Occurrence Of Critical Heat Flux Research Articles

Characteristics of Pool Boiling Bubble Dynamics in Bead Packed Porous Structures

Spherical glass and copper beads have been used to create bead packed porous structures for an investigation of two-phase heat transfer bubble dynamics under geometric constraints. The results demonstrated a variety of bubble dynamics characteristics under a range of heating conditions. The bubble generation, growth, and detachment during the nucleate pool boiling heat transfer have been filmed, the heating surface temperatures and heat flux were recorded, and theoretical models have been employed to study bubble dynamic characteristics. Computer simulation results were combined with experimental observations to clarify the details of the vapor bubble growth process and the liquid water replenishing the inside of the porous structures. This investigation has clearly shown, with both experimental and computer simulation evidence, that the millimeter scale bead packed porous structures could greatly influence pool boiling heat transfer by forcing a single bubble to depart at a smaller size, as compared with that in a plain surface situation at low heat flux situations, and could trigger the earlier occurrence of critical heat flux by trapping the vapor into interstitial space and forming a vapor column net at high heat flux situations. The results also proved data for further development of theoretical models of pool boiling heat transfer in bead packed porous structures.

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: • 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. • 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). • 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. • 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. • 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. • 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.

Dry-out CHF correlation for R134a flow boiling in a horizontal helically-coiled tube

An experimental study was carried out to investigate the R134a dry-out critical heat flux (CHF) characteristics in a horizontal helically-coiled tube. The test section was heated uniformly by DC high-power source, and its geometrical parameters are the outer diameter of 10 mm, inner diameter of 8.4 mm, coil diameter of 300 mm, helical pitch of 75 mm and valid heated length of 1.89 m. The experimental parameters are the outlet pressures of 0.30–0.95 MPa, mass fluxes of 60–500 kg m −2 s −1, inlet qualities of −0.36–0.35 and heat fluxes of 7.0 × 10 3–5.0 × 10 4 W m −2. A method based on Agilent BenchLink Data Logger Pro was developed to determine the occurrence of CHF with a total of 68 T-type thermocouples (0.2 mm) set along the tube for accurate temperature measurement. The characteristics of wall temperatures and the parametric effect on dry-out CHF showed that temperature would jump abruptly at the point of CHF, which usually started to form at the front and offside (270° and 90°) of the outlet cross-section. The CHF values decrease nearly linearly with increasing inlet qualities, while they decrease more acutely with increasing critical qualities, especially under larger mass flux conditions. The mass flux has a positive effect on CHF enhancement, but the pressure has negative one. A new dimensionless correlation was developed to estimate dry-out CHF of R134a flow boiling in horizontal helically-coiled tubes under current experimental conditions and compared to calculated results from Bowring and Shah correlations.

Results of acoustic monitoring of ITER divertor vertical target prototype

Acoustic monitoring is a method under development to indicate the occurrence of critical heat flux (CHF) events on plasma facing components exposed to high heat fluxes (HHFs) from plasma wall interaction, in order to be able to stop plasma operation before irremediable damages appear. It is a non-intrusive promising method thanks to the property of acoustic waves to propagate in channels and to the CHF acoustic precursory indicators which have been observed in several previous HHF experimental studies. This method is on a preliminary assessment stage and HHF experiment on relevant mock-up is an opportunity to collect and analyse data for improving the method efficiency. This paper deals with the post-processing of acoustic signals recorded on a European ITER divertor qualification prototype, on which an unexpected loss of flow accident occurred and caused a cooling channel rupture. The acoustic data have been analysed in the same way as for a CHF scenario, in order to seek precursory indicators of this kind of event.

CFD modelling of subcooled boiling—Concept, validation and application to fuel assembly design

The paper describes actual Computational Fluid Dynamics (CFD) approaches to subcooled boiling and investigates their capability to contribute to fuel assembly design. In a prototype version of the CFD code CFX a wall-boiling model is implemented based on a wall heat flux partition algorithm. It can be shown, that the wall boiling model is able to calculate the cross sectional averaged vapour volume fraction of vertical heated tubes tests with good agreement to published experimental data. The most sensitive parameters of the model are identified. Needs for more detailed experiments are established which are necessary to support further model development. The model is applied for investigation of the phenomena inside a hot channel of a fuel assembly. Here the essential phenomenon is the critical heat flux. Although subcooled boiling represents only a preliminary state towards the critical heat flux occurrence, essential parameters like swirl, cross flow between adjacent channels and concentration regions of bubbles can be determined. By calculating the temperature of the rod surface the critical regions can be identified which may later on lead to departure from nucleate boiling and possible damage of the fuel pin. The application of up-to-date CFD with a subcooled boiling model for the simulation of a hot channel enables the comparison and the evaluation of different geometrical designs of the spacer grids of a fuel rod bundle.

