http://chineseinput.net/에서 pinyin(병음)방식으로 중국어를 변환할 수 있습니다.
변환된 중국어를 복사하여 사용하시면 됩니다.
원자로 용기 외벽 냉각을 통한 차세대 원전 관통부의 건전성 평가를 위한 지속 가열 실험 연구
강경호(K. H. Kang),박래준(R. J. Park),민병태(B. T. Min),김상백(S. B. Kim),이기영(K. Y. Lee),박종균(J. K. Park) 대한기계학회 2002 대한기계학회 춘추학술대회 Vol.2002 No.5
Under the same induction heating power histories, two tests named as KNGR-SUS-DRY and KNGR-SUS-EXT<br/> were performed varying the cooling conditions at the outer surface of the lower head vessel. The maximum heat flux<br/> imposed to the inner surface of the lower head vessel was about 0.35 MW/m2 in both the tests. In the KNGR-SUS-DRY<br/> test performed without the external vessel cooling, the inner surface of the lower head vessel was ablated by the<br/> thickness of 45 mm, which indicated the total ablation of the welding material. Although the welding material was<br/> ablated in all, the penetration tube was not ejected outside the lower head vessel due to the thermal expansion of the<br/> lower head vessel and the penetration tube. In the KNGR-SUS-EXT test performed with the external vessel cooling,<br/> however, the thickness of the ablation is about 10 mm at most.
복합혼합날개를 장착한 5×5 봉다발에서 부수로 혼합 및 임계열유속 실험 연구
강경호(K. H. Kang),신창환(C. H. Shin),추연준(Y. J. Choo),윤영중(Y. J. Youn),박종국(J. K. Park),문상기(S. K. Moon),천세영(S. Y. Chun) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5
Experiments were performed to determine the thermal (or turbulent) diffusion coefficient (TDC) and to investigate the critical heat flux (CHF) performance in the 5×5 rod bundle with 5 unheated rods which are supported by Hybrid Mixing Vane. In this study, HFC-134a fluid was used as working fluid and the fluid temperature were measured in the important subchannels. To determine the TDC value, the measured fluid temperatures were compared with the predicted values obtained from the MATRA code. The best optimized value of β was found to be 0.02 by considering prediction statistics, i.e., average and standard deviations of the differences between the experimental results and code calculations. Using the best optimized value of β as 0.02, the MATRA code predicts the test results of the fluid temperature within ±1.0 % of error. According to the experimental results on CHF of 5 non-heating guide tubes, the case with non-heating guide tube showed a little good performance in terms of CHF.
FIRST ATLAS DOMESTIC STANDARD PROBLEM (DSP-01)FOR THE CODE ASSESSMENT
김연식,최기용,강경호,박현식,조석,백원필,김경두,SUK K. SIM,EO-HWAK LEE,SEYUN KIM,김주성,TONG-SOO CHOI,CHEOL-WOO KIM,SUK-HO LEE,SANG-IL LEE,KEO HYOUNG LEE 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.1
KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for AccidentSimulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basisaccidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed andsuccessfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from theATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermalhydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. Forthe first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 wasselected by considering its technical importance and by incorporating comments from participants. Twelve domesticorganizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. ThisATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participantsprior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was alsoused by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison resultsbetween the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations forcode users as well as for developers are suggested.
2층으로 성층화된 풀 내에서의 자연대류 열전달과 고화현상에 대한 연구
김종태(J. Kim),강경호(K. S. Kang),김상백(S. B. Kim),김희동(H. D. Kim) 한국전산유체공학회 2001 한국전산유체공학회지 Vol.6 No.1
The natural convection heat transfer and solidification in a stratified pool arc studied. The flow and heat transfer characteristics in a heat generating pool are compared between single-layered and double-layered pools. And local Nusselt number distributions on outer walls are obtained to consider thermal loads on a vessel wall. The cooling and solidification of Al₂O₃/Fe melt in a hemispherical vessel are simulated to study the mechanism of heat transfer and temperature distribution. A unstructured mesh is chosen for this study because of the non orthogonality originated from the boundaries of double-layered pool. Interface between the layers is modeled to be fixed. With this assumption mass flux across the interface is neglected, but shear force and heat flux are considered by boundary conditions. The colocated cell -centered finite volume method is used with the Rhie-Chow interpolation to compute cell face velocity- To prevent non- physical solutions near walls in case body force is large the wall pressure is extrapolated by the way to include body force. The numerical solutions calculated by current method show that averaged downward heat flux of the double-layered pool increases compared to single-layered pool and maximum temperature occurs right below the interface of the layers.