• Korean Journal of Materials Research
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• v.3 no.5
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• pp.443-450
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• 1993

Self-alignment gatc Schottky contact structure on Si- implanted GaAs was formed by plasma enhanced chemical vapor dcposirion. Tungsten nitride thin films (ahclut 1600$\AA$) \vcre dopositcd on GaAs at $350^{\circ}C$ in order to fahricarc GaAs 1Cs and ttwn rapidly annealed at $750^{\circ}C$ to $900^{\circ}C$. Thermal charac tcristics of PECVD)-$W_{67}N_{43}$/GaAs structure were investigated by X-ray diffraction, photolumintesccnce. and optical deep level transient specrroscopy. Results revealed that $W_{67}N_{33}$ gate was more thermally sta ble with GaAs substrate than W gate and Si atoms implanted In $W_{67}N_{33}$/GaAs structure became morr active than those In W/GaAs after annealing. I-V characteristics of $W_{67}N_{33}$/GaAs diod c exhibired a nearly ideal diode behavior. The termal stability of $W_{67}N_{33}$/GaAs diode was better than that of W/GaAs diode with the post annealing at temperatures from 800 to $900^{\circ}C$ for 20s without As overpressure.

• Transactions on Electrical and Electronic Materials
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• v.8 no.2
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• pp.53-57
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• 2007

In this study, a novel slurry containing ceria as the abrasive particles was analyzed in terms of its frictional, thermal and kinetic attributes for interlayer dielectric (ILD) CMP application. The novel slurry was used to polish 200-mm blanket ILD wafers on an $IC1000_{TM}$ K-groove pad with in-situ conditioning. Polishing pressures ranged from 1 to 5 PSI and the sliding velocity ranged from 0.5 to 1.5 m/s. Shear force and pad temperature were measured in real time during the polishing process. The frictional analysis indicated that boundary lubrication was the dominant tribological mechanism. The measured average pad leading edge temperature increased from 26.4 to $38.4\;^{\circ}C$ with the increase in polishing power. The ILD removal rate also increased with the polishing power, ranging from 400 to 4000 A/min. The ILD removal rate deviated from Prestonian behavior at the highest $p{\times}V$ polishing condition and exhibited a strong correlation with the measured average pad leading edge temperature. A modified two-step Langmuir-Hinshelwood kinetic model was used to simulate the ILD removal rate. In this model, transient flash heating temperature is assumed to dominate the chemical reaction temperature. The model successfully captured the variable removal rate behavior at the highest $p{\times}V$ polishing condition and indicates that the polishing process was mechanical limited in the low $p{\times}V$ polishing region and became chemically and mechanically balanced with increasing polishing power.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.419-426
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• 2012

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.8
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• pp.1119-1127
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• 2010

In this paper, the design and nonlinear simulation of a multistable electromagnetic microactuator, which provides four stable equilibrium positions within its operating range, have been discussed. Quadstable actuator motion has been made possible by using both X- and Y-directional bistable structures with snapping curved beams. Two pairs of the curved beams are attached to an inner frame in both X- and Y-directions to realize independent bistable behavior in each direction. For the actuation of the actuator at the micrometer scale, an electromagnetic actuation method in which Lorentz force is taken into consideration was used. By using this method, micrometer-stroke quadstability in a plane parallel to a substrate was possible. The feasibility of designing an actuator that can realize quadstable motion by using the electromagnetic actuation method has been thoroughly clarified by performing nonlinear static and dynamic analyses and electrothermal coupled-field analysis of the multistable microactuator.

• Fire Science and Engineering
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• v.20 no.3 s.63
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• pp.65-70
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• 2006

From the beginning of the construction stage, the fire protection regulation for the nuclear power plants conforms to the design requirements for the acquisition of the license permit. This regulation is based on the plant status of the normal operation, but it is not enough to be used as an application standard for fire protection at the transient mode of the plant and the outage time for refueling as well as for the plant decommissioning. While the advanced fire protection requirement that has been developed in America recently suggests the performance-based requirement and management rule applicable to the overall life time of the plant, it simply represents the conceptual application. It means that it can not be treated as appropriate standards because it does not deal with the qualitative and quantitative approach in specific ways. By the way, with the use of the performance-based fire risk analysis, the dynamic behavior of the heat and smoke at the fire compartment of the nuclear power plants can be analyzed and the thermal effect to the safety-related equipment and cables can be evaluated as well. At this paper, it suggests the ways to change the applicable fire protection regulations and the required evaluation items for the fire risk resulted from the plant configuration change with an intent to introduce the state-of-the-art quantitative fire risk analysis technology at the domestic nuclear power plants.

• Fire Science and Engineering
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• v.24 no.3
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• pp.131-138
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• 2010

Experimental and numerical studies were conducted to investigate the thermal and chemical characteristics of heptane fires in a full-scale ISO 9705 room. Representative fire conditions were considered for over-ventilated fire (OVF) and under-ventilated fire (UVF). Fuel flow rate and doorway width were changed to create OVF and UVF conditions. Detailed comparisons of temperature and species concentrations between experimental and numerical data were presented in order to validate the predictive performance of FDS (Fire Dynamic Simulator). The OVF and UVF were explicitly characterized with distributions of temperature and product formation measured in the upper layer, as well as combustion efficiency and global equivalence ratio. It was shown that the numerical results provided a quantitatively realistic prediction of the experimental results observed in the OVF conditions. For the UVF, the numerically predicted temperature showed reasonable agreement with the measured temperature. The predicted steady-state volume fractions of $O_2$, $CO_2$, CO and THC also agreed quantitatively with the experimental data. Although there were some limitations to predict accurately the transient behavior in terms of CO production/consumption in the UVF condition, it was concluded that the current FDS was very useful tool to predict the fire characteristics inside the compartment for the OVF and UVF.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.