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Literatura académica sobre el tema "Nuclear boiling"

Autor: Grafiati

Publicado: 14 de diciembre de 2024

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Artículos de revistas sobre el tema "Nuclear boiling"

Wilczek, Frank. "Nuclear and subnuclear boiling". Nature 395, n.º 6699 (septiembre de 1998): 220–21. http://dx.doi.org/10.1038/26107.

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Guo, Zhong De y Shu Fang Zhang. "The Pure Heat Conversion Coefficient Analysis Method for Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit". Advanced Materials Research 383-390 (noviembre de 2011): 6514–18. http://dx.doi.org/10.4028/www.scientific.net/amr.383-390.6514.

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Pure heat conversion coefficient, which is one important parameter for the advance boiling water reactor nuclear power unit, is defined, and the coefficient reflects energy grade values of heaters. According to the structural characteristics of the thermodynamic system of the advanced boiling water reactor nuclear power plant, four sorts of auxiliary steam-water components are categorized. Via strict deduction and demonstration, the general matrix of the coefficient is deduced, so the thermal economic analysis of advance boiling water reactor nuclear power unit can be done with one of extraction steam efficiency. In this way, a new method of thermal economic quantitative analysis for this unit is offered.

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Giustini, Giovanni. "Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review". Inventions 5, n.º 3 (14 de septiembre de 2020): 47. http://dx.doi.org/10.3390/inventions5030047.

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The boiling process is utterly fundamental to the design and safety of water-cooled fission reactors. Both boiling water reactors and pressurised water reactors use boiling under high-pressure subcooled liquid flow conditions to achieve high surface heat fluxes required for their operation. Liquid water is an excellent coolant, which is why water-cooled reactors can have such small sizes and high-power densities, yet also have relatively low component temperatures. Steam is in contrast a very poor coolant. A good understanding of how liquid water coolant turns into steam is correspondingly vital. This need is particularly pressing because heat transfer by water when it is only partially steam (‘nucleate boiling’ regime) is particularly effective, providing a great incentive to operate a plant in this regime. Computational modelling of boiling, using computational fluid dynamics (CFD) simulation at the ‘component scale’ typical of nuclear subchannel analysis and at the scale of the single bubbles, is a core activity of current nuclear thermal hydraulics research. This paper gives an overview of recent literature on computational modelling of boiling. The knowledge and capabilities embodied in the surveyed literature entail theoretical, experimental and modelling work, and enabled the scientific community to improve its current understanding of the fundamental heat transfer phenomena in boiling fluids and to develop more accurate tools for the prediction of two-phase cooling in nuclear systems. Data and insights gathered on the fundamental heat transfer processes associated with the behaviour of single bubbles enabled us to develop and apply more capable modelling tools for engineering simulation and to obtain reliable estimates of the heat transfer rates associated with the growth and departure of steam bubbles from heated surfaces. While results so far are promising, much work is still needed in terms of development of fundamental understanding of the physical processes and application of improved modelling capabilities to industrially relevant flows.

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Kim, Kang Seog, Andrew Ward, Ugur Mertyurek, Mehdi Asgari y William Wieselquist. "Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor". Journal of Nuclear Engineering 5, n.º 3 (1 de agosto de 2024): 260–73. http://dx.doi.org/10.3390/jne5030018.

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The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.

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Podowski, Michael Z. y Raf M. Podowski. "Mechanistic Multidimensional Modeling of Forced Convection Boiling Heat Transfer". Science and Technology of Nuclear Installations 2009 (2009): 1–10. http://dx.doi.org/10.1155/2009/387020.

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Due to the importance of boiling heat transfer in general, and boiling crisis in particular, for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive efforts have been made in the past to develop a variety of methods and tools to evaluate the boiling heat transfer coefficient and to assess the onset of temperature excursion and critical heat flux (CHF) at various operating conditions of boiling channels. The objective of this paper is to present mathematical modeling concepts behind the development of mechanistic multidimensional models of low-quality forced convection boiling, including the mechanisms leading to temperature excursion and the onset of CHF.

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Bang, In Cheol, Jacopo Buongiorno, Lin-Wen Hu y Hsin Wang. "ICONE15-10030 Measurement of Key Pool Boiling Parameters in Nanofluids for Nuclear Applications". Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_11.

