Critical Heat Flux Experiments Research Articles

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.

Study on the CHF correction factor under rolling condition in vertical tube

The marine reactor is one of the hot topics in recent years. Several papers have been published about critical heat flux (CHF) under rolling condition in a qualitatively way but none about quantitative analysis, and no CHF correction factor was given. This paper presented some non-dimensional numbers which show CHF mechanism under rolling condition based on the departure from nucleate boiling (DNB) and Dry-out mechanistic models, and used the data obtained from the Korea scholar Jin-Seok Hwang who carried out a CHF experiment under rolling condition in vertical tube with CFC-R134a as refrigerating agent to get the appropriate CHF discipline under rolling condition. And the paper analyzed the influence mechanism of the CHF under the rolling motion.

Face-centred cubic CuO nanocrystals for enhanced pool-boiling critical heat flux and higher thermal conductivities

In this study, we established a face-centered cubic CuO nanocrystals (fcc.CuO), and their nano liquids (NLs), and apply them to pool boiling critical heat flux (pbCHF) and thermal conductivity (TC) applications. Additionally, we compared them to commercial CuO nonmaterials. The fcc.CuO was fabricated by a solution chemical method using a specific hydrophilic surfactant, such as polyvinyl pyrrolidone, as the capping agent for controlling the shape, and size. We investigated two kinds of nanoliquids which are fcc. CuO-NL and CuO-NL with 0.06 wt%, 0.03 wt%, 0.006 wt%, and 0.003 wt% for application of pbCHF, and TCs. The pbCHF of fcc.CuO-NL(0.06 wt%) was shown 121% compared to CuO-NL, and 132% higher compared with de-ionized (DI) water, respectively. The thermal conductivity of fcc.CuO-NL is 7.2% compared with DI water, owing to the effect of deposition of fcc.CuO with regular shape, and size. The contact angles (CAs) measurements shown higher hydrophilic nature due to the formation of the nano template layer on a Nichrome wire surface. Besides that, we evaluated the bubble site density, bubble growth, and departure studies during critical heat flux experiments using a speed camera. In conclusion, the developed fcc.CuO-NLs exhibited higher pbCHFs and Thermal conductivities with lower concentrations, respectively.

Elucidating the Impact of Flow on Material-Sensitive Critical Heat Flux and Boiling Heat Transfer Coefficients: An Experimental Study with Various Materials

Steady-state internal flow boiling critical heat flux (CHF) experiments were carried out at 10℃ inlet subcooling and atmospheric pressure (84 kPa at an altitude of about 5300 feet) with mass flux ranging from 200 to 2,000 kg/m2-s on seven tube materials, including the accident-tolerant fuel cladding candidate FeCrAl (C26M) alloy. At a relatively low flow rate, the sensitivity of CHF to different materials was appreciable (the average difference among the tested materials was 19% at 200 kg/m2-s). As the absolute CHF differences stayed constant for higher flow rates, the increased flow rate led to reduced relative CHF differences among the tested materials (average relative difference of 15% at 1,000 kg/m2-s). FeCrAl alloy claddings that were oxidized under simulated Light Water Reactor (LWR) conditions yielded CHF comparable to as-received FeCrAl alloy cladding, with less than 1.8% difference. Experimental CHF results imply marginal material sensitivity impacts on the departure from nucleate boiling ratio in steady-state nuclear reactor cores. Yet, for accident progression analysis, material sensitivity may hold importance, as boiling behavior variation at low mass flux is appreciable. Macroscopic material parameters (surface wettability, surface roughness, and thermal properties) do not provide a suitable explanation for observed material-sensitive CHFs. Yet, surface roughness and thermal effusivity revealed power and linear relations, respectively, to the nucleate boiling heat transfer coefficients at low mass flux. At high mass flux, material sensitivity on nucleate boiling heat transfer significantly decreased due to the increased shear stress on boiling surface overwhelming material-related boiling implications.

