Um dos parâmetros termo-hidráulicos de segurança mais importantes no projeto e operação de reatores a água pressurizada é o fluxo de calor crítico (FCC). O FCC ocorre quando se atinge uma região de instabilidade na mudança de mecanismo de transferência de calor de uma parede aquecida para um fluido, aumentado drasticamente a temperatura da parede. Transientes em um reator nuclear podem afetar a taxa de geração de calor ou a fluxo de refrigerante no núcleo, prejudicando a retirada de calor das varetas combustíveis. Conhecer o FCC nestas condições é essencial para evitar danos às varetas e, consequentemente, a liberação de material radioativo. O objetivo deste trabalho é analisar o FCC para o LABGENE (Laboratório de Geração Nucleoelétrica) por meio do planejamento experimental e da simulação de seções de teste em condições de operação utilizando o código COBRAIIIc/MIT-1 e a correlação EPRI para o FCC. Considerou-se primeiramente seções de teste 3×3 de dois tamanhos distintos e os resultados para 100 pontos experimentais foram mostrados por meio de superfícies de resposta, a fim de melhor visualizar e analisar o comportamento de FCC para cada condição. Dois pontos importantes são os valores máximo e mínimo do FCC encontrados. O valor máximo (1,038 MBtu/hr.ft2 ou 3,27 MW/m2) indica o fluxo de calor necessário para a realização dos experimentos e o mínimo (0,162 MBtu/hr.ft2 ou 0,51 MW/m2) indica a pior condição de operação, a qual estaria mais próxima do ponto de ebulição. As simulações e modificações no código foram verificadas utilizando o banco de dados da Universidade de Columbia. Foram selecionados 2718 pontos experimentais referentes a seções de teste 5×5 com perfil de potência uniforme. Os resultados foram apresentados pela razão entre o valor predito e o valor experimental (DNBR) e os limites de tolerância unilateral 95/95 foram calculados, estando dentro dos valores esperados. / One of the most important thermal-hydraulic safety parameters for pressurized water reactor design and operation is the critical heat flux (CHF). The CHF occurs when a region of instability reached in the change of heat transfer mechanism from a hot wall to a fluid is reached, dramatically increasing the wall temperature. Transients in a nuclear reactor can affect the heat generation rate or the coolant flow in the core, impairing the removal of heat from the fuel rods. Knowledge of the CHF on these conditions is essential to prevent fuel rod damages and therefore the release of radioactive material. The main goal of this work is to analyze the CHF for LABGENE (Nuclear-electrical Generation Laboratory) by an experimental design and test sections simulation in operating conditions by using COBRAIIIc/MIT-1 code and the EPRI correlation for CHF. 3x3 test sections were initially considered for two different heights and outcomes for 100 experimental points were shown by means of response surfaces in order to better visualize and analyze the behavior of CHF for each condition. Two important points are the maximum and minimum values of the CHF found. The maximum value (1.038 MW/m2 or 3.27 MBtu/hr.ft2) indicates the power required for the experiments and the minimum one (0.162 MBtu/hr.ft2 or 0.51 MW/m2) indicates the worst operation condition, which would be closer to the boiling point. Code simulations and modifications were verified using the CHF database of Columbia University. 2718 data points pertaining to test sections 5×5 with uniform power profile were selected. The results were presented by the ratio between the predicted value and the experimental value (DNBR) and the limits of unilateral tolerance 95/95 were calculated, being within the expected values.

