scholarly journals Heat Transfer Correlations for Supercritical Water in Vertically Upward Tubes

2020 ◽  
Author(s):  
Huixiong Li ◽  
Xiangfei Kong ◽  
Xianliang Lei ◽  
Qian Zhang ◽  
Changjiang Liao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Heat Transfer Correlations

Comparison of Three-Rod Bundle Data With Existing Heat-Transfer Correlations for Bare Vertical Tubes

2010 ◽  
Author(s):  
Krysten King ◽  
Amjad Farah ◽  
Sahil Gupta ◽  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Annular Channel ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Nuclear Systems ◽  
Heat Transfer Correlations ◽  
Vertical Tubes ◽  
Bare Tube

Many heat-transfer correlations exist for bare tubes cooled with SuperCritical Water (SCW). However, there is very few correlations that describe SCW heat transfer in bundles. Due to the lack of extensive data on bundles, a limited dataset on heat transfer in a SCW-cooled bundle was studied and analyzed using existing bare-tube correlations to find the best-fit correlation. This dataset was obtained by Razumovskiy et al. (National Technical University of Ukraine “KPI”) in SCW flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within a range from 800 to 3000 kg/m2s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2. The objective of this study is to compare bare-tube SCW heat-transfer correlations with the data on 1- and 3-rod bundles. This work is in support of SuperCritical Water-cooled Reactors (SCWRs) as one of the six concepts of Generation-IV nuclear systems. SCWRs will operate at pressures of ∼25MPa and inlet temperatures of 350°C.

Analysis of Updated SuperCritical Water Heat Transfer Correlations for Vertical Bare Tubes

2010 ◽  
Author(s):  
Sarah Mokry ◽  
Sahil Gupta ◽  
Amjad Farah ◽  
Krysten King ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Thermophysical Properties ◽  
Vertical Tube ◽  
Large Set ◽  
Conservative Approach ◽  
Bulk Fluid ◽  
Heat Transfer Correlations ◽  
Vertical Tubes ◽  
Bare Tube

In support of developing SuperCritical Water-cooled Reactors (SCWRs), studies are currently being conducted for heat-transfer at supercritical conditions. This paper presents an analysis of heat-transfer to SuperCritical Water (SCW) flowing in bare vertical tubes as a first step towards thermohydraulic calculations in a fuel-channel. A large set of experimental data, obtained in Russia, was analyzed. Two updated heat-transfer correlations for forced convective heat transfer in the normal heat transfer regime to SCW flowing in a bare vertical tube were developed. It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) with high coolant temperatures (350–625°C). 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. The analyzed experimental dataset was obtained for SCW flowing upward in a 4-m-long vertical bare tube. The data was collected at pressures of about 24 MPa for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values for mass flux ranged from 200–1500 kg/m2s, for heat flux up to 1250 kW/m2 and inlet temperatures from 320–350°C. The Mokry et al. correlation was developed as a Dittus-Boelter-type correlation, with thermophysical properties taken at bulk-fluid temperatures. Alternatively, the Gupta et al. correlation was developed based on the Swenson et al. approach, where the majority of thermophysical properties are taken at the wall temperature. An analysis of the two updated heat-transfer correlations is presented in this paper. Both correlations demonstrated a good fit (±25% for Heat Transfer Coefficient (HTC) values and ±15% for calculated wall temperatures) for the analyzed dataset. Thus, these correlations can be used for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for SCW heat exchangers, for future comparisons with other independent datasets and for the verification of computer codes for SCWR core thermohydraulics.

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.

Development of Supercritical Water Heat-Transfer Correlation for Vertically-Upward Internally-Ribbed Tubes

2014 ◽  
Author(s):  
Weiqiang Zhang ◽  
Huixiong Li ◽  
Qing Zhang ◽  
Yifang Zhang ◽  
Tai Wang
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Large Deviation ◽  
Experimental Studies ◽  
Great Influence ◽  
Dimensionless Numbers ◽  
Experimental Database ◽  
Heat Transfer Correlations

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed. This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry. The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.

