• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Journal of Fisheries and Marine Sciences Education
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• v.14 no.2
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• pp.177-190
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• 2002

An experimental study was carried out to make clear heat transfer characteristics in flow boiling of binary mixtures of refrigerants R134a and R123 in a uniformly heated horizontal tube. Experiments were run at a pressure of 0.6 MPa both for pure fluids and mixtures in the ranges of heat flux $10{\sim}50{kW/m}^2$, vapor quality 0~100% and mass flux 150-600 $kg/m^2s$. Heat transfer coefficients of mixtures were reduced compared to the interpolated values between pure fluids both in the low quality region where the nucleate boiling is dominant and in the high quality region where the convective evaporation is dominant. Total pressure drop during two-phase flow boiling in a horizontal tube consists of the sum of two components, that is, the frictional pressure drop and pressure drop due to acceleration. The frictional pressure drop is the most difficult component to predict, and makes the most important contribution to the total pressure drop. On the other hand, the acceleration pressure drop resulting from the variation of the momentum flux caused by phase change is generally small as compared to the frictional pressure drop. There is no significant difference in measured pressure drop between mixtures and pure fluids. The correlation of Martinelli and Nelson predicted most of the present data both for pure and mixed refrigerants within 30%.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.6
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• pp.1518-1528
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• 1995

The deformation and departure processes of a bubble attached to the wall are studied experimentally and numerically to understand the phenomena of the nucleate boiling heat transfer enhancement under DC electric field. An air-bubble is injected in a dielectric liquid with different electric fields generated by changing three types of electrode system (Type 1,2 and 3) in the bubble generator. Experimental variables are the electric field strength and the distance and the shape of the electrodes under DC electric field. From experimental results, it is observed that the bubble under Dc electric field is elongated in the same direction as the electric field and the contact angle increases. For the parallel plate electrode which generates a uniform electric field, bubble departure volume doesn't seem to decrease within our experimental range. However, when a needle is raised a few millimeters from the lower electrode to make a nonuniform electric field around the needle, bubble departure volume decreases continuously with the increase of an applied voltage. The reduction effect of bubble departure volume is the most effective under a strong nonuniform electric field generated with Type 3. As the nonuniformity of the electric field due to the shape of a electrode increases, the terminal velocity and the acceleration of a bubble increase largely. For the comparison with visualization results, the deformation of a bubble attached to the electrode is carried out by a numerical method. Numerical results show good agreement qualitatively with experimental results.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.3
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• pp.237-244
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• 2011

An experimental study was carried out to investigate the effect of the inclination angle on the nucleate pool boiling of saturated water at atmospheric pressure. We considered an annulus with a gap of 5 mm and a bottom opening. The inner tube of the annulus was heated, and the outer diameter and the length of the tube were 25.4 mm and 500 mm, respectively. The inclination angle was varied from horizontal to vertical. The results were compared to those for an annulus with a larger gap and a single tube. In the small-gap annulus, the effect of the inclination angle on the heat transfer was not significant. However, an early onset of the critical heat flux was observed at 80 kW/$m^2$ when the annulus was horizontal. Liquid agitation and bubble coalescence were considered to be the major heat-transfer mechanisms.

• Journal of Mechanical Science and Technology
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• v.17 no.10
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• pp.1507-1519
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• 2003

The two-phase flow patterns for both non-loop and loop type oscillating capillary tube heat pipes (OCHPs) were presented in this study. The detailed flow patterns were recorded by a high-speed digital camera for each experimental condition to understand exactly the operation mechanism of the OCHP. The design and operation conditions of the OCHP such as turn number, working fluid, and heat flux were varied. The experimental results showed that the representative flow pattern in the evaporating section of the OCHP was the oscillation of liquid slugs and vapor plugs based on the generation and growth of bubbles by nucleate boiling. As the oscillation of liquid slugs and vapor plugs was very speedy, the flow pattern changed from the capillary slug flow to a pseudo slug flow near the annular flow. The flow of short vapor-liquid slug-train units was the flow pattern in the adiabatic section. In the condensing section, it was the oscillation of liquid slugs and vapor plugs and the circulation of working fluid. The oscillation flow in the loop type OCHP was more active than that in the non-loop type OCHP due to the circulation of working fluid in the OCHP. When the turn number of the OCHP was increased, the oscillation and circulation of working fluid was more active as well as forming the oscillation wave of long liquid slugs and vapor plugs in the OCHP. The oscillation flow of R-142b as the working fluid was more active than that of ethanol and the high efficiency of the heat transfer performance of R -142b was achieved.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.7
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• pp.1713-1721
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• 1994

Intermetallic-matrix composites(IMCs) have the potential of combing matrix properties of oxidation resistance and high temperature stability with reinforcement properties of high specific strength and modulus. One of the major limiting factors for successful applications of these composite at high temperatures is the formation of interfacial reactions between matrix and ceramic reinforcement during composite process and during service. The purpose of the present investigation is to develop a better understanding of the nature of creep fracture mechanisms in a $Ni_{3}Al$ composite reinforced with both $TiB_{2}$ and SiC particulates. Emphasis is placed in the roles of the products of the reactions in determining the creep lifetime of the composite. In the present study, creep rupture specimens were tested under constant ranging from 180 to 350 MPa in vacuum at $760^{\cric}C$. The experimental data reveal that the stress exponent for power law creep for the composite is 3.5, a value close to that for unreinforced $Ni_{3}Al$. The microstructural observations reveal that most of the cavities lie on the grain boundaries of the $Ni_{3}Al$ matrix as opposed to the large $TiB_{2}/Ni_{3}Al$ interfaces, suggesting that cavities nucleate at fine carbides that lie in the $Ni_{3}Al$ grain boundaries as a result of the decomposition of the $SiC_{p}$. This observation accounts for the longer rupture times for the monolicthic $Ni_{3}Al$ as compared to those for the $Ni_{3}Al/SiC_{p}/TiB_{2} IMC$. Finally, it is suggested that creep deformation in matrix appears to dominate the rupture process for monolithic $Ni_{3}Al$, whereas growth and coalescence of cavities appears to dominate the rupture process for the composite.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.4
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• pp.275-281
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• 2016

Subcooled boiling under low pressure was numerically investigated using computational fluid dynamics(CFD). The wall boiling model was used for simulating the subcooled boiling; this model requires sub-models consisting of bubble departure diameter, nucleation site density and bubble departure frequency. The CFD code CFX provides the default models based on experimental data. Because these models are mostly developed under high pressure conditions, it would not be predicted well in low pressure conditions. Thus in this study, CFD validation for subcooled boiling under low pressure was analyzed. The numerical results were compared with experimental data from published paper. Simulations were performed with mass flux ranging from 250 to $750kg/m^2s$, heat flux ranging from 0.37 to $0.77MW/m^2$ and constant outlet pressure of 0.11 MPa. Employing the empirical correlation developed under low pressures could increase the accuracy of numerical analysis.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.6
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• pp.383-389
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• 2016

Based on a surface design with rectangular cavities and channels, we investigated the effects of gravity and capillary pressure on pool-boiling Critical Heat Flux (CHF). The microcavity structures could prevent liquid flow by the capillary pressure effect. In addition, the microchannel structures contributed to induce one-dimensional liquid flow on the boiling surface. The relationship between the CHF and capillary flow was clearly established. The driving potentials for the liquid supply into a boiling surface can be generated by the gravitational head and capillary pressure. Through an analysis of pool boiling and visualization data, we reveal that the liquid supplement to maintain the nucleate boiling condition on a boiling surface is closely related to the gravitational pressure head and capillary pressure effect.