• Journal of computational fluids engineering
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• v.15 no.3
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• pp.24-31
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• 2010

boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and gas(vapour) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 MPa and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit for lqiuid phase ($y^+_{w,l}$) was examined from 101 to 313 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., $y^+_{w,l}$ > 300 at the tube exit.

• Journal of Ocean Engineering and Technology
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• v.35 no.1
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• pp.50-58
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• 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.

• 한국전산유체공학회:학술대회논문집
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• 2010.05a
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• pp.320-325
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• 2010

A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and vapor(steam) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 MPa and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit ($y_{w}^{+}$) was examined from 64 to 172 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., $y_{w}^{+}$ > 100.

• Journal of Power System Engineering
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• v.17 no.2
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• pp.21-28
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• 2013

Phenomenon of droplet impingement with high temperature wall needs to be investigated because atomization process of droplet and cooling process of the wall by the impingement are very important in industry, thus studies concerned with temperature of piston wall have been conducted in spray characteristics analysis of diesel engine. Hence, in this study, we defined $DT_{sat}(=T_w-T_{sat})$ superheat degree of the wall by difference between $T_w$ considering surface temperature of piston in the actual engine and $T_{sat}$ saturation temperature of the fuel and then investigated spray behavior of wall impinging with variance of the boiling process. In this study, in order to analyze wall impingement of droplet in accordance with difference of boiling condition, calculational conditions were set as $DT_{sat}=40K$(nucleate boiling), $DT_{sat}=140K$(transition boiling), and $DT_{sat}=240K$(film boiling). As a result, it can be found that fuel vapor increases and droplet mass decreases in the order of the nucleate boiling, transition boiling, and film boiling.

• Journal of Advanced Marine Engineering and Technology
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• v.18 no.3
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• pp.26-33
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• 1994

A new reliable method to prediet the axial vapor fraction distribution from the measured probability density of the liquid bulk temperature is suggested in this paper. And also the actual quality of the subcooled boiling flow is easily calculated from the liquid bulk temperature. When the heat generating rate is reached to the CHF value, the sharp wall temperature increasing by the wall temperature fluctuation is occurred under the CHF condition. This paper presents the simple wall temperature fluctuation model of transition boiling by the repeating process of overheating and quenching, when the coalescent bubble passes slowly near the wall. Experiments for the subcooled R-113 flow are carride-out in the range of(0.9399~4.461)${\times}10^6$kg/$m^2$hr mass velocity and 10~3$0^{\circ}C$ intel subcooling condition.

• Journal of Advanced Marine Engineering and Technology
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• v.13 no.4
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• pp.40-53
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• 1989

Boiling heat transfer phenomena is widely applied to BWR and electrical heating system because of its high heat transfer coefficient. In these systems, steady state heat transfer is dependent on nucleate boiling. When the heat generating rate is sharply increased or the cooling capacity of coolant is sharply decreased, sharp wall temperature rise is occurred under the critical heat flux(CHF) condition. This paper presents the simple wall temperature fluctuation model of transition mechanism in the repeating process of overheating and quenching, when coalescent bubble passes relatively slowly on the wall and simultaneously the transition from nucleate boiling to film boiling is carried at especially onset of the CHF state. The values calculated by the present model are resulted comparatively good with the measured.

• Journal of Advanced Marine Engineering and Technology
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• v.10 no.2
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1860-1873
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.15 no.4
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• pp.254-260
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• 2003

The boiling characteristics for R134a are studied to clarify the hysteresis at low temperature on enhanced tubes of a flooded evaporator. Initial boiling conditions, refrigerant temperature, and inlet temperature of the chilled water are considered as the key parameters of the experiments. Unlike previous studies of the boiling heat transfer with uniform heat flux and uniform wall temperature, the wall temperature was varied along the tube. In, this study, it was found that the hysteresis of the temperature overshoot (705) at the onset of nucleate boiling initially at the inlet section of the tube. It is also concluded that the abnormal operation can be avoided during the low temperature boiling if the refrigeration system is started with LMTD larger than $3.4^{\circ}C$ at initial stage and larger than $1.0^{\circ}C$ at normal stage.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.236-255
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• 1999

An improved mechanistic model was developed to predict a convective boiling critical heat flux (CHF) in the vertical round tubes with uniform heat fluxes. The CHF formula for subcooled and low quality boiling was derived from the local conservation equations of mass, energy and momentum, together with appropriate constitutive relations. The model is characterized by the momentum balance equation to determine the limiting transverse interchange of mass flux crossing the interface of wall bubbly layer and core by taking account of the convective shear effect due to the frictional drag on the wall-attached bubbles. Comparison between the present model predictions and experimental CHF data from several sources shows good agreement over a wide range of How conditions. The present model shows comparable prediction accuracy with the CHF look-up table of Groeneveld et al. Also the model correctly accounts for the effects of flow variables as well as geometry parameters.