Critical Heat Flux Experiments Research Articles

A knowledge gap analysis for transient CHF prediction within RELAP5-3D

We compare several experimental pool and flow boiling critical heat flux experiments with corresponding computational models developed with RELAP5-3D. Encompassed within this paper are several sensitivity analyses for the boiling heat transfer correlations, critical heat flux (CHF) prediction within RELAP5-3D and the thermal properties of the cladding for experiments in the Transient Reactor Test Loop (TRTL) and Transient Reactor Test (TREAT) facilities. The sensitivity analyses were consistent in determining that there is a large contribution from the critical heat flux multiplier which in this application, accounts for the effects of power transient critical heat flux. Comparison to experimental results of both the TREAT CHF-SERTTA experimental campaign and a University of New Mexico (UNM) pool boiling experimental campaign demonstrate both the under-prediction of transient CHF and nucleate boiling heat transfer. To address the gaps in knowledge regarding the transient heating effect for both boiling heat transfer and CHF prediction, we present an experimental test matrix for the TRTL facility that will isolate the power transient effect under PWR relevant operational conditions. The test matrix will incorporate 6 power pulses with a constant energy deposition and a timescale that spans over several orders of magnitude, including a steady state test with an exponential period of 10 s.

Critical heat flux, heat transfer and pressure drop at high mass fluxes for R123 in a single microchannel

This paper presents new experimental data on pressure drop, heat transfer coefficient, and critical heat flux during flow boiling of low-pressure fluid R123 in microchannels. High mass velocities were used, above the values found in the literature in tests with this fluid. Experiments were conducted in a single smooth and horizontal microchannel, with an internal diameter of 1.1 mm and heating sections of 46 and 100 mm. The mass velocities used were between 1500 and 7500 kg/m2s. Saturation temperature ranged from 40.1 to 85.1 oC. The outlet vapor quality varied between -0.54 and 0.53 for the CHF results, between 0 and 0.5 for the pressure drop results, and between 0 and 0.35 for the HTC results, respectively. Record critical heat flux values for flow boiling of R123 in smooth microchannels were achieved, reaching 1.44 MW/m2. An analysis was made of the influence of an upstream pressure drop in avoiding instabilities during critical heat flux experiments. The highest heat transfer coefficient values during convective boiling of R123 in smooth microchannels were also obtained, with values close to 43.3 kW/m2K. The saturation temperatures used were among the highest for flow boiling of R123 in smooth microchannels found in the literature. Correlations for frictional pressure drop, heat transfer coefficient, and critical heat flux were evaluated in predicting the experimental data. An analysis was made on the influence of the void fraction correlation when estimating both the accelerational and frictional pressure drop.

Prediction of critical heat flux in non-uniform heated rod bundle based on modified CHF empirical correlations

The accurate estimation of critical heat flux (CHF) continues to be a significant challenge in the study of radial and axial non-uniform heated rod bundles under NHR-200Ⅱ operating conditions. The present work develops the subchannel analysis code CTF algorithm that incorporates modified CHF empirical correlations to forecast CHF of non-uniform heated rod bundles preliminarily. The article selected experimental data from HTRF, which includes both uniform heated and non-uniform heated rod bundle CHF experiments, for preliminary validation and analysis of this method. The results demonstrated that the predicted results of modified Barnett correlation have the good agreement with the experimental values. Additionally, the modified Barnett correlation is used to predict the CHF of rod bundles under various power distribution. In different radial power distribution, the critical power of rod bundles decreases with the increase of radial power factor ratio. In different axial power distribution, the critical power in inlet and outlet peak distribution is lower than uniform heated rod bundle. Regarding the case of radial uniform heating and axial inlet peak distribution, CHF locations often occur near the central section of the heated rod height. In the axial intermediate peak distribution, CHF positions often emerge in the intermediate and downstream sections of the heated rod to the outlet. In a typical operating scenario of the NHR-200II, the CHF will occur towards the exit of the hot central subchannel, particularly near the radial heated rod exhibiting the maximum thermal power. The occurrence of CHF can be attributed to factors such as lower mass flux, higher local equilibrium quality, and heat flux.

