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.

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 to Supercritical Pressure Water Flowing in a 2×2 Rod Bundle

2014 ◽  
Author(s):  
Han Wang ◽  
Qincheng Bi ◽  
Linchuan Wang ◽  
Haicai Lv ◽  
Laurence K. H. Leung
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Heat Flux ◽  
Pressure Drop ◽  
Wall Temperature ◽  
Mass Flux ◽  
Heat Fluxes ◽  
Outer Diameter ◽  
Square Channel ◽  
Rod Bundle

An experiment has recently been performed at Xi’an Jiaotong University to study the wall temperature and pressure drop at supercritical pressures with upward flow of water inside a 2×2 rod bundle. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 350–1000 kg/m2s and heat flux on the rod surface of 200–1000 kW/m2. According to the experimental data, it was found that the circumferential wall temperature distribution of a heated rod is not uniform. The temperature difference between the maximum and the minimum varies with heat flux and/or mass flux. Heat transfer characteristics of supercritical water in bundle were discussed with respect to various heat fluxes. The effect of heat flux on heat transfer in rod bundles is similar with that in tubes or annuli. In addition, flow resistance reflected in the form of pressure loss has also been studied. Experimental results showed that the total pressure drop increases with bulk enthalpy and mass flux. Four heat transfer correlations developed for supercritical pressures water were compared with the present test data. Predictions of Jackson correlation agrees closely with the experimental data.

Numerical Investigation of Heat and Mass Flux Effects on Heat Transfer Characteristics of Supercritical Water in an Upward Flow Vertical Tube

2014 ◽  
Vol 592-594 ◽  
pp. 1667-1671
Author(s):  
T. Vinoth ◽  
K. Karuppasamy ◽  
D. Santhosh Kumar ◽  
R. Dhanuskodi
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Heat Flux ◽  
Supercritical Water ◽  
Mass Flux ◽  
Vertical Tube ◽  
Upward Flow ◽  
Heat Transfer Characteristics ◽  
Ansys Fluent ◽  
Transfer Characteristics

In the present work, the heat transfer characteristics of supercritical pressure water are numerically investigated in an upward flow vertical smooth tube. The numerical simulations are carried out by using Ansys-Fluent solver. The objective of the present work is to investigate the effect of heat flux and mass flux on heat transfer characteristics in supercritical water. In order to perform numerical simulation, experimental data of Mokryet al.[2] is considered. Various simulations were carried out for the inlet parameters of temperature 350°C, pressure 240bar; heat flux values ranging from 190 to 884kW/m2and mass flux values ranging from 498 to 1499kg/m2s. Based on the available parameters of heat flux and mass flux, they are segregated as groups with heat flux to mass flux ratios of 0.39 and 0.67. According to computational data, the heat transfer enhancement and heat transfer deterioration phenomenon of supercritical water were analyzed and based on the comparison with experimental data; their occurrence and mechanism were addressed.

Comparison of Selected Forced-Convection Supercritical-Water Heat-Transfer Correlations for Vertical Bare Tubes

2010 ◽  
Author(s):  
Amjad Farah ◽  
Krysten King ◽  
Sahil Gupta ◽  
Sarah Mokry ◽  
Wargha Peiman ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Heat Exchangers ◽  
Power Plants ◽  
Heat Fluxes ◽  
Operating Conditions ◽  
Transfer Coefficients ◽  
Heat Transfer Correlations ◽  
Bare Tube

This paper presents an extensive study of heat-transfer correlations applicable to supercritical-water flow in vertical bare tubes. A comprehensive dataset was collected from 33 papers by 27 authors, including more than 125 graphs and wide ranges of parameters. The parameters ranges were as follows: pressures 22.5–34.5 MPa, inlet temperatures 85–350°C, mass fluxes 250–3400 kg/m2s, heat fluxes 75–5,400 kW/m2), tube heated lengths 0.6–27.4 m, and tube inside diameters 2–36 mm. This combined dataset was then investigated and analyzed. Heat Transfer Coefficients (HTCs) and wall temperatures were calculated using various existing correlations and compared to the corresponding experimental results. Three correlations were used in this comparison: Bishop et al., Mokry et al. and modified Swenson et al. The main objective of this study was to select the best supercritical-water bare-tube correlation for HTC calculations in: 1) fuel bundles of SuperCritical Water-cooled Reactors (SCWRs) as a preliminary and conservative approach; 2) heat exchangers in case of indirect-cycle SCW Nuclear Power Plants (NPPs); and 3) heat exchangers in case of hydrogen co-generation at SCW NPPs from SCW side. From the beginning, all these three correlations were compared to the Kirillov et al. vertical bare-tube dataset. However, this dataset has a limited range of operating conditions in terms of a pressure (only one pressure value of 24 MPa) and one inside diameter (only 10 mm). Therefore, these correlations were compared with other datasets, which have a much wider range of operating conditions. The comparison showed that in most cases, the Bishop et al. correlation deviates significantly from the experimental data within the pseudocritical region and actually, underestimates the temperature at most times. On the other hand, the Mokry et al. and modified Swenson et al. correlations showed a relatively better fit within the most operating conditions. In general, the modified Swenson et al. correlation showed slightly better fit with the experimental data than other two correlations.