Comparison of Heat-Transfer Correlations Obtained in Supercritical-Water-Cooled Bundles

2017 ◽  
Author(s):  
Hanqing Xie ◽  
Hakim Maloufi ◽  
Andrew Zopf ◽  
William Anderson ◽  
Christian Langevin ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Annular Channel ◽  
Bulk Fluid ◽  
Efficient Operation ◽  
Normal Heat ◽  
Flow Geometry ◽  
Heat Transfer Correlations ◽  
Bare Tube

SuperCritical Water-cooled Reactor (SCWR) as one of the six Generation-IV nuclear-power-reactor concepts will have increased thermal efficiency compared to that of current Nuclear Power Plants (NPPs) equipped with water-cooled reactors by operating the reactor coolant at supercritical conditions: Coolant pressure of about 25 MPa, inlet temperatures between 300–350°C, and outlet temperatures between 550–625°C. The major flow geometry inside the reactor core is the bundle flow geometry. For safe and efficient operation of an SCWR heat transfer coefficients should be calculated with minimum uncertainties. Unfortunately, the vast majority of experimental datasets were obtained in vertical bare tubes cooled with SCW. Experiments in a bundle flow geometry are even more complicated and expensive compared to that in bare tubes. Due to this very few experiments have been performed in bundles. According to the abovementioned, the vast majority of heat-transfer correlations are based on bare-tube data, and only one currently known correlation is based on a 7-element bundle cooled with SCW (the so-called, Dyadyakin and Popov correlation (1977)). Rods in this bundle are equipped with four helical ribs to enhance the heat transfer. However, the authors have not provided any dataset(s) associated with this bundle and correlation. In the current paper a number of bare-tube heat-transfer correlations obtained in SCW and the Dyadyakin and Popov correlation were compared with two datasets obtained in an annular channel with the heated central rod and 3-element bundle. The central rod in this annular channel and rods in the 3-element bundle have the same heated length as those in the 7-element bundle tested by Dyadyakin and Popov in 1977, and are also equipped with four helical ribs. The comparison showed that the Jackson correlation (2002) is the most accurate one in predicting Heat-Transfer-Coefficient (HTC) profiles in the annular channel at normal heat-transfer regime. The Dittus and Boelter correlation (1930) is the most accurate in predicting HTC profiles in the 3-element bundle at normal heat-transfer regime. No one correlation is capable to follow closely HTC profiles at the deteriorated heat-transfer regimes in both flow geometries. Aloo, it should be mentioned that bare-tube heat-transfer correlations, which have thermophysical properties based on bulk-fluid and wall temperatures, might have problems with convergence at high heat fluxes, i.e., above the heat flux at which the deteriorated heat-transfer regime starts in bare tubes.

Developing Heat-Transfer Correlations for Supercritical CO2 Flowing in Vertical Bare Tubes

2012 ◽  
Author(s):  
Sahil Gupta ◽  
Donald McGillivray ◽  
Prabu Surendran ◽  
Liliana Trevani ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Supercritical Carbon Dioxide ◽  
Supercritical Fluid ◽  
Supercritical Water ◽  
Fast Reactor ◽  
Supercritical Co2 ◽  
Critical Parameters ◽  
Heat Transfer Correlations ◽  
Supercritical Carbon

This paper presents an analysis of three new heat-transfer correlations developed for supercritical carbon dioxide (CO2) flowing in vertical bare tubes. A large set of experimental data was obtained at Chalk River Laboratories (CRL) AECL. Heat-transfer tests were performed in upward flow of CO2 inside 8-mm ID vertical Inconel-600 tube with a 2.208-m heated length. Data points were collected at outlet pressures ranging from 7.4 to 8.8 MPa, mass fluxes from 900 to 3000 kg/m2s, inlet fluid temperatures from 20 to 40°C, and heat fluxes from 15 to 615 kW/m2; and for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The objective of the present experimental research is to obtain reference dataset on heat transfer in supercritical CO2 and improve our fundamental knowledge of the heat-transfer processes and handling of supercritical fluids. In general, heat-transfer process to a supercritical fluid is difficult to model, especially, when a fluid passes through the pseudocritical region, as there are very rapid variations in thermophysical properties of the fluid. Thus, it is important to investigate supercritical-fluid behaviour within these conditions. In general, supercritical carbon dioxide was and is used as a modelling fluid instead of supercritical water due to its lower critical parameters compared to those of water. Also, supercritical carbon dioxide is proposed to be used as a working fluid in the Brayton gas-turbine cycle as a secondary power cycle for some of the Generation-IV nuclear-reactor concepts such as a Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR) and Molten-Salt-cooled Reactor (MSR). In addition, supercritical carbon dioxide was proposed to be used in advanced air-conditioning and geothermal systems. Previous studies have shown that existing correlations deviate significantly from experimental Heat Transfer Coefficient (HTC) values, especially, within the pseudocritical range. Moreover, the majority of correlations were mainly developed for supercritical water, and our latest results indicate that they cannot be directly applied to supercritical CO2 with the same accuracy as for water. Therefore, new empirical correlations to predict HTC values were developed based on the supercritical CO2 dataset. These correlations calculate HTC values with an accuracy of ±30% (wall temperatures with accuracy of ±20%) for the analyzed dataset.

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.

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