abstract: Microchannel heat sinks can possess heat transfer characteristics unavailable in conventional heat exchangers; such sinks offer compact solutions to otherwise intractable thermal management problems, notably in small-scale electronics cooling. Flow boiling in microchannels allows a very high heat transfer rate, but is bounded by the critical heat flux (CHF). This thesis presents a theoretical-numerical study of a method to improve the heat rejection capability of a microchannel heat sink via expansion of the channel cross-section along the flow direction. The thermodynamic quality of the refrigerant increases during flow boiling, decreasing the density of the bulk coolant as it flows. This may effect pressure fluctuations in the channels, leading to nonuniform heat transfer and local dryout in regions exceeding CHF. This undesirable phenomenon is counteracted by permitting the cross-section of the microchannel to increase along the direction of flow, allowing more volume for the vapor. Governing equations are derived from a control-volume analysis of a single heated rectangular microchannel; the cross-section is allowed to expand in width and height. The resulting differential equations are solved numerically for a variety of channel expansion profiles and numbers of channels. The refrigerant is R-134a and channel parameters are based on a physical test bed in a related experiment. Significant improvement in CHF is possible with moderate area expansion. Minimal additional manufacturing costs could yield major gains in the utility of microchannel heat sinks. An optimum expansion rate occurred in certain cases, and alterations in the channel width are, in general, more effective at improving CHF than alterations in the channel height. Modest expansion in height enables small width expansions to be very effective. / Dissertation/Thesis / M.S. Mechanical Engineering 2011

Durante um acidente em uma usina nuclear, a integridade do vaso pressurizado deve ser assegurada. Em resposta a um possível derretimento do combustível nuclear, a atual geração de usinas possui um sistema para a injeção de água potável na cavidade do vaso pressurizado com intuito de resfriar sua parede, prevenindo danos a sua estrutura e evitando o vazamento de material radioativo. Esse estudo considerou o uso de água marinha como refrigerante para inundar a cavidade do vaso pressurizado combinado com a fixação de um estrutura porosa em forma de grade em sua parede externa como meio de aprimorar a margem de segurança durante a contenção de acidentes. Experimentos de longa duração para a ebulição em piscina de água marinha artificial foram conduzidos em uma superfície circular de cobre plana com 30 mm de diâmetro. Foi encontrado um fluxo de calor crítico de 1; 6 MW/m2 sob pressão atmosférica. Esse valor é significantemente maior que aquele obtido (1; 0 MW/m2) nas mesmas condições experimentais. Foi verificado que os depósitos de sais marinhos podem aumentar a molhabilidade e a capilaridade da superfície de teste, aprimorando assim o fluxo crítico. Combinando a água marinha e a fixação da estrutura porosa sobre a superfície de teste, verificou-se um melhora no coeficiente de transmissão de calor e no fluxo de calor crítico de até 110 % (2; 1 MW/m2), quando comparado a água destilada na superfície limpa, sem a instalação da estrutura. Após os experimentos, foi identificado que muitos dos poros presentes nas superfícies da estrutura porosa encontravam-se bloqueados devido ao aglutinamento de sais marinhos. Isso levou a conclusão que o aumento no valor do fluxo crítico observado para a água marinha artificial ocorreu devido, principalmente, a separação das fases líquida e gasosa do fluido na região próxima a superfície de teste, efeito proporcionado pela forma de grade da estrutura porosa, e ao aumento da molhabilidade e capilaridade da superfície devido a formação dos depósitos marinhos. / During a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1; 6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1; 0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2; 1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged because of sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.