Orientador: Prof. Dr. Pedro Carajilescov / Dissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia, 2015. / Fluxo Crítico de Calor (FCC) consiste no principal limite termohidráulico de reatores do tipo PWR, que representa a opção nuclear brasileira. Trata-se, ainda, de um fenômeno de entendimento limitado. Em projetos, a estimativa de seu valor é realizada apenas por correlações empíricas, resultando valores aproximados, com elevadas incertezas. O presente projeto consiste no desenvolvimento de um método computacional para o cálculo do FCC, através de conceitos de Redes Neurais Artificiais, programado na linguagem Fortran, utilizando para treinamento e teste os dados das chamadas "Look up Tables" (LUT). Considerou-se a faixa de variação dos dados das tabelas, com a pressão variando de 1 e 21 MPa, fluxo de massa, na faixa de 50 a 8000 kg m-2 s-1 e título do escoamento entre - 0,5 a 0,9. Comparando os resultados da RN com a LUT, a média da razão dos valores resultou em 0,993, com o erro médio quadrático de 13,3%. Com a rede neural foi realizado o estudo paramétrico do FCC, para observar a influência dos parâmetros operacionais tais como pressão, fluxo de massa e título termodinâmico. Observa-se o aumento do FCC com o aumento do fluxo de massa e a atenuação do FCC com o aumento da pressão e título, como esperado. Porém algumas tendências imprevistas ocorreram, as quais podem ser atribuídas à incerteza dos dados, ou por fatos desconhecidos do fenômeno. A aplicação da rede neural em geometrias de feixe de varetas com arranjo quadrado apresentou bons resultados pelo método de balanço de energia (HBM) e a correção de PEI com erro médio quadrático de até 20,08%. Pelo método da substituição direta (DSM), foram elaborados diversos métodos de correção para adaptar os valores da rede neural à geometria de feixe de varetas. Os resultados não foram satisfatórios, pois apresentaram erro médio quadrático elevado, sendo o menor erro médio quadrático alcançado de 19,92%, utilizando uma rede neural com o espectro de parâmetros de entradas restritos e fator de correção multivariável. A correlação de EPRI com a correção de PEI apresentou resultado de erro médio quadrático de 18,73%, sendo menor que todos os métodos desenvolvidos nesse projeto. Portanto, o método de rede neural, desenvolvido nesse trabalho, não se revelou satisfatório para aplicação em feixe de varetas. / The critical heat flux (CHF) is one of the principal thermal hydraulic limits of PWR type nuclear reactors. To date, the CHF phenomenon is not well understood. So, for design purpose, the CHF is usually estimated by empirical correlation, resulting in approximate values, with high uncertainties. As an alternative to traditional methods, the present work consists in the development of an artificial neural network (ANN) to estimate the CHF, based on Look Up Table CHF data, published by Groeneveld (2006). Three parameters were considered in the development of the ANN: the pressure in the range of 1 to 21 MPa, the mass flux in the range of 50 to 8000 kg m-2 s-1 and the thermodynamic quality in the range of - 0,5 to 0,9. Comparing the ANN predictions with the data of the Look Up Table, it was observed an average of the ratio of 0.993 and a root mean square error (rms) of 13.3%. With the developed ANN, a parametric study of CHF was performed to observe the influence of each parameter in the FCC. It was possible to note that the CHF decreases with the increase of pressure and thermodynamic quality, while CHF increases with the mass flow rate, as expected. However, some erratic trends were also observed which can be attributed to either unknown aspect of the FCC phenomenon or uncertainties in the data. The ANN application in square array of rods bundle demonstrated nice result for the heat balance method (HBM) with the PEI correction resulting in rms of 20,08%. A few methods of correction were developed for the direct substitution method (DSM) to adapt the ANN in rod bundle geometry. The results wasn¿t satisfactory, because the best rms reached was 19.92%, using the ANN with restricted input range and multivariable correction factor. EPRI correlation with PEI correction results in rms of 18.73%, being better than all of developed methods in this project. Therefore, the ANN method, developed in this work, does not seem to be satisfactory for the application in rod bundle.

Graduation these deals with investigation of critical heat flux in pressurized water nuclear reactors. Theoretical part covers fundamental terms from area hydrodynamics of two-phase flow and critical heat flux. Here are also mentioned the individual approaches to description of physical process of heat transfer crisis. Practical part is devoted to systems design of measuring stand for critical heat flux in vertical canal allowing visualization of two-phase flow.

