• Journal of Advanced Marine Engineering and Technology
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• 제11권1호
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• pp.72-79
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• 1987

When the wall temperature is very high, a stable vapor film covers the heat transfer surface. The vapor film creates a strong thermal resistance when heat is transferred to the liquid though it. This phenomenon, called "film boiling" is very important in the heat treatment of metals, the design of cryogenic heat exchangers, and the emergency cooling of nuclear reactors. In the practical engineering problems of the transient cooling process of a high temperature wall, the wall temperature history, the variation of the heat transfer coefficients, and the wall superheat at the rewetting points, are the main areas of concern. These three areas are influenced in a complex fashion such factors as the initial wall temperature, the physical properties of both the wall and the coolant, the fluid temperature, and the flow state. Therefore many kinds of specialized experiments are necessary in the creation of precise thermal design. The object 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 block was 240 mm high and 79 mm O.D.. The coolant flowed throuogh the center of a 10 mm diameter channel in the copper block. In the copper block, three sheathed thermocouples were placed in a line perpendicular to the flow. These thermocouples were used to take measurements of the temperature histories of the copper block.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.489-494
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• 1997

A comparative transient analyses were performed for oxide and nitride cores or a large (3000 MWt), pool-type, liquid-metal-cooled reactor (LMR). The study was focused on three representative accident initiators with failure to scram : the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected fast transient overpower (UFTOP). The margins to fuel melting and sodium boiling have been evaluated for these representative transients. The results show that there is an increase in safety margin with nitride core which maintains the physical dimensions of the oxide core.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2464-2476
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• 2021

Direct-contact heat transfer of a single saturated droplet upon colliding with a heated wall in the regime of film boiling was experimentally investigated using high-resolution infrared thermometry technique. This technique provides transient local wall heat flux distributions during the entire collision period. In addition, various physical parameters relevant to the mechanistic modelling of these phenomena can be measured. The obtained results show that when single droplets dynamically collide with a heated surface during film boiling above the Leidenfrost point temperature, typically determined by droplet collision dynamics without considering thermal interactions, small spots of high heat flux due to localized wetting during the collision appear as increasing We_n. A systematic comparison revealed that existing theoretical models do not consider these observed physical phenomena and have lacks in accurately predicting the amount of direct-contact heat transfer. The necessity of developing an improved model to account for the effects of local wetting during the direct-contact heat transfer process is emphasized.

• 대한기계학회논문집
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• 제17권6호
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• pp.1529-1540
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• 1993

본 연구에서는 금속 열처리를 위한 고온면의 막비등 급냉각에 관한 1차적 연구로서 과냉과도 비등곡선의 정확한 형상과 냉각제의 냉각조건이 강재의 과냉과도 비등열전달에 미치는 영향, $A_1$변태점 부근의 $A_1$냉각속도와 상변태열량광의 관계, 그리고 상변태열이 냉각곡선에 미치는 영향 등을 규명하고자 한다.

• Nuclear Engineering and Technology
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• 제40권6호
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• pp.477-488
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• 2008

The objective of the paper is to analyze the thermally induced density wave oscillations in water cooled boiling water reactors. A transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain. A two-phase flow was simulated with a one-dimensional homogeneous equilibrium model. The model treats the boundary conditions naturally and takes into account the compressibility effect of the two-phase flow. The axial variation of the heat flux profile can also be handled with the model. Unlike the method of characteristics analysis, the present numerical model is computationally inexpensive in terms of time and works in a Eulerian coordinate system without the loss of accuracy. The model was validated against available benchmarks. The model was extended for the purpose of studying the flow-induced density wave oscillations in forced circulation and natural circulation boiling water reactors. Various parametric studies were undertaken to evaluate the model's performance under different operating conditions. Marginal stability boundaries were drawn for type-I and type-II instabilities in a dimensionless parameter space. The significance of adiabatic riser sections in different boiling reactors was analyzed in detail. The effect of the axial heat flux profile was also investigated for different boiling reactors.

• 한국분무공학회지
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• 제3권2호
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• pp.24-33
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• 1998

The characteristics of liquefied butane spray are expected to be different from conventional diesel fuel spray, because a kind of flash boiling spray is expected when the back pressure is below the saturation vapor pressure of the butane(0.23MPa at $25^{\circ}C$). An accumulator type pintle injector and its fuel delivery system has been simulated in ruder to give injection pressure, needle lift and rate of fuel injected. The governing equation were solved by finite difference metho. The injection duration was controlled by solenoid valve. Spray behaviors such as a transient spray tip penetration, spray angle and SMD were calculated based on the empirical correlations in case that the back pressure is both above the vapor pressure of the butane and below that of butane. When the back preassure is below the vapor pressure of the fuel, conventional correlation is modified to represent the effect of flash boiling.

• 전기의세계
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• 제13권4호
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• pp.16-24
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• 1964

The dynamic equations of the void fraction and the water velocity in boiling region of the BWR reactor core are derived. And these equations are approximated to be able to set on an PACE analog computor. The transient analysis and the frequency response obtained by analog computer are compared with other by digital computor.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1995년도 춘계학술발표회 초록집
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• pp.185-198
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• 1995

This is an experimental study conducted to visualize the nucleate boiling phenomena and flow regimes occurring inside the liquid pool in a closed two-phase thermosyphon. To meet this purpose, an annular-type thermosyphon was designed and manufactured using a glass tube and a stainless steel tube, being assembled axisymmetrically. The heat to be supplied to the working fluid is generated within a very thin layer of stainless steel tube wall by applying a high frequency electromagnetic field through the induction coil, axisymmetrically set around the evaporator zone. Some important results were as follows ; 1) Considering the structural complexity of the tested thermosyphon, it showed good performance for the range of heat flux 2＜ q" ＜25kW/$m^2$ and saturation vapor pressure, 0.1＜Pv＜1.1bar 2) different type of nucleating boiling regimes were observed as described below, -Pulse boiling regime : Flow pattern changed cyclically with time during 1 cycle of pulse boiling process. The onset of Nucleation was followed by expulsive growing of vapor bubble, resulting in the so called blow-up phenomenon, massive expulsion of large amount of liquid around the bubble. -Transient : Some spherical vapor bobbles were observed growing out from 2~3 nucleating sites, that was dispersed at the lower part of the heated tube wall in the liquid pool. But the rest upper region above the nucleating sites were filled with churns or bubbles of vapor. -Continuous nucleate boiling regime : The whole zone of evaporator was filled with lots of spherical vapor bubbles, and the bubbles showed tendency to decrease in diameter as the heat flux increased.ased.

• 한국가시화정보학회지
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• 제18권3호
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.645-650
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• 1997

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). in the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow(ULOF) and the unprotected transient overpower (UTOP). Margins to fuel molting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events.