CFD Modeling of Subcooled Flow Boiling for Nuclear Engineering Applications (original) (raw)
Related papers
CFD modelling of subcooled flow boiling for nuclear engineering applications
2005
In this work a general-purpose CFD code CFX-5 was used for simulations of subcooled flow boiling. The subcooled boiling model, available in a custom version of CFX-5, uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data [1] of highpressure flow boiling in a vertical pipe covering a wide range of conditions (relevant to the pressurized water reactor). In general, a good agreement with the experimental data has been achieved. To adequately predict the lateral distribution of two-phase flow parameters, the modelling of two-phase flow turbulence and non-drag forces under wall boiling conditions have been also investigated in the paper.
CFD Simulation of Subcooled Flow Boiling at Low Pressure
2001
An increased interest to numerically simulate the subcooled flow boiling at low pressures (1 to 10 bar) has been aroused in recent years, pursued by the need to perform safety analyses of research nuclear reactors and to investigate the sump cooling concept for future light water reactors. In this paper the subcooled flow boiling has been simulated with a multidimensional two-fluid model used in a CFX-4.3 computational fluid dynamics (CFD) code. The existing model was adequately modified for low pressure conditions. It was shown that interfacial forces, which are usually used for adiabatic flows, need to be modeled to simulate subcooled boiling at low pressure conditions. Simulation results are compared against published experimental data [1] and agree well with experiments.
ICONE19-44115 CFD Simulation of Boiling Flow Experiment in a Rectangular Vertical Channel
The Proceedings of the International Conference on Nuclear Engineering (ICONE)
Precise and reliable experimental data on velocity profiles and velocity fluctuation within a boundary layer during adiabatic as well as diabatic flow condition plays a key role in better understanding of subcooled flow boiling phenomenon. Such measurements are of paramount importance to development and correct implementation of two-fluid numerical models, which are used for prediction purposes of the dominant flow boiling mechanisms within different flow cross-sections, flow patterns and for different fluids. In this context PTV experimental measurements inside vertical square channel carried out at Texas A&M University have been adopted in order to characterize the phenomenon with local governing mechanisms in the near wall region. The main objective of this work is to critically assess some dominant mechanisms for velocity field development and turbulence intensity distribution in the near wall region through the comparison of the experimental measurements and numerical results. The extension of such analysis would improve prediction models for fluid flow performance behavior, heat transfer and critical heat flux prediction in light water reactors. Moreover, this work presents the basis for further study of near-wall turbulence and flow boiling mechanisms. For calculation purposes the ANSYS CFX code has been used, which already possess some flow boiling modeling capabilities. A successful simulation of a flow boiling would have a significant impact on the development of empirical models for heat transfer and critical heat flux prediction. Moreover, it would enhance system efficiencies and improve both reliability and safety issues in systems with forced convective boiling applications.
Parameter Study of Boiling Model for CFD Simulation of Multiphase-Thermal Flow in a Pipe
Journal of Ocean Engineering and Technology, 2021
The demand for eco-friendly energy is expected to increase due to the recently strengthened environmental regulations. In particular, the flow inside the pipe used in a cargo handling system (CHS) or fuel gas supply system (FGSS) of hydrogen transport ships and hydrogen-powered ships exhibits a very complex pattern of multiphase-thermal flow, including the boiling phenomenon and high accuracy analysis is required concerning safety. In this study, a feasibility study applying the boiling model was conducted to analyze the multiphase-thermal flow in the pipe considering the phase change. Two types of boiling models were employed and compared to implement the subcooled boiling phenomenon in nucleate boiling numerically. One was the "Rohsenow boiling model", which is the most commonly used one among the VOF (Volume-of-Fluid) boiling models under the Eulerian-Eulerian framework. The other was the "wall boiling model", which is suitable for nucleate boiling among the Eulerian multiphase models. Moreover, a comparative study was conducted by combining the nucleate site density and bubble departure diameter model that could influence the accuracy of the wall boiling model. A comparison of the Rohsenow boiling and the wall boiling models showed that the wall boiling model relatively well represented the process of bubble formation and development, even though more computation time was consumed. Among the combination of models used in the wall boiling model, the simulation results were affected significantly by the bubble departure diameter model, which had a very close relationship with the grid size. The present results are expected to provide useful information for identifying the characteristics of various parameters of the boiling model used in CFD simulations of multiphase-thermalflow, including phase change and selecting the appropriate parameters.