<p>Two-phase jet impingement combines the latent heat absorbed by boiling heat transfer with the strong forced convection of an impinging jet. It is a compact and highly effective heat transfer method that is capable of high heat transfer coefficients and high boiling critical heat flux limits. This makes it a suitable technology for electronics immersion cooling applications when configured as a submerged jet of a dielectric coolant. Previous studies have focused on the heat-flux-controlled nucleate boiling performance of an impingement jet up to the critical heat flux. Exploration of other boiling regimes that occur under temperature-controlled surfaces is of fundamental importance to fully understand the design space. It has been shown for free jets that a high and consistent heat flux can be dissipated over a wide range of surface superheats in the transition boiling regime when the surface is temperature-controlled. This effect is strongest in the stagnation zone, directly beneath the jet. Literature that studies this so-called “shoulder effect”, or heat flux shoulder, is scarce and almost completely focused on applications in metals processing using free jets of water as the coolant. It has been hypothesized previously that the impinging subcooled liquid delays and disrupts the start of film boiling, thereby dissipating heat flux levels comparable to that during nucleate boiling. To exploit operation in this unique transition boiling regime for potential applications in immersion cooling of electronics, the occurrence of this shoulder effect, as well as means for estimating the shoulder heat flux across different operating conditions, must be investigated for submerged jets and dielectric coolants. </p> <p>In this work, temperature-controlled submerged jet impingement is experimentally characterized using HFE-7100. A copper heater sized to be completely covered by the jet stagnation zone is increased in surface temperature throughout the transition boiling regime via a PID controller, which allows for steady-state temperature-controlled data to be acquired in this regime. The boiling curves, including critical heat flux and shoulder heat flux, are measured for jet velocities from 0.5-3 m/s and inlet subcooling from 5-30 K. The shoulder effect is shown to exist in these conditions. High-speed imaging is used to relate the flow behavior to the boiling thermal measurements and shows that the shoulder heat flux effect is an enhanced film heat transfer in the film-like mode of transition boiling. Trends and dependencies on inlet subcooling and jet velocity are measured and used to assess available predictive tools. It is observed that there is a proportionality between the critical heat flux and the shoulder heat flux. This implies a mechanistic similarity between the two effects. With further data to correlate, this similarity can potentially be used to predict the shoulder heat flux leveraging existing correlations for the critical heat flux, widening the design space of two-phase jet impingement systems. </p>

<p>The effect of pressure transients on the onset of intermittent dryout in upward annular flow was experimentally investigated in order to resolve the conflict between the observations drawn from two major data sets in the literature. A delay in time to the onset of dryout at the test section exit relative to the time predicted based on steady-state data was observed in the R-12 experiments of Celata et al (1988; 1991). Steady-state prediction methods were sufficient to predict the upstream progression of a pre-existing dryout front in the water experiments of Lyons and Swinnerton (1983). Steady state and pressure transient dryout experiments were performed using water with outlet pressures of 2 to 6 MPa and mass fluxes of 1000 to 2500 kg/m2/s in an electrically heated 1.32 m long 4.6 mm ID vertical Inconel 600 tube with depressurisation rates of up to 1.0 MPa/s. Transient experiments were performed with a small margin to dryout and with post-dryout initial conditions in order to test the hypothesis that these initial conditions influenced the onset of dryout during transients. The results of a comparison between the steady dryout data and two dryout prediction methods---the Biasi et al (1967) correlation and the 2006 CHF look-up table (Groeneveld et al, 2007)---were used to develop correction factor correlations to reduce systematic error when these methods were used to predict the transient time to dryout. These modified methods yielded mean predicted dryout delays of -0.1 and 1.5 s respectively with standard deviations of approximately 3 s. There was no statistically significant variation between the pre- and post-dryout initial conditions. Based on this result it was concluded that the initial conditions did not affect the observed time to dryout. The mean wall temperature exhibited a discontinuous decrease as the heat flux approached 92 to 95% of the dryout value. It was postulated that this was caused by a heat transfer regime change from liquid film evaporation to droplet evaporation based on the observations of Hewitt (1970), Doroschuk et al (1970) and Groeneveld (2011). For the range of conditions of the present work the onset of intermittent dryout (Groeneveld, 1986) was caused by deterioration of droplet evaporation heat transfer. Celata et al (1988) noted that in their pressure transient experiments the decrease in saturation temperature drove a rapid increase in the heat flux to the fluid. This was caused by the release of stored thermal energy as the test section wall cooled. Celata et al (1991) stated that the systematic dryout delay was observed for depressurisation rates greater than 0.2 MPa/s. Using Celata et al's (1988) pressure transient data it was concluded that the stored thermal energy transient did not influence the onset of intermittent dryout when rho_w c_pw L_w *(dT_sat/dt)<0.3*q''_a.</p> / Doctor of Philosophy (PhD)