<p>Despite worldwide interest in a number of applications involving cryogenic fluids that are crucial to future space exploration, there is presently a lack of a large, reliable cryogenic pool boiling critical heat flux (CHF) database that can be used for assessment of accuracy of available predictive tools - model and correlations – or development of new tools. This shortcoming is a primary motivation for the present study, prompting compilation of a new consolidated cryogenic pool boiling CHF database from world literature. The database is used to assess accuracy of previous models and correlations, which are segregated according to ability to predict key operating parameters, such as pressure, surface orientation, and subcooling. A new correlation is constructed which shows very good predictive accuracy, evidenced by a mean absolute error of 16.95%, based on Earth gravity data which comprise a large fraction of the consolidated database. Using a limited subset of datapoints for three cryogens and a reduced gravity range of 0 to 0.7466, the new correlation is further modified with a reduced gravity multiplier to tackle reduced gravity conditions. The modified correlation has a mean absolute error of 17.47%, slightly higher than for Earth gravity alone. Overall, the new correlations are proven far more accurate than all prior models and correlations and therefore constitute new powerful tools for design of cryogenic space systems. It is shown CHF is very sensitive to pressure, increasing with increasing pressure up to maximum before decreasing appreciably toward critical pressure. CHF is also shown to be strongly influenced by surface orientation, being highest for horizontal surfaces and decreasing monotonically with increasing orientation angle, and increasing fairly linearly with increased subcooling.</p><p>Additionally, CHF models and correlations are assessed using amassed quenching CHF data that showed overpredictions of data. A new correlation is formulated which includes the effects of surface material and heater thickness to achieve high predictive accuracy for complied quenching CHF database. The new correlation has a mean absolute error and root mean square error of 10.79% and 16.12%, respectively, based on a compiled database. Analysis of complied quenching data showed that CHF is sensitive to the surface material, increasing with increasing thermal conductivity but, the influence of surface material becomes weak with increasing thermal conductivity. CHF is also strongly influenced by heater thickness, increasing with increased heater thickness till it reaches the asymptotic thickness. </p>

<p>Heat sinks have the capability of increasing operating heat flux limits for improved thermal management in the immersion cooling of electronics using dielectric fluids. However, even for arrays of simple, straight fins, the generation of vapor between and along fins during pool boiling lead to performance effects that are not well understood. Further investigation of the heat-flux-dependent variation of boiling modes that can manifest along the fin height is required. Although methods for the prediction of fin boiling heat transfer exist that incorporate a variable heat transfer coefficient determined from a flat surface, they have been developed and assessed for single, isolated fins under the assumption that the sides of the fin at any location behave like that of a flat surface. As a result, when applied to fin arrays, these methods may not always be accurate for the full range of heat flux operation along boiling curve up to the critical heat flux, due to the fins interfering with each other when arranged in arrays of differing spacing and height. To establish when the fins in an array can be described as isolated and having the flat surface boiling behavior, pool boiling experiments are performed using copper heat sinks in two fluids with vastly different properties: HFE-7100 and water. The spacing and height of the longitudinal fins are varied across a range from much larger to less than half of the scale of the capillary length scale of both fluids, <em>L</em>_<em>b</em>. High-speed visualizations enable the identification of different boiling regimes to identify correspondence between flow observations and the boiling performance, such as when there is bubble confinement from fin interference. Trends in the pool boiling data are also compared, noting changes in superheat at various heat fluxes to establish when fin height or spacing affects boiling behavior. The experimental boiling performance is compared to predictions developed assuming isolated fins so as to identify the spacings and heights for which the fin arrays follow this behavior. Overall, the data from both fluids strongly support a hypothesis that <em>L</em>_<em>b</em> is the key length scale. Heat transfer from fin array heat sinks with heights and spacings above <em>L</em>_<em>b</em> are shown to be accurately predicted in both fluids. However, spacings smaller than <em>L</em>_<em>b</em> lead to bubble confinement which affects the superheat, particularly at low heat fluxes, while heights shorter than <em>L</em>_<em>b</em> are unable to support multiple boiling regimes along the fin sidewall. This work identifies the capillary length as the key length scale at which confinement and height effects need to be considered for accurate predictions of immersion cooling applications.</p>

In nuclear power plants, during operation at high temperatures, there can occur a steam build-up around the fuel rods at an increasing rate. This inhibits the water’s effectivity in removing heat from the rods. When this occurs and reaches a critical point, the temperature in the fuel rods surges, which can lead to them being damaged. This is called critical heat flux (CHF). During operation, the reactor always keeps a safety margin to the point where CHF can occur. The margin to CHF is one of the factors that limits the nuclear power plant’s ability to produce electricity. With the help of the tests that Westinghouse runs, the safety margin to the CHF can be more accurately determined, so that the reactor can safely be closer to the critical point.Westinghouse uses rods that are heated with electricity instead of nuclear fuel. In Westinghouse’s test facility, a problem has been identified, where the uppermost part of the rod has a risk of breaking due to the high temperatures. The temperatures are so high due to the rod being unable to conduct the large amount of current (max. 300kW) through the grid plate, situated at the top. The rod has a tapered end and is hammered into the grid plate’s tapered holes during assembly. The rod’s end is hollow and is attached from above using screws.To find a solution, two theoretical concepts have been developed and an eventual change of rod material has been evaluated. The purpose of the concepts is to limit the risk of problems occurring due to heat increase during operation. Both concepts have reduced hole size and length, to avoid hollow areas around the warmest part of the construction. For concept evaluation, Pugh’s concept selection method has been used. The most appropriate concept has been evaluated to be a reduced hole width, and a deeper hole with a thread insert.