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.

Investigation on Heat Transfer Characteristics in the Test Section With Non-Uniform Heat Flux Distribution Under Natural Circulation

2017 ◽  
Author(s):  
Qiang Wang ◽  
Puzhen Gao ◽  
Xianbing Chen ◽  
Zhongyi Wang ◽  
Ying Huang
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Heat Flux ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Test Section ◽  
Flux Distribution ◽  
Natural Circulation ◽  
Heat Transfer Characteristics ◽  
Transfer Characteristics

Natural circulation served as an indispensable part of nuclear, attracted much more attentions in recent years. It does not need a pump to provide power. The operating principle of natural circulation caused its complexity in analysis process. It was still a difficult issue to reveal the law of natural circulation accurately. Many experiments and calculations had to be conducted to study the basic physical regulation. This paper concentrated upon the heat transfer characteristics in the test section with two different types of heat flux distribution. The two types of heating flux distribution in the test section were linear and chopped cosine along axial direction. Based on a natural circulation experimental facility, physical models and mathematic models were established. RELAP5 code was used to calculate the thermal hydraulic state of natural circulation loop. The variation of heat transfer coefficient along flow direction was different. It was tightly related to heat flux. Some relevant experiments were conducted in many different conditions and steady sate experimental data were achieved to verified theoretical calculations. Experimental data, such as water temperature, wall temperature and flow rate were recorded when the system is stable. The heat transfer coefficients were calculated according to the experimental data. The factors that affected the heat transfer characteristics of natural circulation were analyzed by comparing the heat transfer coefficient under different conditions. The heat transfer coefficient was calculated according to the empirical correlations as well. After a series of analysis, the results indicated heat transfer coefficient had an obvious difference, which influenced ability of natural circulation. Comparing with experimental data, the evaluation of different empirical correlations was conducted in two test sections. Some empirical correlations turned out to be suitable for the estimation of heat transfer in experiment facility. The increase of heat flux could enhance heat transfer process in the two test section under low pressure. Average heat transfer coefficient increased with the decrease of inlet subcooling degree. The system pressure effected the heat transfer characteristics of natural circulation as well. The increase of mass flux would promote heat transfer while the level was different. RELAP5 had a great agreement with experimental data in single phase flow. Natural circulation ability was influenced by the position of average heat source center, which was slightly different in the research objects. The research would lend strong empirical support to the guideline of experiment and subsequence study in natural circulation.

Flow and Heat Transfer Characteristics of a Natural Circulation Evaporative Cooling System for Electronic Components

2004 ◽  
Vol 126 (3) ◽  
pp. 317-324 ◽  
Author(s):  
Hiroshi Honda ◽  
ZhengGuo Zhang ◽  
Nobuo Takata
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Test Section ◽  
Cooling System ◽  
Natural Circulation ◽  
Electronic Components ◽  
Heat Transfer Characteristics ◽  
Transfer Characteristics ◽  
Flow And Heat Transfer ◽  
Test Loop

Experiments were conducted to study the flow and heat transfer characteristics of a natural circulation liquid cooling system for electronic components. The test loop consisted of a horizontal test section, a horizontal evaporator, a vertical tube, a horizontal condenser, a rubber bag attached at the exit of the condenser, a downcomer, a mass flow meter, and a liquid subcooler. The loop height H was set at either 250 or 450 mm. FC-72 was filled in the test loop up to some level of loop height and the upper part was filled with air. During the operation of the cooling system, the rubber bag expanded and stored the mixture of generated vapor and air. Thus the inner pressure was maintained at atmospheric pressure. In the test section, a silicon chip with dimensions of 10×10×0.5 mm3 was attached at the bottom surface of a horizontal duct with dimensions of 10×14 mm2. A smooth chip and four chips with square micro-pin-fins with 150 to 300 μm in fin height were tested. The duct height s was set at 10 mm for most of the experiments. The cases of s=1 and 25 mm were also tested for one of the micro-pin-finned chips. For each H, the average flow rate of FC-72 was correlated well as a function of the static pressure difference between the two vertical tubes. All chips showed the boiling curve similar to that for pool boiling except that the critical heat flux was lower for the natural circulation loop. For all chips tested, the maximum allowable heat flux qmax increased monotonically with increasing liquid subcooling ΔTsub. Comparison of the results for s=1, 10 and 25 mm revealed that the highest qmax was obtained with s=10 mm. The values of qmax for s=1 and 25 mm were 36–46% and 87–90% of that for s=10 mm, respectively. The maximum value of qmax=56 W/cm2 was obtained by one of the micro-pin-finned chips at s=10 mm and ΔTsub=35 K.

