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Doney, George Daniel. "Acoustic boiling detection." Thesis, Massachusetts Institute of Technology, 1994. http://hdl.handle.net/1721.1/28110.
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Aziz, S. "Forced convection film boiling on spheres." Thesis, University of Oxford, 1986. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.371536.
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Samaroo, Randy. "The effects of geometric, flow, and boiling parameters on bubble growth and behavior in subcooled flow boiling." Thesis, The City College of New York, 2016. http://pqdtopen.proquest.com/#viewpdf?dispub=10159915.
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Air bubble injection and subcooled flow boiling experiments have been performed to investigate the liquid flow field and bubble nucleation, growth, and departure, in part to contribute to the DOE Nuclear HUB project, Consortium for Advanced Simulation of Light Water Reactors (CASL). The main objective was to obtain quantitative data and compartmentalize the many different interconnected aspects of the boiling process — from the channel geometry, to liquid and gas interactions, to underlying heat transfer mechanisms.
The air bubble injection experiments were performed in annular and rectangular geometries and yielded data on bubble formation and departure from a small hole on the inner tube surface, subsequent motion and deformation of the detached bubbles, and interactions with laminar or turbulent water flow. Instantaneous and ensemble- average liquid velocity profiles have been obtained using a Particle Image Velocimetry technique and a high speed video camera. Reynolds numbers for these works ranged from 1,300 to 7,700.
Boiling experiments have been performed with subcooled water at atmospheric pres- sure in the same annular channel geometry as the air injection experiments. A second flow loop with a slightly larger annular channel was constructed to perform further boiling experiments at elevated pressures up to 10 bar. High speed video and PIV measurements of turbulent velocity profiles in the presence of small vapor bubbles on the heated rod are presented. The liquid Reynolds number for this set of experiments ranged from 5,460 to 86,000. It was observed that as the vapor bubbles are very small compared to the injected air bubbles, further experiments were performed using a microscopic objective to obtain higher spatial resolution for velocity fields near the heated wall. Multiple correlations for the bubble liftoff diameter, liftoff time and bub- ble history number were evaluated against a number of experimental datasets from previous works, resulting in a new proposed correlations that account for fluid prop- erties that vary with pressure, heat flux, and variations in geometry.
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Zeng, Yi. "Effect of peripheral wall conduction in pool boiling." Thesis, University of Ottawa (Canada), 1985. http://hdl.handle.net/10393/22397.
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Gao, Qi. "A boiling water reactor simulator for stability analysis." Thesis, Massachusetts Institute of Technology, 1996. http://hdl.handle.net/1721.1/39053.
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Lange, Carsten. "Advanced nonlinear stability analysis of boiling water nuclear reactors." Doctoral thesis, Saechsische Landesbibliothek- Staats- und Universitaetsbibliothek Dresden, 2009. http://nbn-resolving.de/urn:nbn:de:bsz:14-qucosa-24954.
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This thesis is concerned with nonlinear analyses of BWR stability behaviour, contributing to a deeper understanding in this field. Despite negative feedback-coefficients of a BWR, there are operational points (OP) at which oscillatory instabilities occur. So far, a comprehensive and an in-depth understanding of the nonlinear BWR stability behaviour are missing, even though the impact of the significant physical parameters is well known. In particular, this concerns parameter regions in which linear stability indicators, like the asymptotic decay ratio, lose their meaning. Nonlinear stability analyses are usually carried out using integral (system) codes, describing the dynamical system by a system of nonlinear partial differential equations (PDE). One aspect of nonlinear BWR stability analyses is to get an overview about different types of nonlinear stability behaviour and to examine the conditions of their occurrence. For these studies the application of system codes alone is inappropriate. Hence, in the context of this thesis, a novel approach to nonlinear BWR stability analyses, called RAM-ROM method, is developed. In the framework of this approach, system codes and reduced order models (ROM) are used as complementary tools to examine the stability characteristics of fixed points and periodic solutions of the system of nonlinear differential equations, describing the stability behaviour of a BWR loop. The main advantage of a ROM, which is a system of ordinary differential equations (ODE), is the possible coupling with specific methods of the nonlinear dynamics. This method reveals nonlinear phenomena in certain regions of system parameters without the need for solving the system of ROM equations. The stability properties of limit cycles generated in Hopf bifurcation points and the conditions of their occurrence are of particular interest. Finally, the nonlinear phenomena predicted by the ROM will be analysed in more details by the system code. Hence, the thesis is not focused on rendering more precisely linear stability indicators like DR. The objective of the ROM development is to develop a model as simple as possible from the mathematical and numerical point of view, while preserving the physics of the BWR stability behaviour. The ODEs of the ROM are deduced from the PDEs describing the dynamics of a BWR. The system of ODEs includes all spatial effects in an approximated (spatial averaged) manner, e.g. the space-time dependent neutron flux is expanded in terms of a complete set of orthogonal spatial neutron flux modes. In order to simulate the stability characteristics of the in-phase and out-of-phase oscillation mode, it is only necessary to take into account the fundamental mode and the first azimuthal mode. The ROM, originally developed at PSI in collaboration with the University of Illinois (PSI-Illinois-ROM), was upgraded in significant points: • Development and implementation of a new calculation methodology for the mode feedback reactivity coefficients (void and fuel temperature reactivity) • Development and implementation of a recirculation loop model; analysis and discussion of its impact on the in-phase and out-of-phase oscillation mode • Development of a novel physically justified approach for the calculation of the ROM input data • Discussion of the necessity of consideration of the effect of subcooled boiling in an approximate manner With the upgraded ROM, nonlinear BWR stability analyses are performed for three OPs (one for NPP Leibstadt (cycle7), one for NPP Ringhals (cycle14) and one for NPP Brunsbüttel (cycle16) for which measuring data of stability tests are available. In this thesis, the novel approach to nonlinear BWR stability analyses is extensively presented for NPP Leibstadt. In particular, the nonlinear analysis is carried out for an operational point (OP), in which an out-of-phase power oscillation has been observed in the scope of a stability test at the beginning of cycle 7 (KKLc7_rec4). The ROM predicts a saddle-node bifurcation of cycles, occurring in the linear stable region, close to the KKLc7_rec4-OP. This result allows a new interpretation of the stability behaviour around the KKLc7_rec4-OP. The results of this thesis confirm that the RAM-ROM methodology is qualified for nonlinear BWR stability analysesDie vorliegende Dissertation leistet einen Beitrag zum tieferen Verständnis des nichtlinearen Stabilitätsverhaltens von Siedewasserreaktoren (SWR). Trotz der Tatsache, dass in diesem technischen System nur negative innere Rückkopplungskoeffizienten auftreten, können in bestimmten Arbeitspunkten oszillatorische Instabilitäten auftreten. Obwohl relativ gute Kenntnisse über die signifikanten physikalischen Einflussgrößen vorliegen, fehlt bisher ein umfassendes Verständnis des SWR-Stabilitätsverhaltens. Das betrifft insbesondere die Bereiche der Systemparameter, in denen lineare Stabilitätsindikatoren, wie zum Beispiel das asymptotische Decay Ratio (DR), ihren Sinn verlieren. Die nichtlineare Stabilitätsanalyse wird im Allgemeinen mit Systemcodes (nichtlineare partielle Differentialgleichungen, PDG) durchgeführt. Jedoch kann mit Systemcodes kein oder nur ein sehr lückenhafter Überblick über die Typen von nichtlinearen Phänomenen, die in bestimmten System-Parameterbereichen auftreten, erhalten werden. Deshalb wurde im Rahmen der vorliegenden Arbeit eine neuartige Methode (RAM-ROM Methode) zur nichtlinearen SWR-Stabilitätsanalyse erprobt, bei der integrale Systemcodes und sog. vereinfachte SWR-Modelle (ROM) als sich gegenseitig ergänzende Methoden eingesetzt werden, um die Stabilitätseigenschaften von Fixpunkten und periodischen Lösungen (Grenzzyklen) des nichtlinearen Differentialgleichungssystems, welches das Stabilitätsverhalten des SWR beschreibt, zu bestimmen. Das ROM, in denen das dynamische System durch gewöhnliche Differentialgleichungen (GDG) beschrieben wird, kann relativ einfach mit leistungsfähigen Methoden aus der nichtlinearen Dynamik, wie zum Beispiel die semianalytische Bifurkationsanalyse, gekoppelt werden. Mit solchen Verfahren kann, ohne das DG-System explizit lösen zu müssen, ein Überblick über mögliche Typen von stabilen und instabilen oszillatorischen Verhalten des SWR erhalten werden. Insbesondere sind die Stabilitätseigenschaften von Grenzzyklen, die in Hopf-Bifurkationspunkten entstehen, und die Bedingungen, unter denen sie auftreten, von Interesse. Mit dem Systemcode (RAMONA5) werden dann die mit dem ROM vorhergesagten Phänomene in den entsprechenden Parameterbereichen detaillierter untersucht (Validierung des ROM). Die Methodik dient daher nicht der Verfeinerung der Berechnung linearer Stabilitätsindikatoren (wie das DR). Das ROM-Gleichungssystem entsteht aus den PDGs des Systemcodes durch geeignete (nichttriviale) räumliche Mittelung der PDG. Es wird davon ausgegangen, dass die Reduzierung der räumlichen Komplexität die Stabilitätseigenschaften des SWR nicht signifikant verfälschen, da durch geeignete Mittlungsverfahren, räumliche Effekte näherungsweise in den GDGs berücksichtig werden. Beispielsweise wird die raum- und zeitabhängige Neutronenflussdichte nach räumlichen Moden entwickelt, wobei für eine Simulation der Stabilitätseigenschaften der In-phase- und Out-of-Phase-Leistungsoszillationen nur der Fundamentalmode und der erste azimuthale Mode berücksichtigt werden muss. Das ROM, welches ursprünglich am Paul Scherrer Institut (PSI, Schweiz) in Zusammenarbeit mit der Universität Illinois (USA) entwickelt wurde, ist in zwei wesentlichen Punkten erweitert und verbessert worden: • Entwicklung und Implementierung einer neuen Methode zur Berechnung der Rückkopplungsreaktivitäten • Entwicklung und Implementierung eines Modells zur Beschreibung der Rezirkulationsschleife (insbesondere wurde der Einfluss der Rezirkulationsschleife auf den In-Phase-Oszillationszustand und auf den Out-of-Phase-Oszillationszustand untersucht) • Entwicklung einer physikalisch begründeten Methode zur Berechnung der ROM-Inputdaten • Abschätzung des Einflusses des unterkühlten Siedens im Rahmen der ROM-Näherungen Mit dem erweiterten ROM wurden nichtlineare Stabilitätsanalysen für drei Arbeitspunkte (KKW Leibstadt (Zyklus 7) KKW Ringhals (Zyklus 14) und KKW Brunsbüttel (Zyklus 16)), für die Messdaten vorliegen, durchgeführt. In der Dissertationsschrift wird die RAM-ROM Methode ausführlich am Beispiel eines Arbeitspunktes (OP) des KKW Leibstadt (KKLc7_rec4-OP), in dem eine aufklingende regionale Leistungsoszillation bei einem Stabilitätstest gemessen worden ist, demonstriert. Das ROM sagt die Existenz eines Umkehrpunktes (saddle-node bifurcation of cycles, fold-bifurcation) voraus, der sich im linear stabilen Gebiet nahe der Stabilitätsgrenze befindet. Mit diesem ROM-Ergebnis ist eine neue Interpretation der Stabilitätseigenschaften des KKLc7_rec4-OP möglich. Die Resultate der in der Dissertation durchgeführten RAM-ROM Analyse bestätigen, dass das weiterentwickelte ROM für die Analyse des Stabilitätsverhaltens realer Leistungsreaktoren qualifiziert wurde
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Chen, Xiangbin. "Direct numerical simulation of nuclear boiling on nanopatterned surface." Electronic Thesis or Diss., Sorbonne université, 2024. https://accesdistant.sorbonne-universite.fr/login?url=https://theses-intra.sorbonne-universite.fr/2024SORUS289.pdf.