No description available.

<p>An experimental investigation of the physical feasibility of thermal creep failure of the Calandria Vessel under a severe accident load is presented in this thesis. Thermal creep failure is postulated to occur if film boiling is instigated in the Shield Tank Water surrounding the Calandria Vessel. The objective of this experimental study is to measure the Critical Heat Flux (CHF) for a representative geometry in environmental conditions similar to those existing in the CANDU Calandria Vessel and Shield Tank Water.<br />Two geometries of downwards facing surfaces are studied. The first is termed the ‘confined’ study in which bubble motion is demarcated to the heated surface. The second is termed the ‘unconfined’ study where individual bubbles are free to move along the heated surface and vent in any direction.<br />The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The level of confinement is adjusted by varying the inclination angle. Data has been obtained for Bond Numbers (Bo) 0, 1.5, 3, 3.6 and 11.8 with corresponding qCHF 596, 495, 295, 223, and 187 kW/m2, respectively. A correlation relating the CHF to level of confinement is stated. The CHF results are in good agreement with Theofanous et. al. (1994), as is the observation that a transition angle is observed in the correlation. The transition angle in this study is found to be ~5.5°. The obtained nucleate boiling curves are compared to Su et. al. (2008) data for similar Bo and excellent agreement is achieved in the medium to high heat flux regions.<br />The unconfined study consists of a downward facing plate in a pool of subcooled water. The obtained nucleate boiling curve is compared with the Stephan-Andelsalam correlation and agreement is not observed. There were visibly different trends in the convective heat transfer coefficient with a mean difference of 31%. The experimental data is compared to data obtained by Nishikawa et. al. (1984) and is found to be in acceptable agreement. The power requirement to instigate film boiling was not met, meaning that the CHF is greater than 1 MW/m2. Visual observations are made and an argument is based on the premise that the phenomenon of dryout for a downwards facing surface is similar to that of an upwards facing surface. The theory and current acceptance of CHF for an upwards facing surface is discussed—in particular Zuber’s “Hydrodynamic Limit” of 1.1 MW/m2, Dhir (1992) and recent experimental evidence from Theofanous et. al. (2002). These three studies were found to be in agreement with results presented here.<br />The experimental evidence presented herein supports the statement that thermal creep failure of the Calandria Vessel is physically unreasonable under analyzed severe accident loads.</p> / Master of Applied Science (MASc)

This thesis investigates pool boiling heat transfer from bare and graphene-coated NiCr wires in a saturated liquid of FC 770, a fluorocarbon fluid. Of particular interest was the effect of graphene-oxide platelets, dip-coated onto the heater surface, in enhancing the nucleate boiling heat transfer (BHT) rates and the critical heat flux (CHF) value. In the course of the pool boiling experiment, the primary focus was on the reduction mechanism of graphene oxide. The transition from hydrophilic to hydrophobic behavior of the graphene oxide-coated surface was captured, and the attendant effects on surface wettability, porosity and thermal activity were observed. A parametric sensitivity analysis of these surface factors was performed to understand the CHF and BHT enhancement mechanisms. In the presence of graphene-oxide coating, the data indicated an increase of 50% in CHF. As the experiment continued, a partial reduction of graphene oxide occurred, accompanied by (a) further enhancement in the CHF to 77% larger compared to the bare wire. It was shown that the reduction of graphene oxide progressively altered the porosity and thermal conductivity of the coating layer without changing the wettability of FC 770. Further enhancement in CHF was explained in terms of improved porosity and thermal activity that resulted from the partial reduction of graphene-oxide. An implication of these results is that a graphene-oxide coating is potentially a viable option for thermal management of high-power electronics by immersion cooling technology. / Master of Science

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.

