Strength and Deformation Characteristics of Hamaoka sand by Torsional Shear Test

황성춘,박춘식 國立 昌原大學校 産業技術硏究所 1999 産技硏論文集 Vol.13 No.-

미소변형률(10??이하)에서 전단파괴 이후까지 Hamaoka 모래 공시체의 단조 및 반복 비틀림 단순전단시험을 실시하여 강도.변형특성을 조사하였다. 그 결과 다음과 같은 결론을 얻었다. 단조비틀림전단시험에서 얻어진 최대전단계수(Gmax)와 반복비틀림전단시험에서 얻어진 Gmax는 거의 같았다. 또, 단조비틀림전단시험에서 얻어진 할선전단계수(Geq)의 변형률의존성은 반복비틀림전단시험에서 얻어진 할선전단계수(Gsec)의 변형률의존성에 비해 컸다. 한편, 단조비틀림전단시험의 정규화한 최대내부마찰각은 평면변형률시험의 그것과 매우 유사하였다.

American Medical Association 5판을 근거로 한 한국형 말초신경계 장애 평가기준 개발

박동식,김병조,황선철,송인욱,손병철,장성호,정두신 순천향대학교 의학연구소 2009 Journal of Soonchunhyang Medical Science Vol.14 No.3

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심초음파로 결정된 응급 심낭천자술의 천자부위

김성환,황성오,이강현,조준휘,강구현,문중범,이승환,윤정한,최경훈,김영식 대한응급의학회 2000 대한응급의학회지 Vol.11 No.3

Background: The aim of this study was to determine whether the conventional subcostal approach is suitable for emergency pericardiocentesis in patients with cardiac tamponade or impending cardiac tamponade. Methods: This study was a prospective, observational study conducted at the emergency department of a tertiary hospital, Patients who had symptomatic pericardial effusion and who needed emergency pericardiocentesis in the emergency department were included in this study. We measured the epicardium-to-pericardium distance at the subcostal, parasternal, and apical area with two-dimensional echocardiography to determine the appropriate puncture site for pericardiocentesis. An epicardium-to-pericardium distance of more than 1.0 cm was considered as the primary safety factor in determining the Puncture site for pericardiocentesis. The skin-to-pericardium distance was considered as secondary safety factor. Results: Ninety-five consecutive patients(55 males and 40 females; total mean age: 53 year old) with cardiac tamponade or impending cardiac tamponade were enrolled in this study. The puncture site for pericardiocentesis, as determined by echocardiography, was the subcostal area in 43 patients(45%), the apical area in 40 patients(42%), the left parasternal area In 11 patients(12%), and the right parastemal area in one patient(1%). Pericardiocentesis failed in 2 patients(2%) with the subcostal approach and in one patient(1%) with the apical approach. The average epicardium-to-pericardium distance was 31 ±21 mm in patients with the subcostal approach and 21±8 mm in patients with other approaches. There were no differences in the amount of pericardial fluid and in the intraperical pressure among patient groups according to puncture site. There were two procedure related complications: a puncture of the right ventricle with the subcostal approach and a ventricular tachycardia with the apical approach.

평면변형압축시험에 의한 각종 모래의 변형특성 이방성

박춘식,황성춘 國立 昌原大學校 産業技術硏究所 2000 産技硏論文集 Vol.14 No.-

공중낙하법에 의해 만든 등방압밀 모래공시체를 미소변형률 측정장치를 사용한 평면변형률압축시험을 실시하여 미소변형률에서 파괴후까지의 강성률에 대한 이방성을 연구하였다. 세계 각국의 주요 연구기관에서 사용되고 있는 7종류의 연구용 표준사 공시체를 멤브레인의 관입에 의한 오차와 변위를 외부에서 측정함으로 하여 생기는 오차(bedding error) 등의 영향을 제거하여 측정한 최대주응력방향의 변형률과 최소주응력방향의 변형률을 각각 0.0001%에서 10%까지 넓은 범위에 걸친 응력-변형률 관계를 얻었다. 그 결과 최대 영률 ??은 퇴적면과 최대주응력 σ₁이 이루는 각도 δ에 관계없이 일정하였다. 그러나, 정규화한 ??은 모래의 종류에 따라 달랐다. 또, 강성률의 변형률 수준과 응력 수준에 대한 의존성은 δ가 감소함에 따라 증가하였다. Anisotropy of stiffness, from extremely small strains to post-failure strains, of isotropically consolidated air-pluviated sands in plane strain compression was studied by using the newly developed instrumentation for small strain measurements. Seven types of sand of the world-wide origins were tested, which have been extensively used for research purposes. Stress-strain relationships for a wide range of strain from about 0.0001% to 10% were obtained with measuring axial and lateral strains locally free from the effects of bedding and membrane penetration errors at the specimen boundaries. It was found that the maximum Young's modulus ?? was irrespective of the angle δ of the σ₁direction relative to the bedding plane. However, the normalized ?? was varied with the types of sand. Furthermore, the dependency of the strain and stress level on the stiffness was increased as δ decreased.