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Les dynamiques de l'ébullition nucléée sont intimement influencées par les interactions entre les domaines fluide et solide, en particulier dans des conditions de petits angles de contact et de phénomènes d'interface complexes. Cette thèse présente le développement et l'application de méthodes numériques avancées pour simuler ces interactions avec une précision accrue. Au cœur de ce travail se trouve la mise en œuvre d'une méthode de convection Volume de Fluide (VOF) conservatrice de masse, intégrée de manière fluide avec une approche de domaine solide intégré. Cette technique est spécifiquement conçue pour relever les défis de la capture précise des dynamiques d'ébullition à de petits angles de contact.Le cadre numérique est construit sur la plateforme Basilisk, développée par Stéphane Popinet, en utilisant un modèle à fluide unique où la méthode VOF capture efficacement les dynamiques d'interface fluide. Un modèle de force de surface continue est employé pour représenter avec précision les effets de tension de surface, tandis que le modèle solide intégré assure un couplage robuste entre les domaines fluide et solide. Pour améliorer encore la fidélité de la simulation, le modèle de changement de phase de Leon Malan est intégré, incorporant une méthode d'interface de convection en deux étapes et une équation conservatrice d'énergie pour gérer les complexités de la transition de phase. De plus, un modèle de résistance thermique interfaciale développé par Lubomír Moravcík est implémenté, quantifiant la résistance thermique à l'interface fluide-solide.Un exercice de validation rigoureux est réalisé, démontrant une forte concordance avec les données expérimentales de référence et les modèles théoriques établis. Les contributions clés de ce travail incluent l'amélioration des techniques de modélisation du changement de phase, une compréhension approfondie des dynamiques des microlayers, et des insights sur l'interaction entre la tension de surface, les forces visqueuses, et le transfert de chaleur dans l'ébullition nucléée. Cette recherche constitue une base solide pour les études futures, y compris les simulations tridimensionnelles et l'investigation des effets de la rugosité de surface et de la distribution initiale de la température sur les dynamiques de l'ébullitionNucleate boiling dynamics are intricately influenced by the interactions between fluid and solid domains, particularly under conditions of small contact angles and complex interface phenomena. This thesis presents the development and application of advanced numerical methods to simulate these interactions with enhanced precision. Central to this work is the implementation of a mass-conservative Volume of Fluid (VOF) advection method, seamlessly integrated with an embedded solid domain approach. This technique is specifically designed to address the challenges of accurately capturing the dynamics of boiling processes at small contact angles.The numerical framework is constructed within the Basilisk platform, developed by Stéphane Popinet, utilizing a one-fluid model where the VOF method efficiently captures fluid interface dynamics. A continuous surface force model is employed to accurately represent surface tension effects, while the embedded solid model ensures robust coupling between the fluid and solid domains. To further enhance the simulation's fidelity, Leon Malan's phase-change model is integrated, incorporating a two-step advection interface method and an energy-conservative equation to handle the complexities of phase transition. Additionally, an interfacial heat resistance model by Lubomír Moravcík is implemented, quantifying the thermal resistance at the fluid-solid interface.A rigorous validation exercise is performed, demonstrating strong agreement with reference experimental data and established theoretical models. Key contributions of this work include the refinement of phase-change modeling techniques, a deeper understanding of microlayer dynamics, and insights into the interplay between surface tension, viscous forces, and heat transfer in nucleate boiling. This research provides a solid foundation for future studies, including three-dimensional simulations and the investigation of surface roughness and initial temperature distribution effects on boiling dynamics
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Breen, R. J. "PWR safety studies : nucleate boiling heat transfer." Thesis, University of Oxford, 1988. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.236258.
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Kim, Sung Joong Ph D. Massachusetts Institute of Technology. "Pool boiling heat transfer characteristics of nanofluids." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41306.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.Includes bibliographical references (leaves 79-83).
Nanofluids are engineered colloidal suspensions of nanoparticles in water, and exhibit a very significant enhancement (up to 200%) of the boiling Critical Heat Flux (CHF) at modest nanoparticle concentrations (50.1% by volume). Since CHF is the upper limit of nucleate boiling, such enhancement offers the potential for major performance improvement in many practical applications that use nucleate boiling as their prevalent heat transfer mode. The nuclear applications considered are main reactor coolant for PWR, coolant for the Emergency Core Cooling System (ECCS) of both PWR and BWR, and coolant for in-vessel retention of the molten core during severe accidents in high-power-density LWR. To implement such applications it is necessary to understand the fundamental boiling heat transfer characteristics of nanofluids. The nanofluids considered in this study are dilute dispersions of alumina, zirconia, and silica nanoparticles in water. Several key parameters affecting heat transfer (i.e., boiling point, viscosity, thermal conductivity, and surface tension) were measured and, consistently with other nanofluid studies, were found to be similar to those of pure water. However, pool boiling experiments showed significant enhancements of CHF in the nanofluids. Scanning Electron Microscope (SEM) and Energy Dispersive Spectrometry (EDS) analyses revealed that buildup of a porous layer of nanoparticles on the heater surface occurred during nucleate boiling. This layer significantly improves the surface wettability, as shown by measured changes in the static contact angle on the nanofluid-boiled surfaces compared with the pure-water-boiled surfaces. It is hypothesized that surface wettability improvement may be responsible for the CHF enhancement.
by Sung Joong Kim.
S.M.
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Mason, VerrDon Holbrook. "Chemical characterization of simulated boiling water reactor coolant." Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/10945/28026.
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Burns, Chad (Chad D. ). III. "Optimization algorithms in boiling water reactor lattice design." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82443.
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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013."June 2013." Cataloged from PDF version of thesis.
Includes bibliographical references (pages 32-33).
Given the highly complex nature of neutronics and reactor physics, efficient methods of optimizing are necessary to effectively design the core reloading pattern and operate a nuclear reactor. The current popular methods for optimization are Simulated Annealing and the Genetic Algorithm; this paper explores the potential for a new method called Greedy Exhaustive Dual Binary Swaps (GEDBS). The mandatory trade-off in computation is accuracy for speed; GEDBS is an exhaustive search and tends toward longer runtimes. While GEDBS performed acceptably for the criterion administered in this paper (local peaking and k, on a Boiling Water Reactor (BWR) fuel lattice) the exhaustive nature of GEDBS will inevitably lead to combinatorial explosion for the addition of the potential dozens of factors that commercial application mandates. This issue may be resolved with the addition of metaheuristics to reduce the search space for GEDBS, or by an increasing computation.
by Chad Bums.
S.B.
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Uhle, Jennifer Lee. "Boiling heat transfer characteristics of steam generator U-tube fouling." Thesis, Massachusetts Institute of Technology, 1997. http://hdl.handle.net/1721.1/17499.
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Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1997.Includes bibliographical references (leaves 109-111).
The boiling heat transfer characteristics of steam generator u-tube fouling deposits were identified by developing a boiling heat transfer model and determining its accuracy through the comparison of calculated and experimental results. Magnetite deposits were fabricated in the laboratory and were characterized using a variety of techniques. Heat transfer measurements were then taken, so that the effect of deposit parameters, including pore size distribution, porosity, permeability and thickness, as well as the effect of mass flux, heat flux and steam quality were investigated. The model predictions were consistent with the experimental results, differing by an average of ±17.5%. Over the range of parameters studied, pore size distribution dominated the deposit heat transfer. It was found that some fabricated deposits improved the heat transfer of the u-tubes, whereas others hindered it. The data were consistent with that of fouled u-tubes pulled from CANDU steam generators. The conditions of the heat transfer measurements and the fabricated deposits were similar to those of US and Canadian steam generators. Therefore, the conclusions drawn in this study are presumed to apply to the steam generators used in the Canadian and US industries.
by Jennifer L. Uhle.
Ph.D.
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Truong, Bao H. (Bao Hoai). "Effects of surface parameters on boiling heat transfer phenomena." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76925.
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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 148-156).
Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified in pool boiling. However, no such work has been done for flow boiling. Using a cylindrical tube pre-coated with Alumina nanoparticles coated via boiling induced deposition, CHF of water was found to enhance up to 40% compared to that of the bare tube. This confirms that nanoparticles on the surface is responsible for CHF enhancement for flow boiling. However, existing theories failed to predict the CHF enhancement and the exact surface parameters attributed to the enhancement cannot be determined. Surface modifications to enhance critical heat flux (CHF) and Leidenfrost point (LFP) have been shown successful in previous studies. However, the enhancement mechanisms are not well understood, partly due to many surface parameters being altered at the same time, as in the case for nanofluids. Therefore, the remaining objective of this work is to evaluate separate surface effect on different boiling heat transfer phenomena. In the second part of this study, surface roughness, wettability and nanoporosity were altered one by one and respective effect on quenching LFP with water droplet was determined. Increase in surface roughness and wettability enhanced LFP; however, nanoporosity was most effective in raising LFP, almost up to 100°C. The combination of the micro posts and nanoporous coating layer proved optimal. The nanoporous layer destabilizes the vapor film via heterogeneous bubble nucleation, and the micro posts provides intermittent liquid-surface contacts; both mechanisms increase LFP. In the last part, separate effect of nanoporosity and surface roughness on pool boiling CHF of a well-wetting fluid, FC-72, was investigated. Nanoporosity or surface roughness alone had no effect on pool boiling CHF of FC-72. Data obtained in the literature mostly for microporous coatings showed CHF enhancement for well wetting fluids, and existing CHF models are unable to predict the enhancement.
by Bao Hoai Truong.
Ph.D.
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Kossolapov, Artyom. "Transient flow boiling CHF under exponentially escalating heat inputs." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119047.
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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 41-42).
Reactivity initiated accidents (RIAs) are a potential concern for nuclear reactor safety. In RIA scenarios, following the insertion of positive reactivity, e.g. by an unanticipated extraction of the control rods, the reactor power may increase exponentially. The period of the exponential rise, τ, depends on the amount of positive reactivity inserted, as well as the fuel composition. During such an event, boiling of the water coolant can provide not only an effective way of heat removal, but also a stabilizing, negative reactivity feedback. However, the reactor power could reach extremely high levels and lead to a boiling crisis, e.g. by departure from nucleate boiling (DNB), in turn leading to fuel damage. The aim of the current work is to improve the understanding of transient DNB phenomena. This goal was achieved by running experiments on a specially designed flow boiling platform, which includes high speed video (HSV) and high speed infrared (HSIR) diagnostics. Specifically, the IR radiation recorded by the HSIR camera was analyzed with dedicated post processing algorithms that enable measurements of the time-dependent temperature and heat flux distributions on the boiling surface. Experiments were performed on a flat heater in upward flow conditions at atmospheric pressure. This work explores the effects of flow velocity, liquid subcooling and exponential power escalation period on critical heat flux (CHF). The results show that, for our flow conditions, the CHF value does not depend on the escalation period for periods longer than 100 ms, and is essentially the same as in steady-state boiling. For shorter periods, CHF increases as the escalation period decreases, and the effect of flow velocity becomes less important at short periods. Larger subcooling was shown to increase the CHF at all conditions. For extreme cases of 50 K and 75 K of subcooling the entire heating surface was covered by tiny bubbles. Those bubbles had a very short (less than 50 [mu]s) lifetime and were quenched right after the nucleation. Such behavior prevented bubbles from coalescing and resulted in a very efficient heat transfer mechanism. CHF was observed at much higher values compared to steady boiling conditions, when nucleation site density and bubble size were large enough for the bubble to start coalescing. An interesting effect was observed for very short periods (5, 10 and 20 ms) and low subcoolings (10 K). At those conditions the boiling surface experiences CHF during the growth of the first generation of bubbles. Therefore, the points of ONB and CHF are almost coincident, with CHF delayed only by the time required for adjacent bubbles to coalesce and the microlayer underneath them to evaporate.
by Artyom Kossolapov.
S.M.
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Grover, David J. (David Joseph). "Modeling water chemistry and electrochemical corrosion potential in boiling water reactors." Thesis, Massachusetts Institute of Technology, 1997. http://hdl.handle.net/1721.1/39772.
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O'Hanley, Harrison Fagan. "Separate effects of surface roughness, wettability and porosity on boiling heat transfer and critical heat flux and optimization of boiling surfaces." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/78208.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.B.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 2012.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 157-161).
The separate effects of surface wettability, porosity, and roughness on critical heat flux (CHF) and heat transfer coefficient (HTC) were examined using carefully-engineered surfaces. All test surfaces were prepared on nanosmooth indium tin oxide - sapphire heaters and tested in a pool boiling facility in MIT's Reactor Thermal Hydraulics Laboratory. Roughness was controlled through fabrication of micro-posts of diameter 20[mu]m and height 15[mu]m; intrinsic wettability was controlled through deposition of thin compact coatings made of hydrophilic SiO₂ (typically, 20nm thick) and hydrophobic fluorosilane (monolayer thickness); porosity and pore size were controlled through deposition of layer-by-layer coatings made of SiO₂ nanoparticles. The ranges explored were: 0 - 15[mu] for roughness (Rz), 0 - 135 degrees for intrinsic wettability, and 0 - 50% and 50nm for porosity and pore size, respectively. During testing, the active heaters were imaged with an infrared camera to map the surface temperature profile and locate distinct nucleation sites. It was determined that wettability can play a large role on a porous surface, but has a limited effect on a smooth non-porous surface. Porosity had very pronounced effects on CHF. When coupled with hydrophilicity, a porous structure enhanced CHF by approximately 50% - 60%. However, when combined with a hydrophobic surface, porosity resulted in a reduction of CHF by 97% with respect to the reference surface. Surface roughness did not have an appreciable effect, regardless of the other surface parameters present. Hydrophilic porous surfaces realized a slight HTC enhancement, while the HTC of hydrophobic porous surfaces was greatly reduced. Roughness had little effect on HTC. A second investigation used spot patterning aimed at creating a surface with optimal characteristics for both CHF and HTC. Hydrophobic spots (meant to be preferential nucleation sites) were patterned on a porous hydrophilic surface. The spots indeed were activated as nucleation sites, as recognized via the IR signal. However, CHF and HTC were not enhanced by the spots. In some instances, CHF was actually decreased by the spots, when compared to a homogenous porous hydrophilic surface.
by Harrison Fagan O'Hanley.
S.B.
S.M.
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Fridström, Richard. "Response of the Gamma TIP Detectorsin a Nuclear Boiling Water Reactor." Thesis, Uppsala University, Applied Nuclear Physics, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-126969.
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In order to monitor a nuclear boiling water reactor fixed and movable detectors are used, such as the neutron sensitive LPRM (Local Power Range Monitors) detectors and the gamma sensitive TIP (Traversing Incore Probe) detectors. These provide a mean to verify the predictions obtained from core simulators, which are used for planning and following up the reactor operation. The core simulators calculate e.g. the neutron flux and power distribution in the reactor core. The simulators can also simulate the response in the LPRM and TIP detectors. By comparing with measurements the accuracy of the core simulators can be quantified. The core simulators used in this work are PHOENIX4 and POLCA7. Because of the complexity of the calculations, each fuel assembly is divided axially into typically 25 nodes, which are more or less cubic with a side length of about 15 cm. Each axial segment is simulated using a 2D core simulator, in this work PHOENIX4, which provides data to the 3D code, in this case POLCA7, which in turn perform calculations for the whole core. The core simulators currently use both radial pin weights and axial node weights to calculate the gamma TIP detector signal. A need to bring forward new weight factors has now been identified because of the introduction of new fuel designs. Therefore, the gamma TIP detector response has been simulated using a Monte Carlo code called MCNPX for a modern fuel type, SVEA-96 Optima2, which is manufactured by Westinghouse. The new weights showed some significant differences compared to the old weights, which seem to overestimate the radial weight of the closest fuel pins and the axial weight of the node in front of the detector. The new weights were also implemented and tested in the core simulators, but no significant differences could be seen when comparing the simulated detector response using new and old weights to authentic TIP measurements.
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Ko, Yu-Chih Ph D. Massachusetts Institute of Technology. "Conceptual design of an annular-fueled superheat boiling water reactor." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/76976.
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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2011."September 2010." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 219-225).
The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron moderation throughout the reactor core. A single fuel design is used in the core. Each annular fuel element, or fuel tube, is cooled externally by boiling water and internally by steam. Fuel pellets are made of low enrichment U0 2, somewhat higher than the traditional BWR fuel enrichment. T91 and Inconel 718 are selected as candidates for the cladding material in view of their excellent physical properties and corrosion resistance. The fuel-cladding gap is filled with pressurized helium gas, like the existing lighter water reactor fuels. The ASBWR fuel assembly contains sixty annular fuel elements and one square water rod (occupying a space of four fuel elements) in an 8 by 8 square array. Annular separators and steam dryers are utilized and located above the core in the reactor vessel. Reactor internal pumps are used to adjust the core flow rate. Cruciform control rods are used to control the reactivity of the core, but more of them may be needed than a traditional BWR in view of the harder spectrum. The major design constraints have been identified and evaluated in this work. The ASBWR is found promising to achieve a power density of 50 kW/L and meet all the main safety requirements. This includes a limit on the minimum critical heat flux ratio, maximum fuel and cladding operating temperatures, and appropriate stability margin against density wave oscillations. At the expected superheated steam of 520 °C, the plant efficiency is above 40%, which is substantially greater than the efficiency of 33 to 35% that today's generation of LWRs can achieve. In addition to generating electricity, the ASBWR may also be useful for liquid fuel production or other applications that require high temperature superheated steam. The uncertainties about this design include the performance of cladding materials under irradiation, the attainment of desirable heat transfer ratio between the external and internal coolant channels throughout the fuel cycle, and the response to the traditional transients prescribed as design basis events.
by Yu-Chih Ko.
Ph.D.
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Truong, Bao H. (Bao Hoai). "Determination of pool boiling Critical Heat Flux enhancement in nanofluids." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41689.
Full textAbstract:
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, June 2007."May 2007."
Includes bibliographical references (leaves 51-53).
Nanofluids are engineered colloids composed of nano-size particles dispersed in common fluids such as water or refrigerants. Using an electrically controlled wire heater, pool boiling Critical Heat Flux (CHF) of Alumina and Silica water-based nanofluids of concentration less than or equal to 0.1 percent by volume were measured. Silica nanofluids showed CHF enhancement up to 68% and there seems to be a monotonic relationship between nanoparticle concentration and magnitude of enhancement. Alumina nanofluids had CHF enhancement up to 56% but the peak occurred at the intermediate concentration. The boiling curves in nanofluid were found to shift to the left of that of water and correspond to higher nucleate boiling heat transfer coefficients in the two-phase flow regime. SEM images show a porous coating layer of nanoparticles on wires subjected to nanofluid CHF tests. These coating layers change the morphology of the heater's surface, and are responsible for the CHF enhancement. The thickness of the coating was estimated using SEM and was found ranging from 3.0 to 6.0 micrometers for Alumina, and 3.0 to 15.0 micrometers for Silica. Inductively Coupled Plasma Spectroscopy (ICP-OES) analyses were also attempted to quantify the mass of the particle deposition but the results were inconsistent with the estimates from the SEM measurement.
by Bao H. Truong.
