A new heat transfer correlation for supercritical water flowing in a vertical tube - an hitherto approach

2021 ◽  
Vol 125 ◽  
pp. 105314
Author(s):  
Anand Sundaravel ◽  
Suresh Sivan ◽  
Santhosh Kumar Deenadayalan
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Vertical Tube ◽  
Heat Transfer Correlation

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.

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 Analysis of Supercritical Water Heat Transfer in Vertical Tube With Different Obstacles

2009 ◽  
Author(s):  
Bo Zhang ◽  
Jianqiang Shan ◽  
Jing Jiang
Keyword(s):  
Heat Transfer ◽  
Critical Point ◽  
Supercritical Water ◽  
Transfer Characteristic ◽  
Vertical Tube ◽  
Bulk Temperature ◽  
Transfer Enhancement ◽  
Heat Transfer Characteristics ◽  
Water Reactors ◽  
The Heat Transfer Coefficient

Supercritical Water Reactors (SCWRs) are essentially water reactors operating at pressure and temperature above critical point. The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. To obtain better heat transfer characteristics, increasing the fluctuation using obstacles is the conventional method. Heat transfer characteristic in vertical tube with different obstacles is numerically investigated under supercritical condition. Numerical simulation is carried out with commercial CFD code Fluent 6.1 and adaptive grid. The results show that The RNG k-ε model with enhanced wall treatment can obtain a reliable result; the blockage ratio and the local temperature have large influence on the heat transfer enhancement. The influence region and decay trend of obstacles are also studied and compared with existing correlations.

Numerical Investigation on Heat Transfer to Supercritical Water Flowing Upward in a 4-M Long Bare Vertical Circular Tube

2020 ◽  
Author(s):  
Dong Yang ◽  
Qixian Wu ◽  
Lin Chen ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Numerical Investigation ◽  
Supercritical Water ◽  
Wall Temperature ◽  
Nuclear Power ◽  
Circular Tube ◽  
Vertical Tube ◽  
Nuclear Power Reactor ◽  
Flow Acceleration

Abstract Thermal efficiency and safety of Generation-IV nuclear-power-reactor concept - Supercritical Water-cooled Reactor (SCWR) depend on solid knowledge of specifics of SCW thermophysical properties and heat transfer within these conditions. As a preliminary, but conservative approach to uncover these specifics is analysis of experimental data obtained in bare tubes including numerical investigation. This paper presents the numerical investigation, based on computational fluid dynamics, of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm). The flow and heat-transfer mechanism of SCW in the vertical tube under the influence of buoyancy and flow acceleration are analyzed. Results of numerical simulation predict the experimental data with reasonable accuracy. The results indicated that in the region of q/G > 0.4 kJ/kg, the wall temperature distribution tends to be non-linear, and heat transfer may deteriorate. When Tb < Tpc < Tw, internal wall temperature shows peaks, which corresponds to heat-transfer deterioration. Meanwhile the position, where the deterioration occurs is continuously moved forward to the inlet as the heat flux increases. Velocity changes near the wall show an M shape according to mass conservation for the density change.

Development of supercritical water heat-transfer correlation for vertical bare tubes

2011 ◽  
Vol 241 (4) ◽  
pp. 1126-1136 ◽  
Author(s):  
Sarah Mokry ◽  
Igor Pioro ◽  
Amjad Farah ◽  
Krysten King ◽  
Sahil Gupta ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Heat Transfer Correlation

Numerical Verifications on Heat Transfer to Supercritical Water Flowing Upward in a 4-m Long Bare Vertical Tube

2021 ◽  
Author(s):  
Dong Yang ◽  
Jiaxiang Chen ◽  
Yongchang Feng ◽  
Lin Chen
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Turbulent Viscosity ◽  
Convection Heat Transfer ◽  
Vertical Tube ◽  
Thermal Capacity ◽  
Operating Conditions ◽  
Wall Region ◽  
Nuclear Power Reactor

Abstract Thermal efficiency and safety of Generation-IV nuclear-power-reactor conceptSupercritical Water-cooled Reactor (SCWR) are largely dependent on the coupled SCW thermophysical properties and heat transfer performance in the supercritical region. This paper presents the numerical investigation of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm) based on computational fluid dynamics. Numerical model for SCW was established in this analysis and forced-convection heat transfer was studied at different operating conditions. The data were collected at pressure of about 24 MPa, inlet temperatures from 320 to 350 ?, mass flux from 1000 to 1500 kg/m2s and heat flux up to 1500 kW/m2. Results of numerical simulation predict the experimental data with reasonable accuracy. A dimensional analysis was conducted to derive the general form of an empirical supercritical water heat-transfer correlation. The decrease of turbulent viscosity due to the decrease of density leads to a lower turbulent diffusion and turbulent kinetic energy, which inhibits heat transfer. The increased wall temperature and localized heat transfer deterioration as the liquid in the core of the tube is isolated for the low-density fluid adheres to the near-wall region, which is characterized by low thermal capacity.

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.

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