Experimental investigation of helical fin and inclination effect on critical heat flux

Recently, diverse marine applications of nuclear power are being considered to meet the carbon reduction goals in the maritime industries. One of the distinctive characteristics of marine nuclear reactors is the limit in space and it requires high power density. To address this problem, tight lattice bundle with a helical shaped non-circular fuel rods can be employed. However, the critical heat flux (CHF) mechanisms for such alternative fuel rods have not been adequately investigated experimentally. In particular, no open-access CHF experiment under inclined and motion conditions in accordance with marine conditions has yet been reported. Thus, in this study, the CHF on a helical-finned heater was measured under static vertical and inclined conditions. Considering the reactor operating conditions, a high-pressure experiment was performed using refrigerant R134a. The fin and inclination effects were evaluated based on a comparison between the results of this study and the CHF data for a bare-rod under inclined condition obtained from a previous study. In addition, the inclination effect was analyzed phenomenologically based on the wall temperature behavior. The findings in this work can contribute to designing a marine nuclear reactor core by quantifying the effects of inclination on the CHF.

Steady-state and power transient critical heat flux experiments of FeCrAl and Zircaloy claddings under saturated pool boiling

FeCrAl and Zircaloy cladding tubes are vertically placed in the saturated pool boiling chamber to perform steady-state and power-transient critical heat flux (CHF) experiments. Three FeCrAl-variants, including FeCrAl-C26M, FeCrAl-B126Y, and FeCrAl-B136Y, are tested to compare their pool boiling thermal performances with traditional Zircaloy claddings including Zirlo, Zr705, and Zircaloy-4. In the steady-state pool boiling experiments, it was confirmed that FeCrAl alloys have higher pool boiling CHF values than Zircaloys. Among these tested materials, FeCrAl-C26M has the highest pool boiling CHF, but the FeCrAl-C26M nucleate boiling heat transfer coefficient is less than that of FeCrAl-B136Y. It reconfirms that CHF is enhanced on the post-CHF oxidized samples. However, as the cycling number of post-CHF oxidization increases, the CHF enhancement becomes less appreciable due to the limited physics contributions of the oxide layer. It is noteworthy that the finalized CHF enhancement can vary a lot between different cladding materials. For example, in three Zircaloy variants, their saturated pool boiling CHF results converge to 980 kW/m2 on the post-CHF oxidized samples. In three FeCrAl variants, their pool boiling CHF results are finalized to 1400 kW/m2. This CHF enhancement difference could be possibly attributed to the chemical constituents of oxide layers presented on cladding surfaces. The Fuchs-reactivity initiated accident transients are simulated by direct current programmable power supply to characterize thermal responses of different claddings to various pulse-width power transients. It is found that the heat transfer coefficients show monotonically increasing parametric trend with respect to faster power-transient rates, i.e., shorter pulse widths. Different from the steady-state pool boiling, the transition boiling regime is phenomenologically reflected in the power-transient pool boiling curve. This potentially implies that power-transient CHF featured by surface temperature overshoot may indicate transient thermal-safety margins not so accurately as the power-transient maximum heat flux. The difference gap of power-transient maximum heat flux between various cladding materials can be gradually closed by the progressive increasing of power transient rates.

Rationalization and evaluation of a Phenomenon-driven improved bundle CHF model under PWRs condition with different axial power distributions

In this paper, based on the bubble crowding bundle critical heat flux (CHF) mechanistic model proposed earlier by the author, which is applicable to the pressurized water reactors (PWRs) condition, more reasonable processes are considered, and a phenomenon-driven improved model was proposed. Firstly, the improved model describes the relationship between the latest visualization experimental results within PWRs ranges and bubble crowding concept, and clarifies the theoretical basis of the model. Secondly, based on the grid-effect CHF experiment, a more mechanistic turbulence mixing intensity factor was proposed in the process of model construction. Finally, the validation of the improved model for uniform axial power distribution (APD) CHF shows good prediction ability, and the previous tendentious problem about quality was eliminated. The model’s assessment of non-uniform APD CHF is performed for the first time, which shows that the improved model has good potential for predicting non-uniform APD CHF, and the CHF model needs to consider the upstream heating effect.

