• 한국태양에너지학회 논문집
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• 제37권1호
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• pp.15-23
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• 2017

Critical heat flux refers to the sudden decrease in boiling heat transfer coefficient between a heated surface and fluid, which occurs when the phase of the fluid near the heated surface changes from liquid to vapor. For this reason, critical heat flux is an important factor for determining the maximum limit and safety of a boiling heat transfer. Recently, it is reported that the nanofluid is used as a working fluid for the critical heat flux enhancement. However, it could be occurred nano-flouling phenomena on the heat transfer surface due to nanoparticles deposition, when the nanofluid is applied in a heat transfer system. In this study, we experimentally carried out the effects of the nano-fouling phenomena in oxidized multi-wall carbon nanotube and oxidized graphene nanofluid systems. It was found that the boiling heat flux decreased by hourly 0.04 and $0.03kW/m^2$, also the boiling heat transfer coefficient decreased by hourly 11.56 and $10.72W/m^2{\cdot}K$, respectively, in the thermal fluid system using oxidized multi-wall carbon nanotube or oxidized graphene nanofluid.

• Design & Manufacturing
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• 제13권1호
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• pp.42-47
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• 2019

Nucleate boiling heat transfer is one of the most important phenomenon in the various industries. Especially, critical heat flux (CHF) refers to the upper limit of the pool boiling heat transfer region. Therefore, many researchers have found that CHF can be significantly increased by adding very small amounts of nanoparticles. In this study, the CHF and heat transfer coefficient were tested under the pool boiling state using copper and multi wall carbon nanotube nanoparticles. The results showed that two different types of nanoparticles deposited on the surface of two specimens made of the same material increased the heat flux in the nanoparticles with high conductivity, and there was no difference in the critical heat flux when the same material nanoparticles were deposited on the two different specimen surfaces.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.124-133
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• 2017

In-vessel retention by passive external reactor vessel cooling under severe accident conditions is a viable approach for retention of radioactive core melt within the reactor vessel. In this study, a new and versatile coating technique known as "cold spray" that can readily be applied to operating and advanced reactors was developed to form a microporous coating on the outer surface of a simulated reactor lower head. Quenching experiments were performed under simulated in-vessel retention by passive external reactor vessel cooling conditions using test vessels with and without cold spray coatings. Quantitative measurements show that for all angular locations on the vessel outer surface, the local critical heat flux (CHF) values for the coated vessel were consistently higher than the corresponding CHF values for the bare vessel. However, it was also observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit vary appreciably along the outer surface of the test vessel. Nonetheless, results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward-facing boiling by > 88%.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.139-146
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• 2006

An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This non-monotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.161-161
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• 1999

• Nuclear Engineering and Technology
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• 제44권4호
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• pp.429-436
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• 2012

Critical heat flux (CHF) is the thermal limit of a phenomenon in which a phase change occurs during heating (such as bubbles forming on a metal surface used to heat water), which suddenly decreases the heat transfer efficiency, thus causing localized overheating of the heating surface. The enhancement of CHF can increase the safety margins and allow operation at higher heat fluxes; thus, it can increase the economy. A very interesting characteristic of nanofluids is their ability to significantly enhance the CHF. Nanofluids are nanotechnology-based colloidal dispersions engineered through the stable suspension of nanoparticles. All experiments were performed in round tubes with an inner diameter of 0.01041 m and a length of 0.5 m under low pressure and low flow (LPLF) conditions at a fixed inlet temperature using water, 0.01 vol.% $Al_2O_3$/water nanofluid, and SiC/water nanofluid. It was found that the CHF of the nanofluids was enhanced and the CHF of the SiC/water nanofluid was more enhanced than that of the $Al_2O_3$/water nanofluid.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.235-240
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• 2001

The CHFG (Critical Heat Flux in Gap) test results have been evaluated to quantify the critical power in hemispherical narrow gaps and a new correlation has been developed. The CHFG test results have shown that increases in the gap thickness and pressure lead to an increase in critical power. The pressure effect on the critical power was found to be much milder than predictions by CHF correlations of other researches. From the CHFG test results, a new correlation on critical power in the hemispherical gap has been developed using the non-dimensional parameters as follows: $$\frac{qCHF}{{\rho}g^hfg}{\cdot}4\sqrt{\frac{{\rho}_g^2}{g{\sigma}{\Delta}{\rho}}=\frac{0.1042}{1+0.1375({\rho}g/{\rho}l)^{0.21}(D/s)}$$ The developed correlation has been expanded to apply the spherical geometry using the Siemens/KWU's correlation.

• 설비공학논문집
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• 제27권1호
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• pp.1-7
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• 2015

A boiling heat transfer system is used in a variety of industrial processes and applications, such as refrigeration, power generation, heat exchangers, cooling of high-power electronics components, and cooling of nuclear reactors. The critical heat flux (CHF) is the thermal limit during a boiling heat transfer phase change; at the CHF point, the heat transfer is maximized, followed by a drastic degradation beyond the CHF point. Therefore, Enhancement of CHF is essential for economy and safety of heat transfer system. In this study, the CHF and heat transfer coefficient under the pool-boiling state were tested using multi-wall carbon nanotubes (MWCNTs) CM-95 and CM-100. These two types of multi-wall carbon nanotubes have different sizes but the same thermal conductivity. The results showed that the highest CHF increased for both MWCNTs CM-95 and CM-100 at the volume fraction of 0.001%, and that the CHF-increase ratio for MWCNT CM-100 nanofluid with long particles was higher than that for MWCNT CM-95 nanofluid with short particles. Also, at the volume fraction of 0.001%, the MWCNT CM-100 nanofluid indicated a 5.5% higher CHF-increase ratio as well as an approximately 23.87% higher heat-transfer coefficient increase ratio compared with the MWCNT CM-95 nanofluid.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.217-242
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• 2011

A variety of Generation III/III+ water-cooled reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world in efforts to solve the future energy supply shortfall. Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. Phase change by boiling and condensation in the reverse process is a highly efficient heat transport mechanism that accommodates large heat fluxes with relatively small driving temperature differences. This mode of heat transfer is encountered in a wide spectrum of nuclear systems,and thus it is necessary to determine the thermal limit of water-cooled nuclear energy conversion in terms of economic and safety. Such applications are being advanced with the introduction of new technologies such as nanotechnology. Here, we investigated newly-introduced nanotechnologies relevant to boiling and condensation in general engineering applications. We also evaluated the potential linkage between such new advancements and nuclear applications in terms of advanced nuclear thermal-hydraulics.