Critical Heat Flux Correlation Research Articles

Prediction of Critical Heat Flux in Tube Bundles With Crossflow

Abstract Evaporators and boilers with flow across horizontal tube bundles are widely used in industries including refrigeration and chemical processing. As yet there is a lack of a well-verified correlation for predicting local or average CHF (critical heat flux) in such bundles. The available data and correlations are reviewed. The available methods for predicting bundle mean CHF were found to be too conservative. More realistic values of bundle average CHF are proposed based on published test data. The author's well-verified correlations for CHF on a single tube with the subcooled and saturated flow at zero quality across it are compared with test data for tubes in tube bundles together with other correlations. The author's correlations agree well with all data while the others fail for data other than their own. The results of the comparison are presented and discussed. The needs for further research are identified.

The effect of solid nitrogen on nucleate boiling heat transfer in a slush nitrogen pool

The nucleate boiling heat transfer in a slush nitrogen pool has been studied across a heat-flux range from 5 to 400 kW/m2 and a system pressure range from 12.5 kPa (triple point pressure) to 101 kPa (atmospheric pressure). The effect of solid nitrogen on the heat transfer process was found to vary with the heat flux and the pressure. Experimental results reveal that the nucleate boiling heat transfer coefficient (HTC) of slush nitrogen is higher than that of subcooled liquid nitrogen at the low heat flux, which can be attributed to the fact that the solid nitrogen particles (with their latent heat of melting) absorb heat in the thermal boundary layer and form a sharper temperature gradient near the boiling surface. However, at the high heat flux, the HTC of slush nitrogen is lower than that of subcooled liquid nitrogen, which can be attributed to the invasion and inhibition occurring between the solid nitrogen particles and the bubbles during boiling, leading to the accumulation of vapor phase with low thermal conductivity on the boiling surface, thereby the heat transfer will be suppressed. As a result, a dual effect for the nucleate boiling heat transfer in slush nitrogen can be observed. Furthermore, the critical heat flux (CHF) of slush nitrogen is obtained and compared with the CHF of subcooled liquid nitrogen. According to the CHF model, the bubble period is the dominant parameter in determining the CHF. A calibrated CHF correlation for slush nitrogen is developed based on the hypothesis that the bubble period is prolonged as the effective viscosity of slush nitrogen increases with the solid volume fraction. The CHF values of slush nitrogen with the solid volume fraction of 10%, 20% and 30% are 0.96, 0.91 and 0.84 times those of subcooled liquid nitrogen, respectively, at the pressures higher than 50 kPa. Consequently, the solid volume fraction of slush nitrogen and the degree of subcooling of subcooled liquid nitrogen are found to have opposite effects on CHF.

Prediction of Critical Heat Flux during Downflow in Fully Heated Vertical Channels

Boiling with downflow in vertical channels is involved in many applications such as boilers, nuclear reactors, chemical processing, etc. Accurate prediction of CHF (Critical Heat Flux) is important to ensure their safe design. While numerous experimental studies have been done on CHF during upflow and reliable methods for predicting it have been developed, there have been only a few experimental studies on CHF during downflow. Some researchers have reported no difference in CHF between up- and downflow, while some have reported that CHF in downflow is lower or higher than that in upflow. Only a few correlations have been published that are stated to be applicable to CHF during downflow. No comprehensive comparison of correlations with test data has been published. In the present research, literature on CHF during downflow in fully heated channels was reviewed. A database for CHF in downflow was compiled. The data included round tubes and rectangular channels, hydraulic diameters 2.4 mm to 15.9 mm, reduced pressure 0.0045 to 0.6251, flow rates from 15 to 21,761 kg/m2s, and several fluids with diverse properties (water, nitrogen, refrigerants). This database was compared to a number of correlations for upflow and downflow CHF. The results of this comparison are presented and discussed. Design recommendations are provided.

