• Journal of Advanced Marine Engineering and Technology
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• v.38 no.5
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• pp.581-588
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• 2014

Extensive researches toward pool boiling characteristics have been widely investigated. However, the correct understanding of its boiling crisis by Critical Heat Flux (CHF) phenomenon during steady and transient heat transfer as a fundamental database for designing heat generation systems is still need to be clarified. The pool boiling CHFs were investigated to clarify the generalized phenomena of transition to film boiling at transient condition. The CHFs were measured on 1.0 mm diameter horizontal cylinder of platinum for exponential heat generation rates with various periods for saturated liquids at atmospheric pressure. The incipience of boiling processes was completely different depending on pre-pressurization. Also, the dependence of pre-pressure in transient CHFs changed due to the wettability of boiling liquids. The trend of typical CHFs were clearly divided into the first, second and third groups for long, short and intermediate periods, respectively. By the effect of pre-pressurization, the boiling incipience mechanism was replaced from that by active cavities entraining vapor to that by the HSN in originally flooded cavies.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1205-1214
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• 2009

The critical heat flux (CHF) levels that occurred due to exponential heat inputs for varying periods to a 1.0-mm diameter horizontal cylinder immersed in various liquids were measured to develop an extended database on the effect of high subcoolings for quasi-steady-state and transient maximum heat fluxes. Two main mechanisms of CHF were found. One mechanism is due to the time lag of the hydrodynamic instability (HI) which starts at steady-state CHF upon fully developed nucleate boiling, and the other mechanism is due to the explosive process of heterogeneous spontaneous nucleation (HSN) which occurs at a certain HSN superheat in originally flooded cavities on the cylinder surface. Steady-state CHFs were divided into three regions for lower, intermediate and higher subcooling at pressures resulting from HI, transition and HSN, respectively. HSN consistently occurred in the transient boiling CHF conditions that correspond to a short period. It was also found that the transient boiling CHFs gradually increased, then rapidly decreased and finally increased again as the period became shorter.

• Journal of computational fluids engineering
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• v.16 no.2
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• pp.88-93
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• 2011

A numerical study of subcooled onset of nucleate boiling (ONB) in a micro-channel under pulsed heating using volume of fluids (VOF) model was conducted. The VOF simulation adopting the existing experimental condition is compared to the experimental data. The time to ONB was determined when the void fraction at the microheater surface first appeared. The theoretical superheat for homogeneous nucleation relatively predicts the transient ONB results of convective flow of water well based on local temperature distribution. It was found that once heat load increases at the heater, transient flow boiling starts to occur faster.

• Journal of Advanced Marine Engineering and Technology
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• v.19 no.2
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• pp.20-26
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• 1995

The very rapid cooling problem from $820^{\circ}$C to $20^{\circ}$C on the surface of the steel by thermal conduction including the latent heat of phase transformation of steel and by transient boiling heat transfer of water are considered to principal problem in quenching. The transient boiling process of water at the surface of specimen during the quenching process were experimentally analyzed. Then the heat flux was numerically calculated by the numerical method of inverse heat condition problem. In this report, the simulation program to calculate the cooling curves for large rolls was made using the subcooled transient boiling curve as a boundary condition. By this simulation program, the cooling curves of rolls from D=50mm to D=200mm were calculated and the effects of agitation of circulation of water also investigated.

• 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.

• Journal of Advanced Marine Engineering and Technology
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• v.15 no.3
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• pp.57-65
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• 1991

The quenching of steels by water is one of the important problems in engineering for the applications of heat treatment or continuous casting process, but the fundamental researches by the theoretical approaches have not been satisfactorily improved yet. The very rapid cooling problems by the thermal conduction including the latent heat of phase transformation in steel and the transient boiling heat transfer of water on the surface of the steel covering from $850^{\circ}C$ to $20^{\circ}C$ are the key problems of heat treatment. The present quenching experiments are performed for the cylindrical specimens of carbon steel, S45C of diameters (12-30). Nonlinear transient heat conduction and transient boiling heat transfer problem of water on the surface of specimens is analyzed by the numerical method of inverse heat conduction problem. The conditions for the calculation are that the initial temperature of specimens is $820^{\circ}C$ and the cooling water in bath are $20^{\circ}C$,$40^{\circ}C$,$60^{\circ}C$,$80^{\circ}C$,$95^{\circ}C$ with no agitation.

• Transactions of the Korean Society of Mechanical Engineers
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• v.12 no.1
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• pp.44-47
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• 1988

The present experiments of transient temperature variation and temperature fluctuation were conducted by bead type thermistors in pool boiling without a heat surface. The experimentally obtained conclusion were as follows. (1) The high transient temperature Variatian of 11.64.deg. C for a duration of 0.08 sec and a temperature freguency having a duration of 5.6*10$^{-2}$ sec in a flashing were measured. (2) The highest variation of transient temperature was shown at saturated temperature of 80.deg. C and superheat of 7.9.deg. C. (3) Temperature frequency was found to increase with superheat below 60.deg. C of saturated temperature, but above 60.deg. C it was relative to lower superheat and was found to decrease with higher superheat..

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.63-74
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• 2011

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.3 no.1
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• pp.51-60
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• 1991

The aim of this study is to investigate the heat transfer characteristics in the transient cooling process of a high temperature wall. The slow transient cooling experiment was carried out with a copper block of high thermal capacity. The results of these experiments are as follows. 1. Temperature histories measured by the thermocouple, which is 0.99, 2.00, 2.99mm from the heat transfer surface showed monotonous during the cooling process. These variation are the curves of typical temperature histories in film-boiling, transition-boiling, and nucleate-boiling regions. 2. The temperature histories were measured by thermocouple installed in the copper block. The variations of the surface heat fluxes and surface temperature were computed from the numerical solution method TDMA from the measured temperature histories for radial position one dimensional heat transfer inverse problem. The boiling curves were found by the computed temperature histories. 3. The rewetting point which starts to change from film boiling to nucleate boiling is not connected with the mass velocity and it were found that the temperature of rewetting point indicated about $100^{\circ}C$. 4. The heat flux of rewetting point was about $10^5Kcal/m^2h$, at that time, the heat transfer coeficient indicated about $1000Kcal/m^2h^{\circ}C$ irrelevent to mass velocity. 5. The wall superheat decreases as the pressure increases. But I found that rewetting point appeared under higher condition in the wall temperature.

• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.274-285
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• 2003

Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Loop for Transient Analysis). Two test sections are made of copper to maintain Bi below 0.1. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of both the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the two test sections. The measured heat transfer coefficients for the test section with diameter 120 mm lie within the bounding values from the laminar film boiling analysis, while those for diameter 294 mm are found to be greater than the numerical results on account of the Helmholtz instability. There is little difference observed between the film boiling heat transfer coefficients measured from the two test sections. In addition, the higher thermal conductivity of copper results in the higher minimum heat flux in the tests. For the test section of diameter 120 mm, the Leidenfrost point is lower than that for the test section of diameter 294 mm. Destabilization of film boiling propagates radially inward for the 294 mm test section versus radially outward for the 120 mm Test Section.