Experimental investigation on heat transfer to supercritical water in a three-rod bundle with spacer grids

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.

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Nonuniform heat transfer of supercritical water in a tight rod bundle – Assessment of correlations

Annals of Nuclear Energy ◽

10.1016/j.anucene.2017.07.011 ◽

2017 ◽

Vol 110 ◽

pp. 570-583 ◽

Cited By ~ 9

Author(s):

Han Wang ◽

Qincheng Bi ◽

Linchuan Wang ◽

Laurence K.H. Leung

Keyword(s):

Heat Transfer ◽

Supercritical Water ◽

Rod Bundle

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Numerical Flow Simulation of Spacer Grids With Mixing Vanes in a 5 × 5 PWR Rod Bundle

Volume 5: Fuel Cycle and High and Low Level Waste Management and Decommissioning; Computational Fluid Dynamics (CFD), Neutronics Methods and Coupled Codes; Instrumentation and Control ◽

10.1115/icone17-75354 ◽

2009 ◽

Cited By ~ 1

Author(s):

Moyse´s Alberto Navarro ◽

Andre´ Augusto Campagnole dos Santos

Keyword(s):

Heat Transfer ◽

Pressure Drop ◽

Pressure Loss ◽

Considerable Increase ◽

Local Heat Transfer ◽

Local Heat ◽

Spacer Grid ◽

Spacer Grids ◽

Rod Bundle ◽

Vane Angle

The spacer grids exert great influence on the thermal hydraulic performance of the PWR fuel assembly. The presence of the spacers has two antagonistic effects on the core: an increase of pressure drop due to constriction on the coolant flow area and increase of the local heat transfer downstream the grids caused by enhanced coolant mixing. The mixing vanes, present in most of the spacer grid designs, cause a cross and swirl flow between and in the subchannels, enhancing even more the local heat transfer at the cost of more pressure loss. Due to this important hydrodynamic feature the spacer grids are often improved aiming to obtain an optimal commitment between pressure drop and enhanced heat transfer. In the present work, the fluid dynamic performance downstream a 5 × 5 rod bundle with spacer grids is analyzed with a commercial CFD code (CFX 11.0). Eleven different split vane spacer grids with angles from 16° to 36° and a spacer without vanes were evaluated. The computational domain extends from ∼10 Dh upstream to ∼50 Dh downstream the spacer grids. The standard k-ε turbulence model with scalable wall functions and the total energy model were used in the simulations. The results show a considerable increase of the average Nusselt number and secondary mixing with the angle of the vane up to ∼20 Dh downstream the spacer, reducing greatly the influence of the vane angle beyond this region. As expected, the pressure loss through the spacer grid also showed considerable increase with the vane angle.

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Experimental investigation of the pressure drop and friction factor of supercritical water in a 2 × 2 rod bundle

Annals of Nuclear Energy ◽

10.1016/j.anucene.2021.108732 ◽

2022 ◽

Vol 166 ◽

pp. 108732

Author(s):

Gang Wu ◽

Han Wang ◽

Qincheng Bi

Keyword(s):

Experimental Investigation ◽

Pressure Drop ◽

Friction Factor ◽

Supercritical Water ◽

Rod Bundle

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NUMERICAL INVESTIGATION OF HEAT TRANSFER TO SUPERCRITICAL WATER IN A 2 × 2 ROD BUNDLE WITH TWO CHANNELS

Heat Transfer Research ◽

10.1615/heattransres.2017019599 ◽

2018 ◽

Vol 49 (2) ◽

pp. 103-118

Author(s):

Ibrahim Tahir ◽

Waseem Siddique ◽

Inamul Haq ◽

Kamran Qureshi ◽

Anwar Ul Haq Khan

Keyword(s):

Heat Transfer ◽

Numerical Investigation ◽

Supercritical Water ◽

Rod Bundle

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Experimental investigation of heat transfer from a 2×2 rod bundle to supercritical pressure water

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2014.04.036 ◽

2014 ◽

Vol 275 ◽

pp. 205-218 ◽

Cited By ~ 40

Author(s):

Han Wang ◽

Qincheng Bi ◽

Linchuan Wang ◽

Haicai Lv ◽

Laurence K.H. Leung

Keyword(s):

Heat Transfer ◽

Experimental Investigation ◽

Supercritical Pressure ◽

Rod Bundle ◽

Pressure Water ◽

Supercritical Pressure Water

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

18th International Conference on Nuclear Engineering: Volume 2 ◽

10.1115/icone18-29991 ◽

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.

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