Dependence of critical heat flux in vertical flow systems on dimensional and dimensionless parameters using machine learning

The critical heat flux (CHF) associated with the departure from nucleate boiling (DNB) determines the design and safety aspects of two-phase flow boiling systems. Despite the availability of several predictive tools, within the thermal engineering community, the pursuit of an accurate and robust CHF model remains a significant challenge. In this unique study, we extracted a substantial database from literature to develop machine-learning (ML) models to predict CHF for vertical flows commonly employed in the process industry. The extensive database encompasses 15,006 experimental data points gathered from diverse sources covering wide range of geometric and operational ranges of D, L, P, G, h, and x. This database is then employed to develop five ML – based models, namely Artificial Neural Networks (ANN), K-Nearest Neighbors (KNN), Adaptive Boosting (AdaBoost), Extreme Gradient Boosting (XGBoost), and Random Forests (RF). The selection of optimal input features is conducted by exploring various combinations of both dimensional and dimensionless parameters to identify the most influencing inputs for CHF prediction. The models incorporating features D, L, P, G, h, x, L/D, ρr, We, and BO achieve CHF predictions with Mean Absolute Errors (MAEs) of 8.85 % for the ANN model and 10.39 % for the XGBoost model. The optimized ML models outperformed even the highly reliable CHF correlations and lookup tables. The feature importance analysis of input parameters highlights strong dependence of CHF on dimensionless numbers such as Weber number and Boiling number.

Predicting Critical Heat Flux in Rectangular Channels: Which Correlation to Trust?

The PALLAS-reactor is an advanced nuclear reactor dedicated to the production of medical isotopes and (medical) nuclear technological research currently under construction in The Netherlands RELAP5 is the primary tool used for modeling the accident scenarios relevant for licensing the PALLAS-reactor. The Groeneveld Look-Up Tables (LUT) are the main Critical Heat Flux (CHF) model of RELAP5. These have been extensively validated against experiments with round tubes. However, literature suggests that the LUT over-predict CHF inside the narrow, rectangular fuel channels of research reactors such as the PALLAS-reactor. As literature abounds with CHF correlations of limited applicability and low degrees of validation and confidence, correlations that can accurately predict CHF in rectangular channels are needed. The goal of this work is twofold. First, it aims at selecting a number of promising CHF correlations and to assess these against experiments representative of the PALLAS-reactor design. Secondly, this work ascertains whether the LUT are indeed optimistic with respect to CHF in rectangular channels. To this purpose, a literature review has short-listed the Sudo-Kaminaga (S-K), the Kim-modified S-K, and the Hall–Mudawar (H-M) correlations. These correlations are implemented in the SPECTRA code developed at NRG and assessed against two experiments in rectangular channels. The Kim-modified S-K correlation fits the experiments better than the original S-K scheme for high mass fluxes, while predicting higher CHF values at low flow rates. Interestingly, the H-M correlation predicts the experiments accurately, although the correlation was developed from round tubes data. At high mass fluxes, the LUT predictions converge towards the experiments, thereby suggesting that the effect of the channel geometry becomes less important at high flow rates. At low mass fluxes, on the other hand, the LUT markedly fail to match the experiments. This work contributes to accurately estimating CHF in narrow rectangular channels, which is crucial for licensing research reactors such as the PALLAS-reactor, thus assisting in the betterment of many people’s lives.

Numerical investigation on pool boiling heat transfer with trapezoidal/inverted trapezoidal micro-pillars using LBM

Pool boiling with high-efficient heat transfer process is crucial to chemical engineering, nuclear engineering, electronic cooling, and other energy conversion applications. Surface modification technology has attracted more attention in recent years as a simple and reliable boiling enhancement technique. Most of the studies using surface modification technology have focused on the enhancement effect of rectangular micro-structures, only a few researchers have adequately investigated the enhancement effects on pool boiling heat transfer and mechanisms of other complex micro-structures (like circle, trapezoid, or inverted trapezoid). In the present study, the pool boiling processes on surfaces with trapezoidal/inverted trapezoidal micro-pillars were numerically investigated using lattice Boltzmann method (LBM). The bubble dynamics and heat transfer characteristics during the pool boiling processes on surfaces with different pillar bottom angles (30° ≤ θ ≤ 135°) were compared under different wall-superheat conditions. The orthogonal tests were adopted on micro-pillars to optimize the geometric parameters, including the bottom angle of the micro-pillar, the bottom width of the micro-pillar, the number of the micro-pillar, and the height of the micro-pillar. A new CHF correlation on the mixed wettability surface with trapezoidal/inverted trapezoidal micro-pillars was proposed. It can be found that the main mechanism of nucleate boiling heat transfer on the micro-pillars is the evaporation of the microlayer, and the pillar bottom angle reveals a remarkable effect on the pool boiling performance on the mixed wettability surfaces. The surface of θ = 135° has the minimum wall-superheat at the onset of nucleate boiling (ONB). Within a wide range of wall-superheat conditions, the heat transfer efficiency of pool boiling firstly increases with the increasing of the bottom angle of the micro-pillar and subsequently decreases. The surface of θ = 60° has the highest critical heat flux (CHF). The surface of θ = 90° has the highest heat transfer efficiency in the film boiling process. The results of orthogonal tests suggested that the bottom angle of the micro-pillar is the most significant influential factor, and the optimal geometrical combinations were proposed. The new CHF correlation has acceptable accuracy in predicting the CHF on surfaces with trapezoidal/inverted trapezoidal pillars, which provides guidance for the design and optimization of micro-structured surface modification in various pool boiling applications.