Combined effect of a corrodible material and acidic fluid on flow boiling crisis

The heat transfer surface condition is one of the factors that greatly affects not only boiling heat transfer but also boiling crisis. In particular, in the case of a metal that is easily corroded, oxide formed on the surface of the metal during boiling may affect the critical heat flux (CHF). The oxide has conflicting effects, as it can enhance the CHF by improving the surface wettability but also simultaneously reduce it due to low thermal conductivity. In this research, CHF experiments were carried out under DI water and acidic fluid conditions using SA508 carbon steel, a metal susceptible to corrosion. In the case of pre-oxidized SA508, a 25 to 53 μm magnetite layer was formed on the surface preventing additional oxidation, resulting in a constant CHF. In addition, surface observations directly confirmed that the pre-oxidized layer remained even after a long period of boiling. Conversely, for fresh SA508, two oxide layers were found to form after the experiment. As the boiling time increased, the CHF of the fresh SA508 decreased, which is considered to be related to an increase in the thickness of the magnetite layer.

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.

Critical heat flux of saturated flow boiling for R245fa under hypergravity

• The gravitational effect on saturated flow boiling CHF for R245fa was measured. • Hypergravity enhances flow boiling CHF. • When approaching CHF, raising gravity level restrains wall temperature fluctuation. • Increasing mass flux diminishes gravitational effect on flow boiling CHF. Critical heat flux (CHF) is a very important design parameter for safe operation of heat dissipation applications onboard flight vehicles. Saturated flow boiling CHF experiments were conducted, aimed at investigating the influence of hypergravity, were conducted. The working fluid was R245fa refrigerant flowing inside a horizontal copper tube of 2.43 mm inner diameter. A turntable was used to simulate the hypergravity environment. Two mass fluxes (156.7, 236 kg/m 2 s) with different inlet vapor qualities (0.1 – 0.67) and gravity levels (1 g – 2.77 g ) were tested under the inlet pressure of 250.5 kPa. Besides, the limiting heat flux q lim , under which the vapor quality at the tube outlet equals one, was defined for theoretical analysis, and the gravitational effect on boiling curves was provided. The results show that with decrease in inlet vapor quality, CHF increases and outlet vapor quality at the CHF decreases. However, for low mass flux of 156.7 kg/m 2 s and normal gravity, the increase in CHF is weakened with decreasing inlet vapor quality, and the deviation between the CHF and the corresponding limiting heat flux increases. Raising gravity level raises saturated CHF as well as the corresponding outlet vapor quality, and thus narrows the above deviation. For high mass flux (236 kg/m 2 s), the CHFs under different gravity levels (from 1 g to 1.75 g ) are almost the same for the same inlet vapor quality. They increase approximately linearly with decreasing inlet vapor quality, and gradually deviate from the q lim line, corresponding to a decrease of the outlet vapor quality when CHF occurs.

Experimental investigation on critical heat flux enhancement by mixing-vanes spacer grid in a vertical annular geometry

In this paper, critical heat flux (CHF) experiment was carried out in a vertical single-rod channel using R-134a as working fluid to study the effect of mixing-vanes spacer grid (MVSG) on CHF enhancement. The single-rod channel had a rectangular geometry with a uniformly heated rod centrally installed. The test conditions were pressure from 1.8 to 2.7 MPa, mass fluxes from 600 to 2100 kg·m−2·s−1 and subcooling temperature from 10 to 40 ℃, which was determined by Katto’s fluid-to-fluid modeling method with reference to the operation condition of Pressurized Water Reactors (PWRs). The effects of flow parameter and downstream distance on CHF enhancement by MVSG were investigated. In addition, visualization study on flow regime during CHF occurrence and bubble behavior when passing though MVSG was observed by a high-speed camera, which contributed to improving better understanding on the mechanism of CHF enhancement by MVSG.

Experimental study on the CHF enhancement effect of nanofluids on the oxidized low carbon steel surface

Studies about using nanofluids to enhance the Critical Heat Flux (CHF) of In-vessel Retention (IVR) strategy in the third-generation reactor have been conducted extensively and show a significant CHF enhancement effect. However, low carbon steel SA508 used in the reactor vessel is easy to oxidize and the oxidation can lead to changes in the surface properties which may affect the CHF enhancement effect of nanofluids. In this study, pool boiling CHF experiments with low carbon steel SA508 surfaces were conducted in distilled water and nanofluids under different boiling time to investigate the CHF enhancement effect of nanofluids under low carbon steel surface oxidization condition. CHF in distilled water increases rapidly with boiling time due to the rapid surface oxidation during the boiling and the increase ratio can be nearly 2 due to the surface oxidation. CHF in nanofluids is stable and independent of boiling time. The difference between CHF in nanofluids and CHF in distilled water decrease to 17% under the longest boiling time conditions due to the surface oxidation. The deposition layer of nanoparticles on the surface leads to the capillary wicking and decrease in the nucleation site and thus, CHF is enhanced. This study is of great significance for exploring the actual effect of nanofluids on the CHF enhancement of IVR strategy.