Experimental study for critical heat flux in 2x2 rod bundles at high pressure conditions

Critical heat flux (CHF) phenomena has been a topic of study for over fifty years yet remains a serious concern across industries and especially in water-cooled nuclear reactor design. CHF data in the high-pressure range is limited, and the effects of bundle geometry and non-uniform heat flux profile are challenging to quantify. In this study, CHF experiments were performed in upward flowing water in a 2 × 2 rod bundle with a cosine profile heat flux. Data were collected between 16.5 and 18 MPa at two mass flux conditions with three inlet subcooling conditions. Under these conditions, average CHF was shown to decrease with increasing pressure. However, pressure had less of an effect than decreasing the inlet subcooling or the mass flux, both of which reduce the CHF value considerably. Interestingly, correlations that have been developed for lower pressure continued to predict CHF occurrences with moderate accuracy outside their ranges of validity. Nearly all predicted values were within 20% of experimental values.

Investigation on the critical heat flux in typical 5 by 5 rod bundle at conditions prototypical of PWR based on CFD methodology

The Critical Heat Flux (CHF) is crucial for the safety of power equipment working with boiling heat transfer, such as the heat exchanger and fuel assembly. The wall heat partition model with consideration for the heat transferred to the vapor can be used along with the Eulerian two-fluid model to predict the CHF. In our previous work, the model was validated to be applicable to predict the CHF in a 2 by 2 rod bundle without strong secondary flow. While strong secondary flow occurred in the reactor core due to the existence of MV (Mixing Vane) spacers. In this paper, the model was validated by the CHF experiment in a typical 5 by 5 rod bundle with 3 different types spacers under Pressurized Water Reactor (PWR) condition. The effects of secondary flow induced by MV spacers on CHF were analyzed. The CFD calculation showed good agreement against the experimental measurements with the deviation of heating power 5.06%, and the deviation of axial location 4 mm. As for the CHF location in the cross section, the result obtained by CFD calculation was on the outer rod which was consistent with the experiment, while the results by system code and subchannel codes showed CHF occurred at the inner rod.

Flow boiling CHF experiment with sudden expansion tubes

Abstract Deformed cladding in nuclear power plants can partially block the path of coolant flow and thus alter the heat transfer characteristics between the cladding surface and coolant. In this research, critical heat flux(CHF) in a sudden expansion flow path, was experimentally investigated. Flow boiling CHF experiments were carried out under atmospheric pressure with deionized water for vertical upward flow, with four expansion ratio cases investigated with mass flux ranging from 50 to 250 kg/m2 s. CHF was detected at two different locations. In experiments with either a low expansion ratio(e 1.5) and high mass flux(> 150 kg/m2 s), CHF was detected at the sudden expansion point. Quality at the sudden expansion point was almost zero, which implies that nucleated bubbles under the effect of sudden expansion flow induced CHF at the sudden expansion point. The reason for this transition of CHF location was suggested that local bubble stagnation induced by a balance between drag and buoyancy forces on the bubbles resulted in local flooding in the downward flow.

Review of pool boiling critical heat flux (CHF) and heater rod design for CHF experiments in TREAT

Preventing the occurrence of a departure from nucleate boiling (DNB) event is an important aspect of nuclear safety in pressurized water reactors (PWRs). This phenomenon is partially governed by the cladding-to-coolant heat transfer under transient irradiation conditions, such as during a reactivity-initiated accident (RIA). Currently, there are large uncertainties about cladding-to-coolant heat transfer under these rapid transient conditions. This effort aims to elucidate the mechanisms of CHF under in-pile fast transient irradiation conditions using the Transient Reactor Test (TREAT) facility. These experiments will be carried out under pool boiling conditions, within experimental capsules that will be inserted into the TREAT reactor. A heater rodlet made from stainless steel 304 with tailored natural boron content will be usedin these experiment capsules to induce CHF when submitted to a TREAT power pulse. We will investigate the impacts of the presence of an oxide layer, radiation-induced surface activation (RISA), heat transfer time constant, and rapid surface heating effects on the CHF phenomenon.We review the parameters with significant effects governing the predictions of pool boiling CHF. Only CHF influencing parameters that are highly important, among them coolant subcooling and pressure, as well as oxide layer thickness, RISA, and rapid heating effects were included in this literature assessment.Preliminary neutronics and thermal hydraulic results from the design of the experimental apparatus are also presented in this paper. To aid in the modeling approach, the energy deposition and occurrence of CHF were identified as the most crucial key Figures of Merit (FoMs).In the heater rod design, boron concentrations between 0.1 and 2.09 wt% were considered. Further, a self-shielding study was performed to determine whether an instrumented borated tube could be used in place of a solid borated rod. This study determined that the inner region of the rod can be excluded or instrumented without heat generation penalties. Lastly, a thermal-hydraulics sensitivity study determined the lowest limiting boron concentration needed to induce CHF in water with different degrees of subcooling. Additionally, the value of CHF is known to increase during a rapid transient. Therefore, a CHF multiplier sensitivity study determined what multipliers would inhibit CHF for varying degrees of subcooling of two chosen power coupling factors (PCFs). The current borated tube rodlet geometry configuration achieved a maximum CHF multiplier value of 7.8 using a 1400 MJ power pulse in TREAT. Although this was the power pulse with the greatest energy deposition considered for this study, the TREAT facility is capable of pulses up to ~2500 MJ. This provides a significant margin in energy capacity that was not included within the scope of the calculations carried out to date.The application of the heater rod design was successfully demonstrated in initial experiments in December 2019. The results of these experiments will be explored in future publications.

