• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.134.1-134
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.81.2-81.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.82.2-82.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.442-443
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• 2005

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3928-3942
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• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• 수산해양교육연구
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• 제9권1호
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• pp.83-97
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• 1997

Diesel engine is widely used for ship and industry source of power because of its high thermal efficiency and reliability and durability. However it lead to air pollution due to exhaust gas, and it is important to develop diesel engine of lower air-pollution to decrease the hazardous exhaust gas emissions. As one of the ways, the study for practically using the high pressure of fuel injection and variable injection timing system is being processing. The high pressure injection, which is said to be an effective means for reducing both NOx and particulate emissions, and great improvements in combustion characteristics have been reported by many researchers. In this study, electronic-hydraulic fuel injection system and hydraulic fuel injector system have been applied to the D.I. test engine for high pressure injection and variable injection timing. The injection pressure and injection rate depending upon accumulator pressure were measured with strain gage and Bosch injection rate measuring system before fitting the system into test engine, and analyzed the characteristics of the injection system. The combustion characteristics with this injection system has been analyzed with data concerning heat release rate, pressure rising rate, ignition point, ignition delay and maximum pressure value.

• 환경정책연구
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• 제14권2호
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• pp.3-25
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• 2015

도심재생하천 내부의 수리적 특성이 열환경 변화에 미치는 영향을 알아보고자 하천의 물리적 구조와 수리적 특성, 기상요소, 온열환경을 분석하였다. 연구 대상지는 복원된 도시하천으로서 큰 의의를 가지는 청계천을 선정하였다. 유형은 물리적 구조별 녹피율의 차이에 따라 유형 I(0.0%)과 유형 II(20.2%)로 구분하였다. 여울이 끝나는 Ba지점에서 수온은 $0.2^{\circ}C$ 감소, 유속은 0.7~0.9m/s 증가, 용존산소량은 0.5~0.6mg/L 증가하였으며, 기온은 평균 $1.1{\sim}1.4^{\circ}C$ 감소, 상대습도는 평균 6.6~8.7% 증가, 풍속은 불규칙한 변화를 보였다. 습구흑구온도지수 분석결과, 유형에 따른 상류부의 하천 내부와 수면 외부의 값의 차이는 미비하였으나, 유형 I의 하류부에서 감소폭은 $0.3{\sim}0.6^{\circ}C$, 유형 II는 $0.8^{\circ}C$ 등이었다. 기온저감의 효과는 수직높이로 유형 I의 경우 120cm, 유형 II는 140cm까지 영향이 있었으며, 거리별 기온저감의 효과는 유형 I의 경우 여울이 끝나는 Ba지점 이후로 증가하였고, 유형 II의 경우 지속해서 감소하였다.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2006년도 하계학술발표대회 논문집
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• pp.188-193
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• 2006

In present study, the thermal-hydraulic characteristics of the subchannels are investigated as measuring single-phase heat transfer coefficients and the cross sectional velocity field using LDV in the downstream of support grid in $6{\times}6$ rod bundles. Support grid with mixing vanes make enhancing heat transfer in rod bundles by generating turbulent flow. But this turbulent flow only is reserved in a short distance. Support grid with LSVF mixing vanes keep the turbulent flow a long distance. The experiments are performed at the nominal Reynolds number 30,000 and 50,000. The heat transfer coefficients are measured using heated and unheated copper sensor. In this study, the comparison of local heat transfer coefficients for LSVF mixing vane and split mixing vane is represented.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.915-920
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• 2001

A spent fuel problem has prevented the nuclear power from claiming to be a completely clean energy source. The nuclear transmutation technology to incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(Hybrid Power Extraction Reactor) is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Some major feature of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth(Pb-Bi) is adopted as a coolant and Spallation target material. HYPER is a subcritical reactor which needs an external neutron source. 1GeV proton beam is irradiated to Lead-bismuth(Pb-Bi) target inside HYPER, and spallation neutrons are produced. When proton beams are irradiated, much heat is also deposited in the Pb-Bi target and beam window which separates Pb-Bi and accelerator vacuum. Therfore, an effective cooling is needed for HYPER target. In this paper, we performed the thermal-hydraulic analysis of HYPER target using FLUENT code, and also calculated thermal and mechanical stress of the beam window using ANSYS code.