Assessment of FLUENT Code as a Tool for SCW Heat Transfer Analysis

2013 ◽  
Author(s):  
Amjad Farah ◽  
Glenn Harvel ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Supercritical Water ◽  
High Pressures ◽  
Stokes Equations ◽  
Heat Transfer Analysis ◽  
Model Parameters ◽  
Transfer Coefficients ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients

Generation-IV SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. The behavior of supercritical water however, is not well understood and most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations which do not capture the multi-dimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. Computational Fluid Dynamics (CFD) is a numerical approach to model fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system. In this work, the CFD code, FLUENT-12, is used with associated software such as Gambit and NIST REFPROP to predict the Heat Transfer Coefficients at the wall and corresponding wall temperature profiles inside vertical bare tubes with SuperCritical Water (SCW) as the cooling medium. The numerical results are compared with experimental data and 1-D models represented by existing empirical correlations. Analysis of the individual heat-transfer regimes is conducted using an axisymmetric 2-D model of tubes of various lengths and composed of different nodalizations along the heated length. Wall temperatures and heat transfer coefficients were analyzed to select the best model for each region (below, at and above the pseudocritical region). Two turbulent models were used in the process: k-ε and k-ω, with variations in the sub-model parameters such as viscous heating, thermal effects, and low-Reynolds number correction. Results of the analysis show a fit of ±10% for the wall temperatures using the SST k-ω model in the deteriorated heat transfer regime and less than ±5% for the normal heat transfer regime. The accuracy of the model is higher than any empirical correlation tested in the mentioned regimes, and provides additional information about the multidimensional effects between the bulk-fluid and wall temperatures.

Preliminary Investigation of Heat-Transfer Correlation for Upward Flow of CO2 at Supercritical Pressure

2013 ◽  
Author(s):  
Eugene Saltanov ◽  
Igor Pioro ◽  
Glenn Harvel
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Preliminary Investigation ◽  
Convection Heat Transfer ◽  
Theoretical Models ◽  
Supercritical Pressure ◽  
Operational Parameters ◽  
Bulk Fluid ◽  
Wide Range

The development of Generation IV nuclear reactors is an ongoing worldwide interdisciplinary effort. Canada is involved with the development of the SuperCritical Water-cooled Reactor (SCWR). One of the numerous engineering challenges associated with this development is to ensure intensive and stable heat transfer to SuperCritical Water (SCW) in a core. It is very important to develop sophisticated theoretical models to improving understanding of physical processes behind forced-convective heat transfer and its deterioration in near-critical region. However, not all turbulent models implemented in Computational Fluid Dynamics (CFD) codes are applicable to heat transfer at supercritical pressures. Additionally these codes should be first tuned on the basis of experimental data and, after that, used in similar calculations. Therefore, there is still a great reliance on 1D heat-transfer correlations for preliminary calculations. Performing experiments in SCW is extremely expensive due to the high temperatures and pressures involved. Therefore, it is reasonable to study the general properties of SuperCritical Fluids (SCF) by running experiments with modelling fluids; for example, carbon dioxide, which is widely used as a modeling fluid. In view of this, there is a need to compile a large database that will include experimental data on forced-convection heat-transfer to SuperCritical (SC) CO2 within a wide range of operational parameters. To date, a part of this database has been produced. It contains the bulk-fluid and wall temperatures of CO2 flowing upward in a vertical bare tube. CO2 is at supercritical pressure and the bulk-fluid and wall temperatures are below or above the pseudocritical temperature at the inlet of the test-section. Therefore, the objectives of this paper are: 1) to propose a correlation in a standard non-dimensional form, which will generalize data within ±25%; 2) to compare this correlation with the most recent, as well as previous 1D correlations for SC CO2 and SCW (appropriate scaling is performed).

Assessment of Heat-Transfer Correlations Against Experimental Data Obtained With Supercritical Water in Vertical Annuli

2012 ◽  
Author(s):  
Zhendong Yang ◽  
Qincheng Bi ◽  
Han Wang ◽  
Gang Wu ◽  
Laurence K. H. Leung
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Test Section ◽  
High Mass ◽  
Outer Diameter ◽  
Transfer Coefficients ◽  
Heat Transfer Coefficients ◽  
Normal Heat ◽  
Transfer Region

Eleven correlations proposed for supercritical heat-transfer coefficients were assessed against a set of experimental data obtained recently with supercritical water flow in a vertical annular test section at Xi’an Jiaotong University. The inner heated rod of the test section had an outer diameter of 8 mm, while the outer unheated tube had an inner diameter of 16 mm (resulting in a gap size of 4 mm). The experiment covered pressure range from 23 to 28 MPa, mass-flux range from 350 to 1000 kg/m2s, and heat-flux range from 200 to 1000 kW/m2. The assessment shows relatively good agreement between predicted and experimental heat-transfer coefficients for several correlations. Some discrepancies have been observed at the region where deteriorated heat transfer, and are attributed to the modified Dittus-Boelter formulation that captures mainly the normal heat-transfer region. Overall, the Dittus-Boelter correlation is shown applicable only for the normal heat-transfer region, and significantly overpredicts the heat-transfer coefficient at the deteriorated heat-transfer region. The correlation of Bishop et al. appears valid for the current experimental database, particularly for high mass fluxes.

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