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Tõke, Jan. "Boiling Patterns of Iso-asymmetric Nuclear Matter". Journal of Physics: Conference Series 420 (25 de marzo de 2013): 012100. http://dx.doi.org/10.1088/1742-6596/420/1/012100.

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Stojanovic, Andrijana, Srdjan Belosevic, Nenad Crnomarkovic, Ivan Tomanovic y Aleksandar Milicevic. "Nucleate pool boiling heat transfer: Review of models and bubble dynamics parameters". Thermal Science, n.º 00 (2021): 69. http://dx.doi.org/10.2298/tsci200111069s.

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Understanding nucleate pool boiling heat transfer and, in particular the accurate prediction of conditions that can lead to critical heat flux, is of the utmost importance in many industries. Due to the safety issues related to the nuclear power plants, and for the efficient operation of many heat transfer units including fossil fuel boilers, fusion reactors, electronic chips, etc., it is important to understand this kind of heat transfer. In this paper, a comprehensive review of analytical and numerical work on nucleate pool boiling heat transfer is presented. In order to understand this phenomenon, existing studies on boiling heat transfer coefficient and boiling heat flux are also discussed, as well as characteristics of boiling phenomena such as bubble departure diameter, bubble departure frequency, active nucleation site density, bubble waiting and growth period and their impact on pool boiling heat transfer.

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Baldwin, Michael, Andre LeClair, Alok Majumdar, Jason Hartwig, Vishwanath Ganesan y Issam Mudawar. "Modeling of cryogenic heated-tube flow boiling experiments of nitrogen and methane with Generalized Fluid System Simulation Program". IOP Conference Series: Materials Science and Engineering 1301, n.º 1 (1 de mayo de 2024): 012158. http://dx.doi.org/10.1088/1757-899x/1301/1/012158.

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Abstract Accurate modeling of cryogenic boiling heat transfer is vital for the development of extended-duration space missions. Such missions may require the transfer of cryogenic propellants from in-space storage depots or the cooling of nuclear reactors. Purdue University in collaboration with NASA has assembled a database of cryogenic flow boiling data points from heated-tube experiments dating back to 1959, which has been used to develop new flow boiling correlations specifically for cryogens. Computational models of several of these experiments have been constructed in the Generalized Fluid System Simulation Program (GFSSP), a network flow code developed at NASA’s Marshall Space Flight Center. The new Purdue-developed correlations cover the full boiling curve: onset of nucleate boiling, nucleate boiling, critical heat flux, and post-critical heat flux boiling. These correlations have been coded into GFSSP user subroutines. The fluids modeled are nitrogen and methane. Predictions of wall temperature are presented and compared to the test data.

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Kurskii, A. S., V. M. Eshcherkin, V. V. Kalygin, M. N. Svyatkin y I. I. Semidotskii. "Boiling water vessel reactors for nuclear district heating". Atomic Energy 111, n.º 5 (19 de febrero de 2012): 370–76. http://dx.doi.org/10.1007/s10512-012-9506-9.

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Tesis sobre el tema "Nuclear boiling"

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 analyses
Die 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'ébullition
Nucleate 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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Libros sobre el tema "Nuclear boiling"

J, Moody F., ed. The thermal-hydraulics of a boiling water nuclear reactor. 2^a ed. La Grange Park, Ill., USA: American Nuclear Society, 1993.

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(Bonn, Germany) Kerntechnische Gesellschaft. SWR 1000: Ein zukunftsweisendes Reaktorkonzept : Tagungsbericht = SWR 1000 : a reactor concept for the future : proceedings. Bonn: INFORUM, 1998.

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Hawthorne, J. R. Experimental assessments of Gundremmingen RPV archive material for fluence rate effects studies. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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Shih, Chunkuan y K. Tien. The development and application of Kuosheng (BWR/6) nuclear power plant TRACE/SNAP model. Washington, DC: U.S. Nuclear Regulatory Commission, Division of Systems Analysis, Office of Nuclear Regulatory Research, 2014.

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Nikolaevich, Andreenko Sergeĭ, ed. Peregruzochnye mashiny kanalʹnykh i͡a︡dernykh ėnergeticheskikh reaktorov. Moskva: Ėnergoatomizdat, 1986.

