Supercritical Water Choking Flow Experiments Through a Convergent-Divergent Test Section

2020 ◽  
Author(s):  
Alberto Teyssedou ◽  
Altan Muftuoglu ◽  
Akila Hidouche
Keyword(s):  
Critical Temperature ◽  
Temperature Difference ◽  
Supercritical Water ◽  
Test Section ◽  
Critical Mass ◽  
Fluid Temperature ◽  
Flow Data ◽  
Bulk Fluid ◽  
Choked Flow ◽  
Discharge Pressure

Abstract The heat transport system of GEN-IV Supercritical Water-Cooled Nuclear Reactors (SCWRs) will operate at pressures close to 25 MPa and outlet temperatures of up to 898 K (625°C). The design and safety analyses of this type of reactors still necessitate amongst others, experimental information and validation of critical (choked) flows models of water above the thermodynamic critical state. Up to now, choked flow data were collected at atmospheric discharge pressure conditions, without changing the discharge pressure to verify the occurrence of choking flow; in most of the cases, using fluids different from water. This paper presents experimental supercritical water choking flow data collected by using a convergent-divergent test section by changing the discharge pressure to verify the occurrence of choked flow. The critical mass flux is presented as a function of the temperature difference between a pseudo-critical temperature and the bulk fluid temperature. This representation allows us to assess similar experiments performed by using different test sections. Hence, a comparison of actual data with those previously obtained using 1.0 mm and 1.4 mm diameter sharp-edged orifices, shows peculiar differences. The actual experiments were limited by very low values of choking mass flow rates. Furthermore, in some cases, it was observed the presence of an increase in the discharge pressure that seems to indicate the existence of shock-wave structures. We are also able to estimate a pseudo-critical temperature difference below which choking flow systematically occurs.

Forced Convection Heat Transfer From a Cylinder in Carbon Dioxide Near the Thermodynamic Critical Point

1971 ◽  
Vol 93 (3) ◽  
pp. 290-296 ◽  
Author(s):  
J. R. Green ◽  
E. G. Hauptmann
Keyword(s):  
Heat Transfer ◽  
Critical Temperature ◽  
Forced Convection ◽  
Temperature Difference ◽  
Forced Flow ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Transfer Rates ◽  
Heated Cylinder ◽  
Increasing Temperature

In an attempt to determine the heat transfer rates in forced flow normal to a heated cylinder and to provide some insight into the mechanisms in heat transfer in the critical region, heat transfer rates have been measured for both free and forced flow of supercritical carbon dioxide normal to a horizontal heated cylinder. The 0.006-in-dia cylinder was held at various constant temperatures by a feedback circuit. The effects of bulk fluid temperature, bulk fluid pressure, and surface temperature were studied for a range of bulk fluid temperatures and pressures from 0.8 to 1.4 times the critical temperature and pressure, and free-stream velocities from 0 to 3 fps. The temperature difference between the heated cylinder and the bulk fluid was varied from 1 to 300 deg F. Several photographs of the flow field are presented. In a supercritical fluid the heat transfer rate increases smoothly and monotonically with increasing temperature difference, increasing velocity, and increasing pressure. In fluid with the bulk temperature below the pseudo-critical temperature the heat transfer coefficient shows large peaks when the cylinder temperature is near the pseudo-critical temperature. The heat transfer coefficient decreases with increasing temperature difference when the bulk fluid temperature is above the pseudo-critical temperature. Supercritical forced convection does not exhibit the characteristic maximum in heat transfer rate shown in forced-flow nucleate boiling. Heat transfer rates at larger temperature differences are very similar in forced-flow film boiling and supercritical forced-flow heat transfer. With this horizontal constant-temperature cylinder, no “bubble-like” or “boiling-like” mechanisms of heat transfer were observed in supercritical free or forced convection.