Helium pool boiling critical heat flux under various magnetically controlled gravity levels

The characterization of Critical Heat Flux (CHF) under magnetic compensation of gravity in liquid helium (LHe) is particularly important for the design of the future very high magnetic field superconducting magnets. Through a series of experimental sessions, the critical heat flux in LHe pool boiling has been determined under various gravity conditions ranging from 0.03g to 2.2g using magnetic forces. It is the largest range of gravity variation, from reduced to hyper-gravity, ever studied in a single facility so far. The study was conducted in a very high magnetic field up to 30 T. The magnetic compensation of gravity does not allow for perfect resulting gravity homogeneity. Nevertheless, for the first time, the magnetic field configuration is considered in the experimental results analysis. The relative resulting gravity inclination with respect to the heat exchange surface is also taken into account. The experimental results are compared with the well-known CHF correlation in LHe pool-boiling and show a very good behavior agreement. For our experimental gravity level, the CHF is ranging from 3000 W/m2 to 12000 W/m2.

Coupling dynamic thermal analysis and surface modification to enhance heat dissipation of R410A spray cooling for high-power electronics

With high critical heat flux (CHF) and heat transfer coefficient (HTC), spray cooling is considered as one of the most promising thermal management technologies for high-power electronic devices. To increase its cooling performance, a closed-loop experimental rig was constructed to study the effects of spray and system parameters on heat transfer enhancement by R410A. The best cooling performance can be achieved under optimal subcooling degree of 17 °C and nozzle diameter of 0.56 mm. When the compressor frequency reaches the upper limit of 90 Hz, maximum CHF and HTC on flat surface are 301.6 W/cm2 and 91.7 kW/(m2·K). To further improve CHF, mechanism of heat transfer enhancement by square pin finned surface was revealed in terms of droplet splashing. With fin width of 0.5 mm and height of 3 mm, CHF as high as 522.1 W/cm2 and peak HTC of 407.0 kW/(m2·K) are reached, while maintaining the cooling surface temperature lower than 55.6 °C. Compared to flat surface, CHF and HTC are enhanced by around 73.4% and 3.5 times, respectively. Based on the experimental data, CHF correlation applicable to pin finned surface was obtained with precision of ±12.4% by introducing fin height and width.

Temperature-dependent wicking dynamics and its effects on critical heat flux on micropillar structures in pool boiling heat transfer

Boiling heat transfer is a vital process that needs to be incorporated into next-generation cooling systems due to its high heat transfer performance. The critical heat flux (CHF), which defines boiling performance, needs to be enhanced to expand the operating limits of heat transfer applications. The crucial factor enhancing CHF is wicking performance, i.e., liquid-supply capacity during the boiling process. This study investigated the relationship between wicking performance and CHF enhancement using micropillar structures of various roughness (diameter: 4–20 um and gap: 10–40 um). To measure wicking performance near the boiling condition, we increased the surface temperature from 20 °C to 95 °C. At the highest temperature (95 °C), the D04 G10 sample (i.e., highly-roughness) showed an enhanced wicking coefficient (6.8 mm/s0.5), 48% higher than the value measured at room temperature. For the boiling tests, the D04 G10 sample (∼164 W/cm2) showed a notable CHF enhancement of 85%, compared to the plain surface (∼89 W/cm2), owing to the strong liquid flow induced by wicking, which delayed vapor film formation. Based on these results, we proposed a new CHF correlation that accounts for the wicking performance near the boiling condition, which is accurate within an error of 9% with experimental results.