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U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering. y Oak Ridge National Laboratory, eds. Boiling-water reactor internals aging degradation study: Phase 1. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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U.S. Nuclear Regulatory Commission. Division of Reactor Controls and Human Factors, ed. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors. Washington, DC: Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1995.

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Capítulos de libros sobre el tema "Nuclear boiling"

Todreas, Neil E. y Mujid S. Kazimi. "Pool Boiling". En Nuclear Systems Volume I, 625–64. Third edition. | Boca Raton : CRC Press, 2021- |: CRC Press, 2021. http://dx.doi.org/10.1201/9781351030502-12.

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Todreas, Neil E. y Mujid S. Kazimi. "Flow Boiling". En Nuclear Systems Volume I, 665–742. Third edition. | Boca Raton : CRC Press, 2021- |: CRC Press, 2021. http://dx.doi.org/10.1201/9781351030502-13.

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Zohuri, Bahman y Nima Fathi. "Convective Boiling". En Thermal-Hydraulic Analysis of Nuclear Reactors, 375–412. Cham: Springer International Publishing, 2015. http://dx.doi.org/10.1007/978-3-319-17434-1_14.

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Zohuri, Bahman. "Convective Boiling". En Thermal-Hydraulic Analysis of Nuclear Reactors, 455–99. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-53829-7_14.

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Anglart, Henryk. "Boiling Crisis". En Thermal Safety Margins in Nuclear Reactors, 253–72. Boca Raton: CRC Press, 2024. http://dx.doi.org/10.1201/9781003255000-9.

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Masterson, Robert E. "The Boiling Water Reactor". En Nuclear Reactor Thermal Hydraulics, 69–94. Boca Raton : CRC Press, [2019]: CRC Press, 2019. http://dx.doi.org/10.1201/b22067-3.

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Anglart, Henryk. "Boiling Heat Transfer". En Thermal Safety Margins in Nuclear Reactors, 209–52. Boca Raton: CRC Press, 2024. http://dx.doi.org/10.1201/9781003255000-8.

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Mesquita, Amir Z., Antonella L. Costa, C. Pereira, Maria A. F. Veloso y P. A. L. Reis. "Experimental Investigation of the Onset of Subcooled Nucleate Boiling in an Open-Pool Nuclear Research Reactor". En Film and Nucleate Boiling Processes, 183–97. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp49339t.

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Kondo, Koichi, Yasuo Ota, Hiroshi Ono, Masahiko Kuroki, Yuji Koshi y Masayoshi Tahira. "Actual Operation Control of Boiling Water Reactor". En Nuclear Reactor Kinetics and Plant Control, 129–66. Tokyo: Springer Japan, 2012. http://dx.doi.org/10.1007/978-4-431-54195-0_7.

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Actas de conferencias sobre el tema "Nuclear boiling"

Povolny, Antonin y Martin Cuhra. "Two-Phase CFD of Channel Boiling for Boiling Transition Problems". En 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30970.

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In order to ensure safety of nuclear installations, thermohydraulics has developed many ways how to predict the behavior of coolant in a heated boiling channel. Accuracy of these predictions can be improved using three-dimensional Computational Fluid Dynamics (CFD) method, which is based on first principles of fluid mechanics. Even though when using CFD, there is a struggle between the accuracy and low computation costs, in many cases CFD can provide feasible improvement of accuracy compared to more traditional approaches. In this research, the focus is set on channel boiling problems, especially those associated with boiling transitions. The phenomenon of critical heat flux (CHF) is investigated using two-phase CFD computation and is compared to experimental data. There is also comparison with other computation methods. When experiment provides some set of data, CFD calculation provides description of the whole flow behavior that provides significantly more information and is of great value during the design process when it gives the understanding of undergoing effects. Besides CHF, general ability of CFD to predict changes in boiling patterns in two-phase channel boiling flows is discussed.

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Yi, Liao, Wang Cong y Chen Lei. "Modular Boiling Water Reactor Concept". En 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81196.

