• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 1998년도 추계 학술대회논문집
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• pp.121-126
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• 1998

A numerical optimization procedure is developed to find the inlet velocity profile that yields the most uniform epitaxial layer in a vertical MOCVD reactor. It involves the solution of fully elliptic equations of motion, temperature, and concentration; the finite volume method based on SIMPLE algorithm has been adopted to solve the Navier-Stokes equations. The overall optimization process is highly nonlinear and has been efficiently treated by the sequential linear programming technique that breaks the non-linear problem into a series of linear ones. The optimal profile approximated by a 6th-degree Chebyshev polynomial is very successful in reducing the spatial non-uniformity of the growth rate. The optimization is particularly effective to the high Reynolds number flow. It is also found that a properly constructed inlet velocity profile can suppress the buoyancy driven secondary flow and improve the growth-rate uniformity.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1158-1161
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• 2003

For underwater milling of parts of nuclear reactor, a waterproof main spindle system was developed. which used a servo meter. Particularly, a waterproof system is available to cope with emergencies such as an electricity failure so that it prevents hazards from cutting radioactive materials. A developed spindle was designed to be capable of horizontal and vertical cutting and structural analysis was conducted with a FEM tool(Design Space) when the forces were loaded in each axis-direction.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.291-298
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• 1997

In advanced nuclear reactors, the passive containment cooling has been suggested to enhance the safety. The passive cooling has two mechanisms, air natural convection and oater cooling with evaporation. To confirm the coolability of PCCS, many works have been performed experimentally and numerically. In this study, the water cooling test was performed to obtain the evaporative heat transfer coefficients in a scaled don segment type PCCS facility which have same configuration with AP600 prototype containment. Air-steam mixture temperature and velocity, relative humidity and well heat flux are measured. The local steam mass flow rates through the vertical plate part of the facility are calculated from the measured data to obtain evaporative heat transfer coefficients. The measured evaporative heat transfer coefficients are compared with an analytical model which use a mass transfer coefficients. From the comparison, the predicted coefficients show good agreement with experimental data however, some discrepancies exist when the effect of wave motion is not considered. Finally, a new correlation on evaporative heat transfer coefficients are developed using the experimental values.

• 한국전산유체공학회지
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• 제18권4호
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• pp.53-60
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• 2013

The boron transport model has been implemented into the CUPID code to simulate the boron transport phenomena of the PWR. The boron concentration conservation was confirmed through a simulation of a conceptual boron transport problem in which water with a constant inlet boron concentration injected into an inlet of the 2-dimensional vertical flow tube. The step wise boron transport problem showed that the numerical diffusion of the boron concentration can be reduced by the second order convection scheme. In order to assess the adaptability of the developed boron transport model to the realistic situation, the ROCOM test was simulated by using the CUPID implemented with the boron transportation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2047-2052
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• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• 한국반도체및디스플레이장비학회:학술대회논문집
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• 한국반도체및디스플레이장비학회 2003년도 추계학술대회 발표 논문집
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• pp.1-6
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• 2003

Flow modulation chemical vapor deposition process has been reported as an alternative way to obtain low resistivity, low residual chlorine content and good step-coverage titanium nitride film. Flow and concentration characteristics in a vertical FMCVD reactor are analyzed by using computational fluid dynamics method. The results show that 1.0 second as Cl reduction period is too short and there is still $TiCl_4$ gas above the holder at the end of the period. Time variation of $TiCl_4$ gas concentration on the holder shows that at least 3.0 second is necessary as Cl reduction time for the sake of film characteristics.

• 한국안전학회지
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• 제11권2호
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• pp.79-88
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• 1996

A hypothetical accident in the containment vessel of liquid metal reactor could cause a pressure, temperature rise, and a strong aerosol release. The computer codes relating to the modelization of these accident make it necessary to use various input parameter, among which is the dynamic shape factor of aerosols produced. Combustion experiments of sodium spray fire carried out in a closed vessel, which was vertical cylinder made of 1.2m in diameter and 1.8m hight with a volume of 1.7$m^3$. The results of theoretical analysis presented here was compared to data obtained from experiments. The experimental results were summarized as follows. 1) The aerodynamic diameter and geometric diameter of aerosols are decreasing with increasing of injection pressure and injection temperature of sodium 2) The dynamic shape factor of aerosol is proportional to the aerodynamic diameter for a given particle. 3) The correspondence between the aerodynamic diameter and geometric diameter can be as $D_{ae}=0.70 D_{ge}$. 4) Peak pressure rose with increase in pressure and temperature of injection sodium, being more sensitive to the injection pressure than the injection temperature.

• 유기물자원화
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• 제8권4호
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• pp.129-137
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• 2000

본 연구에서는 도시고함수율 유기성폐기물(MWWT)의 처리를 위하여, 음식물 쓰레기의 처리에 사용되었던 컨테이너 호기성 퇴비화 시스템과, 계분 및 축분처리에 사용되었던 직립 상향 기계식 교반기를 조합하여 개발하였다. 호기성 처리 공정은 운전이 쉽다는 점과 운전비용이 저렴하다는 점, 자연에서 이용하기 가장 편리한 안정된 부산물을 만들 수 있다는 장점에도 불구하고, 염분의 농축, 과도한 부형제의 사용에 의한 경제적 부담의 증가 및 설비의 대형화 등의 장애물을 가지고 있다. 개발된 공정은 이러한 문제점을 보완하고자 하수슬러지는 초기의 수분조절 및 고속발효를 위해 직립 밀폐형 반응조에서 5일간, 음식물쓰레기는 탈염과 수분조절을 위해 밀폐형 무교반 필터에서, 침출수는 폐수처리 장치에서 처리하였다. 실험결과 국내 퇴비에 품질규격을 만족하면서 부자재의 사용량을 3% 이내로 조절할 수 있어 기존공정의 문제점을 해결할 수 있는 가능성을 확인하였다.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.