S.B.
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Mowrey, James A. "Control system modeling for a boiling water reactor." Thesis, Georgia Institute of Technology, 1995. http://hdl.handle.net/1853/17083.
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Maurer, John H. (John Howard). "Surveillance strategy for a four year operating cycle in commercial boiling water reactors." Thesis, Massachusetts Institute of Technology, 1996. http://hdl.handle.net/1721.1/39051.
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Correll, Sam. "Flow film boiling on levitated molten drops and vapour explosion triggering." Thesis, University of Oxford, 1989. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.352915.
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Richenderfer, Andrew Jonathan. "Experimental study of heat flux partitioning in pressurized subcooled flow boiling." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119033.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 133-137).
Understanding of subcooled flow boiling and the critical heat flux (CHF) is of the utmost importance for both safety and profitability of pressurized water nuclear reactors since they are major factors in the determination of the reactor power rating. Motivated by the emergence of a new wall boiling model by Gilman [3] and previous experimental insights from Phillips [12], a first-of-a-kind experimental investigation of pressurized steady-state subcooled flow boiling was conducted using state-ofthe- art diagnostics to gain a unique insight of the relevant mechanisms, including the partitioning of the wall heat flux. Conditions up to 10 bar pressure, 2000 kg/m²s mass flux and 20 K subcooling were explored. High-speed infrared thermometry tools were developed and used to measure the local time-dependent 2-D temperature and heat flux distributions on the boiling surface. These distributions were analyzed to determine fundamental boiling heat transfer parameters such as the nucleation site density, growth and wait times, nucleation frequency, departure diameter as well as the partitioning of the wall heat flux. While established mechanistic models can capture the trends of growth time and wait time with relatively good accuracy, this work reveals current models do not accurately predict the activation and interaction of nucleation sites on the boiling surface. This is a major roadblock, since boiling curves and CHF values obtained in nominally identical environments can be significantly different depending upon the nucleation site density which in turn is determined by the surface properties. The role of evaporation in the partitioning of the heat flux increases monotonically as the average heat flux increases, up to a maximum value of 70%, and is the dominant mechanism at high heat fluxes. At low and intermediate heat fluxes single-phase heat transfer is the dominant mechanism. Traditional heat partitioning models fail to capture these physics, but newer models with a comprehensive and physically consistent framework show promise in predicting the wall heat transfer. The data and understanding produced by this work will be essential for the development and validation of these modeling tools.
by Andrew Jonathan Richenderfer.
Ph. D.
24
Lenci, Giancarlo. "Prediction of departure from nucleate boiling in PWR fast power transients." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/80659.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 89-91).
An assessment is conducted of the differences in predicted results between use of steady state versus transient Departure from Nucleate Boiling (DNB) models, for fast power transients under forced convective heat exchange conditions. Theoretical DNB models based on liquid film thickness variation are adapted and modified from existing studies into a generalized formulation to allow implementation into a reactor simulation code. The formulation is validated using experimental data available at low pressure. An application is performed at Pressurized Water Reactor (PWR) operating conditions, simulating rod ejection accidents. The transient DNB model is applied to PWR rod ejection accident cases computed by the reactor dynamics code SIMULATE-3K. Rod power profiles deriving from pin power reconstruction are used in a subchannel simulation done with VIPRE to obtain local pin parameters. Results show that a significant delay exists for the occurrence of transient DNB compared to quasi steady-state DNB and in some cases DNB does not occur, even if predicted by quasi steady-state methods. Most modem codes for PWR thermal hydraulic simulation use quasi steady-state approaches to predict DNB, thus applying a steady-state correlation to time-dependent cases. However, according to the transient DNB model used in this work, a time lag exists between DNB as predicted by steady-state correlations, and effective transient DNB. During that time lag, the liquid film between the wall and the bubbly layer thins until the heated surface is eventually dried out. Such DNB prediction by steady state models is a conservative estimate. This work assesses the consequences of the use of more accurate models for predicting transient DNB, which are desirable to get better knowledge of design margins, to allow optimization of plant safety and efficiency.
by Giancarlo Lenci.
S.M.
25
Hu, Rui Ph D. Massachusetts Institute of Technology. "Stability analysis of the boiling water reactor : methods and advanced designs." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/62693.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 278-285).
Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been considerable concern about the effects of such oscillations when coupled with neutronic feedback. The current trend of increasing reactor power density and relying more extensively on natural circulation for core cooling may have consequences for the stability characteristics of new BWR designs. This work addresses a wide range of issues associated with the BWR stability: 1) flashing-induced instability and natural circulation BWR startup; 2) stability of the BWRs with advanced designs involving high power :densities; 3) modeling assumptions in stability analysis methods; and 4) the fuel clad performance during power and flow oscillations. To capture the effect of flashing on density wave oscillations during low pressure startup conditions, a code named FISTAB has been developed in the frequency domain. The code is based on a single channel thermal-hydraulic model of the balance of the water/steam circulation loop, and incorporates the pressure dependent water/steam thermodynamic properties, from which the evaporation due to flashing is captured. The functionality of the FISTAB code is confirmed by testing the experimental results at SIRIUS-N facility. Both stationary and perturbation results agree well with the experimental results. The proposed ESBWR start-up procedure under natural convection conditions has been examined by the FISTAB code. It is confirmed that the examined operating points along the ESBWR start-up trajectory from TRACG simulation will be stable. To avoid the instability resulting from the transition from single-phase natural circulation to two-phase circulation, a simple criterion is proposed for the natural convection BWR start-up when the steam dome pressure is still low. Using the frequency domain code STAB developed at MIT, stability analyses of some proposed advanced BWRs have been conducted, including the high power density BWR core designs using the Large Assembly with Small Pins (LASP) or Cross Shape Twisted (CST) fuel designs developed at MIT, and the Hitachi's RBWR cores utilizing a hard neutron spectrum and even higher power density cores. The STAB code is the predecessor of the FISTAB code, and thermodynamic properties of the coolant are only dependent on system pressure in STAB. It is concluded that good stability performance of the LASP core and the CST core can be maintained at nominal conditions, even though they have 20% higher reactor thermal power than the reference core. Power uprate does not seem to have significant effects on thermal-hydraulic stability performance when the power-to-flow ratio is maintained. Also, both the RBWR-AC and RBWR-TB2 designs are found viable from a stability performance point of view, even though the core exit qualities are almost 3 times those of a traditional BWR. The stability of the RBWRs is enhanced through the fast transient response of the shorter core, more flat power and power-to-flow ratio distributions, less negative void feedback coefficient, and the core inlet orifice design. To examine the capability of coupled 3D thermal-hydraulics and neutronics codes for stability analysis, USNRC's latest system analysis code, TRACE, is chosen in this work. Its validation for stability analysis and comparison with the frequency domain approach, have been performed against the Ringhals 1 stability tests. Comprehensive assessment of modeling choices on TRACE stability analysis has been made, including effects of timespatial discretization, numerical schemes, thermal-hydraulic channel grouping, neutronics modeling, and control system modeling. The predictions from both the TRACE and STAB codes are found in reasonably good agreement with the Ringhals 1 test results. The biases for the predicted global decay ratio are about 0.07 in TRACE results, and -0.04 in STAB results. However, the standard deviations of decay ratios are both large, around 0.1, indicating large uncertainties in both analyses. Although the TRACE code uses more sophisticated neutronic and thermal hydraulic models, the modeling uncertainty is not less than that of the STAB code. The benchmark results of both codes for the Ringhals stability test are at the same level of accuracy. The fuel cladding integrity during power oscillations without reactor scram is examined by using the FRAPTRAN code, with consideration of both the stress-strain criterion and thermal fatigue. Under the assumed power oscillation conditions for high burn-up fuel, the cladding can satisfy the stress-strain criteria in the ASME Code. Also, the equivalent alternating stress is below the fatigue threshold stress, thus the fatigue limit is not violated. It can be concluded that under a large amount of the undamped power oscillation cycles, the cladding would not fail, and the fuel integrity is not compromised.
by Rui Hu.
Ph.D.
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Caraghiaur, Diana. "On drops and turbulence in nuclear fuel assemblies of Boiling Water Reactors." Doctoral thesis, KTH, Reaktorteknologi, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-107115.
Full textAbstract:
The study aims to develop the understanding of the mechanistic-type approach to quantify drop deposition in nuclear fuel assemblies of Boiling Water Reactors. This includes the effect of spacers. Spacers have a complex geometry to serve their purposes, but optimization of them alone can improve the thermal limit parameters in nuclear fuel assemblies. Thus, a mechanistic model might prove useful to increase the safety of the reactor as well as economic competitiveness of the nuclear power plant. In this thesis, measurement techniques, such as mobile pressure rod and Laser Doppler Velocimetry are developed and tested to provide local data of the flow around spacers. It is shown experimentally that the effect of spacer on the flow differs depending on the placement of the subchannel in the rod bundle. Partly, because the spacer part differs, but also due to a global velocity profile development. Very few studies in the literature indicate this effect. It is shown that single subchannel models using Computational Fluid Dynamics (CFD) can predict the average velocity increase downstream of the spacer; however, they are not capable of calculating the spacer effect on turbulence parameters. The single subchannel CFD model has limited capability to predict the pressure development inside the spacer part, mainly because cross-flows are not taken into consideration. The deposition of drops in annular two-phase flow is still a scientific challenge. Only empirical correlations are used nowadays to quantify this process. Empirical coefficients are needed for each spacer type to calculate the deposition increase due to obstacle. The discussion about the deposition starts with the phenomenological description. The important input parameter, namely drop size, is carefully analysed, and a new correlation is proposed to calculate the mean drop diameter. The correlation is constructed on a larger experimental data base. Lagrangian Particle Tracking model is tested in its capability to calculate deposition. Additionally, a Eulerian-type model is developed and tested. Turbulent parameters of drops are tightly related to the turbulence of the gas phase and the inertia of the drops. Several approaches are discussed about how to calculate the root-mean-square fluctuating velocities of drops. Both, Lagrangian Particle Tracking and the Eulerian-type of models show good capability in calculating the obstacle effect on deposition, providing improvements are made in prediction of drop size. The effect of increased drop concentration plays a large role and it must be taken into consideration if good quantitative approaches are envisaged.QC 20121207
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BOADU, HERBERT ODAME. "CONTINUOUS-TIME OPTIMAL CONTROL OF A SIMULATED BOILING WATER NUCLEAR (BWR) POWER PLANT." Diss., The University of Arizona, 1985. http://hdl.handle.net/10150/188087.