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Modular boiling water reactor (MBWR) can be considered as a small sized economic simplified boiling water reactor (ESBWR). It has the advantage of easier fabrication, transportation and construction. In this paper, a 65MWe MBWR core was designed with natural circulation, passive safety features, high power density and an 18 months fuel cycle. The MBWR core consists of 104 fuel assemblies with 4.6 w/o U-235, the assemblies were divided into 3 batches based on the depletion level, the batches shuffled at the end of each cycle. The core converged to equilibrium after 8 fuel cycles. A steady-state equilibrium fuel cycle depletion analysis was performed over a 540 day cycle using the HELIOS and PARCS software. The control blades insertion patterns were chosen to minimize axial and radial power peaking and provide uniform burnup throughout the cycle. At the end of the equilibrium cycle, 16% of total control blade worth remained inserted and the average assembly burnup is 21.318 GWd/MTHM. Thermal hydraulic analysis was also performed to insure the core’s safety feature.

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Ren, Tingting, Changqi Yan, Meiyue Yan y Shengzhi Yu. "CFD Analysis on Wall Boiling Model During Subcooled Boiling in Vertical Narrow Rectangular Channel". En 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81554.

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Two-fluid model is a common method to simulate the subcooled flow boiling heat transfer, in which the wall boiling model is mainly used for the partition of wall heat flux and the mass transfer between two phases on the wall. The model determines the amount of vapor phase and predicts the cross-sectional void fraction in the channel, nucleate site density and bubble departure diameter play an important role in the accurate prediction of wall boiling model. Eulerian two-fluid model coupled with Rensselaer Polytechnic Institute (RPI) wall boiling model is employed to simulate the heat transfer characteristics and boiling phenomena in vertical narrow rectangular channels by using FLUENT code. Based on the experimental data of subcooled boiling in vertical narrow rectangular channel, different combinations of nucleate site density and bubble departure diameter correlations are used to calculate under different conditions of heat flux and inlet subcooling. Comparing the calculated heat transfer coefficients along the vertical height with experimental results, it can be found that these two parameters have a significant effect on the subcooled boiling heat transfer in narrow rectangular channels. Different parameter combinations lead to differences in wall heat flux distribution, different heat flux and inlet subcooling also have different effects on these models, which eventually lead to different evaporative heat flux, thus affecting the prediction of void fraction.

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Liu, Wei, Masanori Monde y Y. Mitsutake. "Characteristics of Transient Boiling Heat Transfer". En 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22741.

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In this paper, one dimensional inverse heat conduction solution is used for a measurement of pool boiling curve. The experiments are performed under atmospheric pressure for copper, brass, carbon steel and gold. Boiling curves, including unsteady transition boiling region, are found can be traced fairly well from a simple experiment system by solving inverse heat conduction solution. Boiling curves for steady heating and transient heating, for heating process and cooling process are compared. Surface behavior around CHF point, transition boiling and film-boiling regions are observed by using a high-speed camera. The results show the practicability of the inverse heat conduction solution in tracing boiling curve and thereby supply us a new way in boiling heat transfer research.

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Karoutas, Zeses E., Yixing Sung, Yutung R. Chang, Gennady A. Kogan y Paul F. Joffre. "Subcooled Boiling Data From Rod Bundles". En 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49492.

