• Particle and aerosol research
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• v.5 no.3
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• pp.95-109
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• 2009

A low temperature plasma process has been widely used for semiconductor fabrication and can also be applied for the preparation of solar cell, MEMS or NEMS, but they are notorious in the point of particle contamination. The nano-sized particles can be generated in the low temperature plasma process and they can induce several serious defects on the performance and quality of microelectronic devices and also on the cost of final products. For the preparation of high quality thin films of high efficiency by the low temperature plasma process, it is desirable to increase the deposition rate of thin films with reducing the particle contamination in the plasmas. In this paper, we introduced the studies on the generation and growth of nanoparticles in the low temperature plasmas and tried to introduce the recent interesting studies on nanoparticle generation in the plasma reactors.

• Journal of Advanced Marine Engineering and Technology
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• v.11 no.1
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• pp.72-79
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• 1987

When the wall temperature is very high, a stable vapor film covers the heat transfer surface. The vapor film creates a strong thermal resistance when heat is transferred to the liquid though it. This phenomenon, called "film boiling" is very important in the heat treatment of metals, the design of cryogenic heat exchangers, and the emergency cooling of nuclear reactors. In the practical engineering problems of the transient cooling process of a high temperature wall, the wall temperature history, the variation of the heat transfer coefficients, and the wall superheat at the rewetting points, are the main areas of concern. These three areas are influenced in a complex fashion such factors as the initial wall temperature, the physical properties of both the wall and the coolant, the fluid temperature, and the flow state. Therefore many kinds of specialized experiments are necessary in the creation of precise thermal design. The object of this study is to investigate the heat transfer characteristics in the transient cooling process of a high temperature wall. The slow transient cooling experiment was carried out with a copper block of high thermal capacity. The block was 240 mm high and 79 mm O.D.. The coolant flowed throuogh the center of a 10 mm diameter channel in the copper block. In the copper block, three sheathed thermocouples were placed in a line perpendicular to the flow. These thermocouples were used to take measurements of the temperature histories of the copper block.

• Korean Journal of Materials Research
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• v.17 no.5
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• pp.268-272
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• 2007

Inconel 617 is a candidate tube material for high temperature gas-cooled reactors(HTGR). The microstructure and mechanical properties of Inconel 617 were studied after exposure at high temperature($1050^{\circ}C$). The dominant oxide layer was Cr-oxide. The internal oxide and Cr-depleted region were observed below the Cr-oxide layer. The depth of Cr-depleted zone and internal oxide increased with exposure time. The major phases of carbides are $M_{23}C_6\;and\;M_6C$. The composition of $M_{23}C_6\;and\;M_6C$ were determined to be Cr-rich and Mo-rich, respectively. $M_6C$ carbide is more stable than $M_{23}C_6$ at high temperature. From the results of high temperature compression test, there were no significant changes in hardness and yield strength upon increasing exposure time.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.379-380
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• 2002

Steam generators (S/G) of pressurized water reactors are large heat exchangers that use the heat from the primary reactor coolant to make steam in the secondary side for driving turbine generators. Reciprocating sliding wear experiments have been performed to examine the wear properties of Incoloy 800 and Inconel 690 steam generator tubes in high temperature water. In present study, the test rig was designed to examine the reciprocating and rolling wear properties in high temperature (room temperature - $300^{\circ}C$) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of steam generator tube materials. To investigate the wear mechanism of material, the worn surfaces were observed using scanning electron microscopy. At $290^{\circ}C$, wear rate of Inconel 690 was higher than that of Incoloy 800. It was assumed to be resulted from the oxide layer property difference due to the a\loy composition difference. Between 25 and $150^{\circ}C$ the wear loss increased with increasing temperature. Beyond $150^{\circ}C$, the wear loss decreased with increasing temperature. The wear loss change with temperature were due to the formation of wear protective oxide layer. From the worn surface observation, texture patterns and wear particle layers were found. As test temperature increased, the proportion of particle layer increased.

• Journal of Mechanical Science and Technology
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• v.20 no.12
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• pp.2061-2078
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• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Corrosion Science and Technology
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• v.22 no.4
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• pp.265-272
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• 2023

The aim of this study was to develop a new in-situ observation method and instrument in micro-scale to investigate the mechanism of stress corrosion cracking (SCC) initiation of Ni-base alloys in a high temperature water environment of pressurized water reactors (PWRs). A laser confocal microscope (LCM), an autoclave with diamond window view port, and a slow strain-rate tester with primary water circulation loop system were components of the instrument. Diamond window, one of the core components of the instrument, was selected based on its optical, chemical, and mechanical properties. LCM was used to observe the specimen in micro-scale, considering the experimental condition of a high-temperature primary water environment. Using in-situ method and instrument, it is possible to observe oxidation and deformation of specimen surface in micro-scale through the diamond window in a high-temperature primary water in real-time. The in-situ method and instrument developed in this work can be utilized to investigate effects of various factors on SCC initiation in a high-temperature water environment.

