Diameter effect on critical heat flux

The critical heat flux look-up table (CHF LUT) is widely used to predict CHF for various applications, including design and safety analysis of nuclear reactors. Using the CHF LUT for round tubes having inside diameters different from the reference 8 mm involves conversion of CHF to 8 mm. Different authors [Becker, K.M., 1965. An Analytical and Experimental Study of Burnout Conditions in Vertical Round Ducts, Aktiebolaget Atomenergie Report AE 177, Sweden; Boltenko, E.A., et al., 1989. Effect of tube diameter on CHF at various two phase flow regimes, Report IPE-1989; Biasi, L., Clerici, G.C., Garriba, S., Sala, R., Tozzi, A., 1967. Studies on Burnout, Part 3, Energia Nucleare, vol. 14, pp. 530-536; Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. AECL-UO critical heat flux look-up table. Heat Transfer Eng., 7, 46-62; Groeneveld et al., 1996; Hall, D.D., Mudawar, I., 2000. Critical heat flux for water flow in tubes - II subcooled CHF correlations. Int. J. Heat Mass Transfer, 43, 2605-2640; Wong, W.C., 1996. Effect of tube diameter on critical heat flux, MaSC dissertation, Ottawa Carleton Institute for Mechanical and Aeronautical Engineering, University of Ottawa] have proposed several types of correlations or factors to describe the diameter effect on CHF. The present work describes the derivation of new diameter correction factor and compares it with several existing prediction methods.     

Experimental study of the effect of non-circular flow geometry on the critical heat flux

An experimental study of the effect of flow geometry (circular, rectangular, triangular, and dumb-bell shaped) on the critical heat flux (CHF) was performed using R-134a as a coolant. The CHF is affected by the following geometric parameters: hydraulic-equivalent diameter, heated length, gap size, channel shape, and curvature. It may also be affected by the thermal conductivity of the wall material and wall thickness. The effect of flow geometry on CHF is influenced by flow parameters. The effect of these parameters on CHF was examined, and recommendations for predicting the CHF in non-circular geometries have been made.     

Numerical study on effect of gap width of narrow rectangular channel on critical heat flux enhancement

According to the flow passage characteristic of narrow rectangular channel and liquid film dry-out mechanics of annular flow critical heat flux (CHF), an annular flow CHF analytical model for narrow rectangular channel has been achieved. This model may be used to predict the CHF behavior of boiling two-phase flow in narrow rectangular channel with gap width of not being less than 0.0005 m (the equivalent diameter of this channel is 0.001 m). Through analyzing and calculating, when the inlet dimensionless gap width of narrow rectangular channel is within 30-85, the enhancement of CHF in channel is obvious. At the same time, according to the characteristic of two-phase flow, the new determinant laws of CHF in boiling two-phase flow system have been derived. Through analyzing and calculating, it is substantial that this determinant laws is appropriate. The best dimensionless gap width of heat flux enhancement has been achieved to be 45-75.     

A comparative study of correlations of critical heat flux of pool boiling

Nucleate pool boiling is desirable for many engineering systems. One challenge task for designing a system with nucleate pool boiling is to estimate the critical heat flux (CHF), which needs an accurate pool boiling CHF correlation. A few evaluations of pool boiling CHF correlations were reported, which used limited experimental data or covered limited correlations, resulting in inconsistent results. Therefore, it is difficult to determine which one is more appropriate for a given application. In this paper, a database containing 600 data points of pool boiling CHF of 12 pure liquids on plain surfaces having orientation angles of 0°?180° is compiled from 40 published papers. The reduced pressure is from 0.0001 to 0.98, and the 13 fluids are water, helium, nitrogen, hydrogen, R113, FC-72, FC-87, HFE-7100, ethanol, benzene, hexane, pentane, and methanol. With the database, 21 pool boiling CHF correlations are assessed. The most accurate one has a mean absolute deviation of 27.1%, indicating a need for developing more accurate correlations for engineering applications. Besides, the factors affecting the accuracy of correlations are analyzed and some valuable conclusions are obtained. The work lays a valuable foundation for the further study of pool boiling CHF correlations and provides a guide for choosing proper correlations for given applications. Several topics worthy of attention for future studies are suggested.     

On the definition of critical heat flux margin

This paper reviews the current definition of critical heat flux (CHF) margins and discusses their differences.     

