• Journal of Korean Living Environment System
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• v.16 no.2
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• pp.89-100
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• 2009

The present study was designed to observe the thermal dynamics of core and skin temperatures during sub-maximal treadmill exercise; to investigate the effect of regional body cooling during short rest after the treadmill exercise on the thermal dynamics. Three conditions (No cooling, Head/Hand cooling, Leg cooling) were simulated in a climatic chamber at 24±1℃ and 50±5%RH. Subjects performed two bouts of treadmill exercise at a rate of 80%HR_max followed by rest. Body cooling with a hood, long gloves, and a blanket that circulated water set at 15℃ was assigned during two bouts of rest. The results showed that (1) rectal temperature (T_re) did not show significant difference between three conditions; (2) Skin temperatures had specific features, depending on body regions. In particular, the initial fall phenomena of skin temperatures at the onset of exercise were noteworthy in the chest, thigh, calf, and finger tip. Of these, the most significant initial fall was found in finger temperature (T_fing). (3) During the period of the initial fall in skin temperatures, T_re gradually increased. (4) The magnitude of the fall of T_fing at the onset of 2^nd running was on average 4.8, 5.1 and 3.4℃ for Control, HH cooling, and Leg cooling, respectively (p<0.05). The initial drop of T_fing at the onset of running was maintained for an average of 8.1, 7.9 and 6.3 minutes for Control, HH cooling, and Leg cooling, with no significant differences. In conclusion, the initial fall phenomena at the onset of treadmill exercise reflected non-thermal factors, as opposed to internal thermal status. The magnitude of the initial fall in T_fing was affected by legs cooling. Therefore, the initial fall phenomenon should be considered when interpreting the thermal status of the shell during heavy works/exercises that assigned with intermittent regional body cooling.

• Journal of Biomedical Engineering Research
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• v.38 no.4
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• pp.163-167
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• 2017

The purpose of this study is to perform integrated body temperature cooling of the arteriovenous anastomosis site. In the arteriovenous anastomosis site, heart cooling was performed using the palm of the hand, Cooling was achieved by using Peltier and copper plates to cool the palm of the hand with the heat transferred. The control range of the conducted heat is adjustable from 25 degrees to 30 degrees. The experimental environment was to place the treadmill in the house, The temperature in the house was set at 40 degrees and the experimenter treadmill at a speed of 5 Km. The subjects were exercised until the body temperature reached about $39^{\circ}C$. As a method to lower the body temperature after the experiment, the data of the body temperature was obtained by the general rest, onehand cooling, two-hand cooling. Experiment result better than normal rest when Two hands cooling and an average decrease of 0.66 degrees. if you develop a cooling glove with Peltier, it will be an epoch-making athletic assistant to achieve thermal fatigue.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2444-2448
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• 2007

The multi channel of DDI which is the core part of the LCD-TV has been propelled. When multi channel in DDI is introduced, it brings a thermal problem because of the increased power. To solve the thermal problem of the DDI it needs to be investigated each at the package level and module level. It is important to extract the junction temperature(Tj) of DDI clearly from the system level. The objective of this research is to construct a compact model. The compact model is to reduce LCD module including DDI. When the compact model is used, it will be able to easily handle the boundary condition and accurately predict the temperature. Consequently, the temperature of DDI can be calculated easily at the system level. Through this research,we also proposed the cooling plan of DDI for a protection of overheating. The cooling plan was utilized in DDI design.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.531-540
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• 1998

The DYLAM-3 code which overcomes the limitation of event tree/fault tree was applied to LOOP (Loss of Off-site Power) in the mid-loop operation employing HEPs (Human Error Probabilities) supplied by the ASEP (Accident Sequence Evaluation Program) and the SEPLOT (Systematic Evaluation Procedure for Low power/shutdown Operation Task) procedure in this study. Thus the time history of core uncovery frequency during the mid-loop operation was obtained. The sensitivity calculations in the operator's actions to prevent core uncovery under LOOP in the mid-loop operation were carried out. The analysis using the time dependent HEP was performed on the primary feed & bleed which has the most significant effect on core uncovery frequency. As the result, the increment of frequency is shown after 200 minutes duration of simulation conditions. This signifies the possibility of increment in risk after 200 minutes. The primary feed & bleed showed the greatest impact on core uncovery frequency and the recovery of the SCS (Shutdown Cooling System) showed the least impact. Therefore the efforts should be taken on the primary feed & bleed to reduce the core uncovery frequency in the mid-loop operation. And the capability of DYLAM-3 in applying to the time dependent concerns could be demonstrated.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2006.05a
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• pp.311-316
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• 2006

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Journal of Magnetics
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• v.11 no.4
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• pp.160-163
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• 2006

We have prepared crystalline Ag nanoparticles with an average size of 4 nm in diameter by using an inductively coupled plasma reactor equipped with the liquid nitrogen cooling system. Our magnetic data show that the nano-sized effect of Ag nanoparticles on the magnetic properties is ferromagnetic, instead of a diamagnetic component of the Ag bulk and a superparamagnetic component of magnetic nanoparticles. We have also studied the magnetic properties of Ag-Cu nanocomposites with an opposite concentration profile between surface and core. These comparisons indicate that the ferromagnetic component strongly depends on the surface of Ag nanoparticles, while the paramagnetic component is strongly affected by the outer oxide layer, with the background of a diamagnetic component from the core of Ag.

• Proceedings of the KSME Conference
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• 2000.11b
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1431-1441
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• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.