Study on CHF characteristics in narrow rectangular channel under complex motion condition

In present work, the critical heat flux (CHF) characteristics in narrow rectangular channel were theoretically investigated under complex motion condition. A three-field analysis program was developed combining with a unified form of additional forces caused by motions, and the program was validated by CHF experiment data with a better prediction accuracy. The CHF characteristics of rectangular channel were evaluated based on two types of situations: (1) oscillating inlet flow and (2) constant inlet flow with different motions including heaving, rolling and the coupled motion of heaving and rolling. The results show that the oscillation of the inlet flow has a great influence on the CHF, the increases in the amplitude and period of the inlet flow oscillation cause the dryout to occur in advance. When the inlet flow is fixed, various motion conditions have little effect on CHF for high mass flux, and the reduction of CHF does not exceed 5%. However, the CHF is obviously reduced for the low mass flux due to the significant oscillation or thinning effect of the outlet liquid film caused by motions.

A study on the effect of oxidation on critical heat flux in flow boiling with downward-faced carbon steel

To execute in-vessel retention external reactor vessel cooling (IVR-EVRC) successfully, critical heat flux (CHF) plays a key role in securing the thermal and structural integrity of the reactor pressure vessel (RPV). In real-world applications, the RPV outer surface is exposed and oxidized. In this study, we performed a downward-face flow boiling CHF experiment using a carbon steel plate with several boiling cycles under water. Because the direction of gravity was inverse with the normal of the wall, it was difficult to remove bubbles, and in the local area, some local burning could repeat several times. After polishing the surface with sandpaper, the heat was supplied to the surface by cartridge heaters in a step-by-step manner until CHF was reached, after which the surface was cooled down. Using the same surface, without any further surface treatment, this heating cycle was repeated five times. The CHF value increased with the number of cycles. When the experiment was repeated for different mass fluxes, the CHF value also increased with the number of cycles. We suggest that this increasing tendency in CHF is associated with the increasing level of surface oxidation. After every repetition of the heating cycle, surface oxidation level increased, fewer bubbles were observed on the surface, resulting in higher CHF values. After the experiment, surface roughness increased and contact angle decreased. The liquid-vapor mixture area decreased with increasing heat flux in one experiment, whereas it decreased with boiling time, under the same heat flux, which delayed CHF. The gradual oxidation process of carbon steel could be beneficial for real-world applications of IVR-ERVC.

Oxidation effect on pool boiling critical heat flux enhancement of Cr-coated surface for accident-tolerant fuel cladding application

The oxidation of an accident-tolerant fuel (ATF) cladding alters the surface roughness, which, in turn, affects the capillary wicking characteristics and critical heat flux (CHF) during the boiling process. To evaluate the effect of oxidation on the boiling performance of ATF coating, this study fabricated Cr-coated surfaces by varying the substrate temperature, resulting in various sizes of deposited particulate nanostructures by using a direct-current magnetron sputtering technique. To consider the oxidation effect, sputtered test specimens were oxidized at 400 °C dry air for 20 days, forming a Cr2O3 oxide layer with a thickness of approximately 0.1 μm. Pool boiling CHF experiments were conducted in deionized water under atmospheric pressure. The resultant CHF on the fresh and oxidized Cr-coated surfaces was enhanced by up to 36% and 48%, respectively. A higher CHF enhancement was achieved with a lower substrate temperature during sputtering, which resulted in a higher nanoscale surface roughness. The spreading velocities on each specimen were measured through the characterization experiments of liquid droplet spreading, which can be interpreted as the capillary wicking momentum. The pool boiling CHF was predicted with spreading velocity and roughness parameters. The predicted CHF showed a good agreement with the measured CHF. This study suggests that the oxidation of Cr-coated ATF specimens has a positive effect on the enhanced CHF, which can be confirmed through the altered surface roughness and capillary wicking momentum.

Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties

In this work we analyze the differences between previous transient critical heat flux (CHF) experiments using iron-chromium-aluminum (FeCrAl), Inconel 600, and stainless steel 316 (SS316) alloy test sections and best-estimate modeling results from widely-used nuclear engineering systems and subchannel analysis tools. FeCrAl is an Accident Tolerant Fuel (ATF) candidate cladding material. The thermal hydraulic performance and safety characteristics of FeCrAl are being evaluated to determine viability as a cladding material in Light Water Reactors (LWRs). In this study, the results of the CHF experiments conducted at atmospheric pressure and fixed inlet coolant temperature and mass flux are compared to models built in the fifth version of the Reactor Excursion and Leak Analysis Program (RELAP5-3D) and CTF, the modernized version of COBRA-TF developed by the Consortium for Advanced Simulation of LWRs (CASL). Results from RELAP5-3D and CTF showed differences from the experiments and from each other in predicting CHF. In the Inconel 600 case, both computational tools overpredicted CHF, which led to an underprediction in the tube outer surface temperature. In the SS316 and FeCrAl cases, CHF was underpredicted by the codes, leading to an overprediction of the tube outer surface temperature.To understand the discrepancies in CHF and post-CHF predictions, studies were performed using RELAP5-3D and RAVEN to determine the sensitivity of CHF and peak test section temperature, an analog to peak cladding temperature (PCT), to heat transfer coefficients, a CHF multiplier, and uncertainties in the thermal conductivity and volumetric heat capacity. We found that CHF depends most strongly on the CHF multiplier and thermophysical properties. A combination of these factors that produced the best match to the experiment based on CHF, PCT, and the total energy deposited into the tube was determined. The best match parameters were able to provide best-estimate predictions of the CHF and integral heat flux, but were still conservative when predicting the PCT. The best match set of parameters developed in this paper are intended only as a demonstration of an approach that could be applied in the future with a larger set of experiments to produce more accurate models of CHF and post-CHF behavior.

The Study of Critical Heat Flux in Upflow Boiling Vertical Round Tube under High Pressure

The Critical Heat Flux (CHF) prediction under high pressure condition, even close to the vicinity of the critical pressure of water, is an important issue. Although there are many empirical CHF correlations, most of them have covered the pressure under 15MPa. In this study, based on the CHF experiment database of upflow boiling in vertical round tube from 15MPa to the vicinity of the critical pressure of water, the Katto, Bowring, Hall-Mudawar, Alekseev correlations, and Groeneveld LUT-2006 are comparatively studied. With an error analysis of the predicted CHF to the experiment database, the prediction capability and the applicability of these correlations are evaluated and the parametric trends of CHF varying with pressure from 15MPa to critical pressure are proposed. Simultaneously, according to the characteristics of Departure from Nucleate Boiling (DNB) type CHF under high pressure condition, the constitutive correlations of Weisman & Pei model are proposed. The prediction results of three entrainment and deposition correlations of Kataoka, Celata, and Hewitt corresponding to the Dry-Out (DO) type CHF are analyzed. Based on the two improved models above, a comprehensive CHF mechanistic model under high pressure condition combining the DNB and DO type CHF is established. The verification based on the experiment database of upflow boiling in vertical round tube and the parametric trends analysis of CHF varying with thermal-hydraulic and geometric parameters are carried out. Findings of this study have a positive effect on further development of CHF prediction method for universal CHF mechanism, especially under high pressure region.