Prediction of critical heat flux using ANFIS

The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by Vaziri et al. (2007) to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.     

Experimental research progress on critical heat flux of Chinese PWR

One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of Φ10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR.     

Experimental study on critical heat flux on a downward-facing stainless steel disk in confined space

Critical heat flux (CHF) is experimentally studied on a relatively large downward-facing surface with a heated stainless steel disk diameter of D = 300 mm in confined space at atmospheric pressure using water as the working fluid. The bulk working fluid is subcooled. The gap size s can be adjusted to 0.9, 2.2, 2.6, 3.0, 3.2, 5.0, 7.0, 10.0, 13.0, 15.6, 19.5, 25.0, 36.0, 51.0 and 77 mm. We found that the average CHF under the present condition is approximately 0.25 MW/m² which is only about 23% of which occurs on an upward-facing surface without confined space in pool boiling. The CHF increases with the increase of the gap size when the gap size is smaller than 7 mm and it is a function of Bond and Jakob numbers when the gap size is larger than 7 mm.     

A compilation of rod array critical heat flux data sources and information

The geometric characteristics of rod array test sections employed in critical heat flux (CHF) tests with water coolant, and the ranges of the operating parameters for the tests, are presented for 126 test sections. The corresponding 4277 CHF data points have been stored on a magnetic tape for ease of reference and analysis. A versatile computer program associated with the data library has been used to determine the distributions of the data with respect to geometric and operating parameters. The dependence of CHF on operating parameters and the importance of subchannel conditions are shown through the use of some of the data. Tables are given for CHF data with a Freon coolant, for CHF data from test sections which only simulate a rod array, and for CHF data for transient situations.     

底部封闭矩形通道临界热流密度的理论研究

分析研究了在底部封闭矩形通道内逆流汽液两相流条件下的临界热流密度的发生机理。研究表明,临界热流密度与流入矩形通道内的最大下降液体流量相对应,并且临界热流密度可通过求解动量方程、包络线和能量方程得到。通过与日本数土幸夫建立的模型、经验关联式和实验数据比较,该模型可在精度±30 %范围内预测底部封闭矩形通道条件下的临界热流密度。     

Anomalies and other concerns related to the critical heat flux

This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following:

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Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon.

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Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X).

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Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve.

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Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope.

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Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical.

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CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions.

The mechanisms responsible for these trends and the implications for bundle geometries are discussed.Concerns regarding the reported uncertainty of predicted CHF values and the range of application of CHF prediction methods are also discussed.     

Development of critical heat flux correlation for In-vessel retention

ABSTRACT

In-vessel retention (IVR) is a strategy for severe accident management in which the lower head of the reactor vessel is submerged in a water-flooded reactor cavity. Critical heat flux (CHF) data for IVR are important for estimating cooling capacity of the reactor vessel. The existing CHF data for IVR which were obtained for the specific geometries and thermal-hydraulic conditions of actual plants are difficult to be applied to plants with other specifications. Hence, the purpose of this study is to develop CHF correlations applicable to various pressurized water reactor plants in a wide range of thermal outputs based on newly obtained CHF data. A rectangular test section with a cross-section of 150 mm × 150 mm and length of 600 mm was used for simulating a cooling channel. The thermal-hydraulic conditions expected in actual plants were studied, and the results were used in the experiment. The effects of parameters such as pressure, mass flux, thermodynamic quality, and angle on CHF were investigated . Based on these results, we developed a CHF correlation formula that can be applied to a wider range than previously, up to a maximum heat flux of 3000 kW/m², and that predicts CHF with an error of ± 10%.     

Prediction of critical heat flux in vertical pipe flow

A previously developed semi-empirical model for adiabatic two-phase annular flow is extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients.     

Parametric trends analysis of the critical heat flux based on artificial neural networks

Parametric trends of the critical heat flux (CHF) are analyzed by applying artificial neural networks (ANNs) to a CHF data base for upward flow of water in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. Katto's and Groeneveld et al. dimensionless parameters are used to train the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS errors of 8.9%, 13.1% and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local conditions hypothesis, respectively. The parametric trends of the CHF obtained from those trained ANNs show a general agreement with previous understanding. In addition, this study provides more comprehensive information and indicates interesting points for the effects of the tube diameter, the heated length, and the mass flux. It is expected that better understanding of the parametric trends is feasible with an extended data base.     