Power transient critical heat flux in a narrow rectangular channel under downward flow

Power transient critical heat flux (CHF) experiments were conducted under downward flow conditions in a narrow rectangular channel simulating a sub-channel consisting of plate-type fuels for a research reactor. The test section is designed as a sub-channel with a wetted width of 66.6 mm, heated width of 62 mm, gap size of 2.35 mm, and length of 640 mm. In the experiment, a mass flux of 3750 kg/m2s with a downward flow, an inlet temperature of 37 °C, and outlet pressure of 170 kPa were maintained constantly at a given heat flux transient condition. The heat flux was exponentially increased with the power increase rate ranging from 0.03 to 1 s−1 for the simulation of power transients in the reactivity-initiated accidents of the research reactor. It was found that the CHF increased significantly as the power increase rate increased at a given flow condition. The power transient CHF was analyzed using a dimensionless velocity consisting of parameters representing the liquid-sublayer thinning mechanism that is expected in the vicinity of the heated wall. Finally, we developed a power transient CHF correlation based on the experimental data set consisting of the present data, data of Kataoka et al., and the R-12 data of Celata et al. The developed correlations showed an average error of 0.24% and root mean square error of 3.09% against the experimental data set for power increase rates of 0.03–1 s−1.

CFD investigation of the cold wall effect on CHF in a 5 × 5 rod bundle for PWRs

In this study, a set of mathematic models suitable for flow boiling simulation and CHF prediction in rod bundle channels of Pressurized Water Reactor (PWR), including Eulerian two-phase model combined with wall boiling model, were established using CFD method. The models were validated against the international flow boiling and Critical Heat Flux (CHF) experiment benchmark. The results show that the error between the calculated CHF and the experimental data is less than 10%. Then a 3-D model of 5 × 5 rod bundles with cold wall was built and the detailed three-dimensional thermal–hydraulic characteristics were investigated, especially emphasising on the cold wall effect. The existance of cold wall would affect the coolant thermal mixing intensity, leading to a certain degree influence of the CHF. Results show that under the same inlet conditions, the calculated CHF value after introducing the cold wall is higher by 11.4%, while the overall power decreases due to the existence of cold wall. Besides, the pressure drop and temperature in the channel would decrease, while the distribution of void fraction basically remain unchanged, which is still mainly affected by the spacer grids and mixing vanes. Compared with the modified W-3 formula, the error of the CHF value calculated by CFD method is less than 10%, and the CFD method can show the three-dimensional distribution characteristics and evolution process of physical parameters.

Sensitivity analysis of in-pile CHF experiments in the TREAT facility: Characterization of impacts of fuel system thermal properties

Understanding fuel-to-coolant heat transfer mechanisms under transient heating irradiation conditions is an important aspect of pressurized-water reactor (PWR) safety. The work here aims to elucidate the potential impacts of critical heat flux (CHF) events on nuclear fuel systems, utilizing the heat transfer time constant (HTTC) as the fundamental basis. A set of thermophysical material and heat transfer coolant properties was established to represent the HTTCs of two Transient Reactor Test (TREAT) facility experimental fuel/cladding designs. Pool and flow boiling conditions were used to model these fuel systems within TREAT experimental capabilities, and a power pulse was used to represent a transient insertion of energy. Using Sobol sensitivity analysis and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code, an investigation was conducted to identify the most important HTTC parameters pertaining to the output of the peak outer cladding temperature (POCT), maximum fuel centerline temperature, and the time of CHF.The maximum fuel centerline temperature was shown to be highly dependent on the UO2 volumetric heat capacity. Regarding the POCTs, DNB was determined to have a dominant effect on the heat transfer mechanisms of these fuel systems. Further, the HTTC significantly impacted POCT predictions with the UO2 volumetric heat capacity, the film boiling heat transfer coefficient, and the CHF multiplier being most influential regardless of coolant flow conditions. Gap variations were shown to influence the POCT in cases that experienced the DNB event. In non-DNB cases, the HTTC had no significant impact on POCT. For the time of CHF, variations in gap thickness, and the CHF multiplier were overwhelmingly dominant.