Data consolidation, correlations assessment, and new correlation development for pool boiling critical heat flux specific to cryogenic fluids

A common and crucial process for cryogenic fluid management is the need to fill the cryogenic fluids into warmer storage tanks efficiently while minimizing the evaporative and entrainment losses. However, no good predictive tool that can estimate the heat transfer behavior accurately during this fill process is available in the literature. In this study, we start with developing a new generalized correlation set for predicting critical heat flux (CHF), an important parameter on the boiling curve. First, a new pool boiling test facility is built to generate CHF data using a flat plate geometry. Based on our assessment of prior pool boiling data and correlations, 30 CHF datapoints in three different orientations with respect to gravity are obtained and validated against the existing CHF data from published literatures. Second, our experimental data is then combined with worldwide data for cryogenic fluids to generate a consolidate database that includes 2205 datapoints for 6 cryogenic fluids for 3 boiling surfaces (flat plate, cylinder and sphere). Third, a comprehensive assessment of prior pool boiling CHF correlations is conducted and it is observed that no one correlation or correlation form can capture all important parametric effects important to CHF. Fourth, a new correlation model for flat plate is setup that gave a mean absolute error (MAE) of 15.96% based on 1259 datapoints. In addition, multipliers to the flat plate correlation form are used to predict CHF for curved surfaces and gave MAE of 22.38% based on 355 datapoints and MAE of 17.16% based on 177 datapoints for cylinders and spheres, respectively. Values of CHF based on quenching is observed to be smaller than CHF based on boiling for the same test conditions regardless of the boiling surface material. Finally, a parameter to account for the thickness & thermal properties of the quenching surface is generated and added to the original correlation form and gave an MAE of 14.05% based on 245 datapoints.

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). To further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

A mechanistic model of critical heat flux for pool boiling based on supply failure mechanisms depending on the contact angle

Experimental and theoretical analyses are conducted to develop a mechanistic model of critical heat flux (CHF) for application to pool boiling of fluids with contact angles between 0° and 180°. To accurately predict the CHF over the entire range of the contact angle, it is postulated that an unquenchable dry patch occurs when liquid fails to be supplied to the dry patch; A supply failure mechanism is classified into two different mechanisms depending on the wettability of the boiling surface. The first is an evaporation mechanism which is dominant for non-wetting surfaces. This type occurs when the liquid supply rate in the liquid film due to the capillary force becomes lower than the evaporation rate. The second mechanism is a levitation mechanism which is dominant for wetting surfaces. This type occurs when the surface temperature under the dry patch exceeds the Leidenfrost point. To check the validity of the postulate, experiments involving the pool boiling of water and FC-72 are conducted on a titanium surface, which is a non-wetting surface for water but a wetting surface for FC-72. In the experiments, the temperature and heat flux distributions on the boiling surface are measured using infrared thermography for a quantitative characterization of boiling at the CHF. Experimental analyses confirm the postulate by showing that the CHF occurs due to the evaporation mechanism for the pool boiling of water, whereas the levitation mechanism for the pool boiling of FC-72. It is also observed that a large area of liquid film exists for the pool boiling of water, whereas this liquid film hardly exists for the pool boiling of FC-72. This observation indicates that the change in the dominant mechanism is due to the dependence of the liquid film area on the wettability of the boiling surface. After validating the postulate, a CHF correlation is developed that reflects the change of the supply failure mechanism depending on the contact angle. The developed correlation shows a good match to the extensive sets of CHF data previously obtained by various investigators to within 20%.

Development of a critical heat flux correlation based on an annular film dryout mechanistic model under the annular flow conditions

ABSTRACT This study proposes a mechanistic model-based correlation for the critical heat flux to meet the practical requirements for the design and safety analysis of light water reactors. The correlation is based on an annular film dryout mechanistic model for a single tube, which needs numerical integration from the onset of annular flow to the dryout point, expressed as an explicit non-dimensional equation to ensure a wide applicable range of coolant conditions. The correlation predicts a critical heat flux for the annular flow region in the pressure range of 1–16 MPa, the mass velocity range of 500–3000 kg/m2s, and the thermal equilibrium quality range of 0.05–0.7. The prediction by the developed correlation agrees well with that of the existing mechanistic model without any numerical integration of the annular flow conditions.