On the velocity profile of vapour bubbles at critical heat flux

A theoretical model is put forward to describe the flow velocity of the vapour phase near a heated wall when departure from nucleate boiling (DNB) is approached. The model refers to experimental observations which have been put into correlations and presented in an earlier article. It consists in a possible extrapolation of these previously proposed DNB-correlations towards the determination of the maximum critical heat flux in cooling water under forced convection. Comparison is made with the well-known Kutateladze/Zuber correlation, which usually applies to CHF in saturated pool boiling. It is suggested that the present hydrodynamic model could generate analogous results by using a different approach. Moreover, it is shown that the model could equally be applied to determine the DNB limit in both free and forced convection, constituting a link between them.     

The critical heat flux for boiling water in tubes

An analysis of the experimental results obtained by various authors on critical heat flux is carried out by using nondimensional criteria. Recommendations are given for the numerical methods of determining values of the critical heat flux in the case of a steam-water mixture, underheated to saturation in tubes and in ring-shaped and plane slotted channels.     

Effects of mechanical vibration on critical heat flux in vertical annulus tube

The effects of mechanical vibrations on critical heat flux (CHF) are examined in this study at atmospheric pressure in vertical annulus tube under electrically heated condition. Vibration of heating rod was increased as flow regime changed from subcooled region to bubbly region. CHF was increased by mechanical vibration up to 16.4%. Vibration amplitude was one of the effective parameters on CHF enhancement. It seems to come from turbulence increasing and increment of deposition of droplet from the liquid film by vibration. Vibration is an effective method for heat transfer enhancement as well as CHF enhancement.     

Occurrence of critical heat flux during blowdown with flow reversal

A small-scale experiment using Freon-11 at 54°C and 450 kPa in a transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The transients were initiated by simultaneous operation of three air-actuated values. The inner tube of the annulus was uniformly heated over its 0.61-m length while the outer transparent pyrex wall was unheated. Test section instrumentation included seven pressure taps, inlet and outlet capacitance void probes, inlet and outlet turbine flowmeters, inlet and outlet fluid thermocouples, and 22 wall thermocouples. High-speed motion pictures were taken of the lower end of the test section where the intial CHF occurred. From these high-speed pictures, the flow reversal was observed to occur between 60 and 80 ms followed by a rapid thermal excursion at about 400 ms in the lower regions of the test section. Consequently, this measured CHF occurred well after the flow had reversed and re-established itself in the downward direction. At about 300 ms an annuflar flow pattern appeared and was well-developed at 400 ms. Therefore, the early CHF measured were all associated with the transition from bubbly to annular flows. In some cases this early CHF was rewet and the heated section remained in a stable coolable state for a considerable length of time, and experienced a later CHF when the liquid was nearly depleted. This long-term dryout was a function of the liquid volume contained in the system, whereas the early CHF was independent of the system volume. In addition, the early CHF did not show any significant sign of propagation whereas the latter one was observed to propagate smoothly upward.     

A theoretically based critical heat flux prediction for rod bundles at PWR conditions

A theoretically based procedure developed for round tubes has been applied to the prediction of DNB heat fluxes in rod bundles at PWR conditions. State-of-the-art subchannel analysis procedures were used to determine local flows and enthalpies. Very good comparison between DNB predictions and experimental observations are found for rod bundles which both uniform and non-uniform axial heat fluxes.     

A new model for predicting the critical heat flux based on nucleation site density in downward-face boiling

ABSTRACT

Due to the important role critical heat flux (CHF) plays in the boiling field, it is of great significance to study CHF, especially the mechanism of CHF in the nucleate boiling. In this study, a new model to predict CHF both in pool boiling and flow boiling of downward-face was proposed and the relationship between CHF and nucleation site density (NSD) was studied. The model was based on the bubble interaction theory, which assumed that CHF happened due to the coalescing of the bubbles generated on the heating surface and prevented liquid to be supplied. The relationship between NSD and CHF was derived from previous observations in the experiments and simulations. To validate the relationship between NSD and CHF, several experiments with CHF and NSD were chosen and they all showed good agreement with our assumptions. Due to the rarity of experimental data on NSD and CHF, the numerical method was also used to validate. The results also showed an inverse relationship between CHF and NSD.