Progress of the EU activities for the ITER Divertor Inner Vertical Target procurement

F4E undertook the qualification of so-called “additional suppliers” in order to enhance competition among the potential bidders and to secure the procurement of the ITER Divertor Inner Vertical Target. In order to assess the performances of W armoured plasma facing components under the conditions expected in the divertor target strike point region, a total of 36 W monoblock mock-ups were manufactured by Atmostat (F) by diffusion bonding, and by CNIM (F) and Research Instruments GmbH (D) by brazing, from which fifteen mock-ups were High Heat Flux (HHF) tested at IDTF (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facility.The basic HHF testing program foresaw the performance of 5000 cycles at 10 MW/m2and 300+700 cycles at 20 MW/m2. The test results showed some significant improvements, in particular on the issues of W monoblocks macro-cracking and heat sink thermo-mechanical fatigue performances, identified during previous HHF tests campaigns. Some enhanced HHF thermal fatigue testing and critical heat flux experiments were also performed.

Effects of carbon steel surface oxidation on critical heat flux in downward-face pool boiling

In this study, a downward-face pool boiling critical heat flux (CHF) experiment was performed using a carbon steel plate soldered to a copper base. The results of the experiment were analyzed and compared with those of a carbon steel and copper block experiment previously conducted at the same test facility. After polishing the surface using sandpaper, heat was applied to the surface by cartridge heating until CHF was attained. Due to the oxidation of the carbon steel, the surface changed gradually during boiling. The boiling cycle was repeated several times, and the CHF variations were observed. Detailed images of the heating surface behavior during boiling acquired by two synchronized high-speed cameras with different heat fluxes were analyzed. It was found that the more oxidized the surface became, the fewer bubbles were generated and the higher the CHF became; however, the bubble film departure frequency remained the same in all conditions. The contact angle and microscope surface images of the oxidation material were also obtained, and it was determined that the combined effect of the increase in wettability and decrease in nucleation site density was likely the reason that the CHF increased. This gradual oxidation process could be beneficial in actual situations since it can increase the upper limit of in-vessel retention external reactor vessel cooling.

Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel

Steady and transient (with a heating rate of 685 °C/s) internal-flow CHF (Critical Heat Flux) experiments were conducted under atmospheric pressure at a fixed inlet temperature (40 °C or 60 °C) and mass flow (300 kg/m2 s) on Fe-13Cr-6Al alloy, Inconel 600 and Zircaloy-4 tube samples. Multiple experiments were repeated on the same specimen to investigate the effect of surface characteristic changes (i.e., roughness, wettability, and oxide scale morphology) on the occurrence of CHF. Despite notable changes of wettability, roughness, and oxide layer characteristics on samples that had already been subjected to CHF, measured flow CHF remained unchanged throughout repeated experiments for tested materials. This demonstrates that the surface effects on flow CHF are limited in the test conditions. In the steady-state flow boiling condition, Fe-13Cr-6Al alloy demonstrated a 22% and 14% increase in CHF compared to Zircaloy-4 and Inconel 600, respectively. Compared to the 2006 Groeneveld CHF lookup table, Fe-13Cr-6Al alloy gives a 13% increase in the tested flow boiling condition. Material properties are considered primarily responsible for the observed CHF differences among the tested materials. The surface thermal economy parameter (ρcp3/2k) is proposed as an explanation for the observed CHF differences; this parameter is related to material’s ability to avoid an irreversible dry spot formation. The apparent disagreement of Zircaloy-4 CHF with both the look up table predictions and Inconel 600 shows the limitation of departure of nucleate boiling (DNB) evaluations that do not consider cladding materials. The transient Fe-13Cr-6Al CHF is 39% and 23% higher than the lookup table prediction and the steady-state condition experimental result, respectively.

Experimental study of critical heat flux in flow boiling under subatmospheric pressure in a vertical square channel

Critical Heat Flux (CHF) is the maximal limit of heat flux in two-phase nucleate boiling heat transfer; therefore, an understanding of CHF under a wide range of conditions is important for safe system operation. In this work, CHF experiments are conducted over a range of subatmospheric system pressures in a vertical square channel that is heated on one side. The experimental conditions cover a pressure range of 20 kPa to 108 kPa, a mass flux range of 45–190 kg/m2-s, and an inlet subcooling range of 0–14 K. Heat flux is gradually increased until an excursion of the wall temperature occurs, indicating CHF. For the experimental conditions considered, CHF increases with rising system pressure, mass flux, and inlet subcooling, although the effects of mass flux and inlet subcooling are weak. A new correlation for CHF is developed and found to predict the data with an average error of ±15.6%.

