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.

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.

Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle

2009 ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyy ◽  
A. Eu. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Annular Channel ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter

The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.

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.

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.

Heat Transfer to Supercritical Water in Vertical 7-Rod Bundle

2008 ◽  
Author(s):  
V. G. Razumovskiy ◽  
E. N. Pis’mennyy ◽  
A. E. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter ◽  
Heated Length

The results of heat transfer to supercritical water flowing upward in a vertical tight 7-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 700 to 1500 kg/m2s, inlet temperature from 125 to 325°C, outlet temperature up to 379°C and heat flux up to 1.6 MW/m2 (heat flux rate up to 1.5 kJ/kg). Temperature regimes of the bundle cooled by supercritical water were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of the fuel bundles.

Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Malwina Gradecka ◽  
Roman Thiele ◽  
Henryk Anglart
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Water Flow ◽  
Supercritical Water ◽  
Complex Geometry ◽  
Inlet Section ◽  
Fluid Temperature ◽  
Rod Bundle ◽  
Flow And Heat Transfer

This paper presents a steady-state computational fluid dynamics approach to supercritical water flow and heat transfer in a rod bundle with grid spacers. The current model was developed using the ANSYS Workbench 15.0 software (CFX solver) and was first applied to supercritical water flow and heat transfer in circular tubes. The predicted wall temperature was in good agreement with the measured data. Next, a similar approach was used to investigate three-dimensional (3D) vertical upward flow of water at supercritical pressure of about 25 MPa in a rod bundle with grid spacers. This work aimed at understanding thermo- and hydrodynamic behavior of fluid flow in a complex geometry at specified boundary conditions. The modeled geometry consisted of a 1.5-m heated section in the rod bundle, a 0.2-m nonheated inlet section, and five grid spacers. The computational mesh was prepared using two cell types. The sections of the rods with spacers were meshed using tetrahedral cells due to the complex geometry of the spacer, whereas sections without spacers were meshed with hexahedral cells resulting in a total of 28 million cells. Three different sets of experimental conditions were investigated in this study: a nonheated case and two heated cases. The nonheated case, A1, is calculated to extract the pressure drop across the rod bundle. For cases B1 and B2, a heat flux is applied on the surface of the rods causing a rise in fluid temperature along the bundle. While the temperature of the fluid increases along with the flow, heat deterioration effects can be present near the heated surface. Outputs from both B cases are temperatures at preselected locations on the rods surfaces.

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.

Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Laurence K. H. Leung ◽  
Yanfei Rao ◽  
Krishna Podila
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
Supercritical Pressure ◽  
Heat Transfer Correlation ◽  
Computational Tools ◽  
Experimental Heat ◽  
Experimental Heat Transfer ◽  
Transfer Behavior ◽  
Supercritical Water Cooled Reactor

Experimental data and correlations are not available for the fuel-assembly concept of the Canadian supercritical water-cooled reactor (SCWR). To facilitate the safety analyses, a strategy for developing a heat-transfer correlation has been established for the fuel-assembly concept at supercritical pressure conditions. It is based on an analytical approach using a computational fluid dynamics (CFD) tool and the ASSERT subchannel code to establish the heat transfer in supercritical pressure flow. Prior to the application, the CFD tool was assessed against experimental heat transfer data at the pseudocritical region obtained with bundle subassemblies to identify the appropriate turbulence model for use. Beyond the pseudocritical region, where the normal heat transfer behavior is anticipated, the ASSERT subchannel code also was assessed with appropriate closure relationships. Detailed information on the supporting experiments and the assessment results of the computational tools are presented.

Experimental Research of Supercritical Water Heat Transfer in Different Channels

2013 ◽  
Author(s):  
Hong-bo Li ◽  
Meng Zhao ◽  
Han-yang Gu ◽  
Fei Wang ◽  
Jian-min Zhang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Flow Velocity ◽  
Critical Point ◽  
Heat Transfer Enhancement ◽  
Experimental Research ◽  
Supercritical Water ◽  
Annular Channel ◽  
Fluid Temperature ◽  
Transfer Enhancement ◽  
Low Mass

The experimental research of supercritical water heat transfer has been performed on the supercritical water multipurpose test loop (SWAMUP) with tube, annular channel, and bundles. The normal heat transfer, heat transfer deterioration (HTD) and heat transfer enhancement were observed; and the heat transfer experimental data were obtained. The experimental results show that: the first kind of HTD caused by buoyancy effect only occurs with low mass flow velocity and high heat flux when the fluid temperature is below pseudo-critical point in all the tested channels and the second kind of HTD caused by acceleration effect always occurs when the fluid temperature reaches pseudo-critical point in tube and annular channel; the heat transfer enhancement occurs when the fluid temperature reaches pseudo-critical point with low mass flow velocity in tube; and the heat transfer enhancement in bundles is caused by the space grids. It is concluded that the heat transfer in bundles is better than in other tested channels.

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.

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