• 한국자동차공학회논문집
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• 제22권5호
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• pp.20-28
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• 2014

The high velocity oxygen fuel spraying (HVOF) is a kind of surface modification process technology to form the sprayed coating layer after spraying the powder to molten or semi-molten state by the ultra-high speed at the high-temperature heat source and conflicting with a substrate. It is desirable to melt completely the thermal spray powder in order to produce the coating layer with an optimal adhesion, however, because a semi-molten powder in a spray process has the low efficiency and become a factor that degrades the mechanical property by the inducement of pore-forming within the coating layer. To improve the wear resistance, corrosion resistance and heat resistance, in this study, the plungers of high-speed and ultra-high pressure reciprocating hydraulic pumps for oil and water used in ironwork are produced with $420J_2$ and the coating layers of plungers are formed by the powders of WC-Co-Cr and WC-Cr-Ni including the high hardness WC. The surface of these plungers is modified by the super-mirror face grinding machine using variable air pressure developed in this laboratory, and then the characteristics of cross-sectional microstructure, and surface roughness and hardness values between no operation and 100 days-operation are examined and made a comparison. The fine tops and bottoms on surface roughness curve of oil-hydraulic pump plunger sprayed by WC-Cr-Ni are molded more and higher than those of water-hydraulic pump sprayed by WC-Co-Cr because the plunger diameter of oil-hydraulic pump is 0.4 times smaller than that of water-hydraulic pump and the pressure of oil-hydraulic pump exerted on the plunger is operated with the 70 bars higher than that of water-hydraulic pump. As a result, it is found that the values of centerline average surface roughness and maximum height for oil-hydraulic pump plunger are bigger than those of water-hydraulic pump plunger.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.403-408
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• 1996

The subchannel analysis was performed to verify the behavior of hot channel characteristics and obtain the information to support the core thermal-hydraulic behavior at post-trip steam line break with low flow condition. During this postulated accident, buoyancy-induced cross flow occurs, and the coupled nuclear and thermal-hydraulic interactions become important. The code predictions with TORC are in good agreement with the test data. Under such conditions, the mass flow increase in the hot channel by buoyancy-induced cross flow depends on the parameter $GR^{*}\;/\;Re^2$, and buoyancy effect becomes more noticeable as $GR^{*}\;/\;Re^2$ increases.

• 한국지반공학회:학술대회논문집
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• 한국지반공학회 2002년도 봄 학술발표회 논문집
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• pp.711-716
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• 2002

When judging the ground by core-logging, depth of coring might be stopped by coming into view of the moderately weathered rock and also considered as bedrock line. However, highly weathered rock may appear again, if coring more, because there are core-stones in the residual soil and highly weathered rock by the effect of hydraulic-thermal differentiation and does the irregular rock weathering or metamorphic rock region. Therefore, there are room for misunderstanding of diagnosing the moderately weathered rock. Even though the irregular ground where the corestones were come out will show clear geotechnical differences between the ground and the gradually weathered bedrock, nowadays, the construction sites do not take into account the characteristic of core-stone region. In conclusion, to study the failure-strength characteristics of soil layers containing core-stones, we made artificial core-stones and varied percentage of corestones, and measured cohesion and friction factors to adjust them to construction sites containing corestones such as slope, tunnel, and underground.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1504-1513
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• 2019

Recently, ocean nuclear reactors have received attention due to enhanced safety features. The movable and transportable characteristics distinguish ocean nuclear reactors from land-based nuclear reactors. Therefore, for safety/design analysis of the ocean reactor, the thermos-hydraulics must be investigated in the moving system. However, there are no studies reporting the general two-fluid equations that can be used for multi-dimensional simulations of two-phase flows in moving systems. This study is to systematically formulate the multi-dimensional two-fluid equations in the non-inertial frame of reference. To demonstrate the applicability of the formulated equations, we perform a total of six different simulations in 2D tanks with translational and/or rotational motions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.175.1-175
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.96-96
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• 2004

• Nuclear Engineering and Technology
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• 제34권5호
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• pp.455-467
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• 2002

RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2005년도 제17회 워크샵 및 추계학술대회
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• pp.587-590
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• 2005

A detailed geothermal characteristics survey with numerical simulations of the heat transfer in a site for ground source heat pump system is necessary for deploying a shallow geothermal utilization system. Density, specific heat, thermal diffusivity, and thermal conductivity are measured on 91 core samples from a 300 m deep borehole in KIGAM(Korea Institute of Geoscience and Mineral Resources). The heat flow is estimated from the thermal gradient and average thermal conductivity and the correlation between fracture system and hydraulic conductivity is analyzed. From the obtained ground information of the study site the performance of the ground heat pump system can be analyzed with some detailed numerical simulations for seasonal heat pump operation skill and optimal system design techniques.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.66-74
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• 1990

원자력 7, 8호기에 장전된 17$\times$17 최적 핵연료집합체의 열수력학적 특성을 원자력 5, 6호기에 장전된 17$\times$17표준 핵연료집합체와 비교하여 분석하였다. 분석된 열수력학적 특성은 정상상태와 과도출력상태에서의 최소 DNBR, 연료봉 중심온도, 출구 기포율등이며, 아울러 연료봉의 중심이 용융되는 국부선출력과 원자로 운전변수들에 대한 DNBR 민감도 계수도 계산하였다. 사용된 코드는 COBRA-IV-I이며, 임계열속 계산에는 R형 그리드에 대해 수정된 W-3 상관식을, 기포율계산에는 drift-flux model을 이용하였다. 계산결과, DNB가 발생할 확률은 최적 핵연료집합체가 더 높았으나, 연료봉의 중심이 용융되는 국부선출력은 표준 핵연료집합체와 거의 동일한 것으로 나타났다.