Nucleate Boiling Research Articles

Investigation of hybrid nanofluids effects on heat transfer characteristics in VVER-1000 nuclear reactor

In this study, the analysis of heat transfer characteristics of VVER-1000 nuclear reactor by using hybrid nanofluids as a coolants in subchannels in hot fuel assembly have been performed. Firstly, the criticality (keff) and fuel assembly power distributions of the VVER-1000 reactor by using Monte Carlo particle transport (MCNP) code to determine of the heat transfer characteristics with COBRA IV PC code have been examined. Secondly, the axial coolant temperature distributions and Departure from Nucleate Boiling Ratio (DNBR) values in hottest channel for Cu, CuO and TiO2 nanoparticles of the different volume fractions (0.05%, 0.01% and 0.1%) mixed with 0.1 vol % Al2O3 hybrid nanofluids have been analyzed. The obtained results by using hybrid nanofluids have been compared with the pure water results. The keff decreases with the increase of volumetric fractions of hybrid nanoparticles Cu, CuO and TiO2 mixed with 0.1% Al2O3 in the coolant. The maximum temperature has been reached by the püre water coolant at the end of the channel as 612.68 K whereas, the highest temperature change has been obtained in the case of 0.1% Al2O3 + 0.1% TiO2 as 617.35 K when evaluated among nanoparticle themselves. The results of analysis show that the heat transfer enhances with increasing hybrid nanoparticles volume fractions in the subchannel geometry. Also, the obtained minimum DNBR values using hybrid nanofluid haven't been exceed the safe operational limits; therefore the reactor operation would be safer.

Experimental analysis of pool boiling heat transfer on surfaces fabricated by electric discharge coating

The recent advances in the growth of miniature high power devices have led to the increased need for dissipation of large amount of heat generated in those devices. Compared to single phase cooling techniques, boiling provides a more effective two-phase heat dissipation method. Micro-structures in the boiling surface help to enable higher rate of heat transfer by providing air pockets for bubble generation. The present experimental study involves fabrication of micro-structured surfaces by employing Electric Discharge Coating (EDC) process. The surfaces were prepared by depositing a composite coating of titanium and copper microparticles on a copper surface at different current settings. The surfaces obtained were hydrophobic in nature which resulted in a decrease in the surface superheat at boiling incipient. Pool boiling experiments and analysis were performing on it by using deionized water as the boiling fluid. The Critical Heat Flux (CHF) also improved in the modified surfaces due to the globular structures formed with increased cavity sizes and porosity. The average Nucleate Boiling Heat Transfer Coefficient (NBHTC) increased to 206% and the surface superheat decreased by 42% for the best performing surface.

Thermal management of power electronics using nanofluids and nucleate boiling heat transfer technique

Nucleate boiling is an efficient solution for thermal management of power electronics where high-intensity heat flux needs to be transferred in a compact space. Both micro surfaces and nanofluids are used in literature to enhance the performance of nucleate pool boiling heat transfer systems; however, the effect of nanofluids over the microporous surface still needs to explore in several phase change heat transfer applications. The study’s objective is to examine the performance of nanofluid over microporous surfaces in Nucleate Boiling Heat transfer (NBHT) and performs a comparative analysis with plain and micro-porous surfaces. The study analyzed three ceramic (Alumina) nanofluids for nucleate pool boiling heat transfer (NBHT) over plain and microporous surfaces. The literature reported enhanced performance of nanofluids over the plane surface is verified in the study before proceeding to the detailed analysis over the microporous surface. Three different concentrations of 0.005%, 0.0005%, and 0.00005% were tested for the nanofluid. For plane surface, the critical heat flux (CHF) has been increased to 1035 [kW/m2] for Alumina from 783 [kW/m2] for deionized water, with the average increase in heat transfer coefficient (HTC) being 46.5%. The maximum CHF has been reported for micro-porous surfaces with a value of 1110 kW/m2. Out of the tested surfaces and fluids combination, the micro-porous surfaces with deionized water have been concluded as the most efficient surface with the highest heat transfer capacity for thermal management of high-power electronics and other high heat flux applications.