Modelling and experimental validation of heat transfer behavior during trans-critical transients

Though designed to be operated in supercritical conditions, supercritical power systems could experience trans-critical transients in which the pressure transfers between supercritical and subcritical condition during startup, shutdown, or accident condition. In the present work the thermal–hydraulic code ATHLET-SC was upgraded to predict the heat transfer behavior during trans-critical procedures. Prediction model of dispersed droplet flow film boiling (DFFB) heat transfer correlation, critical heat flux (CHF) correlation, and the correlation of minimum film boiling temperature TMFB has been upgraded. It is demonstrated by the validation work that the prediction error is large in the inverted annular film boiling (IAFB) regime. However, steady-state IAFB heat transfer experiment at high-pressure condition is unavailable. Hence, a calibration approach to IAFB heat transfer correlation is developed based on trans-critical transient experiment. As validated, the change of the wall temperature during the trans-critical transient gets well predicted by the corrected Bromley correlation.

Critical Heat Flux Condition and Post-Critical Heat Flux Heat Transfer of Carbon Dioxide at High Reduced Pressures in a Microchannel

Abstract Flow boiling heat transfer around the critical heat flux (CHF) condition at high reduced pressures of carbon dioxide in a 296-μm hydraulic diameter microchannel was experimentally studied. The CHF conditions for developing flow and fully developed flow were measured and compared to established correlations. The post-CHF heat transfer coefficient was obtained for l/d of 3.2, 7.4, and 11.6 for inlet Reynolds numbers, based on the homogeneous two-phase flow model, ranging from 6622 to 32,248. The critical heat flux conditions seemed to peak around a reduced pressure of about 0.5 and gradually decreased with reduced pressure. However, the typical rapid increase in the surface temperature following the CHF condition decreased with increasing pressure, and the post-CHF heat transfer coefficient was appreciably high (up to about 50 kW/m2K) at high reduced pressures. The enhancement in the heat transfer coefficient and CHF condition near the inlet were quantified. The experimental results were compared to established CHF correlations and heat transfer coefficient correlations with some limited success. Thus, the Katto CHF correlation (Katto and Ohno, 1984, “An Improved Version of the Generalized Correlation of Critical Heat Flux for the Forced Convective Boiling in Uniformly Heated Vertical Tubes,” Int. J. Heat Mass Transfer, 27(9), pp. 1641–1648) and the Bishop correlation (Bishop et al., 1964, “Forced-Convection Heat Transfer to Water at Near-Critical Temperatures and Supercritical Pressures,” Westinghouse Electric Corp, Atomic Power Division, Pittsburgh, PA.) for the post-CHF heat transfer coefficient were adjusted to better predict the experimental results. Additionally, an enhancement factor was derived to predict the increase in the heat transfer coefficient in the developing region.

Critical heat flux correlations for tube and annulus geometries under inclination and rolling conditions

In floating nuclear reactors, critical heat flux (CHF) prediction is crucial because the CHF can be influenced under heaving, rolling, and inclination conditions, unlike in land-based reactors. Hence, the existing land-based empirical correlations of the CHF can be invalid under some conditions in marine reactors. Therefore, empirical correlations of the flow boiling CHF were developed in this study to predict the CHF under inclination and rolling motions in tubes and annuli. Available experimental CHF data (616 points) were collected including our previous experimental data (405 points) and their trends under marine conditions were analyzed. The CHF variation was expressed using two types of modified Froude numbers depending on the CHF regime—departure from nucleate boiling and dryout. These parameters were used in correlations for the CHF under inclination. CHF correlations under rolling conditions were developed. Rolling was found to attenuate the CHF variation through the inclination. This observation was reflected in the newly developed rolling CHF correlation. The proposed correlations adequately described the existing CHF experimental data under inclination and rolling within an uncertainty of 9.5%. The developed CHF correlations can be used for the safety analysis and reactor core thermal–hydraulic design of marine reactors.