• Journal of environmental and Sanitary engineering
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• v.4 no.1 s.6
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• pp.43-52
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• 1989

A synthetic wastewater composed of powdered baby food and inorganic nutrients was treated by five 3.6L anaerobic reactors in order to test the relationship between solids retention time (SRT) and media surface ratio, and the removal efficiencies of organic substrate. Of the five reactors, four were semi-continuously fed stirred-tank reactors and one continuously-stirred batch reactor. The influent was 7430mg/L in COD, 7120 mg/L $BOD_L$ and 6350mg/L in $BOD_5$, respectively. Operating temperature was $35{\pm}1^{\circ}C$ and pH in the range 6.9 to 7.2. In this experimental study it was found that a linear relationship existed, within the experiment range, between SRT and media surface ratio, and that SRT and removal efficiency increased with increasing media surface ratio. The substrate removal efficiencies were 82.7 to $88.2\%$ in COD, 82.9 to $88.4\%$ in $BOD_L$ and 83.3 to $88.7\%$ in $BOD_5$, respectively.

• Microbiology and Biotechnology Letters
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• v.23 no.4
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• pp.492-498
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• 1995

Net type polyvinylidene chloride (PVDC) media and cillium type polyethlene polypropylene (PEPP) media were installed in the aereted submerged biofilm reactors. Synthetic phenol wastewater for feed were made to contain 1,480 mg of phenol per liter of water. The organic loading range of reactors were 0.439-0.456 kg COD/m$_{3}$, 0.882 - 0.919 kg COD/m$_{3}$ and 1.199-1.339 kg COD/m$_{3}$. Comparing PVDC to PEPP media, the bacterial number found in biofilm on PEPP were slightly higher. With the low temperature (10$\circ$C), the number of bacteria was some what deceered. Number of bacterial strains identified from PVDC were 23 and those from PEPP were 42. Genera identified in the PVDC media were Flavobacterium (47.8%), Unidentified (17.6%), Pseudomonas (13.0%), Micrococcus (8.7%) and Beggratoa (8.7%). Genera identified in the PEPP media reactor were Pseudomonas (35.7%), Alcaligenes (19.0%), Aeromonas (14.33%) and Micrococcus (11.9%), In the steady, state, a filamentous bacteria, type 1701 was identified in all of the reactors. Paramecium sp. and fungi were present in the PVDC media reactor. While, Difflugia sp, Paramecium sp. and fungi were found in the PEPP media reactor. The low diversity of protozoa was ascribed to high concentration of phenol.

• Corrosion Science and Technology
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• v.7 no.2
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• pp.77-84
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• 2008

Based upon the good compatibility to neutron irradiation and high temperature water environment, austenitic stainless steels are widely used for core internal structural materials of light water reactors. But, recently, intergranular cracking was detected in the stainless steels for the core applications in some commercial PWR plants. Authors studied on the root cause of the intergranular cracking and developed the countermeasure including the alternative materials for these core applications. The intergranular cracking in these core applications are defined as an irradiation assisted mechanical cracking and irradiation assisted stress corrosion cracking. In this paper, the root cause of the intergranular cracking and its countermeasure are summarized and discussed.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2859-2874
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• 2024

This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as thermal boundary conditions were used, and ABAQUS thermal and structural analyses were performed. After full ablation, the temperature of the inner surface in the thinnest section remained high (920 ℃), but the stress remained relatively low (less than 6 MPa). At the outer surface, the stress was as high as 250 MPa; however, the resulting plastic strain was small owing to the low temperature of 200 ℃. Variations in stress, inelastic strain, and temperature with time in the thinnest section suggest that the plastic and creep strains are saturated owing to stress relaxation, resulting in low cumulative damage. Thus, the lower head of the vessel can maintain its structural integrity under LBLOCA with IVR-ERVC conditions. The sensitivity analysis of internal pressure indicates the occurrence of failure in the thinnest section at an internal pressure >9.6 MPa via local necking followed by failure due to high stresses.