Overview on critical heat flux experiment for the reactor fuel assemblies

The critical heat flux (CHF) experiment for the reactor fuel assemblies (FAs) is the most difficult and important one in the reactor thermo-hydraulic study. The FA CHF experiments are performed on the large scale thermo-hydraulic test facilities with large DC power suppliers, complicated test sections and rigid test process. The results are analyzed with subchannel codes to obtain the parameters at the CHF point. Based on the experimental data, the CHF correlation is derived as the tool for the reactor safety analysis and design. After many years’ work, the common understanding of the FA CHF was established and the method of the FA CHF experiment becomes similar. This paper reviewed the progress of the FA CHF experiments, analyzed the modeling of the test sections, introduced requirement of the test facilities, and discussed the hypothesis in the modelling. The benchmark qualification test adopted to verify the test method and test section used in the experiment are discussed in the paper, too.

Flow boiling transient critical heat flux tests with stainless steel and FeCrAl: Transient correlation implementation, model calibration, and sensitivity analysis

Steady-state and transient internal-flow critical heat flux (CHF) experiments were carried out under two flow conditions, at atmospheric pressure, on 316 stainless steel (316-SS) and iron-chromium-aluminum (FeCrAl) tubes. Slow power transients at a low mass flow (300kg/m2−s) with very low subcooling (Xe=−0.0054) generated premature CHFs, which were prevented by faster power transients and high mass flow (1,000kg/m2−s). The measured transient CHFs increased linearly with increasing power transients, compared with the steady-state CHF. Yet, the wall superheat at the CHFs decreased with faster power transients. Transient CHF correlations highlighting heterogeneous spontaneous nucleation were calibrated to the measured CHFs and compared with other existing correlations. Transient CHF multipliers were acquired from pool and flow boiling empirical CHF correlations that were generated. The multipliers were applied to the RELAP5-3D nuclear system code to analyze the discrepancies between the measured data and the predicted CHF and post-CHF behavior, which improved peak cladding temperature predictions by 24.6%. A variance-based global sensitivity study perturbing the experimental uncertainties and cladding material thermal properties showed the diminishing influence of flow boiling heat transfer with increasing power transients, highlighting the significance of the volumetric heat capacity for the cladding integrity during transients. Transient CHF correlations were applied to the most limiting design basis accident: a hot-zero-power reactivity-initiated accident on a generic pressurized water reactor RELAP5-3D model.

Sensitivity analysis of in-pile critical heat flux experiments in TREAT for characterization of RIA power-transient effects

A reactivity-initiated accident (RIA) is one type of postulated design basis accident (DBA) that can cause a departure from nucleate boiling (DNB) event in pressurized-water reactors (PWRs). A DNB occurrence and its consequences depend on the thermophysical properties of the fuel components and coolant, characteristics of the transient energy insertion into the fuel rod, and the onset of the critical heat flux (CHF) phenomenon. To leverage the restart of the Transient Reactor Test (TREAT) Facility, an effort is currently underway to better understand the cladding-to-coolant heat transfer mechanisms and the CHF phenomenon under fast-transient irradiation conditions. This paper characterizes the impact of power transients on the thermal-hydraulic behavior of a TREAT Facility reactor heater rodlet CHF experiment to provide the priority of parameters that need to be investigated for an improved CHF model. Sobol sensitivity analysis methods and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code were used to identify key input parameters on the uncertainty in the prediction of peak outer- and inner-surface temperatures of the heater tube, as well as the time of the DNB event. A series of sensitivity analyses revealed the total energy deposition on the tube and the transient effects of power pulse had large impacts on the maximum temperatures. The CHF multiplier had the largest impact on the time occurrence of CHF. The overall results show the energy deposition rate in the tube is the most influencing factor to the manifestation of CHF and the resulting thermal-hydraulic behaviors of the tube. The multiplier for the CHF, which is interpreted as the predicted CHF value, has the largest Sobol indices for the time of the CHF in all cases, since it directly determines the occurrence of CHF. It is inferred that the uncertainties in the thermal-hydraulic behaviors of fuels increase with respect to the key parameters as the power pulse becomes broader, and an accurate estimation of the energy deposition rate is required to reduce the uncertainty in the evaluation of the integrity of fuel if the CHF is expected to occur near the peak power. The outputs are expected to provide rigorous interpretation of ongoing in-pile CHF experiments in the TREAT Facility reactor regarding thermal-hydraulic behavior of the fuel system aiming for a new transient in-pile CHF model.