Complex or smart fluids, made of evenly and stably suspended nanoparticles, have been an object of considerable research in the last decade due to their promising applications in a number of applications such as micro-electronic cooling, high power demands in nuclear plants, smaller and more efficient heat exchangers, oil recovery, and transportation. Nanofluids consist of typically less than 100nm sized particles dispersed in base fluids. The heat transfer characteristics of several nano-oxide suspensions in pool boiling with a suspended heating NiChrome wire have been analyzed. The pH value of the nanosuspensions is important from the point of view that it determines the stability of the particles and their mutual interactions towards the suspended heated wire. Heat transfer in silica nanofluids at different acidity and base is measured for various ionic concentrations in a pool boiling experiment. Nanosilica suspension increases the critical heat flux by up to 300% compared to conventional fluids. The l0 nm particles possess a thicker double diffuse layer compared to 20 M particles. The catalytic properties of nanofluids decrease in the presence of salts, allowing the particles to cluster and minimize the potential increase in heat transfer. Nanofluids in a strong electrolyte, i.e., in high ionic concentration, allow a higher critical heat flux than in buffer solutions because of the difference in surface area. The formation and surface structure of the deposition affect the thermal properties of the liquid. When there is no particle deposition on the wire, the nanofluid increases CHF by about 50% within the uncertainty limits, regardless of pH of the base fluid or particle size. The extent of oxidation on the wire impacts CHF, and is influenced by the chemical composition of nanofluids in buffer solutions. The boiling regime is further extended to higher heat flux when there is agglomeration on the wire. This agglomeration allows high heat transfer through inter-agglomerate pores, resulting in a nearly 3-fold increase in burnout heat flux. The pool boiling heat transfer has been even higher (- up to 4 times that of the base fluid) for Double Walled Carbon Nanotubes (DWNTs). A comparison study between Single and Double Walled Carbon Nanotube suspensions has been performed. A closed flow loop has been designed and fabricated to study the thermal transport characteristics of nanosilica suspensions in heated flow. The heat transfer coefficient and pressure drop data are provided for laminar and turbulent regimes (2000

A pesquisa realizada tratou do estudo da transferência de calor e do fluxo crítico durante a ebulição convectiva no interior de canais de diâmetro reduzidos a partir de dados levantados em bancadas experimentais construídas para esta finalidade. Extensa pesquisa bibliográfica foi efetuada e os principais métodos disponíveis para previsão de coeficiente de transferência de calor, fluxo crítico e mapas de escoamento foram levantados. Os resultados obtidos foram parametricamente analisados e comparados com os métodos da literatura. Pela primeira vez para microcanais, resultados experimentais foram levantados por um mesmo autor em laboratórios distintos buscando verificar a tendência e comportamentos. Tal comparação tem sua importância destacada em face das elevadas discrepâncias observadas na literatura quando resultados de autores distintos, obtidos em condições similares, são comparados. Os resultados levantados foram utilizados na elaboração de modelos que consideram os padrões de escoamento observados em microcanais. A incorporação dos padrões permitiu o desenvolvimento de modelos mecanísticos para coeficiente de transferência de calor, fluxo crítico e critérios para a caracterização da transição entre macro e microcanais baseados na formação do padrão de escoamento estratificado e na simetria do filme líquido no escoamento anular. / This research comprises an experimental and theoretical study on flow boiling heat transfer and critical heat flux inside small diameter tubes based on data obtained in experimental facilities specially designed for this purpose. A broad literature review was carried out and the main methods to predict the heat transfer coefficient, critical heat flux and flow patterns were pointed out. The experimental results were parametrically analyzed and compared against the predictive methods from literature. For the first time, microchannels experimental results obtained by an unique researcher in distinct laboratories were compared and a reasonable agreement was observed. The importance of such a comparison is high-lighted for flow boiling inside microchannels due to the high discrepancies ob-served when results from independent laboratories obtained under similar experimental conditions are compared. Moreover, the experimental results obtained in the present study were used to develop correlations and models for the heat transfer coefficient and heat flux that takes into account the flow patterns observed in microchannels. The heat transfer coefficient and critical heat flux models were developed based on mechanistic approach. In addition, criteria to characterize macro to microchannel transition were proposed based in the occurrence of the stratified flow pattern and the liquid film symmetry under annular flow conditions.