A hydrodynamic CHF model for downward facing boiling on a coated vessel

A hydrodynamic critical heat flux (CHF) model has been developed in this study for downward facing boiling on the outer surface of a hemispherical vessel with micro-porous layer coatings. The CHF model considers the vapor dynamics and the boiling-induced two-phase motions in three separate regions adjacent to the heating surface. These include a micro-porous coating layer, a liquid sublayer, and a two-phase boundary layer. Mathematical formulation of the CHF model has been made, taking into account the flow paths to an evaporation site within the porous coating layer, the behavior of the liquid sublayer, and the fluid motions of the two-phase boundary layer. It is found that the local CHF limit is reached when the liquid supply from the two-phase boundary layer through the liquid sublayer and the micro-porous coating to the heating surface is not sufficient to prevent local dryout of the heating surface. The dryout process consists of two sequential stages. In the initial stage, the vapor slug that forms by the downward facing boiling process moves toward the micro-porous coating layer owing to the depletion of the liquid sublayer. In the final stage, depletion of the capillary-assisted liquid supply occurs in the micro-porous coating layer, resulting in the dryout of the heating surface. This is followed by an abrupt increase in the local surface temperature, marking the occurrence of CHF. Comparison of the hydrodynamic CHF model with experimental data has been made and the results are found to be quite satisfactory.

Pressure drops for steam and water flow in heated tubes

Measurements of pressure drop for a steam and water flow inside several heated tubes were obtained at two different laboratories (Chalk River Laboratories and École Polytechnique). The test sections used to carry out the experiments were constructed with Inconel-600 tubes, and were cooled internally with an upward flow of water. A reduction in frictional pressure gradient with an increase in heat flux was observed for single-phase flows at the same cross-sectional-average conditions. After boiling was initiated, the two-phase friction multiplier increased with increasing heat flux in the bubbly-flow region, but decreased in the annular-flow region. As the heated surface approached dryout, the two-phase friction multiplier exhibited a maximum value and decreased with increasing thermodynamic quality, until dryout occurred. The decreasing trend was attributed to a reduction in the liquid-film thickness and the termination of liquid entrainment. Following the occurrence of critical heat flux, a significant drop in frictional pressure gradient was noted, caused by the change in the near-wall sublayer to a low-viscosity vapor blanket.

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.

An utilization of liquid sublayer dryout mechanism in predicting critical heat flux under low pressure and low velocity conditions in round tubes

From a theoretical assessment of extensive critical heat flux (CHF) data under low pressure and low velocity (LPLV) conditions, it was found out that lots of CHF data would not be well predicted by a normal annular film dryout (AFD) mechanism, although their flow patterns were identified as annular–mist flow. To predict these CHF data, a liquid sublayer dryout (LSD) mechanism has been newly utilized in developing the mechanistic CHF model based on each identified CHF mechanism. This mechanism postulates that the CHF occurrence is caused by dryout of the thin liquid sublayer resulting from the annular film separation or breaking down due to nucleate boiling in annular film or hydrodynamic fluctuation. In principle, this mechanism well supports the experimental evidence of residual film flow rate at the CHF location, which can not be explained by the AFD mechanism. For a comparative assessment of each mechanism, the CHF model based on the LSD mechanism is developed together with that based on the AFD mechanism. The validation of these models is performed on the 1406 CHF data points ranging over P=0.1–2 MPa, G=4–499 kg m −2 s −1, L/ D=4–402. This model validation shows that 1055 and 231 CHF data are predicted within ±30 error bound by the LSD mechanism and the AFD mechanism, respectively. However, some CHF data whose critical qualities are <0.4 or whose tube length-to-diameter ratios are <70 are considerably overestimated by the CHF model based on the LSD mechanism. These overestimations seem to be caused by an inadequate CHF mechanism classification and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities as well as a new bubble detachment model for LPLV conditions, are needed to improve the model accuracy.

Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one.

Dimensional analysis of critical heat flux in subcooled water flow under one-side heating conditions for fusion application

Critical heat flux (CHF) experiments have been carried out to design a high performance cooling device with pressurized subcooled water flow under one-side heating conditions for plasma facing components. Swirl, screw and hypervapotron tubes are the most efficient geometries to remove high incident heat fluxes. A dimensional analysis shows that CHF phenomenon can be featured by dimensionless groups that are the critical Boiling, the Eckert and the Reynolds numbers, the mass enthalpic quality and the ratio between liquid and vapor densities. CHF for uniformly heated smooth tubes, the maximum wall heat flux for smooth and for swirl tubes under one-side heating conditions are predicted with a reasonable accuracy by correlating these five dimensionless groups. An enhancement factor featuring the difference between one-side and uniform heating is also introduced. CHF occurrence under one-side heating is characterized by a normalized temperature, ratio between the maximal wall temperature and the onset of nucleate boiling temperature, which only depends on the pressure. According to finite element calculations performed for smooth and swirl tubes, temperature and heat flux are distributed along the inner wall of the cooling channel under one-side heating conditions. These distributions are characterized by two dimensionless numbers for each geometry: a peaking factor, which is the ratio between the maximal heat flux at the inner wall of the cooling channel and the incident heat flux, and a full width angle at half maximum of wall heat flux at the inner wall of the cooling channel. This study draws the dimensionless groups featuring CHF phenomenon not only under uniform heating, but also under one-side heating, which involve additional normalized parameters.