<div>Nucleate boiling is an efficient heat transfer mechanism that enables the dissipation of high heat fluxes at low temperature differences. Heat transfer phenomena during nucleate boiling are closely linked to the two-phase flow morphology that evolves in time and based on the operating conditions. In particular, the critical heat flux, which is the upper limit for the nucleate boiling regime, can be triggered by hydrodynamic mechanisms resulting from interactions between the liquid and vapor phases. The aim of this thesis is to characterize the two-phase flow morphology evolution during nucleate boiling at high heat fluxes in two configurations: pool boiling, and confined and submerged two-phase jet impingement. The characterization is performed via non-invasive, high-speed optical based diagnostic tools. </div><div>Experimental characterization of liquid-vapor interfaces during boiling is often challenging because the rapidly evolving vapor structures are sensitive to invasive probes and multiple interfaces can occlude one another along a line of sight. In this thesis, a liquid-vapor interface reconstruction technique based on high-speed stereo imaging is developed. Images are filtered for feature enhancement and template matching is used for determining the correspondence of local features of the liquid-vapor interfaces between the two camera views. A sampling grid is overlaid on the reference image and windows centered at each sampled pixel are compared with windows centered along the epipolar line in the target image to obtain a correlation signal. To enhance the signatures of true matches, the correlation signals for each sampled pixel are averaged over a short time ensemble correlation. The three-dimensional coordinates of each matched pixel are determined via triangulation, which yields a set of points in the physical world representing the liquid-vapor interface. The developed liquid-vapor interface reconstruction technique is a high-speed, flexible and non-invasive alternative to the various existing methods for phase-distribution mapping. This technique also has the potential to be combined with other optical-based diagnostic tools, such as tomographic particle image velocimetry, to further understand the phase interactions.<br></div><div>The liquid-vapor interface reconstruction technique is used to characterize liquid-vapor interfaces above the heated surface during nucleate pool boiling, where the textured interface resulting from the boiling phenomena and flow interactions near the heated surface is particularly suited for reconstruction. Application of the reconstruction technique to pool boiling at high heat fluxes produces a unique quantitative characterization of the liquid-vapor interface morphology near heated surface. Analysis of temporal signals extracted from reconstructions indicate a clear transition in the nature of the vapor flow dynamics from a plume-like vapor flow to a release mode dominated by vapor burst events. Further investigation of the vapor burst events allows identification of a characteristic morphology of the vapor structures that form above the surface that is associated to the square shape of the heat source. Vapor flow morphology characterization during pool boiling at high heat fluxes can be used to inform vapor removal strategies that delay the occurrence of the critical heat flux during pool boiling.</div><div>As compared to pool boiling, nucleate boiling can be sustained up to significantly higher heat fluxes during two-phase jet impingement. The increases in critical heat flux are explained via hydrodynamic mechanisms that have been debated in the literature. The connection between two-phase flow morphology and the extension of nucleate boiling regime is investigated for a single subcooled jet of water that impinges on a circular heat source via high-speed visualization from two synchronized top and side views of the confinement gap. When boiling occurs under subcooled exit flow conditions and at moderate heat fluxes, the regular formation and collapse of vapor structures that bridge the heated surface and the orifice plate is observed, which causes significant oscillations in the pressure drop across. Under saturated exit flow conditions, the vapor agglomerates in the confinement gap into a bowl-like vapor structure that recurrently shrinks, due to vapor break-off at the edge of the orifice plate, and replenishes due to vapor generation. The optical visualizations from the top of the confinement gap provide a unique perspective and indicate that the liquid jet flows downwards through the vapor structure, impinges on the heated surface, and then flows underneath the vapor structure, as a fluid wall jet the keeps the heated surface wetted such that discrete bubbles continue to nucleate. At high heat fluxes, intense vapor generation causes the fluid wall jet to transition from a bubbly to a churn-like regime, and some liquid droplets are sheared off into the vapor structure. The origin of critical heat flux appears to result from a significant portion of the liquid in the wall jet being deflected off the surface, and the remaining liquid film on the surface drying out before reaching the edge of the heater.</div><div>The flow morphology characterizations presented in this dissertation further the understanding of flow and heat transfer phenomena during nucleate boiling. In the pool boiling configuration, the vapor release process was quantitatively described; during two-phase jet impingement, a possible mechanism for critical heat flux was identified. Opportunities for future work include the utilization of image processing techniques to extract quantitative measurements from two-phase jet impingement visualizations. Also, the developed liquid-vapor interface reconstruction technique can be applied to a boiling situation with a simpler liquid-vapor interface geometry, such as film boiling, to generate benchmark data for validation and development of numerical models.</div><div><br></div>