Steady state thermal hydraulic modelling of WWR-S tank-in-pool research reactor for the purpose of its power upgrading

Abstract This paper presents a newly developed steady state core thermal hydraulic model (named SSTH-RR10 model) for upgrading the Egyptian first Research Reactor (ETRR-1), from its original power of 2 MWth to a higher level of 10 MWth, by considering different types of nuclear fuels. The SSTH-RR10 model is capable to predict and calculate, by means of a developed computer program, all the steady state thermal hydraulic parameters for the defined core configuration for each fuel type at 10 MWth. Three different fuel types were investigated: the reference fuel EK-10 rod type, the MTR plate type, and the IRT-4M ducted type. For each fuel type, the distribution of central fuel, clad, and coolant temperatures for average and hot channels of the core were predicted in the axial direction. Power distributions and pressure gradients were predicted as well. Moreover, the program calculates the safety limits and margins against the critical phenomena encountered such as the Onset of Nucleate Boiling (ONB), Departure from Nucleate Boiling (DNB), and the Onset of Flow Instability (OFI). Results of the SSTH-RR10 program for benchmarks of powers of 2 and 10 MWth are verified by comparing it with the published results of the International Atomic Energy Agency (IAEA), and those published for other programs such as PARET code, and very good agreement is found. The safety margins against ONB and DNB were evaluated in which the minimum DNB ratio was found to be about 3.1, which gives a sufficient margin against the DNB. The present work gives confidence in the model results and applications.

Numerical Simulation Study on the Flow and Heat Transfer Characteristics of Subcooled N-Heptane Flow Boiling in a Vertical Pipe under External Radiation

In the top submerged lance (TSL) smelting process, flow boiling may occur in the lance’s inner pipe due to the heat coming from the furnace when liquid fuel is adopted. In the current study, a numerical simulation was carried out by coupling the Eulerian two-fluid model with the improved RPI wall boiling model to investigate the subcooled n-heptane flow boiling in the inner pipe. The effects of inlet velocity and pipe wall emissivity on two-phase flow and heat transfer are elucidated. The results show that, for pipes with inlet velocity ranging from 0.3 m·s−1 to 1.0 m·s−1, an increase in inlet velocity leads to a lower void fraction near the outlet, as well as a lower average velocity and a lower average temperature of each phase. Meanwhile, the Onset of Nucleate Boiling (ONB) position approaches to the outlet, and the total pressure drop of the entire pipe reduces when the inlet velocity increases. However, the opposite trends appear when increasing the pipe wall emissivity. The maximum wall temperature corresponding to the critical heat flux (CHF) point is slightly affected by inlet velocity but significantly affected by pipe wall emissivity. The non-equilibrium effect and the specific components of pressure drop are also further investigated.

Numerical simulation of single bubble growth in vertical rectangular narrow flow channels

The narrow rectangular channels of compact type reactors rely on boiling to generate bubbles to transfer a large amount of heat. On the one hand, the growth behavior of bubbles is closely related to the heat-mass conversion efficiency. On the other hand, the resulting flow instability and critical heat flow also affect operation safety. We set up a test bench with a visible flow channel to investigate the bubble growth in vertical rectangular narrow channels. The bubbles at the ONB (Onset of Nucleate Boiling) site were observed at 1 atm in subcooled flow boiling water. Two typical bubble growth processes were picked out to simulate numerically. A high-speed camera and thermocouples were used to record bubble contours and wall temperatures in the experiment. The heat flow was obtained based on assumptions and theoretical analysis. A numerical simulation of the whole process was conducted, where the bubble contours and temperature fields can be obtained. According to the simulation results of two examples, the microlayer evaporation at the bubble bottom accounts for about 30% of the total evaporation amount.

Numerical Simulation of Departure from Nucleate Boiling in Rod Bundles under High-Pressure Conditions

In subcooled boiling flows beyond a certain heat flux, heat transfer is hampered due to a phenomenon known as Departure from Nucleate Boiling (DNB). Conducting DNB experiments at one-to-one nuclear reactor operating conditions is highly challenging and expensive. Another alternative approach is to use Look-up table data. However, its applicability is limited due to its dependence on rod bundle correction factors. In the present investigation, a state-of-the-art Eulerian-Eulerian two-fluid model coupled with an extended heat flux partitioning model is used to predict DNB in tubes and rod bundles with square and hexagonal lattices (relevant to Pressurized Water Reactors). In this approach, bubble departure characteristics are modeled using semi-mechanistic models based on force balance analysis. The predicted DNB values are compared with experimental and Look-up table data and found out to be within 1.8% to 20%.