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This paper provides single and two phase rod bundle data to support verification of heat transfer models being used in steaming rate and crud model predictions for rod bundles. The effort to summarize this work was supported by the EPRI Robust Fuel Program and is defined in more detail in EPRI report 1003383. Subcooled boiling tests were performed by Combustion Engineering (CE) in early 1980s to provide insight on heavy crud deposits and fuel failures observed on peripheral rods for bundles in Maine Yankee cycle 4. Two 5×5 tests were performed at the Columbia University Heat Transfer Research Facility simulating the peripheral region of adjacent CE 14×14 fuel bundles for two different perimeter strip geometries. The test conditions were at typical reactor pressure, temperature, and heat flux. The rods were 7’ in heated length and were electrically heated with a uniform axial power shape. There were no mixing vanes on the spacer grids. Thermocouples were placed on the hot rod in the center of the test section and on an adjacent rod at 4 different axial levels. Thermocouples were also located in the center of the subchannels at the end of the test section. Boiling curves were generated over a range of test conditions (system pressure, inlet temperature, and flow rate) by plotting rod surface temperature versus heat flux. The boiling curves covered single phase, subcooled boiling, bulk boiling and DNB conditions. The data from the boiling curves were reduced and evaluated with the VIPRE thermal hydraulic code. Clad temperature predictions were made with VIPRE based on available heat transfer correlations for comparison to clad temperature measurements. These heat transfer correlations include the Dittus Boelter correlation for single phase flow, the Jens Lottes, Thom and Chen correlations for two phase flow conditions (subcooled boiling). The VIPRE predictions of the hot rod average surface temperature, based on the Dittus-Boelter correlation with a grid enhancement factor for single-phase forced convection and the Thom correlation for nucleate boiling, gave the best agreement with the rod bundle test data among all the available modeling options. It was concluded that current heat transfer models used in TH codes, are adequate for average steaming rate calculations supporting Axial Offset Anomalies (AOA) evaluations, as long as the appropriate grid enhancement factor is utilized for the spacer grids in the analysis. However, further testing and modeling may be needed to simulate local grid effects and hot spots downstream of spacer grids.

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Luo, Hanwen, Hongbin Wang y Jinbiao Xiong. "Numerical Simulation of Subchannel Flow Boiling Using Five-Component Wall Boiling Model and iMUSIG Model". En 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-134545.

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Abstract Boiling heat transfer of subchannel has significant influence on nuclear reactor thermal-hydraulic performance. Employing the Euler-Euler two-fluid approach in computational multi-fluid dynamic (CMFD) simulations, this study conducts a high-fidelity multi-fluid simulation on subchannels. Utilizing the OECD/NRC NUPEC PWR Subchannel and Bundle Tests (PSBT) benchmark, four subchannels are numerically simulated: a typical central subchannel, a central subchannel with a thimble, a side subchannel, and a corner subchannel. The simulation model integrates a five-component wall boiling model and an inhomogeneous multiple size group (iMUSIG) model in OpenFOAM v6, capturing phenomena like bubble sliding, size variability, coalescence, and breakup. Turbulence is modeled using the standard k-ϵ and Sato models, the latter incorporating an additional turbulent viscosity term for bubble-induced turbulence. The study compares cross-sectional averaged void fractions at measurement planes and analyzes key bubble dynamics parameters, including bubble departure and lift-off diameters, wall heat flux partitioning, and bubble size distribution. The results have shown the promising accuracy of the five-component wall boiling model coupled with the iMUSIG model in predicting the void fraction.

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Wang, Shixian, Kai Wang, Shuichiro Miwa y Koji Okamoto. "Analyzing Flow Rate Impact on CHF Front Behavior During Boiling Crisis in Downward Flow Boiling". En 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-124786.

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Abstract Our research thoroughly examines the Critical Heat Flux (CHF) in downward-facing flow boiling, essential for comprehending the intricacy of the CHF phenomenon. Conducted on a downward-facing copper surface within a flow loop, our research uncovered significant differences in CHF between upstream and downstream regions, with CHF predominantly occurring downstream first. High-speed camera analysis during boiling crisis led to the identification of the ‘CHF front,’ marking the transition from nucleate to film boiling phases. This CHF front typically originates downstream and progresses upstream, significantly influencing final CHF thresholds. This movement, linked to dry spot formation on the heated surface as per dry-out theories, is a key focus of our analysis. Furthermore, our experiments on smooth copper surfaces under varied flow rates revealed that increased flow rates decrease the CHF front velocity, delaying CHF onset and resulting in higher final CHF values. This observed effect is attributed to the enhanced supply of liquid to the micro-layer beneath the bubble, which increases its thickness and thereby prevents the formation of dry spots. While our study offers deep insights, we recognize uncertainties in our data. Future research will focus on various factors for a more comprehensive understanding of CHF front velocity. These findings are crucial for predicting CHF levels in various systems, marking a significant contribution to the field.

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Shimkevich, Alexander L., Michael N. Ivanovsky, Valentine A. Morozov, Kenneth M. Sprouse, Mohamed S. El-Genk y Mark D. Hoover. "Natural Convection Boiling Potassium Flow Loop". En SPACE NUCLEAR POWER AND PROPULSION: Eleventh Symposium. AIP, 1994. http://dx.doi.org/10.1063/1.2950136.