Review on Thermal-Hydraulic Characteristics of Nuclear Reactors Under Ocean Conditions

The thermal-hydraulic characteristics of nuclear reactors under ocean conditions are significant for reactor safety and reliability. A large number of experiments concerning this issue have been done. However, the focus of these experiments has been mainly on dynamical systems in submarines and aircraft carriers. With the development of floating nuclear power plants (FNPPs)—which are used to provide power for remote areas or ocean platforms—more studies on FNPPs are needed. The differences between FNPPs and dynamical systems bring new challenges to research of thermal-hydraulic characteristics in a nuclear reactor under ocean conditions.Many experimental studies on natural circulation and forced circulation under ocean conditions are based on tube and rectangular channels. The effects of ocean motions on friction, flow instability, heat transfer, and critical heat flux (CHF) have been investigated. The intensity of ocean motions and their driving head have a huge impact on pressure drop, heat transfer, and flow instability, especially when the driving head is small. Flow oscillations induced by ocean motions can overlap with thermal-induced flow instability. This resonance effect can cause potential harm to systems. CHF under ocean conditions is mainly dependent on motion types and the CHF mechanism.There are challenges with this research. First, conclusions obtained from a simple channel cannot always extend to rod bundle systems. Second, studies of ocean conditions, which always use one- or two-dimensional movements, are relatively simple and cannot reflect complicated ocean environments. Third, there is a lack of comparison between natural frequency of two-phase flow instability and frequency of ocean waves, and this may cause unwanted problems due to resonance or coupling behaviors. Fourth, CHF experiments that are performed to examine CHF events under ocean conditions are far from real reactor conditions where CHF events occur under high-temperature and high-pressure conditions, with complicated three-dimensional geometry, with open channel environments, and sometimes even under the influence of mixing vane grids [for most small modular reactor designs]. Last, there has been no work on the length effect, nonuniform heating, and mixing effect induced by spacer grids, which differentiate FNPP thermal-hydraulic characteristics from the other applications. A series of suggestions are provided for future work.

Critical heat flux experiments and a post-CHF heat transfer analysis using 2D inverse heat transfer

Abstract The critical heat flux (CHF) is one of the main thermal-hydraulics safety limits in water-cooled reactors. At CHF, a film of vapor is formed on the heated wall and the wall temperature can sharply increase in a short period of time, which may lead to damage of the heater surface. In spite of many studies, the CHF mechanisms are still not well understood due to the complexity of this two-phase phenomenon and its dependence on local thermal-hydraulic conditions. Two-phase boiling experiments have been conducted to observe CHF and post-CHF behavior in prototypical conditions for small modular reactors (mass flux: 560–1570 kg/m2.s, pressure: 7.6–16.4 MPa, inlet subcooling: 160–440 kJ/kg) in a 2 × 2 square rod bundle geometry with a chopped-cosine power profile at the University of Wisconsin-Madison. The fuel rods temperature history during and following CHF were used in the solution of the two-dimension inverse heat transfer problem to estimate the wall temperatures and wall heat fluxes. This analysis showed different transient boiling curves for different flow regimes that differ qualitatively and quantitatively from the typical boiling curve considered in steady-state two-phase heat transfer analysis. The experiments suggest that the inverse heat transfer analysis approach can be used to estimate the post-CHF heat transfer and to better understand the CHF mechanisms at high pressure typical of water-cooled reactors.

Comparison of pool boiling CHF of a polished copper block and carbon steel block on a declined slope

ABSTRACT This study conducts a critical heat flux (CHF) experiment on a carbon steel block, and the block is positioned on slope that is declined at angles of 5° and 10°. The results of the carbon steel block experiment were then analyzed and compared with the results obtained from a copper block experiment that had been conducted previously at the same test facility. The comparison showed that several different types of phenomena had occurred, and the carbon steel block CHF at both 5° and 10° was much lower than that of the copper block. Detailed images of the heating surface of each material were acquired by a high-speed camera under different heat fluxes and analyzed. The carbon steel block surface generates more bubbles compared to the copper block under the same heat flux, which indicates that the carbon steel block should have a large number of nucleation sites. This causes a higher CHF. Finally, several existing theories on CHF mechanisms were also analyzed in an attempt to explain the difference of copper and carbon steel. It seemed that the contact angle alone was not sufficient to explain the large CHF decrease in the carbon steel block.