Full textAbstract:
A suboptimal controller has been developed for a Boiling Water Reactor Nuclear Power Plant, using the DARE P Continuous Simulation Language, which was developed in the Electrical Engineering Department at the University of Arizona. A set of 48 nonlinear first-order differential equations and a large number of algebraic equations has been linearized about the equilibrium state. Using partitioning, the linearized equations were transformed into a block triangular form. The concept of optimal control and a square performance index reflecting the desired plant behavior have been applied on the slow subsystem to develop a suboptimal controller. The obtained feedback law is shown by simulation to be able to compensate for a variety of plant disturbances. A large variety of responses can be obtained by changing the weighting matrices. The control is basically a regulator approach to speed up response during load demand changes. Several simulations are included to demonstrate the control performance. The variables to be controlled have mainly been the average neutron density and the average coolant temperature. Simplifications have been suggested, thus obtaining considerable savings in the computations and ease in design.
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Coyle, Carolyn Patricia. "Synthesis of CRUD and its effects on pool and subcooled flow boiling." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/103652.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 127-132).
This work is dedicated to studying the effects of synthetic CRUD (Chalk River Unidentified Deposits) on pool and subcooled flow boiling parameters. Previous pool boiling studies have demonstrated the potential of porous, hydrophilic surfaces to lead to more efficient boiling. CRUD is a naturally occurring porous, hydrophilic layer that forms on fuel rods during reactor operation. As such, CRUD deposition may have large effects on critical heat flux (CHF) and heat transfer coefficient (HTC). An investigation of such effects was conducted as part of the CASL project by creating well-defined and characterized synthetic CRUD with parameters representative of reactor CRUD on indium tin oxide-sapphire heaters. The effects of synthetic CRUD on boiling heat transfer were then experimentally studied, focusing on heat transfer coefficient (HTC), critical heat flux (CHF), nucleation site density, bubble departure frequency, and bubble departure diameter. These heaters were tested in pool and flow boiling facilities in MIT's Reactor Hydraulics Laboratory. Synthetic CRUD was created using layer-by-layer deposition of 100 nm silica nanoparticles to form porous, hydrophilic thick films. Photolithography was used to manufacture posts that were then dissolved to create characteristic boiling chimneys. Features such as thickness, wettability, pore size, and chimney diameter and pitch were verified to be representative of reactor CRUD. Silica nanoparticles were used as a surrogate for reactor CRUD nanoparticle materials (iron and nickel oxides) since they create more stable films. To ensure accurate modeling, independent of material, 10 nm silica nanoparticle and 10 nm iron oxide nanoparticle boiling tests were conducted and found to be similiar. During testing, IR thermography and high-speed video (HSV) are used to obtain two dimensional temperature profiles of the active heater area to quantify properties such as HTC, nucleation site density, bubble departure frequency, and bubble departure diameter. The bubble parameters follow expected trends with mass flux and heat flux. IR/HSV flow data (Chapter 6) has shown that HTC increases with the presence of chimneys, increasing thickness and increasing chimney diameter. However the HTC is relatively unaffected by the chimney pitch and is decreased by the presence of an LbL layer. The boiling curves and CHF data obtained from pool boiling experiments with iron oxide and silica oxide nanoparticles with and without chimneys also confirm these trends. The largest HTC is observed in the case of uncoated heaters, followed by heaters with chimneys, with heaters with an LbL layer without chimneys having the lowest HTC. From pool boiling data, the benefit of a CRUD layer is observed in the enhancement of CHF. The flow boiling trends are further supported by the combination of measured basic bubble parameters according to the heat flux partitioning model. The statistical significance of these trends varies with mass flux. The data generated here may inform advanced models of boiling heat transfer and/or validate existing models.
by Carolyn Patricia Coyle.
S.M.
29
Latif, Medhat Gamil. "Modelling the simplified boiling water reactor natural circulation loop and its stability." Diss., The University of Arizona, 1993. http://hdl.handle.net/10150/186405.
Full textAbstract:
An integrated model that estimates loop flow rate, heat removal, and stability parameters for the General Electric Simplified Boiling Water Reactor SBWR was developed. The three parameters above used to be calculated individually each by a separate code. The initial approach in loop thermal hydraulic modelling was the steady state solution of the SBWR loop mass, energy, and momentum equations. The power-to-flow map obtained proved to be quite comparable with the Natural Circulation in Boiling Water Reactor (NATBWR) code developed by EPRI, in addition to that of General Electric. At low power levels buoyancy forces are the controlling factor in determining the loop flow rate, while at high power levels two-phase friction losses become the dominating one. Evaluation criteria necessary for comparing different loop geometries performance have been the "minimum critical heat flux ratio (MCHFR)" and the "decay ratio." The predicted flow, from the DFM, at different power levels was used later in a parametric study to answer an important question of which combination of core and riser heights are to be selected that meets both the stability and critical power ratio limits. By modelling bubble time delay through riser in the loop momentum equation, a loop damping coefficient as a measure of loop stability, with higher damping meaning a more stable loop was calculated. Results indicated that during normal operation the SBWR loop is pretty damped. Finally, a detailed code that consists mainly of a fuel pin model, reactor point kinetics for the time dependent reactor normalized power with one group of delayed neutrons, and coolant channel mass, energy, and momentum equations is considered. Reactivity feedbacks from voids and fuel temperature, (Doppler effect), were considered. The loop momentum equation was modified to account for bubble time delay in the riser. After a small perturbation in reactivity, fuel temperature, core average void, and loop flow rate were shown to reach equilibrium values after a period of time equivalent to the transit time of the bubble through the riser. Results from this code matched that of the SBWR safety analysis report.
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Brodeur, David Lester 1963. "A study of US nuclear power boiling water reactor, class IV, operating performance, 1992-1997." Thesis, Massachusetts Institute of Technology, 1998. http://hdl.handle.net/1721.1/49796.
Full textAbstract:
Thesis (Nucl.E. and S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1998.Includes bibliographical references.
The steady improvement of US Nuclear Utility generation capability observed over the past two decades has recently halted and somewhat degraded. For the industry to resume its upward trend in performance a detailed examination must be performed of current performance and new methods developed to continue the improvement. A detailed study of Boiling Water Reactor, Class IV (BWR/4) performance over the past five years was conducted to gain insight to the nature of lost generation capability and develop a methodology to improve capability. Extensive electronic NRC records were used in conjunction with detailed power plant records and engineering experience at PECO Energy's Limerick Generating Station and Peach Bottom Atomic Power Station for this research. Administrative or regulatory shutdowns within the study dominated the lost generation capability and detracted from the goal of analyzing equipment reliability. Nine of two hundred thirty five shutdowns were therefore limited to maximum impact of 30 days lost generation. Balance of Plant system failures were found to initiate 69% of the occurrences of lost generation capability and account for 59% of the capability loss. The failures of these systems were found to be infrequent events which correlated poorly to the aggregate industry experience. Approximately fifty percent of the forced outages were the result of equipment related failures such as weak design or worn parts with the remaining fifty percent the result of human related failures. Only 19% of the failures were noted to be the result of component age related failures while 31% of the failures were related to poor equipment design. The time frame of forced outages with in operating cycles was additionally reviewed. Failures were found to be more frequent in the early phase of the operating cycle following start up from a refueling and approximately 400 to 550 days after start up. The impact of these failures was not great enough to affect the steady state cumulative capability factor of the aggregate BWR/4 utility achieved after one year of operation. Individual utility sites were found to have opposing strong and weak periods of performance within their operating cycles. The loss of generation capacity taken for planned maintenance outages and on line maintenance for minor equipment problems was not found to have a significant impact on aggregate BWR/4 performance. For plants not involve in lengthy shutdowns, the strongest impacts on cumulative capacity were forced outages, initial start up and coast down. The unpredictable and design nature of system failures necessitates a structured effort to improve the combined performance of all systems at a utility. Balance of Plant systems were found to all have a 25% probability of causing a single forced outage lasting slightly less than 5 days in length. The infrequent nature of significant failures necessitates a broad based communication between utilities to maintain an adequate level of awareness of system vulnerabilities and possible improvements. Two specific sites examined had opposing and repeatable strong and weak cycle performance traits. The unique nature of site performance demonstrates the impact that improved communications between utilities could have on transferring strengths and diminishing weaknesses thus improving overall utility performance.
by David Lester Brodeur.
Nucl.E.and S.M.
31
Johnsson, John. "Detailed B-10 depletion in control rods operatingin a Nuclear Boiling Water Reactor." Thesis, Uppsala universitet, Institutionen för materialkemi, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-155416.
Full textAbstract:
In a nuclear power plant, control rods play a central role to control the reactivity ofthe core. In an inspection campaign of three control rods (CR 99) operated in theKKL reactor in Leibstadt, Switzerland, during 6 respectively 7 consecutive cycles,defects were detected in the top part of the control rods due to swelling caused bydepletion of the neutron-absorbing 10B isotope (Boron-10). In order to correlatethese defects to control rod depletion, the 10B depletion has in this study beencalculated in detail for the absorber pins in the top node of the control rods.Today the core simulator PLOCA7 is used for predicting the behavior of the reactorcore, where the retrievable information from the standard control rod follow-up isthe average 10B depletion for clusters of 19 absorber holes i.e. one axial node.However, the local 10B depletion in an absorber pin may be significantly differentfrom the node average depletion that is re-ceived from POLCA7. To learn more, the 10B depletion has been simulated for each absorber hole in the uppermost node usingthe stochastic Monte Carlo 3D simulation code MCNP as well as an MCNP- based2D-depletion code (McScram). It was found that the 10B depletion is significantly higher for the uppermost absorberpins than the node average. Furthermore, the radial depletion in individual absorberpins was found to be much higher than expected. The results are consistent with theexperimental data on control rod defects.
