• Nuclear Engineering and Technology
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• v.20 no.2
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• pp.112-119
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• 1988

The purpose of this work is to obtain the experimental data for the forced How dryout heat flux in a heat generating stratified debris bed which simulates the degraded nuclear reactor core after severe accident. The present observations were mainly focused on the effect of coolant mass flux on the dryout heat flux in the stratified debris bed which consists of several layers with selected particle sizes under constant bed depth and temperature of inlet coolant flow conditions. The following results were obtained from this experimental work: (1) The dryout heat flux in the stratified debris bed increases with increase of upward forcing mass flux of coolant. The similar trend of increase rate of dryout heat flux in the stratified bed was observed in the uniform particle size bed. (2) For the comparison of theoretical values and experimental data, the use of surface area mean diameter as a particle diameter was suitable for the calculation of dryout heat flux.

• Transactions of the KSME C: Technology and Education
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• v.3 no.3
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• pp.201-207
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• 2015

Air cooled condenser for power plant is used at inland area of desert or mountainous area because condenser coolant like sea water is not necessary. However, the performance of air cooled condenser is influenced by ambient condition such as wind speed and air temperature. Therefore, various devices have been designed to improve the performance of air cooled condenser. In this study, the CFD analysis for air cooled condenser was carried out according to wind speed and wind screen configuration. As wind speed increased from 3m/s to 7m/s, the fan flow rate was reduced about 15.76% and the rise of inlet air temperature was 5.55 degree of Celsius. When the suitable wind screen is equipped, the fan flow rate went up about 5.18% and inlet air temperature dropped by 2.08 degree of Celsius in comparison with original case without wind screen at 7m/s wind speed.

• Journal of Advanced Marine Engineering and Technology
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• v.28 no.1
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• pp.65-74
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• 2004

The heat transfer coefficient and pressure drop of $CO_2$(R-744) during gas cooling Process of carbon dioxide in a horizontal tube were investigated experimentally and theoretically. The experiments were conducted without oil in the refrigerant loop. The main components of the refrigerant loop consist of a receiver. a variable-speed pump. a mass flowmeter, an evaporator. and a gas cooler(test section). The main components of the water loop consist of a variable-speed Pump. an constant temperature bath. and a flowmeter. The gas cooler is a counterflow heat exchanger with refrigerant flowing in the inner tube and water flowing in the annulus The test section consists of smooth, horizontal stainless steel tube of 9.53 mm outer diameter and 7.75 mm inner diameter. The length of test section is 6 m. The refrigerant mass fluxes were 200 ~ 300 kg/($m^2{\cdot}s$) and the inlet pressure of the gas cooler varied from 7.5 MPa to 8.5 MPa. The main results were summarized as follows : The predicted correlation can evaluated the R-744 exit temperature from the gas cooler within ${\pm}10%$ for most of the experimental data, given only the inlet conditions. The predicted gas cooley capacity using log mean temperature difference showed relatively food agreement with gas cooler capacity within ${\pm}5%$. The pressure drop predicted by Blasius estimated the pressure drop on the $CO_2$ side within ${\pm}4.3%$. The predicted heat transfer coefficients using Gnielinski's correlation evaluated the heat transfer coefficients on the $CO_2$ side well within the range of experimental error. The predicted heat transfer coefficients using Gao and Honda's correlation estimated the heat transfer coefficients on the coolant side well within ${\pm}10\;%$. Therefore. The predicted equation's usefulness is demonstrated by analyzing data obtained in experiments.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.251-259
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• 2003

The Taguchi method was applied to investigate the effect of design factors on the performance of the shutdown cooling heat exchanger in the SMART-P. This method provided the simulation matrix for the KDESCENT program and an efficient tool for analyzing the simulation results. Levels of the design factors were selected by the effectiveness-NTU method. From 18 runs with the KDESCENT program, it was found that the performance of the system was greatly influenced by the inlet temperature at the shell side and the mass flow rate of the reactor coolant at the tube side. After applying the Taguchi method, we identified the important design factor that should be controlled and designed carefully. This method provides an efficient way to estimate the influence of each design factor on a system performance.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.2
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• pp.1-6
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• 2019

The objective of this study was to investigate performance characteristics of thermal management system(TMS) in a fuel cell electric vehicle with 100kW Fuel Cell(FC) system. In order to build up analytic modelling for TMS, each component was installed and tested under various operating conditions, such as water pump, radiator, 3-Way valve, COD heater, and FC stack etc. and as the results of them, correlations reflecting component's characteristics with flow rate, air velocity were developed. Developed analytic modelling was carried out under various operating conditions on the road. To verify modelling's accuracy, after prediction for optimum coolant flow rate was fulfilled under certain operating conditions, such as FC system, water pump speed, opening of 3-way valve, and pipe resistance, analytic and experimental values were compared and good agreement was shown. In order to predict cold-start operating performance for analytic modelling, coolant temperature variation was analyzed with $-20^{\circ}C$ ambient temperature and duration was predicted to rise in optimum temperature for FC. Because there is appropriate temperature difference between inlet and outlet of FC stack to operate FC system properly, related analysis was performed with respect to power consumption for TMS and heat rejection rate and performance map was depicted along with FC operating conditions.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.273-280
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• 1986

The purpose of this study is to obtain the experimental data of the forced flow dryout heat flux in a heat generating debris bed which simulates the degraded nuclear reactor core after severe accident. An experimental investigation has been conducted of dryout heat flux in an inductively heated bed of steel particles with upward forced flow rising coolant circulation system under atmospheric pressure. The present observations were mainly focused on the effects of coolant mass flux, particle size, bed height, and coolant subcooling on the dryout heat flux The data were obtained when carbon steel particles in the size distribution 1.5, 2.5, 3.0 and 4.0 mm were placed in a 55 mm ID Pyrex glass column and inductively heated by passing radio frequency current through a multiturn work coil encircling the column. Distilled water was supplied with variation of mass flux from 0 to 3.5 kg/$\textrm{cm}^2$ s as a coolant in the tests, while the bed height was selected as 55 mm and 110 mm. Inlet temperature of coolant varied by 2$0^{\circ}C$ and 8$0^{\circ}C$. The principal results of the tests are: (1) Dryout heat flux increases with increase of upward forcing mass flux and particle size; (2) The dryout heat flux at the zero mass flux obviously depends on the Particle size as Previous studies; (3) The forced flow dryout heat flux in the shallow bed is somewhat higher than that in the deep bed,

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Transactions of Materials Processing
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• v.14 no.4 s.76
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• pp.368-374
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• 2005

In the present study we used a diagrammatic analysis of 6 sigma quality control and Taguchi method for injection molding of monitor back-cover, evaluated the influence on the cycle time with part design, mold design, molding process and standardization activity involving design and molding, adopted analysis of sensitivity and effective factors of the part design and molding process conditions for productivity, identified main design molding factors. The contributing factors for the final cycle time could be enumerated as follows; the thickness of hot spot, main nominal part thickness, coolant inlet temperature, melt temperature and cooling line layout, etc.. As a first step, all the critical factors of design process applied to the current monitor housing were investigated through 6 sigma process. Thereafter, the optimal and better critical factors found in the first step were applied to new product design to prove that our process was correct. The Moldflow was used for injection molding simulation, and Minitab software for the statistical analysis, respectively. Finally, the productivity of new design was increased about 33 percents for our specific case.