Experimental Study of Onset of Nucleate Boiling in Vertical Rectangular Channels with Different Flow Path Heights

A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm ∗ 60 mm and 2.8 mm ∗ 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that Δ T ONB (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, Δ T ONB in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.

Implementation and validation of CHF-models in the two-phase porous-media code TWOPORFLOW

The local deterioration of the heat transfer at certain elevation of the fuel rods of a water-cooled reactor may occur either in case of dry-out of the liquid film (in case of BWR) or due to Departure from Nucleate Boiling (DNB, in case of PWR). Under such local conditions, a sudden cladding temperature escalation may take place challenging the integrity of the cladding material i.e., of a safety barrier. For a reliable evaluation of the safety features of a nuclear reactor under nominal or accidental conditions, the use of validated numerical tools using relevant experimental data is mandatory. In this paper, the implementation of different correlations (Biasi, Bowring, and Groeneveld Look-Up Table) for the prediction of the Critical Heat Flux (CHF) in TWOPORFLOW as well as its validation is presented and discussed. Based on the performed investigations it can be stated that the implemented correlations predict the CHF-value and their location of appearance in acceptable agreement to the measured data.

Direct Numerical Simulation of Microlayer Formation and Evaporation Underneath a Growing Bubble Driven by the Local Temperature Gradient in Nucleate Boiling

Direct Numerical Simulation of Microlayer Formation and Evaporation Underneath a Growing Bubble Driven by the Local Temperature Gradient in Nucleate Boiling

Experiments to Understand Bubble Base Growth Mechanism(S) on Hydrophobic Surfaces Under the Influence of Bulk Flow Inertia During Nucleate Boiling Regime

Experiments to Understand Bubble Base Growth Mechanism(S) on Hydrophobic Surfaces Under the Influence of Bulk Flow Inertia During Nucleate Boiling Regime

Molecular Dynamics Study on Enhanced Nucleate Boiling Heat Transfer on Nanostructured Surfaces with Rectangular Cavities

Molecular Dynamics Study on Enhanced Nucleate Boiling Heat Transfer on Nanostructured Surfaces with Rectangular Cavities

Experimental Study on Onset of Nucleate Boiling in Vertical Annular Channel with a Short Heated Rod Under Low Pressure

Experimental Study on Onset of Nucleate Boiling in Vertical Annular Channel with a Short Heated Rod Under Low Pressure

Characterization of The Heat Transfer on Spray Quenching for Different Material Properties

A broad range of water spray applications as a means of two-phase cooling scheme has encouraged researches in the thermal management system to support safety and process efficiency in industries. In the application of above saturation temperature, the cooling process follows the boiling curve where the dissipated heat flux is figured out as a function of the wall temperature. Knowledge on constructing the boiling curve is an essential part in order to define the moving boundary, and boundary value problems occur in metal cooling process analysis involving heat transfer and phase change. The objective of the research was to characterize the boiling parameters on different materials in the regime of film boiling, transition boiling, and nucleate boiling as the basis for its boiling curve construction. To explain the influence of material properties, this work is featuring, firstly, the calculated vapor film thickness in film boiling regime by promoting self-developed analytical model of single droplet and, secondly, the calculated boiling width which indicates a strong combination of surface temperature and heat flux observed as the boiling phenomena. This is obtained by calculating the propagation of wetting front and 100 o C points. This experimental work employed a volumetric spray flux of 4.2, 10 and 13.7 kg/m 2 s to cool a hot metal samples of aluminum alloy AA6082 and nickel heated up to 560 °C. An infrared camera was used to record the temperature drop over time. Heat flux calculation follows the numerical procedure according to 1D energy balance model. Calculated vapor film thickness explains why the HTC tends to increase with the decrease of the surface temperature . Leidenfrost and Departure from Nucleate Boiling (DNB) temperatures are found to be inversely proportional to the heat penetration coefficient of the metal while maximum heat flux and boiling width increase with it.

