CFD investigation of nucleate boiling in non-circular geometries at high pressure

Thakrar, Ronak; Murallidharan, Janani Srree; Walker, S.P.

doi:10.1016/j.nucengdes.2016.08.020

Cited by 11 publications

(3 citation statements)

References 25 publications

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“…If the various adjustable coefficients, most importantly bubble departure diameter values, are carefully selected, the approach enables simulation in challenging geometries and industrially relevant conditions. For example, boiling in unusual geometries, such as non-circular channels [24], which are still challenging from a hydrodynamic point of view, can be reliably predicted, within the inevitable limitations of a largely semi-empirical approach, which requires careful tuning of input parameters in order to obtain reliable estimates of the various quantities of interest, such as void fraction distributions and boiling heat transfer rates.…”

Section: Importance Of the Wall-boiling Modelmentioning

confidence: 99%

“…Recent efforts [27,35] aimed to improve CFD prediction of boiling heat transfer using data generated with semi-mechanistic bubble departure models. In [24] in particular, popular 'force balance' models were used to compute bubble departure diameters and used in CFD simulation of boiling flows. The models were assessed via comparison with established validation cases close to real PWR conditions, showing that approach can return predictions in good agreement with experimental trends Figure 8.…”

Section: Development Of Wall-boiling Modelsmentioning

confidence: 99%

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Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Giustini

2020

Inventions

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The boiling process is utterly fundamental to the design and safety of water-cooled fission reactors. Both boiling water reactors and pressurised water reactors use boiling under high-pressure subcooled liquid flow conditions to achieve high surface heat fluxes required for their operation. Liquid water is an excellent coolant, which is why water-cooled reactors can have such small sizes and high-power densities, yet also have relatively low component temperatures. Steam is in contrast a very poor coolant. A good understanding of how liquid water coolant turns into steam is correspondingly vital. This need is particularly pressing because heat transfer by water when it is only partially steam (‘nucleate boiling’ regime) is particularly effective, providing a great incentive to operate a plant in this regime. Computational modelling of boiling, using computational fluid dynamics (CFD) simulation at the ‘component scale’ typical of nuclear subchannel analysis and at the scale of the single bubbles, is a core activity of current nuclear thermal hydraulics research. This paper gives an overview of recent literature on computational modelling of boiling. The knowledge and capabilities embodied in the surveyed literature entail theoretical, experimental and modelling work, and enabled the scientific community to improve its current understanding of the fundamental heat transfer phenomena in boiling fluids and to develop more accurate tools for the prediction of two-phase cooling in nuclear systems. Data and insights gathered on the fundamental heat transfer processes associated with the behaviour of single bubbles enabled us to develop and apply more capable modelling tools for engineering simulation and to obtain reliable estimates of the heat transfer rates associated with the growth and departure of steam bubbles from heated surfaces. While results so far are promising, much work is still needed in terms of development of fundamental understanding of the physical processes and application of improved modelling capabilities to industrially relevant flows.

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Section: Importance Of the Wall-boiling Modelmentioning

confidence: 99%

Section: Development Of Wall-boiling Modelsmentioning

confidence: 99%

Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Giustini

2020

Inventions

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“…Even though it is hard to obtain fully grid independent results in some cases, it is necessary to discuss the effect of key grid parameters. Thakrar et al (2016) has investigated the nucleate boiling in rectangular geometries at high pressure by use of CFD code STAR-CCM+. Standard k-ε model and Reynolds Stress Transport models were compared to investigate the influence of turbulence.…”

Section: Introductionmentioning

confidence: 99%

Numerical Investigation of the Effect of Grids and Turbulence Models on Critical Heat Flux in a Vertical Pipe

et al. 2018

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Numerical simulation has been widely used in nuclear reactor safety analyses to gain insight into key phenomena. The Critical Heat Flux (CHF) is one of the limiting criteria in the design and operation of nuclear reactors. It is a two-phase flow phenomenon, which rapidly decreases the heat transfer performance at the rod surface. This paper presents a numerical simulation of a steady state flow in a vertical pipe to predict the CHF phenomena. The detailed Computational Fluid Dynamic (CFD) modeling methodology was developed using FLUENT. Eulerian two-phase flow model is used to model the flow and heat transfer phenomena. In order to gain the peak wall temperature accurately and stably, the effect of different turbulence models and wall functions are investigated based on different grids. Results show that O type grid should be used for the simulation of CHF phenomenon. Grids with Y+ larger than 70 are recommended for the CHF simulation because of the acceptable results of all the turbulence models while Grids with Y+ lower than 50 should be avoided. To predict the dry-out position accurately in a fine grid, Realizable k-ε model with standard wall function is recommended. These conclusions have some reference significance to better predict the CHF phenomena of vertical pipe. It can also be expanded to rod bundle of Boiling Water Reactor (BWR) by using same pressure condition.

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CFD model development for two-phase flows

Colombo

Fairweather

Walker

et al. 2019

Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment

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CFD investigation of nucleate boiling in non-circular geometries at high pressure

Cited by 11 publications

References 25 publications

Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Numerical Investigation of the Effect of Grids and Turbulence Models on Critical Heat Flux in a Vertical Pipe

CFD model development for two-phase flows

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