A comparison of critical heat flux in tubes and bilaterally heated annuli

This paper examines the critical heat flux (CHF) behaviour for annular flow in bilaterally heated annuli and compares it with that in tubes and unilaterally heated annuli. It was found that the differences in CHF between bilaterally and unilaterally heated annuli or tubes strongly depend on pressure and quality. The CHF in bilaterally heated annuli can be predicted by tube CHF prediction methods for the simultaneous CHF occurrence at both surfaces, and the following flow conditions: pressure 7–10 MPa, mass flux 0.5–4.0 mg m −2 s −1 and critical quality 0.23–0.9. The effect on CHF of the outer-to-inner surface heat flux ratio, was also examined. The prediction of CHF for bilaterally heated annuli was based on the droplet-diffusion model proposed by Kirillov and Smogalev (1969, 1972). While their model refers only to CHF occurrence at the inner surface, we extended it to cases where CHF occurs at the outer surface, and simultaneously at both surfaces, thus covering all cases of CHF occurrence in bilaterally heated annuli. From the annuli CHF data of Becker and Letzter (1975), we derived empirical functions required by the model. The proposed equations provide good accuracy for the CHF data used in this study. Moreover, the equations can predict conditions at which CHF occurs simultaneously at both surfaces. Also, this method can be used for cases with only one heated surface.

Critical heat flux and flow pattern for water flow in annular geometry

An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for the upward flow. In addition to the experiment, selected CHF correlations for annuli are assessed based on 1156 experimental data from various sources. The Doerffer et al. (1994); Barnett (1966); Jannsen and Kervinen (1963); Levitan and Lantsman (1977) correlations show reasonable predictions for wide parameter ranges, among which the Doerffer et al. (1994) correlation shows the widest parameter ranges and a possibility of further improvement. However, there is no correlation predicting the low-pressure, low-flow CHF satisfactorily.

Analysis of flow instabilities and their role on critical heat flux for two-phase downflow and low pressure systems

A stability analysis of a flow boiling two-phase low pressure and downflow system relative to the occurrence of critical heat flux has been carried out. The problem formulation is based on a time and area averaged one-dimensional drift flux model, with the necessary constitutive equations. A characteristic equation is obtained to predict the onset of flow excursion and density wave oscillations. By non-dimensionalizing the characteristic equation, important groups governing the system stability are determined. The results of the analysis are useful in determining the region of stable operation for downflow in the Westinghouse Savannah River Site Reactor and in avoiding the onset of flow excursions and density wave oscillations. The analytical results for flow excursion are compared with the Babcock and Wilcox flow excursion experimental data with a Savannah River Mark 22 fuel assembly mockup.

A critical heat flux model based on mass, energy, and momentum balance for upflow boiling at low qualities

A mechanistic modeling of critical heat flux (CHF) in upflow boiling at low qualities is performed. The developed model is based on a physical criterion of CHF occurrence and a mechanism limiting the thermal transport between a stagnant bubbly layer and bulk stream. The mechanism can be mathematically formulated by coupling the equation of limiting mixing mass flux, which is derived from momentum balance equations in two regions, with local mass and energy balance equations on the bubbly layer. The resulting form of the model is represented by a general and straightforward CHF formula involving two empirical constants related to the void fraction and the thickness of the bubbly layer. The predictions agree well with the extensive CHF data of water in uniformly heated tubes.

A Study on Limiting Exit Quality of CHF of Forced Convection Boiling in Uniformly Heated Vertical Channels

This paper presents the results from a phenomenological analysis of limited exit quality X°ex, that is, the occurrence of critical heat flux at a nearly constant exit quality X°ex independent of the change of inlet quality under mixed inlet condition. The author’s generalized correlation equations of CHF in uniformly heated channels with subcooled inlet condition are employed for this purpose, and the prediction of X°ex is compared with the existing data of X°ex satisfactorily. As a basis of the above-mentioned analysis, the applicability of the author’s correlation equations to the state near the critical length-to-diameter ratio corresponding to the limiting quality condition is also examined.

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.

An Analysis of Critical Heat Flux in Flow Reversal Transients

A large inlet break Loss of Coolant Accident in a Pressurized Water Reactor (PWR) would cause the flow through the core to reverse within milliseconds. Currently approved methods of analysis conservatively assume that vapor blanketing of core heat transfer surfaces occurs upon this first reduction to zero flow. A coordinated experimental and analytical study has been conducted to determine when and where the vapor blanketing or Critical Heat Flux (CHF) conditions actually do develop in constant pressure rapid flow reversals. The results indicate that first occurrence of CHF is due not to low coolant velocities, but to flow stagnation in the channel interior with associated rapid channel voiding. Calculations indicate that good cooling should persist over large regions of the core for about 1 s longer than is currently assumed.