• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2008년도 하계학술발표대회 논문집
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• pp.1059-1064
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• 2008

The present study is to reduce the thermal phenomena and stack effect of nude elevator of the high-rise building used CFD simulation. Since many High-rise buildings used the curtain-wall glass, thermal phenomena and stack effect can easily occur at hot and cold season, respectively. The simulation has been conducted and verified for the effects of the amount of suppling air to the environment of the inside nude elevator shaft. The results of simulations show that the problems due to the thermal and stack effect will be reduced by enforced ventilation or natural ventilation and those will be presented by temperature and velocity profiles and pressure differences.

• Journal of Welding and Joining
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• 제5권3호
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• pp.28-37
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• 1987

To clarify the debonding phenomena of clad steel, the effect of thermal treatment (temperature, holding time) on the interfacial strength of clad steel was preliminarily investigated. From this study, it was confirmed that the interfacial strength of clad steel was deteriorated by thermal treatment and the amount of strength deteriorated, depending on the condition of thermal treatment, could be evaluated by the following equation. ${\sigma}_{ HT}/{\sigma}_{i}/=A_{0}-A\;exp(-Q/RT)log(t/t_{0})$ This equation implies that temperature has a far strong effect on strength deterioration than tiem. The deterioration of interfacial strength of clad steel after thermal treatment may be derived from the thermal stress caused by the difference in thermal expansion coefficient between component materials and microstructural change along the interface.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1388-1393
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• 2004

Investigators have been perplexed with the thermal phenomena behind the recently discovered nanofluids, fluids with unprecedented stability of suspended nanoparticles although huge difference in the density of nanoparticles and fluid. For example, nanofluids have anomalously high thermal conductivities at very low fraction, strongly temperature-dependent and size-dependent conductivities, and three-fold higher critical heat flux than that of base fluids. Traditional conductivity theories such as the Maxwell or other macroscale approaches cannot explain why nanofluids have these intriguing features. So in this paper, we devise a theoretical model that accounts for the fundamental role of dynamic nanoparticles in nanofluids. The proposed model not only captures the concentration and temperature-dependent conductivity, but also predicts strongly size-dependent conductivity. Furthermore, we physically explain the new phenomena for nanofluids. In addition, based on a proposed model, the effects of various parameters such as the ratio of thermal conductivity of nanofluids to that of a base fluid, volume fraction, nanoparticle size, and temperature on the thermal conductivities of nanofluids are investigated.

• 한국지반공학회논문집
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• 제29권8호
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• pp.65-73
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• 2013

지중 열교환 시스템은 지속적인 에너지 효율의 개선으로 공간 냉난방을 위한 친환경적 에너지 기술로 주목받고 있다. 지중에 매설된 파이프는 내부 유체 순환을 통하여 인접한 지반과 열적 상호작용으로부터 직접적인 열에너지 교환을 수행한다. 하지만, 파이프의 수치모델링에서 열-수리가 연관된 난류해석과 파이프의 긴 세장비에 의한 메쉬사이즈의 부적합성은 열교환 시스템의 적절한 수치해석을 어렵게 하고 있다. 본 논문에서는 파이프 내부 유체흐름에 대한 에너지 보존의 법칙을 적용하여 지배방정식을 유도하였으며, Galerkin수식화와 시간적분을 통하여 열-수리 연동일차원 파이프 요소를 개발하였다. 그리고 제안된 파이프 요소를 기 개발된 다공질 재료를 위한 열-수리-역학(Thermo-Hydro-Mechanical) 해석을 위한 유한요소 프로그램과 결합하였다. 개발된 요소를 이용한 수치해석 결과는 열응답 시험(Thermal Response Test) 결과로부터 주위지반의 유효 열전도도를 평가하기 위하여 사용하는 선형 열원 모델이 인접 파이프간의 열적상호작용과 파이프의 단부효과에 의하여 지반의 열전도도를 과다 평가하는 것으로 보여주었다. 따라서 열응답 시험 해석 결과에 대한 역해석을 적용하여 최적의 수렴성을 보여주는 변환행렬을 제시하였다.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2009년도 추계학술대회 논문집
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• pp.335-338
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• 2009

In this study, in order to investigate the thermal and electrical properties of ZnO arrester block against 60[Hz] AC voltage, the changes in leakage current were measured. The temperature distribution appearing on the ZnO arrester blocks was observed using a forward looking infrared camera. In particular, the correlation between the thermal and electrical properties of a ZnO arrester block was analyzed experimentally. From this analysis, the thermal phenomena resulting from the heat generation and dissipation of the ZnO arrester block were interpreted. The degradation and thermal runaway phenomena of ZnO arrester block are closely related to the temperature limit of the ZnO arrester block. The installation of an additional metal electrode has resulted in the decrease of the leakage current due to the heat dissipation.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.