Impact of core power density on economics of a small integral PWR

Increasing the core power density of nuclear reactors allows higher power generation from the same volume, reducing the cost of the plant. The lower power rating and higher degree of inherent safety for small modular reactors motivate exploring the highest achievable power density to improve their economics and overcome lack of economy of scale. Given the urgency to deploy carbon free energy sources in the near term, the current nuclear fuel supply must be adopted. Assuming standard 17x17 pressurized water reactor (PWR) fuel assemblies with enrichment below 5%, a parametric study on core power density of an integral PWR (iPWR) is performed. The reference core design is rated at 60 kW/L and composed of 77 reduced-height fuel assemblies. The reactivity control during normal operation is assumed to utilize soluble boron and fuel burnable poisons. The commercial STUDSVIK code package is used to design cores ranging from 20 kW/L to 170 kW/L. This range results in respective output powers between 50 and 420 MWe for an assumed 32% plant thermal efficiency. Core safety parameters, such as peaking factors, Minimum Departure from Nucleate Boiling Ratio and maximum fuel temperatures are evaluated for all designs. Simplified economic evaluation of lifetime-levelized unit costs for the fuel cycle, operation and maintenance and capital expenditures was performed as a function of power density. The relatively low core discharge burnup (32 MWD/kgU) leads to high lifetime-levelized fuel cycle unit cost for the reference core. However, increase in power density allows significant reduction in operation, maintenance and capital unit costs. Doubling power density resulted in 80% reduction in total levelized cost of electricity. The unique features of a forced circulation iPWR technology combined with compact steam generator technology could allow achieving high power densities (greater than 125 kW/L). Specific examples on relationship of cost and safety margins are also discussed.

Toward Mechanistic Wall Heat Flux Partitioning Model for Fully Developed Nucleate Boiling

Abstract Mechanistic models developed to predict partial nucleate boiling are not adequate for fully developed nucleate boiling due to differences in the prevailing heat transfer governing mechanisms. In place of the mechanistic model, several empirical correlations and semimechanistic models have been proposed over the years for the prediction of fully developed nucleate boiling as presented in this study but they are unsuitable for use in computational fluid dynamics (CFD) code. Recently, the simulation of fully developed nucleate boiling has become much more practical because of advancement in a computational method that involves the coupling of the interface capturing method (for slug bubbles) with the Eulerian multifluid model (for dispersed spherical bubbles). Nonetheless, there is a need for a mechanistic closure law for the fully developed nucleate boiling phenomenon that would complement this advancement in CFD. Toward this end, a mechanistic wall heat flux partitioning model for fully developed nucleate boiling is proposed in this study. This model is predicated on the hypothesis that a high heat flux nucleate boiling is distinguished by the existence of a liquid macrolayer between the heated wall and the slug or elongated bubbles, and that the macrolayer is interspersed with numerous high frequency nucleate small bubbles. With this hypothesis, the heat flux generated on the heated wall is partitioned into two parts: conduction heat transfer across the macrolayer liquid film thickness and evaporation heat flux of the microlayer of the nucleating small bubbles. The proposed model is validated against experimental data.

Electric Charge-Induced Active Control of Nucleate and Rapid Film Boiling at the Nanoscale: a Molecular Perspective

An effort has been made to understand the nucleate and the rapid film boiling phenomena under the influence of an electric field using a molecular viewpoint. The behavior of water molecules with a solid copper surface during the film boiling process in the presence of an electric field of different intensities has been studied. The molecular reasoning behind the suppression of the Leidenfrost phenomenon upon application of a uniform electric field along the heating substrate is established. Furthermore, the effect of surface characteristics with different wettabilities on film boiling in the presence of an electric field has also been studied. The electric field produces a finger-like water column besides thinning of the water film over a non-wetting surface. A similar phenomenon is also evident over a hydrophilic surface only after reaching a threshold value of electric charge intensity. Molecular simulations have explained the phenomenon of nucleate boiling of water on hydrophilic or non-wetting surfaces. Finally, the ability to control the bubble formation and suppression at a required location using an electric field has also been demonstrated. The water molecules near the surface experience dispersion at a lower electric field and an attraction force at a higher electric field, mimicking bubble nucleation and suppression, respectively.