열공성 경색 환자의 경동맥 초음파소견 및 혈청지방의 분포

여상준,민병국,황성남,석종식,김영백,박승원,권정택,최덕영 중앙대학교 의과대학 의과학연구소 2001 中央醫大誌 Vol.26 No.1

Failure Evaluation Plan of a Reactor Internal Components of a Decommissioned Plant

( Seong Sik Hwang ),( Sung Woo Kim ),( Min Jae Choi ),( Sung Hwan Cho ),( Dong Jin Kim ) 한국부식방식학회 2021 Corrosion Science and Technology Vol.20 No.4

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor’s internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

Area Effect on Galvanic Corrosion of Condenser Materials with Titanium Tubes in Nuclear Power Plants

Hwang, Seong-Sik,Kim, Joung-Soo,Kim, Uh-Chul Korean Nuclear Society 1993 Nuclear Engineering and Technology Vol.25 No.4

Titanium tubes have recently been used in condensers of nuclear power plants since titanium has very good corrosion resistance to seawater. However, when it is connected to Cu alloys as tube sheet materials and these Cu alloys are connected to carbon steels as water box materials, it makes significant galvanic corrosion on connected materials. It is expected from electrochemical tests that the corrosion rate of carbon steel will increase when it is galvanically coupled with Ti or Cu in sea water and the corrosion rate of Cu will increase when it is coupled with Ti, if this couple is exposed to sea water for a long time. It is also expected that the surface area ratios, R$_1$(surface area of carbon steel/surface area of Ti) and R$_2$(surface area of carbon steel/surface area of Cu) are very important for the galvanic corrosion of carbon steel and that these should not be kept to low values in order to minimize the galvanic corrosion on the carbon steel of the water box. Immersed galvanic corrosion tests show that the corrosion rate of carbon steel is 4.4 mpy when the ratio of surface area of Fe/ surface area of Al Brass is 1 while it is 570 mpy when this ratio is 10$^{-2}$ . The galvanic corrosion rate of this carbon steel is increased from 4.4 mpy to 13 mpy at this area ratio, 1, when this connected galvanic specimen is galvanically coupled with a Ti tube. This can be rationalized by the combined effects of R$_1$ and R$_2$ on the polarization curve.

Role of Lead in Electrochemical Reaction of Alloy 600, Alloy 690, Ni, Cr, and Fe in Water

Hwang, Seong Sik,Kim, Joung Soo,Kim, Ju Yup 대한금속재료학회 2003 METALS AND MATERIALS International Vol.9 No.4

It has been reported that lead causes stress corrosion cracking (SCC) in the secondary side of steam generators (SG) in pressurized water reactors (PWR). The materials of SG tubings are alloy 600, alloy 690, or alloy 800, among which the main alloying elements are Ni, Cr, and Fe. The effect of lead on the electrochemical behaviors of alloy 600 and alloy 690 using an anodic polarization technique was evaluated. We also obtained polarization curves of pure Ni, Cr, and Fe in water containing lead. As the amount of lead in the solution increased, critical current densities and passive current densities of alloy 600 and alloy 690 increased, while the breakdown potential of the alloys decreased. Lead increased critical current denqity and the passive current of Cr in pH 4 and pH 10. The instability of passive film of steam generator tubings in water containing lead might arise from the instability of Cr passivity.

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

Hwang, Seong Sik,Kim, Hong Pyo,Kim, Joung Soo,Kasza, Kenneth E.,Park, Jangyul,Shack, William J. Elsevier 2005 Nuclear engineering and design Vol.235 No.23

<P><B>Abstract</B></P><P>A forced outage due to a steam generator tube leak in a Korean nuclear power plant has been reported <ce:cross-ref refid='bib3'>[Kim, J.S., Hwang, S.S., et al., 1999. KAERI Internal Report (Korean). Destructive analysis on pulled tubes from Ulchin unit 1. Korea Atomic Energy Research Institute]</ce:cross-ref>. Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Cracked specimens were prepared using a room temperature cracking technique, and the leak rates and burst pressures of the degraded tubes were determined both at room temperature and at a high temperature. Some tubes with 100% through wall cracks did not show a leakage at 10.8MPa, which is the typical pressure difference of the pressurized water reactors (PWRs) during a normal operation. In some tests, the leak rates of the tubes increased with time at a constant pressure. In a high temperature pressure test at 282°C one specimen showed a very small leakage at 18.6MPa, which stopped after a small increase in the test pressure. Because stress corrosion cracks can develop at relatively low stresses, even 100% through wall cracks can be so tight that they will not leak at a normal operating pressure.</P>

Study on Leak Rate of SCC Degraded Alloy 600 Tubings of PWRs

( Seong Sik Hwang ),( Joung Soo Kim ),( Ken E. Kasza ),( Jang Yul Park ) 한국부식방식학회(구 한국부식학회) 2004 Corrosion Science and Technology Vol.3 No.6