Investigation of Characteristic-Temperature Approaches for Heat-Transfer Correlations at Supercritical Conditions

2013 ◽  
Author(s):  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Wall Temperature ◽  
Characteristic Temperature ◽  
Fluid Temperature ◽  
Transfer Coefficients ◽  
Supercritical Conditions ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients ◽  
Bare Tube

It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) and high coolant temperatures (350–625°C). In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach. A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water, at supercritical pressures. These bare-tube-based correlations are available in various literature sources. There have been a number of methods applied to correlate heat transfer data. The most conventional approach, which accounts for property variations in the data, is to modify the classical Dittus-Boelter equation for forced convection. However, analysis and comparison of these correlations has shown that differences in HTC values can be up to several hundred percent. In general, the familiar correlations of Dittus-Boelter and Bishop et al. have used the bulk-fluid temperature approach for characteristic temperature properties evaluations. However, at high heat fluxes, fluid near the tube-wall will have a temperature close to that of the wall temperature. This might be significantly different from the bulk-fluid temperature. Therefore, another approach can be used based on the wall temperature as the characteristic temperature. The Swenson et al. correlation is based upon this approach. Finally, a third approach has been considered in which the film-temperature is used as the characteristic temperature (Tf = (Tw+Tb) / 2). McAdams et al. based their correlation for annuli on this approach. Therefore, the objective of this paper is to evaluate the three characteristic temperature approaches, (1) Bulk-fluid temperature approach; (2) Wall-temperature approach; and (3) Film-temperature approach, and determine which characteristic temperature method can most accurately predict supercritical water heat transfer coefficients. Both classical correlations and more recently developed correlations are considered in this investigation.

Supercritical Water Heat Transfer in a Vertical Bare Tube

2008 ◽  
Author(s):  
Yevgeniy Gospodinov ◽  
Sarah Mokry ◽  
Pavel Kirillov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Convective Heat Transfer ◽  
Convective Heat ◽  
Supercritical Water ◽  
Wall Temperature ◽  
Test Section ◽  
Bulk Fluid ◽  
Transfer Data ◽  
Normal Heat ◽  
Bare Tube

This paper presents selected results on heat transfer to supercritical water flowing upward in a 4-m-long vertical bare tube. Supercritical water heat-transfer data were obtained at pressures of about 24 MPa, mass fluxes of 200 – 1500 kg/m2s, heat fluxes up to 884 kW/m2 and inlet temperatures from 320 to 350°C for several combinations of wall and bulk-fluid temperatures that were below, at or above the pseudocritical temperature. In general, the experiments confirmed that there are three heat-transfer regimes for forced convective heat transfer to water flowing inside tubes at supercritical pressures: (1) normal heat-transfer regime characterized in general with heat transfer coefficients (HTCs) similar to those of subcritical convective heat transfer far from critical or pseudocritical regions, which are calculated according to the Dittus-Boelter type correlations; (2) deteriorated heat-transfer regime with lower values of the HTC and hence higher values of wall temperature within some part of a test section compared to those of the normal heat-transfer regime; and (3) improved heat-transfer regime with higher values of the HTC and hence lower values of wall temperature within some part of a test section compared to those of normal heat-transfer regime. These new heat-transfer data are applicable as a reference dataset for future comparison with supercritical-water bundle data and for a verification of scaling parameters between water and modeling fluids. Also, these HTC data were compared to those calculated with the original Dittus-Boelter and Bishop et al. correlations. The comparison showed that the Bishop et al. correlation, which uses the cross-section average Prandtl number, represents HTC profiles more correctly along the heated length of the tube than the Dittus-Boelter correlation. In general, the Bishop et al. correlation shows a good agreement with the experimental HTCs outside the pseudocritical region, however, overpredicts the experimental HTCs within the pseudocritical region. The Dittus-Boelter correlation can also predict the experimental HTCs outside the pseudocritical region, but deviates significantly from the experimental data within the pseudocritical region. It should be noted that both these correlations cannot be used for a prediction of HTCs within the deteriorated heat-transfer regime.

Experimental Heat Transfer in an Annular Channel and 3-Rod Bundle Cooled With Upward Flow of Supercritical Water

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyi ◽  
Kh. Sidawi ◽  
I. L. Pioro ◽  
A. Eu. Koloskov
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Annular Channel ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Experimental Heat ◽  
Rod Bundle ◽  
Fuel Bundle ◽  
Experimental Heat Transfer ◽  
Heated Length

There have been relatively few publications detailing heat transfer to supercritical water (SCW) flowing through a channel with a bundle or just with a single rod (annular channel) as compared to heat transfer to SCW in bare tubes. In the present paper, results of experimental heat transfer to SCW flowing upward in an annular channel with a heated rod equipped with four helical ribs and a 3-rod bundle (rods are also equipped with four helical ribs) are discussed. The experimental results include bulk-fluid-temperature, wall-temperature, and heat-transfer-coefficient (HTC) profiles along the heated length (485 mm) for these flow geometries. Data obtained from this study could be applicable as a reference estimation of heat transfer for future fuel-bundle designs.