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Ohtsuka, Masaya, Koji Fujimura, Takuji Nagayoshi, Jun’ichi Yamashita y Yasuyoshi Kato. "Safe and Simplified Boiling Water Reactor (SSBWR)". En 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22565.

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A safe and simplified BWR (SSBWR) has been developed as an innovative future reactor to provide a super-long life core of 20 years and to realize a passive core safety system with infinite grace period. Operability and maintainability can be largely improved by using the super-long life core, cutting the number of active components, and using a one-batch core with no exchange of fuel assemblies, which can also significantly reduce the possibility of nuclear proliferation. Np-237 of MAs (Minor Actinides) can be effectively transmuted using the very hard neutron spectrum of SSBWR and high level radioactive wastes can be reduced.

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Ohtake, Hiroyasu y Yasuo Koizumi. "Boiling Heat Transfer Under Oscillatory Flow Condition". En 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89582.

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Onset of nucleate boiling (ONB), point of net vapor generation (NVG) and critical heat flux (CHF) on subcooled flow boiling under oscillatory flow, focusing on liquid velocity, amplitude and frequency of oscillations were investigated experimentally and analytically. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2, 4 and 6 Hz, respectively; amplitude of oscillatory flow was 25 and 50 % in a ratio of main flow rate, respectively. Temperature at ONB and critical heat flux for oscillatory flow were lower than those for steady flow. The decreasing of liquid velocity by oscillatory caused the ONB and the CHF to decrease. The effects of liquid oscillatory flow on the incipient boiling temperature were examined through a classical stability theory of preexisting nuclei. Heat flux at NVG reduced by superimposed oscillatory flow; the increase of liquid velocity by the oscillatory flow caused the NVG to decrease. The CHF under the oscillatory flow condition was examined analytically by using a liquid sub-layer model and a lumped-heat-capacity model in the heater.

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Informes sobre el tema "Nuclear boiling"

Wheeler, Timothy A. y Huafei Liao. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment. Office of Scientific and Technical Information (OSTI), diciembre de 2014. http://dx.doi.org/10.2172/1323653.

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Anh Bui, Nam Dinh y Brian Williams. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study. Office of Scientific and Technical Information (OSTI), septiembre de 2013. http://dx.doi.org/10.2172/1110336.

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Rosa, M. P. y M. Z. Podowski. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors. Office of Scientific and Technical Information (OSTI), septiembre de 1995. http://dx.doi.org/10.2172/107760.

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Wang, Jy-An John, Hong Wang, Hao Jiang y Yong Yan. CIRFT testing of high-burnup used nuclear fuel rods from pressurized water reactor and boiling water reactor environments. Office of Scientific and Technical Information (OSTI), septiembre de 2015. http://dx.doi.org/10.2172/1214025.

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N. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico. Office of Scientific and Technical Information (OSTI), febrero de 2003. http://dx.doi.org/10.2172/823492.

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Konzek, G. J. y R. I. Smith. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station. Office of Scientific and Technical Information (OSTI), diciembre de 1990. http://dx.doi.org/10.2172/6416764.

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Short, S., A. Luksic y M. Schutz. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors. Office of Scientific and Technical Information (OSTI), junio de 1989. http://dx.doi.org/10.2172/5785023.

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Block, J. A., C. Crowley, F. X. Dolan, R. G. Sam y B. H. Stoedefalke. Nucleate boiling pressure drop in an annulus: Book 4. Office of Scientific and Technical Information (OSTI), noviembre de 1992. http://dx.doi.org/10.2172/10148049.

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Block, J. A., C. Crowley, F. X. Dolan, R. G. Sam y B. H. Stoedefalke. Nucleate boiling pressure drop in an annulus: Book 3. Office of Scientific and Technical Information (OSTI), noviembre de 1992. http://dx.doi.org/10.2172/10148052.

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Block, J. A., C. Crowley, F. X. Dolan, R. G. Sam y B. H. Stoedefalke. Nucleate boiling pressure drop in an annulus: Book 2. Office of Scientific and Technical Information (OSTI), noviembre de 1992. http://dx.doi.org/10.2172/10148055.

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