• Transactions of the Korean Society of Machine Tool Engineers
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• v.14 no.6
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• pp.88-94
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• 2005

A swirl ratio of a charge in the cylinder could be calculated by measuring both the rotary speed of paddle and the intake air flow rate in the swirl measurement apparatus fur several positions of valve lift. The automation of the swirl ratio measurement for a cylinder head is achieved by controlling both the valve lift of cylinder head and a suction pressure of the surge tank, instead of controlling them manually. PID control of the surge tank pressure and positioning a valve lift of the cylinder head are also achieved by using two step motors, respectively. Rotating speed of a paddle are measured using an optical sensor and a counter. Flow rate are measured from ISA 1932 flow nozzle by reading a differential pressure gauge position using IEEE-1394 camera. Time to measure the swirl ratio for a port in the cylinder head is drastically reduced from an hour to 3 minutes by automation control of the apparatus.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2004.11a
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• pp.139-142
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• 2004

The object of the present work is the acoustic emission energy analysis of subambient pressure tri-pad slider. Head/disk interaction during start/stop and constant speed were detected by using acoustic emission (AE) test system The frequency spectrum analysis is performed using the AE signal obtained during the head/disk interaction Natural frequency analysis was performed using Ansys program. Acoustic emission energy was calculated for the slider modes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.1-5
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• 2015

A failure or degradation of reactor upper head penetration is a troublesome problem at Nuclear Power Plants. A flaw in the reactor upper head penetration can result in unplanned plant shutdown for repair, and cause serious economic losses on the plants. Consequently, a detection of flaws is a matter of more importance. Until now, only the base metal, not including J-groove weld, in reactor upper head penetration has been inspected in accordance with 10 CFR 50.55a and ASME code case N-729-1 requirements. Accordingly, it is rather difficult to detect manufacturing defects and repair defects in J-groove weld. This paper presents a case study on the application of Time of Flight Diffraction UT technique to examine the J-groove weld in reactor head penetration using reactor head penetration mockup with artificial flaws. We expect that this study result will offer a way to understand the non-destructive examination technology for J-groove weld in reactor upper head penetration.

• The Korean Journal of Physiology
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• v.7 no.1
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• pp.13-21
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• 1973

An attempt was made to study circulatory and respiratory responses to the passive tilt. Anesthetized dogs were tilted from horizontal to upright $(+90^{\circ})$ and head down $(-90^{\circ})$ position. The arterial blood pressure was decreased in the upright position and was decreased slightly in the head down position comparing to that in the horizontal position. Cardiac index also decreased in the both upright and head down positions. The total systemic vascular resistance was slightly increased in the upright position and was markedly increased in the head down position. The mean pulmonary arterial pressure was significantly decreased in the both upright and head down positions. The total pulmonary vascular resistance was decreased in the both upright and head down positions. Oxygen consumption was slightly decreased in the upright position, whereas it was slightly increased in the head down position. The A-V $O_2$ difference (vol. %) was slightly increased in the upright position and increased in the head down position. From the above results, process of the circulatory compensation to the gravity in the Passive tilting test was discussed. Neuronal cardiovascular regulation to the gravity and tile adaptation of capacitance vessles to hydrostatic stress and oxygen consumption concerning anoxic endurance of the brain were also discussed.

• The Journal of Korean Physical Therapy
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• v.31 no.1
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• pp.1-6
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• 2019

Purpose: This paper proposes proper and effective neck exercises by comparing the deep and superficial cervical flexor muscle activities and thickness according to the pressure level during cranio-cervical flexion exercises between a normal posture group and forward head posture group. Methods: A total of 20 subjects (8 males and 12 females) without neck pain and disabilities were selected. The subjects' craniovertebral angles were measured; they were divided into a normal posture and a forward head posture group. During cranio-cervical flexion exercises, the thickness of the deep cervical flexor neck muscle and the activity of the surface neck muscles were measured using ultrasound and EMG. Results: The results showed that the thickening of the deep cervical flexor was increased significantly to 28 and 30 mmHg in the forward head posture group. The sternocleidomastoid muscle activity increased significantly to 24, 26, 28, and 30 mmHg in the forward head posture group. The anterior scalene muscle activity increased significantly to 26, 28, and 30mmHg in the forward head posture group. A significant difference of 26, 28, and 30 mmHg in the sternocleidomastoid and anterior scalene muscles was observed between two groups. Conclusion: To prevent a forward head posture and maintain proper cervical curve alignment, the use of the superficial cervical flexor muscles must be minimized. In addition, to perform a cranio-cervical flexion exercises to effectively activate the deep cervical flexor muscles, 28 and 30 mmHg for normal posture adults and 28 mmHg for adults with forward head postures are recommended.

• Journal of the Korean Society of Propulsion Engineers
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• v.14 no.5
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• pp.92-100
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• 2010

Combustion stability rating tests of 75-tonf-class thrust chamber for technology demonstration were carried out at a low pressure. Two kinds of mixing heads were used in this study. One of them has injectors of 631 and the other has injectors of 721. Mixing head with injectors of 631 showed a self-oscillation instability at the chamber pressure of 30 bar. Mixing head with injectors of 721 showed that a high frequency combustion stability was maintained under the same pressure and the same mass flow rate. But mixing head with injectors of 721 generated a self-oscillation instability at the chamber pressure of 20 bar and it was found that stability boundary region was changed due to the configuration of a mixing head from these results.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.10 s.253
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• pp.942-949
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• 2006

A lumped model is proposed to predict liquid ejection characteristics of a thermally driven inkjet print head. The model is based on a two-dimensional heat conduction equation, an empirical pressure-temperature equation and a nonlinear hydraulic flow-pressure equation. It has been simulated through the construction of an equivalent R-C circuit, and subsequently analyzed using SIMULINK and a circuit simulation tool, PLECS. Using the model, heating and cooling characteristics of the head are predicted to be in agreement with the IR temperature measurements. The effects of the head geometry on the drop ejection are also analyzed using the nonlinear hydraulic model. The present model can be used as a design tool for a better design of thermal inkjet print heads.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.11
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• pp.2866-2874
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• 2000

This paper is concerned with the optimum design for external reinforcement of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.

• Proceedings of the KSME Conference
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• 2000.11b
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• pp.377-380
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• 2000

The collapsible phenomena of the circular tubes due to the excessive transmural pressure are investigated experimentally. Collapsible tubes are installed in the test section where the external pressure is applied to the test tubes by applying the hydrostatic head. The collapsible circular tubes are made of rubber, whose diameters are 6 and 4.2 mm, respectively. The hydrostatic water head of the upper reservoir is applied to the test section. Pressures at the upstream and downstream sides are measured by the pressure transducers. The collapsible phenomena are observed as the transmural pressure Increases, and also the flutter phenomenon occurs due to the critical transmural pressure.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.782-787
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• 2001

This paper is concerned with the global rupture of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severe accident. While the rise of the temperature cause the thermal softening of RPV material, the rise of the internal pressure could cause failure of the RPV lower head. The global rupture of an RPV is simulated by finite element limit analysis for the collapse pressure and mode and this analysis results have been compared with a variation of the internal pressure of RPV. The finite element limit method is a systematic tool to secure the safety criteria of a nuclear reactor and to evaluate the in-vessel corium retention.