Accuracy of Eulerian–Eulerian, two-fluid CFD boiling models of subcooled boiling flows

Colombo, Marco; Fairweather, Michael

doi:10.1016/j.ijheatmasstransfer.2016.06.098

Cited by 83 publications

(38 citation statements)

References 48 publications

(3 reference statements)

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“…With the aim of predicting the boiling process, the consideration of wall boiling models is crucial to the prediction of the bubble size distribution in the bulk two-phase flow. For two-fluid model, the Rensselaer Polytechnic Institute (RPI) boiling model from Kurul and Podowski (1990) represents the cornerstone for the prediction of boiling flows through its use in safety analysis and CFD codes (Tu and Yeoh, 2002;Tu et al, 2005;Tu, 2006a, 2006b;Yeoh et al, 2008;Krepper et al, 2013;Colombo and Fairweather, 2016). Principally, the RPI boiling model considers the heat flux from the wall being partitioned between the mechanisms responsible for the heat transfer process: single-phase convection, quenching and evaporation.…”

Section: Review Of Models and Methods For Two-phase Flowmentioning

confidence: 99%

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

Yeoh

2019

Exp. Comput. Multiph. Flow

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Thermal hydraulic analysis of nuclear reactor core and its associated systems can be performed using analysis system, subchannel or computational fluid dynamics (CFD) codes to estimate the different thermal hydraulic safety margins. The safety margins and operating power limits under different conditions of the primary as well as secondary cooling system such as the system pressure, coolant inlet temperature, coolant flow rate, and thermal power and its distributions are considered as key parameters for thermal hydraulic analysis. Considering the complexity of rod bundle geometry, boiling heat transfer and different turbulent scales bring about the many challenges in performing the thermal hydraulic analysis to ensure the safe design and operation of nuclear reactor systems under normal and abnormal conditions. A comprehensive review is presented of past, present and future challenges in state-of-the-art thermal hydraulic analysis covering various aspects of experimental, analytical and computational approaches.

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Section: Review Of Models and Methods For Two-phase Flowmentioning

confidence: 99%

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

Yeoh

2019

Exp. Comput. Multiph. Flow

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“…The benchmark calculation shows that the model can predict the high pressure experimental data fairly well. Colombo [11] also adopted the above model to predict the bubble diameter distribution, and employed the heat flux partitioning approach to establish a wall boiling model, which can provide relatively accurate Eulerian-Eulerian CFD (Computational Fluid Dynamics) predicted results except for the average bubble diameter and the mean velocity. Although the bubble size models can provide acceptable simulations, they are only verified partially in the subcooled flow boiling either in high pressure case or in low pressure case.…”

Section: Bubble Behavior In Flow Boilingmentioning

confidence: 99%

An Overall Bubble Diameter Model for the Flow Boiling and Numerical Analysis through Global Information Searching

Zhang

Lin

et al. 2018

Energies

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Bubble diameter is important for the three-dimensional nucleate boiling two-phase flow simulation. However, the bubble diameter is rarely considered as a variable because it is difficult to estimate. In this paper, a novel bubble diameter model is proposed for the boiling flow. The heat transfer growth, the breakup or coalescence of collision and the liquid impacting separation are considered as the factors affecting the bubble diameter, and three bubble diameter sub-models are developed to calculate the overall bubble diameter based on a departure diameter. In the model, the heat transfer growth is calculated after estimating the bubble location. The collision variation is deduced from the interfacial area concentration equation. The liquid impacting variation is calculated through phase interaction force and gas viscous force. The proposed model is conducted through global information searching in the flow boiling simulation and validated using the void fraction and temperature data from the literature. The effects of the heat transfer growth, the collision, and the separating volume on the bubble diameter are discussed.

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“…The predictions of an analytical model of transient heat transfer which developed by Wang and Podowski (2019) showed a good agreement with measurements. Sometimes the two-fluid model considering multiple effects between liquid and vapor phases have higher simulation accuracy, with closure models improved by experimental data for specific problems in a small range (Colombo and Fairweather, 2016;Sadaghiani and Koşar, 2016). However, due to lack of appropriate closure models describing all relevant phenomena and good numerical robustness for geometrically complex systems, applying twofluid model to the engineering filed is very difficult now.…”

Section: Introductionmentioning

confidence: 99%

Numerical Simulation of Flow Boiling in Small Channel of Plate OTSG

et al. 2020

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The plate OTSG (once-through steam generator) with channel diameter of 1-3 mm has high volume-power ratio and powerful resistance to high temperature and high pressure. It can well satisfy the needs of high heat transfer performance and good security of integrated pressure vessel in nuclear power. Heat transfer characteristics of flow boiling in the channel have aroused increasing concerns from scholars in this field. In this paper, based on the experimental results achieved by the researcher in our team, the drift flux model is applied to simulate the flow boiling heat transfer coefficients in the rectangular channel with equivalent diameter of 1.7 mm to further explore the flow boiling mechanism in the channel. The drift velocities and the distribution parameters of drift flux model are obtained by the empirical correlations of the horizontal flow. The simulation boundary conditions comply to the experimental conditions, the simulation resolutions are obtained by using STAR-CCM+. The simulation results indicate that the heat transfer coefficients trends along the flow direction are consistent with the trends of the experimental data. The drift velocities and the distribution parameters have little effect on the heat transfer coefficients of the horizontal small channel. When the drift velocity is 0 and the distribution parameter is 1, compared with the experimental data, the heat transfer coefficients in the single-phase liquid convective heat transfer region of the flow at high pressure are well higher, while those in the region that from bobble flow to the slug flow of the flow increase, even though they are still lower in annular flow region. The error between the predicted and the experimental is from −50 to +50%. Similarly, the model predicted heat transfer coefficients during subcooled flow boiling at low pressure are generally lower than experimental data. And the error between the predicted and the experimental is from −60 to +10%.

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Accuracy of Eulerian–Eulerian, two-fluid CFD boiling models of subcooled boiling flows

Cited by 83 publications

References 48 publications

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

An Overall Bubble Diameter Model for the Flow Boiling and Numerical Analysis through Global Information Searching

Numerical Simulation of Flow Boiling in Small Channel of Plate OTSG

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