique optimal set points for the boiler temperature and near-optimal balance valve positions as a function of outdoor temperature were determined. Simulation results show that the use of optimal set points and balance valve settings up to 20% energy can be saved compared to the conventional outdoor air reset control strategy. Simulation results under operating conditions are presented.

碩士
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.

Satellite District Heating and Cooling (DHC) systems offer an alternative structure to conventional, centralized DHC networks. Both use a piping network carrying steam or water to connect disparate building heating and cooling loads together, providing a platform for improving energy efficiency, reducing emissions, and incorporating alternative means of energy generation. However, satellite DHC networks incorporate thermal production units that are distributed amongst the buildings nodes, which offers greater operational flexibility and reduced capital cost savings for applications using existing building stock. This study was focused on the development of the methodology behind a comprehensive energy model that can assess the practical and financial viability of satellite DHC network scenarios. A detailed scenario application of the model demonstrated significant energy savings and investment potential. Additionally, environmental assessment methods and alternative generation technology were explored in supplementary studies of Deep Lake Water Cooling (DLWC) and building-scale Combined Heat and Power (CHP).

碩士
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.

Thesis (Ph. D.)--University of Wisconsin--Madison, 1985.
Typescript. Vita. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaves 392-401).