Investigation of ATHLET System Code for Supercritical Water Applications

2014 ◽  
Author(s):  
J. Samuel ◽  
G. Lerchl ◽  
G. D. Harvel ◽  
I. Pioro
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
High Pressures ◽  
Computer Code ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Water Reactors ◽  
First Time

SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV nuclear-reactor concepts. They are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. However, codes that have been used to study the phenomena analytically have not been validated for supercritical water. The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid- and transport-properties packages, thus enabling a transition from subcritical to supercritical fluid states. This extension needs to be validated using experimental data. In this work, the applicability of ATHLET code to predict supercritical-water behaviour in various heat-transfer conditions is assessed. Several well-known heat-transfer correlations for supercritical fluids are added to the code and applied for the first time in ATHLET simulations of experiments. A numerical model in ATHLET is created to represent an experimental test section and results for the heat transfer coefficient, bulk fluid temperature, and the tube inside-wall temperature are compared with the experimental data. The results from the ATHLET simulations are promising in the Normal and Enhanced Heat-Transfer Regimes. However, important phenomena such as Deteriorated Heat Transfer are currently not accurately predicted. While ATHLET can be used to develop preliminary design solutions for SCWRs, a significant effort in analysis of experimental work is required to make further advancements in the use of ATHLET for SCW applications.

Experimental Study of Heat Transfer in a 7-Element Bundle Cooled With Supercritical Freon-12

2012 ◽  
Author(s):  
G. Richards ◽  
J. Samuel ◽  
A. S. Shelegov ◽  
P. L. Kirillov ◽  
I. L. Pioro ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Mass Flux ◽  
Test Section ◽  
High Pressures ◽  
Natural Circulation ◽  
Heat Fluxes ◽  
Bulk Fluid ◽  
Test Loop

Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are technical difficulties in testing and experimental costs at high pressures, temperatures and heat fluxes. Also, there are only a few SCW experimental setups currently in the world capable of providing data. SuperCritical Water-cooled nuclear Reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach uses modeling fluids such as carbon dioxide and refrigerants instead of water are practical. In particularly, experiments in supercritical refrigerant-cooled bundles can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97°C. A set of experimental data obtained at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) in a vertically-oriented bundle cooled with supercritical R-12 was analyzed. This dataset consisted of 20 runs. The test section was 7-element bundle installed in a hexagonal flow channel with 3 grid spacers. Data was collected at pressures of approximately 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values of mass flux were ranged from 400 to 1320 kg/m2s and inlet temperatures ranged from 72 to 120°C. The test section consisted of fuel-element simulators that were 9.5 mm in OD with the total heated length of about 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 thermocouples. Analysis of the data has confirmed that there are three distinct heat-transfer regimes for forced convention in supercritical fluids: 1) Normal heat transfer; 2) Deteriorated heat transfer characterized with higher than expected wall temperatures; and 3) Enhanced heat transfer characterized with lower than expected wall temperatures. It was also confirmed that the effects of spacers are evident which was previously observed in sub-critical experimental data. Further analysis needs to be conducted on the deteriorated heat transfer phenomena for low mass flux cases which represent accident scenarios. This can be done by designing a natural circulation experimental test loop.

Study on Deteriorated Heat Transfer in Upward Flow of Supercritical Water in a 1-M Vertical Bare Tube

2021 ◽  
Author(s):  
N. Dort-Goltz ◽  
I. Pioro ◽  
J. McKellar
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Fluid Temperature ◽  
Upward Flow ◽  
Transfer Coefficients ◽  
Safe Operation ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients ◽  
Heated Length

Abstract SuperCritical Water-cooled Reactors (SCWRs) represent potential improvements over traditional water-cooled reactors in many respects, including thermal efficiency. These reactors are still under development, however, thermalhydraulics data needed for their design are lacking. Experimentation is complex and costly. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller hydraulic-equivalent diameters. As a first step, tests have been conducted by others on heat transfer in short, vertical bare tubes cooled with the upward flow of SCW. The objective of this work is to analyze that collected data with particular attention to the Deteriorated Heat Transfer (DHT) regime. The DHT regime is characterized by reduced Heat Transfer Coefficients (HTCs) and consequently increased wall temperatures. As such, it represents a hazard to the safe operation of a Nuclear Power Plant (NPP). The results of this analysis indicate that DHT did occur in each of the tests analyzed, often seen as a gradual decrease in HTC along the heated length, but occasionally as a sharp “dip”. The DHT can occur along the heated length, when the bulk-fluid temperature is close to or within the pseudocritical region. The results also confirmed that the Dittus-Boelter correlation does not adequately predict HTCs within the pseudocritical region. Two other applied correlations (Gupta et al. and Mokry et al.) performed better, but neither was able to predict the occurrence of the DHT. The results of this analysis will be of use to designers and developers of SCWRs, and can help to plan future experiments.

