Investigation Of Critical Heat Flux Research Articles

Experimental Investigation of Critical Heat Flux of Nucleate Pool Boiling of Water and Nanofluid on Platinum Wire Under Hypergravity and Earth Gravity

Experimental Investigation of Critical Heat Flux of Nucleate Pool Boiling of Water and Nanofluid on Platinum Wire Under Hypergravity and Earth Gravity

Numerical investigation of critical heat flux during subcooled flow boiling in a vertical rectangular Mini-channel

The critical heat flux (CHF) of flow boiling in a mini-channel is vital for equipment safety, but the boiling crisis triggering mechanism has been proved unsatisfactory. In this study, the VOSET method is adopted to investigate CHF of subcooled flow boiling in a vertical rectangular mini-channel by considering the conjugated heat transfer and using a reasonable nucleation site density model. Hundred of bubbles are accurately captured, and the evolutions of flow pattern, dry patch, and wall superheating are reproduced. Based on those, the relationship between boiling crisis and dry patch is demonstrated. When the heat flux is lower, some middle-sized isolated bubbles adhere to the wall surface with a dry patch below. This dry patch can be rewetted quickly. Hence, the wall superheating increases slightly along the flow direction and is within control. With the increasing heat flux, the dry patch expands because some big bubbles merge into the elongated bubble, leading to the local heat transfer deterioration, but the local wall superheating rise is still limited due to heat conduction inside the solid wall and the rewetting effect. However, the continuous coalescence of elongated bubbles midstream and downstream causes the appearance of an almost permanent dry patch with a high local wall superheating of 183.04 K near the outlet of the mini-channel when the heat flux rises to 500 kW/m2, which has exceeded the CHF of 375 kW/m2 obtained in this study. This triggering mechanism of the flow boiling crisis is new and different from the traditional ones. Besides, the departure from nucleate boiling (DNB) is found at 325 kW/m2, beyond which the wall superheating increases rapidly with heat flux.

Experimental investigation of critical heat flux, pressure drop and heat transfer coefficient for the flow boiling in a horizontal lobed channel with R-123

ABSTRACT The current study investigates the distribution of local heat transfer coefficient, measurement of pressure loss and CHF in a lobed horizontal channel for flow boiling. Lobed channels with different lobed depths of 2.3, 2.0 and 1.25 mm, each having three different channel lengths (400, 600, and 1000 mm), are used for the experiments. The working substance used is R-123 with different mass flux. The heat flux range used is from 14 to 375 kW/m2. Pressure drop, critical heat flux, and local heat transfer coefficients of the lobed channel are compared with corresponding smooth horizontal tubes.

Investigation of Critical Heat Flux for Plutonium-Based Mixed Oxide Advanced Fuel Bundle Design

Abstract The critical heat flux (CHF) performance of an advanced plutonium-based mixed oxide (MOX) fuel for potential use in a pressure tube heavy water reactor (PT-HWR) has been studied experimentally at Canadian Nuclear Laboratories with an electrically heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile that is representative of the plutonium-based MOX fuel. The measurements of CHF caused by liquid-film dryout were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59 °C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and less even distribution across the fuel element rings, with a relatively higher power in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced or nonconventional fuel designs with radial power profiles deviating from that of natural uranium fuel.

Experimental Investigation of Critical Heat Flux on Different Surfaces at Low Pressure and Low Flow

Steady state flow boiling experiments were carried out on several heated tubes with outer diameter 9.14 mm at outlet pressures 120, 200 and 300 kPa, inlet temperatures 64, 78 and 91 °C and approximately 400, 500, 600 and 800 kg/(m2·s) mass flux entering the vertically aligned test annulus until critical heat flux (CHF) was reached. The tubes were made of Inconel 625 with a length of 400 mm. The Inconel tubes were tested in three different modifications as smooth, abraded with 150 grit sandpaper and bead blasted. Multiple experiments were repeated on the same specimen to investigate the effect of surface characteristic changes (i.e., wettability, roughness and oxide layer morphology) on the occurrence of CHF. Despite the changes in initial wettability, the CHF dependency was not clearly observed, however, the changes in roughness led to an increase in CHF. The total number of 115 experimental runs were collected and the results were also compared with other literature experimental data.

Experimental Investigation of Critical Heat Flux During Impact of a Droplet onto Hot Horizontal Surface

Experimental Investigation of Critical Heat Flux During Impact of a Droplet onto Hot Horizontal Surface

Experimental investigation of critical heat flux on SiO2 thin film deposited copper substrate in DI water at atmospheric pressure

Experimental investigation of critical heat flux on SiO2 thin film deposited copper substrate in DI water at atmospheric pressure

Numerical investigation of critical heat flux in subcooled flow boiling of nanofluids

Nanofluid flow boiling and critical heat flux (CHF) are interesting topics for many researchers and have industrial application for preventing potential damages. On this basis, numerous experimental studies have been performed on the determination of CHF; however, due to the weakness of the existing correlations for this purpose, an accurate numerical study on CHF is yet needed to be reported. In the present study, the subcooled flow boiling of nanofluid has been simulated to predict CHF. For estimating the nucleation site density (NSD), a correlation is used, into which the liquid properties, flow characteristics and surface properties contribute. The studies were performed at different concentrations of alumina nanoparticles. The simulation results indicated error percentages below 10% in CHF estimation, confirming the reliability of the results. No significant difference was observed between the CHF values calculated in the simulation studies considering constant thermal properties for the nanoparticles, as compared to those obtained with variable thermal properties for the nanoparticles. Taking the tube in a horizontal position, rather than the vertical position, increased the CHF value. However, the increase in CHF was negligible.