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Ahmad, Masroor. "Critical heat flux and associated phenomena in forced convective boiling in nuclear systems." Thesis, Imperial College London, 2012. http://hdl.handle.net/10044/1/9181.
Full textAbstract:
In evaporation of a liquid flowing in a tube or nuclear fuel element, there exists a transition (known as "dryout", "burnout", "boiling crisis" or "critical heat flux", CHF) from a high heat transfer coefficient regime to one of greatly reduced heat transfer coefficient. The conditions leading to dryout or CHF and the behaviour of wall temperatures in the ("post dryout or post CHF") region beyond it are of immense importance in nuclear reactor safety. In a nuclear reactor, the clad temperature excursion in the post-dryout region may be unacceptably high and the prediction of the location of dryout and the magnitude of the temperature excursion into the post-dryout region is of great importance. Moreover, the dryout transition and its effects are important not only in nuclear plant but also in many other types of heat transfer equipment. The main focus of work described in this thesis was the improvement and validation of phenomenological models for the prediction of CHF and of heat transfer beyond CHF ("post CHF" or "post dryout" heat transfer). The main focus has been on the process of annular film dryout. In phenomenological modelling of this process the dryout location prediction is sensitive to the boundary value of entrained fraction at churn annular transition, especially at high flow rates. The model was extended to churn flow so that integration of entrainment, deposition and evaporation processes could be started from onset of churn flow. A new correlation for the prediction of entrainment rate in churn flow was presented. The application of the new methodology to experimental data leads to improved predictions of CHF. Another long-standing problem, i.e. effect of heat flux on droplet entrainment, is addressed by analysing the contradictory results of previous experiments by using the annular film dryout model. The capability of phenomenological models to cover the whole range of CHF scenarios, i.e. from subcooled or very low quality to very high quality CHF, was demonstrated by using a possible transition criterion from bubble crowding model (an improved version of the Weisman Pie model) to annular film dryout model. These improved phenomenological models captured trends of CHF data very well (including the Look Up Table data of Groeneveld et al. 2007) and produced improved results over a wide range of system parameters such as pressure, mass flux and critical quality. The implementation of the phenomenological models was pursued by modifying and developing an Imperial College computer code GRAMP. In addition to its application in modelling CHF, the GRAMP code was extended to the post dryout region and predictions for this region compared to a range of data and the results were found to be satisfactory.
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Li, Jinghui Carleton University Dissertation Engineering Mechanical. "The Onset of significant void in boiling flows over a wide range of operation conditions." Ottawa, 1992.
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Phillips, Bren Andrew. "Nano-engineering the boiling surface for optimal heat transfer rate and critical heat flux." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76536.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 130-133).
The effects on pool boiling characteristics such as critical heat flux and the heat transfer coefficient of different surface characteristics such as surface wettability, roughness, morphology, and porosity are not well understood. Layer-by-layer nanoparticle coatings were used to modify the surface of a sapphire heater to control the surface roughness, the layer thickness, and the surface chemistry. The surface was then tested in a water boiling test at atmospheric pressure while imaging the surface with high speed infrared thermography yielding a 2D time dependent temperature profile. The critical heat flux and heat transfer coefficient were enhanced by over 100% by optimizing the surface parameters. It was found that particle size of the nanoparticles in coating, the coating thickness, and the wettability of the surface have a large impact on CHF and the heat transfer coefficient. Surfaces were also patterned with hydrophobic "islands" within a hydrophilic "sea" by coupling the Layer-by-layer nanoparticle coatings with an ultraviolet ozone technique that patterned the wettability of the surface. The patterning was an attempt to increase the nucleation site density with hydrophobic dots while still maintaining a large hydrophilic region to allow for rewetting of the surface during the ebullition cycle and thus maintaining a high critical heat flux. The patterned surfaces exhibited similar critical heat fluxes and heat transfer coefficients to the surfaces that were only modified with layer-by-layer nanoparticle coatings. However, the patterned surfaces also exhibited highly preferential nucleation from the hydrophobic regions demonstrating an ability to control the nucleation site layout of a surface and opening an avenue for further study.
by Bren Andrew Phillips.
S.M.
35
Feng, Bo Ph D. Massachusetts Institute of Technology. "Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76497.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 262-269).
This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel hardens the neutron energy spectrum and results in higher breeding ratios. The state-of-the-art high conversion light water reactor, the Resource-renewable Boiling Water Reactor (RBWR), served as the template core upon which comparative studies between nitride and oxide fuels were performed. A 1/3 core reactor physics model was developed for the RBWR using the stochastic transport code MCNP. The code was coupled with a lumped channel thermalhydraulics 5-channel model for steady-state analyses. The depletion code MCODE, which links MCNP with ORIGEN, was used for all burnup calculations. Select physics parameters were calculated and with the exception of the void coefficients, agreed with reported data. The void coefficients of the coupled core were calculated to be slightly positive using two different methods (10% power increase and 5% flow reduction). The standard RBWR assembly designs, which use tight lattice hexagonal fuel rod arrays, with oxide fuel were then replaced with various nitride fuel assembly designs to determine the potential increase in breeding ratio, the potential to breed with pressurized water, and the potential to improve the critical power ratio with a wider pin pitch. Without changing the assembly geometry or discharge burnup, using nitride fuel resulted in a breeding ratio of 1.14. Using single-phase liquid water, the nitride fuel RBWR assembly resulted in a conversion ratio of 1.00. Another nitride fuel assembly design with boiling water maintained a 1.04 breeding ratio while increasing the pitch-todiameter ratio from 1.13 to 1.20. This modification increased the hot assembly critical power ratio from 1.22 to 1.36, as calculated using the Liu-2007 correlation. A high-porosity nitride fuel is recommended for high burnup conditions, to accommodate the nitride fuel's higher swelling and less favorable mechanical properties compared to the oxide fuel. The high porosity allows additional volume for pressure-induced densification, alleviating swelling and subsequent cladding strain. To predict the performance of high-porosity nitride fuel, fission gas and fuel behavior mechanistic models were developed for high burnup and low-temperature conditions. These models were validated with reported irradiation data and implemented, along with fuel material properties, into the steady-state fuel behavior code FRAPCON-EP. Under simulated RBWR conditions, a fuel density no more than 85% of theoretical density is recommended to maintain satisfactory fuel performance.
by Bo Feng.
Ph.D.
36
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.
Full textAbstract:
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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Guion, Alexandre Nicolas. "Modeling and simulation of liquid microlayer formation and evaporation in nucleate boiling using computational fluid dynamics." Thesis, Massachusetts Institute of Technology, 2017. http://hdl.handle.net/1721.1/112380.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 243-252).
The transport of latent heat makes boiling one of the most efficient modes of heat transfer, allowing a wide range of systems to improve their thermal performance, from microelectronic devices to nuclear power plants. In particular, Boiling Water Reactors (BWR) use boiling as the primary mode of heat transfer in the reactor core to accommodate very high heat fluxes. In Pressurized Water Reactors (PWR) subcooled flow boiling can occur in hot sub-channels. As a bubble grows outside of a surface imperfection during nucleate boiling, viscous stresses at the wall can be strong enough to impede liquid motion and trap a thin liquid layer - referred to as microlayer, underneath the growing bubble. The contribution of microlayer evaporation to overall heat transfer and bubble growth can be large, in particular in the case of water1. In practice, numerical simulations of nucleate boiling resolve the macroscopic interface of the bubble and resort to subgrid models to account for the evaporation of the microlayer at the microscopic scale. The applicability of this subgrid modeling approach relies on the capacity to initialize the microlayer shape and extension, prior to its evaporation. However, existing models of microlayer formation are either physically incomplete2 or purely empirical3. In this work, we first confirm through a sensitivity study the need for accurate modeling of microlayer formation to initialize boiling simulations and to reproduce physical boiling dynamics (a). Then, we build the first generally applicable model for microlayer formation through direct computations of the hydrodynamics of bubble growth at the wall for a wide range of conditions and fluids, including water at 0.101MPa (lab experiments) and 15.5MPa (PWR), capillary numbers Ca [is element of] [0.001; 0.1], and contact angles [theta] [is element of] [10°; 90°] (b). In addition, we modify an existing experimental pool boiling setup to measure with unprecedented accuracy initial bubble growth rates needed to predict microlayer formation (c). Lastly, we develop a numerical procedure based on hydrodynamics theories to obtain mesh-independent results in moving contact line simulations for a wide range of contact angles and viscosity ratios (d). In particular, we use direct computations of the transition to a Landau-Levich-Derjaguin film in forced dewetting to inform the onset of microlayer formation in nucleate boiling. These contributions(a) (b) (c) (d) bridge a significant gap in our understanding of how boiling works and can be modeled at the microscopic scale, which represents a first step in designing surfaces with higher heat transfer performance and in building safer and more efficient energy systems.
by Alexandre Nicolas Guion.
Ph. D.
38
Tetreault-Friend, Melanie. "Systematic investigation of the effects of hydrophilic porosity on boiling heat transfer and critical heat flux." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/95571.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 97-99).
Predicting the conditions of critical heat flux (CHF) is of considerable importance for safety and economic reasons in heat transfer units, such as in nuclear power plants. It is greatly advantageous to increase this thermal limit and much effort has been devoted to studying the effects of surface characteristics on it. In particular, recent work carried out by O'Hanley demonstrated the separate effects of surface wettability, porosity, and roughness on CHF, and found that porous hydrophilic surface coatings provided the largest CHF increase, with a 50-60% enhancement over the base case. In the present study, a systematic investigation of the effects that the physical characteristics of the hydrophilic layers have on heat transfer was conducted. Parameters experimentally explored include porous layer thickness, pore size, and void fraction (pore volume fraction). The surface characteristics are created by depositing layer-by-layer (LbL) thin compact coatings made of hydrophilic SiO₂ nanoparticles of various sizes. A new coating was developed to reduce the void fraction by using polymers to partially fill the voids in the porous layers. All test surfaces are prepared on indium tin oxide - sapphire heaters and tested in a pool boiling facility at atmospheric pressure in MIT's Thermal-Hydraulics Laboratory. Results indicate that CHF follows a trend with respect to each parameter studied and clear CHF maxima reaching up to 114% enhancement are observed for specific thickness and pore size values. ZnO₂ nanofluid-generated coatings are also prepared and their boiling performance is compared to the boiling performance of the engineered LbL coatings. The results highlight the dependence of CHF on capillary wicking and are expected to allow further optimization of the nanoengineered surfaces.
by Melanie Tetreault-Friend.