Improvement of Supercritical Water Heat-Transfer Correlations for Vertical Bare Tubes

2014 ◽  
Author(s):  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Operating Conditions ◽  
Entrance Region ◽  
Fluid Temperature ◽  
Supercritical Conditions ◽  
Bulk Fluid ◽  
Wide Range ◽  
Heat Transfer Correlations ◽  
Region Effect

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. These high temperature, high pressure reactors will have much higher operating parameters compared to current Nuclear Power Plants (NPPs) (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. In support of developing these SCWRs, studies are currently being conducted for heat transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer at supercritical conditions from power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach. A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water at supercritical pressures. There have been a number of methods applied to correlate heat transfer data. The most conventional approach is to modify the classical Dittus-Boelter correlation for forced convection. The Bishop et al. correlation is an example of this type modification with an addition of an entrance-region term. The Mokry et al. correlation (2009) was developed as a Dittus-Boelter-type correlation with thermophysical properties taken at a bulk-fluid temperature. The derived correlation has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in normal and improved heat-transfer regimes. This correlation has an uncertainty of about ±25% for HTC values and about ±15% for calculated wall temperature. However, this correlation does not take into account the entrance-region effect. The objective of this paper is an investigation of the entrance-region effect to be incorporated into the proposed Mokry et al. correlation (2009) in an attempt to further improve its accuracy.

scholarly journals Study of the Performance of a Novel Radiator with Three Inlets and One Outlet Based on Topology Optimization

Micromachines ◽  
2021 ◽  
Vol 12 (6) ◽  
pp. 594
Author(s):  
Tao Zhou ◽  
Bingchao Chen ◽  
Huanling Liu
Keyword(s):  
Topology Optimization ◽  
Heat Exchanger ◽  
Temperature Difference ◽  
Heat Exchangers ◽  
Exchange Capacity ◽  
Numerical Results ◽  
Fluid Temperature ◽  
Channel Depth ◽  
Goal Model ◽  
Flow And Heat Transfer

In recent years, in order to obtain a radiator with strong heat exchange capacity, researchers have proposed a lot of heat exchangers to improve heat exchange capacity significantly. However, the cooling abilities of heat exchangers designed by traditional design methods is limited even if the geometric parameters are optimized at the same time. However, using topology optimization to design heat exchangers can overcome this design limitation. Furthermore, researchers have used topology optimization theory to designed one-to-one and many-to-many inlet and outlet heat exchangers because it can effectively increase the heat dissipation rate. In particular, it can further decrease the hot-spot temperature for many-to-many inlet and outlet heat exchangers. Therefore, this article proposes novel heat exchangers with three inlets and one outlet designed by topology optimization to decrease the fluid temperature at the outlet. Subsequently, the effect of the channel depth on the heat exchanger design is also studied. The results show that the type of exchanger varies with the channel depth, and there exists a critical depth value for obtaining the minimum substrate temperature difference. Then, the flow and heat transfer performance of the heat exchangers are numerically investigated. The numerical results show that the heat exchanger derived by topology optimization with the minimum temperature difference as the goal (Model-2) is the best design for flow and heat transfer performance compared to other heat sink designs, including the heat exchanger derived by topology optimization having the average temperature as the goal (Model-1) and conventional straight channels (Model-3). The temperature difference of Model-1 can be reduced by 37.5%, and that of Model-2 can be decreased by 62.5% compared to Model-3. Compared with Model-3, the thermal resistance of Model-1 can be reduced by 21.86%, while that of Model-2 can be decreased by 47.99%. At room temperature, we carried out the forced convention experimental test for Model-2 to measure its physical parameters (temperature, pressure drop) to verify the numerical results. The error of the average wall temperature between experimental results and simulation results is within 2.6 K, while that of the fluid temperature between the experimental and simulation results is within 1.4 K, and the maximum deviation of the measured Nu and simulated Nu was less than 5%. This indicated that the numerical results agreed well with the experimental results.

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