振動下における飽和流動沸騰の限界熱流束および沸騰気泡挙動に関する実験的検討

振動下における飽和流動沸騰の限界熱流束および沸騰気泡挙動に関する実験的検討

Experimental investigation of critical heat flux under boiling in subcooled swirling flow under conditions of one-sided heating

Results are given of an experimental investigation of critical heat flux in a turbulent swirling flow of subcooled water under one-sided heating. The parameters of flow are varied in the following ranges: the pressure of water at the inlet, 0.7 and 1.0 MPa; mass velocity ρw, 1100 to 9900 kg/(m2 s); flow swirl coefficient k = 0.19 and 0.37; the temperature of water at the inlet—20, 40, and 60°C; and the hydraulic diameter of the channel, 2.16 mm.

High heat flux flow boiling in silicon multi-microchannels – Part III: Saturated critical heat flux of R236fa and two-phase pressure drops

New experimental critical heat flux results for saturated boiling conditions have been obtained for R236fa flowing in a silicon multi-microchannel heat sink composed of 67 parallel channels, 223 μm wide, 680 μm high and with 80 μm thick fins separating the channels. The microchannel length was 20 mm. The footprint critical heat fluxes measured varied from 112 to 250 W/cm 2 and the wall critical heat fluxes from 21.9 to 52.2 W/cm 2 for mass velocities from 276 to 992 kg/m 2s. When increasing the mass velocity, the wall critical heat flux was observed to increase. The inlet saturation temperatures (20.31 ⩽ T sat,in ⩽ 34.27 °C) and the inlet subcoolings (0.4 ⩽ Δ T sub ⩽ 15.3 K) were found to have a negligible influence on the saturated CHF. The best methods for predicting the data were those of Wojtan et al. [L. Wojtan, R. Revellin, J. R. Thome, Investigation of critical heat flux in single, uniformly heated microchannels, Exp. Therm. Fluid Sci. 30 (2006) 765–774] and Revellin and Thome [R. Revellin, J. R. Thome, A theoretical model for the prediction of the critical heat flux in heated microchannels, Int. J. Heat Mass Transfer 50 (in press)]. They both predict the experimental CHF results with a mean absolute error of around 9%. Using the critical vapour quality, an annular-to-dryout transition is also proposed as a limit in a diabatic microscale flow pattern map. Pressure drop measurements were measured and analysed, showing that the homogeneous model could correctly predict the observed trends.

Experimental Investigation of CHF during Countercurrent Flow in Transient Boiling Systems

The objectives of this study are to investigate the effects of the power transient on critical heat flux (CHF), and to develop CHF and maximum heat flux (MHF) correlations in transient boiling systems with a countercurrent flow between the liquid and vapor flow. The test section consists of narrow, vertical, rectangular channels between parallel plates. Rate of change of wall temperature at the CHF point, (dTw/dt)CHF, and a nondimensional transient parameter, kappao =(L2/alphafTsat) (dTw/dt) CHF are introduced to evaluate the effects of power transients on CHF and MHF. Experimental ranges were 738.0-1,968.0 kg/m2s for mass flux, 7.0-17.5 C for inlet subcooling, and 3.0-8.0 mm for channel gap distance. The system pressure was kept constant at 1.0 atm. The experimental results show that CHF and MHF values increase with an increasing rate of change of wall temperature, and the increasing rate of CHF is higher in a wider channel gap distance. CHF and MHF increase linearly with increasing mass flux at the top of the test section, and the linearity decreases at the bottom of the test section. The effect of the inlet subcooling is significant at the top of the test section, therefore the test location closest to liquid inlet has a higher CHF value than the location immediately below it. It was also found that the wider the channel gap distance, the higher was the CHF value obtained for a given rate of change of wall temperature. New CHF and MHF correlations are developed for countercurrent flow in transient boiling systems. The CHF and MHF correlations are in good agreement with the experimental data within 25% error bands, respectively.

The Application of Computational Fluid Dynamics to Critical Heat Flux

The estimation of critical heat flux (CHF) in nuclear reactors is based largely on empirical relations that have a physical limiting conditions, a narrow range of applicability, and are inadequate for transient conditions. It is generally agreed that a more physically based approach is needed. Evidence is presented supporting the importance of boiling-induced fluid flow o the CHF process. Computational fluid dynamics (CFD) is used to model the microscale, transient dynamics of a vapor bubble growing in a subcooled liquid, resulting in qualitative reproduction of vapor blanket growth and CHF. The same CFD techniques are used to evaluate the macroscale thermal diffusion caused by spacers, resulting in qualitative reproduction of previous empirical results. This work forms the basis for a systematic investigation of CHF that could result in improved and less costly procedures for nuclear fuel design. This work is relevant for BWR and PWR reactors.

An Investigation of Critical Heat Flux and Surface Rewet in Flow Boiling Systems

高熱容量テスト部を加熱流路出口に設け、R113の遷移域を含む流動系沸騰曲線を求めた。測定範囲は質量速度357～1465kg/m2s、入口クオリティ0.06～0.62である。定常実験と非定常実験を行い、限界熱流束は非定常実験による極大熱流東とよく一致すること、リウエット開始過熱度は入口クオリティにより大きく変化し、低クオリティ域では120～140Kであるのに対し、高クオリティ域では60～80Kであることを示した。

An investigation of critical heat flux and surface rewet in flow boiling systems

A high thermal capacity copper test section placed at the end of a heating tube is used for investigating the critical heat flux and heat transfer characteristics during rewetting process for Freon-113 upward flow at mass velocities of 357, 627, 1035 and 1465 kg m −2 s −1. The critical heat flux coincides well with the maximum heat flux obtained by the transient cooling test. The wall superheat at the onset of rewetting changes greatly with the inlet quality, and ranges from 120 to 140°C in the subcooled and low quality region and from 60 to 80°C in the high quality region.