Determination of Limiting Heat Flux for The Inception of Nucleate Boiling Regime for Crude Oils

Finding the limiting heat flux above which nucleate boiling starts and below which forced convective heat transfer exists is a crucial task for the accuracy of results from crude oil fouling tests. In this study, crude oils from two sources were tested at bulk temperatures of 100, 120 and 140 °C and different velocities. Heat transfer coefficient increased gradually with bulk temperature indicated lowering of the viscosity at high temperatures which promoted turbulence and enhanced heat transfer. The velocity effects were similar to that of bulk temperatures on maximum heat transfer coefficient while less heat flux was required to achieve the same surface temperature at lower velocities. Deshannavar and Ramasamy’s model to predict maximum heat flux was compared with experimental results and a poor estimation was observed for the crude oils tested.

Measurements of void fraction, liquid temperature and velocity under boiling two-phase flows using thermal-anemometry

Forced convective boiling is of great interest for several applications in the power and process industry, particularly in nuclear plants. Under certain nominal, incidental or accidental conditions, boiling crisis (considering Departure from Nucleate Boiling as well as Dry-Out) may lead to mechanical damage of the heated surface. An accurate prediction of the conditions leading to the occurrence of this phenomenon is then essential. It is believed that such an objective cannot be reached unless a good and accurate description of the associated two-phase flow is provided. In our work, we propose to use thermal anemometry for measuring the void fraction, the liquid temperature and liquid velocity for high pressure and high temperature Freon R134A boiling flow. Experiments have been conducted in a circular tube whose inner diameter and length are 19.2 mm and 3.5 m, respectively. The tube is heated by Joule effect. The sensor (hot-wire d5μm) has been operated with the Constant Current mode (CCA) and the multiple overheating method, which is classical for gas measurements but seems to be very innovative for liquid flows. This method, has been used to access simultaneously the liquid velocity and temperature. To consider for temperature effect on velocity calibration, a new non-dimensional representation of the calibration curve has been proposed. The frequency response of the probe has also been improved using a digital compensation method. The method has been first checked for single-phase flows and has shown that it was possible to get very accurate measurements of both mean and fluctuating liquid and temperature profiles. For boiling flows, a specific two-steps approach has been developed to first measure the void fraction where it is necessary to set a high overheat ratio leading to boiling on the wire surface and secondly to measure the liquid temperature and velocity for the case where boiling on the wire surface is not acceptable due to the multiple overheating method. An innovative method using probability density functions has been adapted from the pioneering work of Delhaye (1969).Some tests have been conducted for boiling flows and the experimental results have been compared to previous ones using thermocouples and optical probes for respectively the liquid temperature and the void fraction. The experimental uncertainties have been carefully analyzed and they are estimated to be close to 0.5 °C for the liquid temperature, ±2% for the void fraction (absolute uncertainty) and ± 5% for the liquid velocity (relative velocity). Those data aim to be used for NEPTUNE_CFD code validation.

Subchannel analysis of thermal-hydraulic performance in rod bundle with spacer grid considering anisotropic turbulent mixing

Combined with the relationship between the time averaged parameters of flow field and the inter-subchannel transverse turbulent mixing velocity, a method for solving the mixing velocity between adjacent subchannels is presented in this paper. Then an anisotropic mixing model, which is based on subchannel-wise inhomogeneous turbulent mixing, is presented to reflect the three-dimensional mixing effect in rod bundle with spacer grid. By comparing the anisotropic mixing model with isotropic mixing model, it is found that the anisotropic mixing model agrees well with the experimental data from literature. In addition, based on the anisotropic mixing model, analysis on the subchannels with different spacer grids is also performed, including configurations with or without the mixing vanes and different arrangements of mixing vanes. It is found that the forced mixing effect between subchannels is mainly caused by the mixing vanes of the spacer grid. Besides, the continuous change of the flow and temperature fields is also caused by mixing vanes, which improves the Departure from Nucleate Boiling Ratio (DNBR) of the rod bundle channel. Also, the effect of different mixing vane arrangements on flow and heat transfer in the rod bundle is analyzed in detail.