Critical heat flux measurement and visualization in R-134a on a vertical single rod with and without mixing-vanes spacer: The characteristics and phenomenon of critical heat flux

In this study, a critical heat flux (CHF) experiment using R-134a as working fluid was carried out in a vertical single rod geometry. A unique feature of this experiment are high speed visualization studies obtained at a viewing port located at the CHF location. The 9.5 mm O.D. heated rod was placed in a 19 x19mm square channel to simulate the Pressurized Water Reactors (PWRs) fuel lattice. Test conditions were selected and scaled to cover a wide range of water conditions relevant to PWRs. The CHF database, experimental trends and comparisons to existing CHF prediction methods are presented. The visualization results of the flow boiling at the CHF location and the CHF mechanism are discussed with a particular emphasis on the impact of grid spacers on the boiling phenomena. Quantitative analysis of bubble layer thickness is carried out, including the relationship between bubble layer thickness and CHF.

Progress in rod bundle CHF in the past 40 years

Critical Heat Flux (CHF) prediction remains one of the most important tasks in thermal hydraulics of Light Water Reactors (LWR’s). During recent years, tremendous progress has been made in understanding and predicting rod bundle CHF. However, due to the unique features associated with rod bundle subchannel systems, including complex and compact anisotropic geometry, non-uniform axial/radial heating, open channel interactions, periodic (positional) mixing vane grids (MVG’s) mixing (Yang et al., 2014), and a broad range of operation conditions (Yang and Han et al., 2019), a precise prediction of CHF in rod bundles, entirely based on analytical methods without the support of any extensive experimental database, is still beyond the present modelling and computational capabilities. For Pressurized Water Reactors (PWR’s), CHF occurs at a relatively low coolant quality and is primarily driven by the Departure from Nucleate Boiling (DNB) type event, which can be triggered by liquid sublayer dryout, bubble crowding, homogeneous nucleation, or other mechanisms causing local starvation of liquid under subcooled flow boiling or low quality saturated boiling conditions, and the dryout event, which describes CHF under a high void situation when liquid film on the fuel surface dries out, usually happens in a high quality saturated annular flow. Boiling Water Reactors (BWR’s) operate at a relatively lower pressure and can reach much higher coolant qualities that correspond to annular two-phase flow conditions. This dominant CHF mechanism, dryout, is based on the gradual thinning and drying out of the liquid film.This paper provides an overview of the progress made in the experimental and analytical investigations of CHF in both PWR and BWR rod bundles during the past 40 years. A complete comprehensive concept map is developed to outline most of the key mechanistic CHF models covering the entire range of flow regimes. The limitation and the potential approaches for the application of mechanistic models to rod bundle CHF prediction are reviewed. This includes the recently proposed, rod bundle specific NHNM (Non-uniform heater Homogeneous Nucleation Model [Yang, 2017]). Various industrial empirical correlations are reviewed together with their corresponding subchannel codes and required rod bundle CHF experiments. The need for clear guidelines and standards for benchmarking new rod bundle CHF test facilities is discussed.Various approaches on investigating, understanding, modeling, and predicting the occurrence of rod bundle CHF are reviewed from mechanistic models, semi-empirical and empirical approaches to advanced CFD simulation. The achievements and limitations in those methods are discussed with suggestions for future advancement.

Experimental study of post-CHF heat transfer in a vertical tubular test section

In this paper, post-CHF (critical heat flux) experiments are performed in a vertical tubular test section with water as the working fluid to study the heat transfer characteristics during the process of formation, growth, and propagation of dry patches. The formation, growth, and propagation of the dry patches are indicated from the measured wall temperatures of the test section. The so-called dry temperature and quench temperature, corresponding to the formation and departure of dry patches, are obtained by analyzing the first-order derivative of the measured wall temperatures. The temperature gradient along the axial direction of the test section wall is also calculated, up to 15 °C/mm. Large temperature gradients result in axial heat conduction as well as thermal stresses in the test section. The convective heat transfer coefficient for the dry patches is calculated based on the saturation temperature of the fluid and compared with three heat transfer correlations in the literature. The computed convective heat transfer coefficient for the dry patches in our experiments is larger than the calculated results of the correlations, especially for conditions of relatively low wall superheats.