Coolability of a corium pool in a debris bed – Critical heat flux (CHF) dependency with tilting angle, debris size and steam flowrate

In case of severe accident in a light water reactor, large debris beds may form inside the core. If, in some part of the debris bed, the temperature increases up to the fuel melting point, it will lead to the formation and expansion of a molten pool which becomes more and more difficult to cool down as it grows. Therefore, it is of primary interest to determine the maximum size of a molten pool surrounded by debris which may be stabilized under water either after reflooding the vessel (as in TMI-2) or in case of a debris bed formed in a flooded reactor pit. One of the key parameters for that issue is the maximum heat flux (CHF) that may be extracted from the pool by water flowing within the debris bed. To determine that CHF, a series of tests have been conducted at IRSN. A heated copper cylinder generates a heat flux at one of its ends which simulates the boundary of the pool and is placed in contact with a debris bed made of monodisperse steel balls. Temperature measurements along the cylinder allow to calculate the heat flux transmitted to the debris bed through the cylinder tip.Several experimental parameters have been investigated:- The tilting angle of the heated surface, to take into account the shape of the molten pool.- The diameter of the balls.- The upwards steam flowrate simulating the steam coming from the part of the core located below the molten pool.- The upwards liquid flowrate simulating the natural circulation of water along the debris bed.As it is usual for CHF experiments, a significant dispersion of the measurements was observed, due to unavoidable evolutions of the contact zone between the copper surface and the balls. However, the impact of that random dispersion could be reduced by performing a large number of tests. This allows to identify clear tendencies on the CHF for the 3 studied parameters, with a rather good degree of confidence.From those tendencies, it was possible to derive a general CHF correlation with the tilting angle and the steam flowrate. When the liquid velocity increases, the CHF increases in most cases, except for some cases with low velocity. However it is more difficult to find a clear correlation with the liquid velocity or flow rate.

Impact of inlet subcooling on flow boiling in microchannels

Inlet subcooling is of paramount importance for the flow boiling heat transfer in microchannels. This work carries out the experimental investigation of the impact of inlet subcooling on the deionized water flow boiling in silicon microchannels (H = 200 μm, W = 80 μm, L = 10 mm). Experiments are conducted with inlet subcooling degrees of 70, 40 and 20 K, mass fluxes of 446–963 kg/m2·s and base heat fluxes of 17.5–407.2 W/cm2. It is found that the boiling hysteresis with explosive boiling occurs in the boiling inception at a high subcooling degree of 70 K, and boiling hysteresis gradually gets gentle as inlet subcooling decreases, which has not been reported in the previous works. At a higher subcooling degree, more heat is required for the bubble nucleation but higher critical heat flux (CHF) can be achieved. Further, as inlet subcooling decreases, the flow reversal is well mitigated and thus the flow boiling becomes more stable. Based on the above observations, we put forward a new dimensionless number called the boiling utilization to evaluate the flow boiling characteristics at different inlet subcooling conditions. It is found that a larger degree of subcooling results in a smaller boiling utilization; and the copper-based microchannel has a higher boiling utilization than that of the silicon-based microchannel. A new CHF correlation considering the influence of inlet subcooling is proposed, which shows good agreement with experiments, and the mean absolute error between the experimental values and predicted values by the proposed correlation is 7.3 %.

Correlation for CHF during subcooled flow across a single cylinder

A correlation is presented for prediction of CHF (critical heat flux) during upward flow of subcooled liquids across single cylinders. The correlation was verified against all available published test data. It includes five fluids (water, R-12, R-113, isopropanol, FC-72), tube diameter 0.25 – 19.1 mm, reduced pressure 0.0046 – 0.3554, liquid velocities 0.01 – 6.8 m/s, CHF 0.14 – 28.7 MW/m2, and quality −0.001 to −0.34. The new correlation predicts 496 data points from 28 data sets from 12 sources with mean absolute deviation (MAD) of 16.4%. The same data were also compared to other published correlations; their MAD ranged from 47% to 319.0%.

Subchannel analysis of critical heat flux in radially Non-uniform heated rod assembly at low flow rate

The effect of radial power distribution and low flow rate on critical heat flux (CHF) was investigated. In this study, based on the CHF experimental results of Columbia University, the research validated the feasibility of the subchannel analysis code CTF to predict CHF. The results show that CTF code using the EPRI empirical CHF correlation can predict the CHF experimental results well. Due to radially non-uniform power distribution, the quality and void fraction of hot subchannels are higher than other subchannels. The influence of the lateral flow will cause the counterflow in low power region under the beginning of two-phase flow at low flow rate. Furthermore, the predicted average CHF of rod assembly increases with the increase of inlet mass flow rate and the decrease of inlet temperature. When CHF happens in the rod assembly, the local CHF is mainly due to the low mass flow rate and high quality.