Microchannel heat exchangers present many advantages, such as reduced size, high thermal efficiency and low fluid inventory; and are increasingly being used for heat transfer in a wide variety of applications including heat pumps, automotive air conditioners and for cooling of electronics.However, the fundamentals of fluid flow and heat transfer in microscalegeometries are not yet fully understood. The aim of this thesis is to contribute to a better understanding of the underlying physical phenomena in single-phase and specially flow boiling heat transfer of refrigerants in small channels. For this purpose, well-characterized heat transfer experiments have been performed in uniformly heated, single, circular, vertical channels ranging from 0.64 to 1.70 mm in diameter and using R-134a, R-22 and R-245fa as working fluids. Furthermore, flow visualization tests have been carried out to clarify the relation between the two-phase flow behavior and the boiling heat transfer characteristics. Single-phase flow experiments with subcooled liquid refrigerant have confirmed that conventional macroscale theory on single-phase flow and heat transfer is valid for circular channels as small as 640μm in diameter. Through high-speed flow boiling visualization of R-134a under non adiabatic conditions seven flow patterns have been observed: isolated bubbly flow, confined bubbly flow, slug flow, churn flow, slug-annular flow, annular flow, and mist flow. Two-phase flow pattern observations are presented in the form of flow pattern maps. Annular-type flow patterns are dominant for vapor qualities above 0.2. Onset of nucleate boiling and subcooled flow boiling heat transfer of R-134a has been investigated. The wall superheat needed to initiate boiling was found as large as 18 ºC. The experimental heat transfer coefficients have been compared to predictions from subcooled flow boiling correlationsav ailable in the literature showing poor agreement. Saturated flow boiling heat transfer experiments have been performed with the 640 μm diameter test section. The heat transfer coefficient has been found to increase with heat flux and system pressure and not to change with vapor quality or mass flux when the quality is less than ∼0.5. For vapor qualities above this value, the heat transfer coefficient decreases with vapor quality. This deterioration of the heat transfer coefficient is believed to be caused by the occurrence of intermittent dryout in this vapor quality range. The experimental database, consisting of 1027 data points, has been compared against predictions from correlations available in the literature. The best results are obtained with the correlations by Liu and Winterton (1991) and by Bertsch et al. (2009). However, better design tools to correctly predict the flow boiling heat transfer coefficient in small geometries need to be developed. Dryout incipience and critical heat flux (CHF) have been investigated in detail. CHF data is compared to existing macro and microscale correlations. The comparison shows best agreement with the classical Katto and Ohno (1984) correlation, developed for conventional large tubes. / QC 20101101

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Cette thèse est motivée par le besoin de compléter les connaissances actuelles des phénomènes ayant lieu lors d’un renvoi d’eau dans un lit de débris radioactifs, opération appelée « renoyage » et qui intervient dans une séquence d’accident grave où un cœur de réacteur nucléaire est dégradé suite à une perte prolongée de refroidissement primaire. Notre étude, de nature expérimentale, vise à mieux caractériser la crise d’ébullition en convection forcée, dans un milieu poreux chauffant. Le cœur du dispositif expérimental est un milieu poreux modèle quasibidimensionnel, composé de 276 cylindres disposés entre deux plaques de céramique distantes de 3 mm, dont l’une, transparente, permet de visualiser les écoulements. Les cylindres, de 2 mm de diamètre, sont des sondes thermo-résistives qui ont une double fonction : elles sont utilisées comme éléments chauffants et comme capteurs de température. Une boucle fluide permet de contrôler le débit d’injection de liquide dans la section test, la température d’injection ainsi que la pression. La section test est placée verticalement, le liquide est injecté par le bas à une température proche de la saturation. Dans une première série d’expériences, la puissance thermique dissipée globalement par un ensemble de cylindres chauffants est augmentée de façon progressive jusqu’à atteindre l’assèchement d’une zone du milieu poreux. Les résultats montrent deux types de phénoménologies dans le déclenchement de la crise d’ébullition. Pour des débits d’injection faibles (densités de flux massique de l’ordre de 4 kg.m^-2.s^-1 maximum), l’atteinte de la puissance d’assèchement se traduit par un lent recul du front diphasique