S.M.
39
Su, Guanyu Ph D. Massachusetts Institute of Technology. "Experimental study of transient pool boiling heat transfer under exponential power excursion on plate-type heater." Thesis, Massachusetts Institute of Technology, 2015. http://hdl.handle.net/1721.1/97863.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.Cataloged from PDF version of thesis.
Includes bibliographical references (page 82).
Conduction and single-phase convective heat transfer are well understood phenomena: analytical models [1] and empirical correlations [2] allow capturing the thermal behavior of plate-type fuels or heaters in contact with a single-phase coolant. On the other hand, transient boiling heat transfer is a scarcely studied and much less understood phenomenon. Although, earlier studies have shown that important features of the boiling curve (i.e. onset of nucleate boiling (ONB), nucleate boiling heat transfer coefficient, and critical heat flux (CHF)) in transient conditions. These parameters significantly differ from those at steady-state. The mechanisms by which these changes occur are not clear. Furthermore, some of the conclusions from different authors are quantitatively or qualitatively in disagreement with each other. This work studied transient pool boiling heat transfer phenomena under exponentially escalating heat fluxes on plate-type heaters, at the time scales of milliseconds typical of Reactivity Initiated Accidents (RIAs) in nuclear reactors. The investigation utilized state-of-the-art diagnostics such as Infrared (IR) thermometry and high-speed video (HSV), to gain insight into the physical phenomena and generate a database that could be used for development and validation of accurate models for transient boiling heat transfer. The tests with exponential power escalation periods ranging from 100 ms to 5 ms and subcoolings of OK (saturation), 25K and 75 K were conducted. The measured pre-ONB heat transfer coefficient agrees well with the theoretical predictions for transient conduction. The ONB and onset of significant void (OSV) temperature and heat flux were found to increase monotonically with decreasing period and increasing subcooling, as expected. The mechanistic ONB model of Hsu was able to predict the measured ONB temperature and heat flux. The transient pool boiling curves were measured up to fully developed nucleate boiling (FDNB). Generally two types of boiling curve were observed: with overshoot (OV) or without overshoot. Data show that, when an OV is present, the OV temperature increases monotonically with decreasing period and increasing subcooling. The present study clears the confusions (eg. the trend of ONB temperature and heat flux versus power period) in previous research, and sheds light to the mechanisms behind transient boiling heat transfer. This can ultimately reduce the uncertainty in both design and safety analyses of the research reactors especially under RIAs.
by Guanyu Su.
S.M.
40
Kim, Sung Joong Ph D. Massachusetts Institute of Technology. "Subcooled flow boiling heat transfer and critical heat flux in water-based nanofluids at low pressure." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/53274.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 285-290).
A nanofluid is a colloidal suspension of nano-scale particles in water, or other base fluids. Previous pool boiling studies have shown that nanofluids can improve the critical heat flux (CHF) by as much as 200%. In this study, subcooled flow boiling heat transfer and CHF experiments were performed with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (< 0.1 % by volume) at atmospheric pressure. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient (HTC) are similar (within ±20%). The HTC increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. The CHF tests were conducted at 0.1 MPa and at three different mass fluxes (1500, 2000, 2500 kg/m2s) under subcooled conditions. The maximum CHF enhancement was 53%, 53% and 38% for alumina, zinc oxide and diamond, respectively, always obtained at the highest mass flux. The measurement uncertainty of the CHF was less than 6.2%. A post-mortem analysis of the boiling surface reveals that its morphology is altered by deposition of the particles during nanofluids boiling. A confocal-microscopy-based examination of the test section revealed nanoparticles deposition not only changes the number of micro-cavities on the surface, but also the surface wettability. A simple model was used to estimate the ensuing nucleation site density changes, but no definitive correlation between the nucleation site density and the heat transfer coefficient data could be found.
(cont.) Wettability of the surface was substantially increased for heater coupons boiled in alumina and zinc oxide nanofluids, and such wettability increase seems to correlate reasonably well with the observed marked CHF enhancement for the respective nanofluids. Interpretation of the experimental data was conducted in light of the governing surface parameters and existing models. It was found that no single parameter could explain the observed HTC or CHF phenomena. The existing models were limited in studying the surface effects, suggesting that more accurate models incorporating surface effects need to be developed. Finally, the research activities performed in this thesis help identify the research gaps and indicate future research directions.
by Sung Joon Kim.
Ph.D.
41
Lucas, Timothy R. "The effect of thermal aging and boiling water reactor environment on Type 316L stainless steel welds." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76921.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 185-191).
The thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels and duplex stainless steels. Spinodal decomposition is largely responsible for the well known "475°C" embrittlement that results in drastic reductions in ductility and toughness in these materials, and this process is operative also in welds of either cast or wrought stainless steels where 6-ferrite is present. While the embrittlement can occur after several hundred hours of aging at 475°C, the process is also operative at lower temperatures, at the 288°C operating temperature of a boiling water reactor (BWR) for example, where ductility reductions have been observed after several tens of thousands of hours. An experimental study has been completed in order to understand how the spinodal decomposition may affect material properties changes in BWR pipe weld metals as well as the effects of the BWR environment on Type 316L stainless steel welds. This thesis also represents the first systematic and rigorous investigation of environmental fracture. In addition, weld metal centerline SCC crack growth rate has been quantified. Material characterization includes SCC crack growth, in-situ fracture toughness, fracture toughness in air, as well as Charpy-V and tensile property evaluation as a function of aging time and temperature. SCC crack growth rate results in BWR normal water chemistry indicate an approximately 2X increase in crack growth rate over that of the unaged material. In-situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40% over the corresponding at-temperature air values. This has been termed "environmental fracture" Detailed analyses of the results strongly suggest that spinodal decomposition is responsible for the degradation in properties measured ex-environment. SCC crack growth rate and fracture toughness have been linked to the microstructural features of the Type 316L weld metal. Analysis of the results also strongly suggests that the in-situ properties degradation is the result of hydrogen absorbed by the material during exposure to the high temperature aqueous environment.
by Timothy R. Lucas.
Ph.D.
42
), Lerch Andrew (Andrew J. "Measurement of near-surface void fraction and macrolayer thickness in boiling water and silica-based nanofluid." Thesis, Massachusetts Institute of Technology, 2008. http://hdl.handle.net/1721.1/44839.
Full textAbstract:
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008.Includes bibliographical references (leaves 48-49).
Nanofluids are engineered fluids that contain a suspension of nanoparticles in a pure substance. Nanoparticles can be any variety of metals, metal oxides, or ceramics. They have been shown to increase heat transfer properties such as thermal conductivity, convective heat transfer, and critical heat flux(CHF). An optical probe used to detect phase was used to measure the void fraction during boiling, from which the macrolayer thickness can be derived. The optical probe was verified to have an error of 11.9% and 10.4% for measuring bubble diameter in water and R-123, respectively, and an error of 5.2% and 7.1% for measuring velocity in water and R-123. The macrolayer dryout theory of CHF was tested by investigating the change in macrolayer thickness for different heat fluxes in de-ionized (DI) water and 0.01% (by volume) SiO₂nanofluid. A current controlled power source heated a sandblasted, stainless steel plate resting in an isothermal bath. The silica nanofluid had a CHF enhancement of 82% over the DI water along with a slightly higher (20% enhancement) heat transfer coefficient. The macrolayer thickness, as measured by the optical probe, at a comparable heat flux was much larger than the DI water, possibly due to the increased wettability of the heater caused by the deposition of nanoparticles on the heater. This trend is in agreement with prediction of existing theory.
by Andrew Lerch.
S.B.
43
Brodeur, David Lester. "A study of US Nuclear Power Boiling Water Reactor, Class IV, operating performance, 1992-1997." Thesis, Monterey, California. Naval Postgraduate School, 1998. http://hdl.handle.net/10945/9035.
Full text
44
Virgen, Matthew Miguel. "Comparison of water boiling models against recent experimental data, with special emphasis on the bubble ebullition cycle." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76939.
Full textAbstract:
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 37).
Using recently collected data which was measured with state-of-the-art techniques, models for nucleation site density, bubble departure diameter, and nucleation frequency were compared against the acquired data. The particular focus of this work is on the ebullition cycle and associated bubble nucleation frequency, looking at the models proposed by M.Z. Podowski. In my analysis, I took the average values for the growth and dwell times directly from the data, rather than from the models for those parameters. The results of those investigations showed that the basic principles approach for considering the parameters of the ebullition cycle held up pretty well with the experimental data, with Ti(t), the temperature curve during the ebullition cycle, corresponding remarkably well with the data curves. However, one parameter which was always overvalued was T(0*) - the predicted temperature of the start of the dwell phase. It was generally 1-2 degrees Celsius higher than the experimental value. For a fully rigorous analysis of the ebullition models in future works, it is recommended that all parameters be predicted rather than pulled from the data, particularly of the growth and dwell times.
by Matthew Miguel Virgen.
S.B.
45
Phillips, Bren Andrew. "Experimental investigation of subcooled flow boiling using synchronized high speed video, infrared thermography, and particle image velocimetry." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/92060.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages [133]-138).
