Supercritical water heat transfer in vertical tubes: A look-up table

2008 ◽  
Vol 50 (2-6) ◽  
pp. 532-538 ◽  
Author(s):  
Matthias F. Loewenberg ◽  
Eckart Laurien ◽  
Andreas Class ◽  
Thomas Schulenberg
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Look Up Table ◽  
Vertical Tubes

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.

Numerical Investigation on Heat Transfer of Supercritical Water With a Variable Turbulent Prandtl Number Model

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Xiangfei Kong ◽  
Dongfeng Sun ◽  
Lingtong Gou ◽  
Siqi Wang ◽  
Nan Yang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Prandtl Number ◽  
Supercritical Fluids ◽  
Supercritical Water ◽  
Viscosity Ratio ◽  
Turbulent Viscosity ◽  
Turbulence Models ◽  
Turbulent Prandtl Number ◽  
Vertical Tubes

Abstract Turbulent Prandtl number (Prt) has a great impact on the performance of turbulence models in predicting heat transfer of supercritical fluids. Unrealistic treatment of Prt may lead to large deviations of the prediction results from experimental data under supercritical conditions. In this study, the effect of Prt on heat transfer of supercritical water was extensively studied by using shear stress transport (SST) k–ω turbulence model, and the results suggested that using the existing Prt models would lead to failures in predicting the heat transfer characteristics of supercritical water under deteriorated heat transfer (dht) conditions. A new variable Prt model was proposed with the Prt varied with pressure, turbulent viscosity ratio, and molecular Prandtl number. The new model was validated by comparing the numerical results with the corresponding experimental data, and it was found that the new variable Prt model exhibited better performance on reproducing the dht of supercritical water in vertical tubes than those of the existing Prt models.

Heat Transfer to Water at Supercritical Parameters in Vertical Tubes, Annular Channels, 3- and 7-Rod Bundles

2013 ◽  
Author(s):  
V. G. Razumovskiy ◽  
E. M. Mayevskiy ◽  
A. E. Koloskov ◽  
E. N. Pis’mennyi ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Experimental Studies ◽  
Heat Fluxes ◽  
Bulk Temperature ◽  
Experimental Setup ◽  
Gaseous State ◽  
Annular Channels ◽  
Wide Range ◽  
Vertical Tubes

The data on deteriorated transfer to supercritical water in vertical tubes and channels simulating coolant flow in fuel assemblies obtained at the same experimental setup during more than dozen of years are considered and compared with some known results of the experimental studies performed by other authors. They involve the data for vast ranges of geometry, mass velocity, heat flux rate, and pressure, in some cases for up- and downward flow, for flow with and without thermoacoustic oscillations. For the first time the data illustrating deterioration of heat transfer in the bundles of fuel elements are presented. An attempt to explain the phenomena of “inlet” peak of wall temperature is made. It is shown that temperature regimes of the tubes cooled with supercritical water in a gaseous state (i.e., at bulk temperature above the pseudocritical temperature) are close to linear, stable and easily reproducible within a wide range of mass and heat fluxes. Some requirements to the experimental setup, coolant quality, test sections etc. that should be followed in studying thermal and hydraulic parameters of supercritical coolant are analyzed.

Heat Transfer to Supercritical Water in Gaseous State or Affected by Mixed Convection in Vertical Tubes

2006 ◽  
Author(s):  
E. N. Pis’menny ◽  
V. G. Razumovskiy ◽  
E. M. Maevskiy ◽  
A. E. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Mixed Convection ◽  
Supercritical Water ◽  
Operating Conditions ◽  
Maximum Effect ◽  
Upward Flow ◽  
Downward Flow ◽  
Gaseous State ◽  
Transfer Data ◽  
Vertical Tubes

The results on heat transfer to supercritical water heated above the pseudocritical temperature or affected by mixed convection flowing upward and downward in vertical tubes of 6.28-mm and 9.50-mm inside diameter are presented. Supercritical water heat-transfer data were obtained at a pressure of 23.5 MPa, mass flux within the range from 250 to 2200 kg/(m2s), inlet temperature from 100 to 415°C and heat flux up to 3.2 MW/m2. Temperature regimes of the tubes cooled with supercritical water in a gaseous state (i.e., supercritical water at temperatures beyond the pseudocritical temperature) were stable and easily reproducible within a wide range of mass and heat fluxes. An analysis of the heat-transfer data for upward and downward flows enabled to determine a range of Gr/Re2 values corresponding to the maximum effect of free convection on the heat transfer. It was shown that: 1) the heat transfer coefficient at the downward flow of water can be higher by about 50% compared to that of the upward flow; and 2) the deteriorated heat-transfer regime is affected with the flow direction, i.e., at the same operating conditions, the deteriorated heat transfer may be delayed at the downward flow compared to that at the upward flow. These heat-transfer data are applicable as the reference dataset for future comparison with bundle data.

scholarly journals Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Thomas Schulenberg ◽  
Hongbo Li
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Physical Models ◽  
Transient Heat Transfer ◽  
Film Boiling ◽  
Critical Conditions ◽  
Boiling Crisis ◽  
Look Up Table ◽  
System Pressure

While supercritical water is a perfect coolant with excellent heat transfer, a temporary decrease of the system pressure to subcritical conditions, either during intended transients or by accident, can easily cause a boiling crisis with significantly higher cladding temperatures of the fuel assemblies. These conditions have been tested in an out-of-pile experiment with a bundle of four heated rods in the supercritical water multipurpose loop (SWAMUP) facility coconstructed by CGNPC and SJTU in China. Some of the transient tests have been simulated at KIT with a one-dimensional (1D) matlab code, assuming quasi-steady-state flow conditions, but time dependent temperatures in the fuel rods. Heat transfer at supercritical and at near-critical conditions was modeled with a recent look-up table of Zahlan (2015, “Derivation of a Look-Up Table for Trans-Critical Heat Transfer in Water Cooled Tubes,” Ph.D. dissertation, University of Ottawa, Ottawa, ON, Canada.), and subcritical film boiling was modeled with the look-up table of Groeneveld et al. (2003, “A Look-Up Table for Fully Developed Film Boiling Heat Transfer,” Nucl. Eng. Des., 225(1), pp. 83–97.). Moreover, a conduction controlled rewetting process was included in the analyses, which is based on an analytical solution of Schulenberg and Raqué (2014, “Transient Heat Transfer During Depressurization From Supercritical Pressure,” Int. J. Heat Mass Transfer, 79(12), pp. 233–240.). The method could well reproduce the boiling crisis during depressurization from supercritical to subcritical pressure, including rewetting of the hot zone within some minutes, but the peak temperature was somewhat under-predicted. Tests with a lower heat flux, which did not cause such phenomena, could be predicted as well. In another test with increasing pressure, however, a boiling crisis was also observed at a heat flux, which was significantly lower than the critical heat flux (CHF) predicted by the CHF look-up table of Groeneveld et al. (2007, “The 2006 CHF Look-Up Table,” Nucl. Eng. Des., 237(15–17), pp. 1909–1922.). The paper is summarizing the physical models and the numerical approach. Comparison with experimental data is used to discuss the applicability of the method for the design of supercritical water-cooled reactors (SCWR).

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