jusqu’à sa stabilisation en haut de la zone chauffée ; en aval de la zone chauffée, l’écoulement est monophasique vapeur. Pour des débits d’injection plus élevés, la crise d’ébullition apparaît autour d’un des éléments chauffants, conduisant à une ébullition en film localisée, tandis qu’un écoulement diphasique liquide-vapeur continue de parcourir l’aval de la section test. Les visualisations de ces expériences permettent d’identifier qualitativement la structure des écoulements. D’autres expériences consistent à mesurer le flux critique local autour d’un cylindre choisi, pour différentes configurations d’écoulements. Le débit d’injection est fixé. Une puissance de chauffe est imposée à une ligne horizontale de cylindres en amont du cylindre choisi. Les résultats montrent que le flux critique sur ce cylindre diminue en fonction de la puissance délivrée à la ligne chauffée. La distance du cylindre étudié à la ligne chauffée semble avoir peu d’influence sur le flux critique. Des visualisations expérimentales sont utilisées pour caractériser l’écoulement diphasique en aval de la ligne chauffée, dans le but de mettre en relation le flux critique local avec des paramètres hydrodynamiques (saturations, vitesses des phases). Les images obtenues sont difficiles à exploiter. Afin de calibrer les paramètres des algorithmes de traitement d’images, nous avons reproduit une cellule d’essai de géométrie identique à l’originale, mais où l’on injecte du gaz par une ligne de cylindres en amont de la section test dans une configuration d’écoulement diphasique isotherme. Dans ce dispositif, le débit d’injection de gaz est contrôlé et mesuré. Les visualisations obtenues servent alors de références auxquelles sont comparées les visualisations d’ébullition convective. / This work is motivated by the need to better understand the phenomena occurring while some water is injected into a heated porous debris bed. This reflooding operation is a part of the planned mitigation procedure used during a Loss Of Coolant Accident (LOCA) that may occur into a nuclear power plant and results into a severe core damage. Our experimental study aims to characterize the boiling crisis that can happen in a boiling flow taking place within a heatgenerating model porous medium. The test section is a two-dimensional model porous medium, composed of an array of 276 cylinders placed between two ceramic plates spaced from one another by 3 mm, one of which is transparent and allows visualizations of the flow. The 2 mm diameter cylinders are Pt100 resistance temperature detectors that perform a dual function: they act as heating elements (heated by Joule effect) and are also used as temperature probes. A fluid loop allows controlling the liquid injection flow rate, its inlet temperature as well as its pressure. The test section is held vertically, the liquid injected from bottom at a temperature close to the saturation temperature. In a first series of experiments, the thermal power applied to a bundle of heating cylinders is progressively increased until a dry zone is detected in the porous medium. Two kinds of phenomenology are observed during these “dryout experiments”. First, at low liquid injection rate (4 kg.m^-2.s^-1 maximum mass flux), reaching the dryout power results into a liquid front receding down to the upper limit of the heated zone, while downstream the heated zone, the porous medium is vapour-saturated. Second, at higher flow rate, the boiling crisis happens at the surface of a single heating element, resulting in a local film boiling, whereas a two-phase flow still go through the whole test section. High-speed visualizations allow characterizing the flow regimes. Other experiments focus on determining the local critical heat flux on a given cylinder, for different upstream flow configurations. The inlet liquid flow rate is fixed. A thermal power is uniformly applied to a line of heating cylinders, upstream the cylinder under study. Results show that the local critical heat flux decreases as the power applied to the heated line increases. The distance from the cylinder under study to the heated line seems not to have a significant effect on the critical heat flux. Visualizations are used to characterize the two-phase flow upstream the heated line, aiming at expressing the critical heat flux as a function of the hydrodynamic parameters (saturations, phase velocities). The image analysis is particularly challenging. In order to calibrate the image processing parameters, we use a second model porous medium with the same geometry as the heat generating one, but where an isothermal two-phase flow is obtained by injecting gas into the liquid flow rather than generated by boiling. The gas injection flow rate is controlled and measured. Isothermal two-phase flow visualizations provide a reference case and are compared to flow boiling visualizations.