Subcooled flow boiling of water was experimentally investigated using high-speed video (HSV), infrared (IR) thermography, and particle image velocimetry (PIV) to generate a unique database of synchronized data. HSV allowed measurement of the bubble departure diameter. IR thermography allowed measurement of wall superheat (local distribution and surface-averaged values), heat transfer coefficient, nucleation site density, and bubble frequency. Particle image velocimetry allowed for the measurement of velocity profiles in the liquid phase for single bubble nucleation events. The tests were performed at pressures of 1.05, 1.5, and 2.0 bar and at subcoolings of 5, 10, and 15 °C. The mass flux values explored were 150-1250 kg/m2/s. The heat flux values explored were 100-1600 kW/m2. As expected, the heat transfer coefficients increased with increasing mass flux in the single-phase convection and partial boiling regions, and converged to a fully-developed boiling curve for high heat fluxes. The bubble departure diameter decreased with increasing mass flux and decreasing heat flux; in accordance with Sugrue's model. The nucleation site density increased with increasing superheat and decreasing mass flux, as predicted by Kocamustafaogullari and Ishii's model. The nucleation site density models under-predicted the nucleation site density for a given wall superheat. Wait time and frequency models did not reproduce the data accurately, and underestimated wait time by an order of magnitude. A new mechanistic model for calculating the wait time was developed that split the wall heat flux into the component that is transferred to the fluid, and the component that is transferred as sensible heat into the heater wall. Significant localized cooling was observed underneath bubbles sliding along the wall after departure from a nucleation site, an effect which should be considered in advanced models of subcooled flow boiling. The sliding bubble thermal effects were found to be insensitive to system conditions and were limited by the thermal conduction within the substrate. Bubble growth front velocities, and regions of flow influence of departing bubbles were measured with PIV. The database generated in this project can be used to inspire or validate mechanistic models and/or CFD simulations of subcooled flow boiling heat transfer.
by Bren Andrew Phillips.
Ph. D.
46
Stehle, Gregory Raymond. "Confinement of Nucleation Sites in Nucleate Pool Boiling Using Atomic Layer Deposition and Constrictive Heaters." Thesis, University of Pittsburgh, 2017. http://pqdtopen.proquest.com/#viewpdf?dispub=10645783.
Full textAbstract:
Boiling heat transfer is a powerful cooling mechanism used in a variety of industries to efficiently dissipate heat by taking advantage of latent heat. Nucleation site interactions have been demonstrated to affect behaviors in the bulk fluid, in the solid substrate and coalescence. Despite extensive studies of multi-site interactions, the conclusions of these studies are not in agreement. Namely, hydrodynamic effects are explained by some studies to promote nucleation while other studies find that, even with thermally isolated heat supplies, the presence of nearby sites diminishes nucleation. The present study identifies superheated fluid as a possible explanation for this variability. Hydrodynamic factors are determined to only promote single site nucleation if there is an appreciable thermal boundary layer present. Even with a thermal boundary layer, the presence of other sites causes competition over the superheated fluid; thus, diminishing the promotive effects of hydrodynamic factors. There have also been studies that have characterized the changing dimensions of the microlayer and the heat transfer that occurs beneath it. However, there is not a complete study of bubble behavior resulting from varying heater areas; specifically heater areas smaller than the microlayer. The present study quantifies the effect of heater diameter on vapor effectiveness and determines the optimal heater diameter. A metric for the coincidence of vapor production and microlayer coverage is proposed. Vapor effectiveness and the coincidence metric are shown to have similar relationships with heater diameter.
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Su, Guanyu Ph D. Massachusetts Institute of Technology. "Thermohydraulics and suppression of nucleate boiling in upward two-phase annular flow : probing multiscale physics by innovative diagnostics." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119035.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 176-181).
In the fuel assemblies of a boiling water reactor (BWR) the steam quality increases along the assembly's length as heat is transferred from the fuel rods to the water coolant. Nucleate boiling is the dominant heat transfer mechanism at low and intermediate steam qualities (typical of the bubbly and slug/churn flow regimes), while forced convective evaporation dominates at higher steam quality in the annular flow regime. The transition of the heat transfer mechanism, also called suppression of nucleate boiling (SNB), affects the local heat transfer coefficient (HTC), the stability of the liquid film, and the entrainment dynamics. To support the efficient design and safe operation of future BWRs with higher power density, a thorough understanding of the thermohydraulic mechanisms and an accurate prediction of the transition conditions for SNB in annular flow is quite desirable. An innovative diagnostic technique combining synchronized infrared thermography and an electrical conductance-based liquid film thickness sensor was utilized here to investigate the details of the SNB phenomena with high spatial and temporal resolutions. The main control parameters of the tests included: the mass flux from 700 to 1400 kg-m⁻²-s⁻¹, steam quality from 0.01 to 0.08, and heat flux from 100 to 2000 kW-m⁻². The system pressure was held close to atmospheric pressure. At each set of conditions, the local distributions of the 2D surface temperature, 2D heat flux, and quasi-2D liquid film thickness were measured. From the measured data, the SNB heat flux, the SNB wall superheat, and the hydrodynamic properties of the disturbance waves were extracted. The experimental observations show for the first time the multiscale interaction of the extremely thin film and small nucleation cavities (on the scale of 10 micron), with the large disturbance waves and their associated temperature oscillations (with wavelengths of ~10 cm). A first of a kind 1D mechanistic model was developed to accurately capture this unique transient effect of the disturbance waves on the local heat transfer. The experimental results also suggest a strong dependency of the SNB heat flux and wall superheat on steam quality, with a second-order, weaker dependency on total mass flux. The same dependency is also found for the disturbance wave properties. A complete set semi-empirical correlations was proposed for predicting the time-averaged film thickness and SNB thermal conditions. Good agreement is found between the semi-empirical correlations and the experimental results. The database generated in this project can be further used for development and validation of CFD models of SNB and two-phase heat transfer in annular flow.
by Guanyu Su.
Ph. D.
48
Arment, Tyrell W. (Tyrell Wayne) 1988. "Departure from nucleate boiling and pressure drop prediction for tubes containing multiple short-length twisted-tape swirl promoters." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/76962.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 150-157).
Previous studies conducted at MIT showed that the power performance of an inverted pressurized water reactor (IPWR) conceptual design, i.e. the coolant and moderator are inverted such that the fuel is the continuous medium and the moderator flows through coolant channels, has potential to outperform a traditional pressurized water reactor (PWR). Similar to the traditional PWR, the IPWR design involves a tradeoff between core pressure drop and the minimum departure from nucleate boiling ratio (MDNBR). In order to increase the power density of the IPWR, Ferroni [231 examined the possibility of inserting multiple short-length twisted-tapes (MSLTTs) in the cooling channels. For a fixed coolant mass flow rate, the swirling flow produced by the MSLTTs allows the IPWR to have a higher operating heat flux while maintaining the design criteria of MDNBR as compared to either the traditional PWR or IPWR without swirl promoters. However, the addition of each twisted-tape increases the core pressure drop which limits the coolant flow rate due to pumping power limitations of existing reactor coolant pumps (RCPs). In order to better characterize the critical heat flux (CHF) enhancement caused by the addition of MSLTTs, this study performed a critical analysis of existing CHF correlations and models. Initially a phenomenological model was sought to describe the mechanisms of CHF for tubes containing MSLTTs; however, the full-length twisted-tape (FLTT) model that was selected for modification was found to have terms that could not be reconciled for the transition from fully developed swirl to decaying swirl. The existing CHF correlations for swirling flow were also found to be unsatisfactory. Therefore, the insights gained through working with the phenomenological model were used to develop a new empirical correlation to describe the departure from nucleate boiling (DNB) using existing swirling flow DNB data as well as an existing swirl decay model. In order to allow for more flexibility in the placement of the MSLThs, an existing FLTT pressure drop correlation was modified to account for the form pressure drop at the entrance to each twisted-tape insert as well as the friction pressure drop in the decaying swirl region downstream from the exit of each MSLTT. A sensitivity analysis of the new pressure drop correlation was also performed to determine if the complete methodology could be simplified. Design insights were presented that help to narrow the design space for the IPWR. These steps should be followed in order to find the maximum power density possible by the IPWR design. Finally, the existing swirl flow CHF data and correlations are presented in the appendices of this thesis.
by Tyrell Wayne Arment.
S.M.
49
Gupta, Atul. "Development of Boiling Water Reactor Nuclear Power Plant Simulator for Human Reliability Analysis Education and Research." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1355347881.
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Sapin, Paul. "Etude expérimentale de l'ébullition en masse dans un milieu poreux modèle." Phd thesis, Toulouse, INPT, 2014. http://oatao.univ-toulouse.fr/12148/1/sapin_partie_1_sur_2.pdf.
Full textAbstract:
Ce travail propose une étude expérimentale de l'ébullition en masse dans une structure poreuse modèle. L'objectif est d'approfondir la compréhension des transferts de chaleur dans un écoulement diphasique avec changement de phase liquide-vapeur en milieu poreux, en liaison avec la problématique de la gestion des accidents graves dans les réacteurs nucléaires. A la suite d'un dysfonctionnement sur le circuit de refroidissement d'un réacteur nucléaire, l'augmentation de la température au sein du cœur provoque l'effondrement des tubes contenant le combustible. Il en résulte la formation d'un lit de débris chaud, assimilable à un milieu poreux dégageant une puissance thermique importante, qui peut être refroidi efficacement par renoyage avec de l'eau. Cela engendre des mécanismes d'ébullition intenses qu'il convient de modéliser proprement pour estimer les chances de succès du renoyage. Notre étude vise à caractériser les échanges de chaleur à l'échelle du pore en fonction des caractéristiques de l'écoulement local. Une partie importante du travail a été consacrée à la mise au point du dispositif expérimental. Le cœur du dispositif est un milieu poreux bidimensionnel formé de cylindres disposés aléatoirement entre deux plaques de céramique. Chaque cylindre est une sonde à résistance de platine, utilisée non seulement pour fournir la puissance thermique désirée mais aussi pour mesurer la température de l'élément : chaque élément chauffant est contrôlé individuellement ou en groupe à l'aide d'un système d'asservissement temps réel. La plaque supérieure étant transparente, la distribution des phases au sein du poreux est obtenue par visualisation haute vitesse. L'acquisition d'images et les mesures thermiques permettent de caractériser l'échange de chaleur effectif local en fonction du régime d'ébullition. Deux configurations principales ont été étudiées. Dans la première, le milieu est initialement saturé en liquide et chauffé jusqu'à l'apparition de la vapeur et l'obtention de différents régimes d'ébullition. Ceci a notamment permis d'établir des courbes de Nukiyama en milieu confiné. Dans la seconde, dite de renoyage, le liquide est injecté dans le milieu sec et surchauffé initialement. Ceci a permis de caractériser la dynamique du renoyage et de visualiser les régimes d'écoulement rencontrés. Les résultats sont discutés en relation avec le modèle macroscopique à non-équilibre thermique local actuellement le plus avancé pour l'étude de ces différentes situations d'ébullition.