Assessment of subcooled boiling wall boundary correlations for two-fluid model CFD
2014
An assessment of the heat and mass transfer wall boundary conditions used for subcooled boiling simulations with a CFD two-fluid model has been performed. This assessment was focused on the wall heat flux partitioning model using the state-of-the-art multidimensional Freon experimental data available in the literature. Various constitutive relations used to close the vapor generation rate at the heated wall were studied in order to obtain the best suited combination(s). The current study was restricted to vertical flows through pipe and annulus geometries. Two Freon data sets from the literature were considered: the first with R12 at about 2.6 MPa pressure and the second with R113 at 269 kPa pressure. In these data sets, the bubble diameter distribution measurements across the ducts were available. Bubble diameter estimation is the largest uncertainty in the boiling two-fluid model predictions and hence using the data with known bubble sizes allowed to focus on the assessment of other parameters to model vapor generation rate at the wall, in particular bubble nucleation site density and bubble departure frequency. From the parametric simulations with different combinations of the nucleation site density and frequency models, the discrepancy in obtaining a consistent frequency model for Freon became evident. The state-of-the-art frequency model under consideration is a function of the wall temperature which is estimated using a well-known empirical nucleate boiling heat transfer coefficient correlation. The frequency model provided inaccurate predictions in cases where the nucleate boiling heat transfer coefficient was inadequate to predict the wall superheat. The frequency model is now improved by using a more recent nucleate boiling heat transfer coefficient correlation with wider applicability, i.e., Freon. The simulations were carried out using the CFD code CFX (version 12). The Freon data used here corresponds to fluid-vapor density ratios ranging from 6 to 76. This assessment validated a consistent site density-departure frequency correlation combination for subcooled boiling simulations which has not been used so far for CFD simulations and an extension of the applicability of the correlation combination to high pressure steam-water conditions due to the favorable scaling of Freon physical properties.
CFD analysis of flow boiling in the ITER first wall
Fusion Engineering and Design, 2012
This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles-Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost.
2021
Boiling flows are widely encountered in several engineering and industrial processes. They have a special interest in nuclear industry, where a Computational Fluid Dynamic (CFD) thermohydraulic investigation becomes very popular for design and safety. Many attempts to model numerically subcooled nucleate boiling flows can be found in the literature, where several interfacial forces acting on bubbles which are interacting on the bulk fluid were neglected, due to the hard convergence of the calculations, or to the bad accuracy of the obtained results. In this paper, a sensitivity analysis is carried out for the interfacial forces acting on bubbles during subcooled nucleate boiling flows. For this purpose, 2D CFD axisymmetric simulations based on an Eulerian approach are performed. The developed models aim to mimic the subcooled nucleate boiling flows in concentric pipes, operating at high pressure. The predicted spatial fields of boiling quantities of interest are presented and commented. The numerical results are compared against the available experimental data, where it is shown that neglecting some interfacial forces like the lift or the wall lubrication forces will yield to good predictions for some quantities but will fail the prediction for others. The models leading to the best predictions are highlighted and proposed as recommendations for future CFD simulations of subcooled nucleate boiling flows.
CFD Analysis of Subcooled Flow Boiling in 4 × 4 Rod Bundle
Applied Sciences
Rod bundle flow is an important research field related to reactor cooling in nuclear power plants. Owing to the rapid development of computerized performance assessments, interest in coolant flow analysis using computational fluid dynamics has garnered research interest. Rod bundle flow research data compared with experimental results under various conditions are thus needed. To address this, a boiling model verification study was conducted with reference to experiments. This study adopts the Reynolds-averaged Navier–Stokes equation, a practical analysis method compared to direct numerical simulation and large eddy simulation, including turbulence modeling, to predict the flow of coolant inside a rod bundle. This study also investigates void behavior in low-pressure subcooled flow boiling using a Eulerian approach (two-fluid model). The rod bundle has a length of 0.59 m and a hydraulic diameter of approximately 14.01 mm. At the cross-section at a height of 0.58 m, near the exit, num...
CFD Simulation of the Departure from Nucleate Boiling
2015
This paper presents an attempt to use multiphase CFD code for prediction of the Departure from Nucleate Boiling (DNB) type of Critical Heat Flux (CHF). Numerical simulations of DNB in boiling flow in vertical tube were performed with the NEPTUNE_CFD V2 code. This code can simulate multicomponent multiphase flow by solving three balance equations for each phase or fluid component. A new set of validated models of physical phenomena in boiling bubbly flow was used in the calculations. Simulated cases were based on data from the Standard tables of CHF in pipes, produced by the Russian Academy of Sciences. It was found out that local DNB criterion based on void fraction equal to 0.8 does not work well with the new set of physical models implemented in NEPTUNE_CFD V2 code. But it was discovered that the criterion for DNB prediction can be based on the ratio of evaporation heat flux and total wall heat flux. Evaporation heat flux is calculated by the extended Kurul and Podowski wall boili...