• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.79-80
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• 2005

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Fashion & Textile Research Journal
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• v.11 no.3
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• pp.474-480
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• 2009

The purpose of this study is to prepare the hardness of polyester(PET) fabric by thermal bonding with low melting component of bicomponent fiber and to describe the change of physical properties of thermal bonded PET fabrics. The PET fabrics were prepared with regular PET fiber as warp and bicomponent fiber as weft. The bicomponent fiber of sheath-core type were composed with a regular PET core and low melting PET sheath. The thermal bonding of PET fabric was carried out in pin tenter from 120 to $195^{\circ}C$ temperature range for 60 seconds. In this study, we investigated the physical properties and melting behavior of PET fiber and the effect of the temperature of the pin tenter on the thermal bonding, mechanical properties. Melting peak of warp showed the thermal behavior of general PET fiber. However, melting peak of weft fiber(bicomponent fiber) showed the double melting peak. The thermal bonding of the PET fabric formed at about temperature of lower melting peak. The optimum thermal bonding conditions for PET fabrics was applied at $190{\sim}195^{\circ}C$ for 60seconds by pin tenter. On the other hand, the tensile strength of the PET fabric decreased with an increasing temperature of thermal bonding.

• Proceedings of the International Microelectronics And Packaging Society Conference
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• 2001.09a
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• pp.147-164
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• 2001

1. "MLS", state of the art MCM-D wiring substrate. 2. High pin-count LSl assembly. 3. Higher speed needs higher packaging density. 4. Wiring substrate, the key of LSl packaging device. 5. "Inter-Layer Transferability", a new index for the performance of wiring substrates. 6. "MLTF package", a core-less flexible package for high pin-count LSl.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Journal of the Korean Society for Nondestructive Testing
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• v.16 no.1
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• pp.46-51
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• 1996

원자로는 압력용기 및 내부 구조물로 구분되어 있다. 내부 구조물들의 경련 열화 현상에 따라 결함이 많이 발생하여 이에 대한 초음파검사가 요구되고 있다. 따라서 원자로 내부 구조물에 대한 초음파검사 현황 및 각각 구조물들의 검사 원리를 기술하였다. 특히 원자로 내부 구조물 중 CRDM, core baffle bolt, core barrel bolt, CRGT-support pin 및 fuel alignment pin에 대한 유럽 및 독일을 중심으로 한 검사 현황 및 검사방법을 간략하게 기술하였다. 이 기술에 대한 지침안(guideline)이 독일, 프랑스, 일본을 중심으로 하여 마련되고 있다.

• Transactions of Materials Processing
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• v.15 no.9 s.90
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• pp.648-653
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• 2006

In this study, the characteristics of Wire Electrical Discharge Machining(WEDM) of the high-hardened mold steel were investigated. WEDM experiments have been carried out based on parameter of wire diameter, pulse on time, pulse off time, feed rate and cycle etc. From the results, the optimized WEDM cycle of RIGOR steel has been revealed as $5{\sim}7$ times. Also, geometrical accuracy of the Core Pin is dependent on WEDM wire radius machining condition and wire chattering.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.500-519
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• 2010

AEGIS is a lattice physics code incorporating the latest advances in lattice physics computation, innovative calculation models and efficient numerical algorithms and is mainly used for light water reactor analyses. Though the primary objective of the AEGIS code is the preparation of a cross section set for SCOPE2 that is a three-dimensional pin-by-pin core analysis code, the AEGIS code can handle not only a fuel assembly but also multi-assemblies and a whole core geometry in two-dimensional geometry. The present paper summarizes the major calculation models and part of the verification/validation efforts related to the AEGIS code.

• Transactions of Materials Processing
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• v.14 no.6 s.78
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• pp.570-576
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• 2005

Powder Injection Molding(PIM) has recently been recognized as an advanced manufacturing technology for low-cost mass production of metal or ceramic parts of complicated geometry With this regards, design technology of dental scaler tip PIM mold, which has complex shape and small core pin (diameter=0.6mm), with the help of computer-aided analysis of powder injection molding process was developed. Computer-aided analysis for dental scaler tip mold was implemented by finite element method with non-Newtonian fluid, modified Cross model viscosity, PvT data of powder/binder mixture. Compter-aided analysis results, such as filling pattern, weldline formation, air vent position prediction were compared with experimental result, and eventually have been shown good agreement. The core pin (diameter=0.6mm) deflection analysis of dental scaler tip PIM mold during PIM filling process was also investigated before mold fabrication.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1024-1036
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• 2018

In this paper, McCARD code was verified using various models listed in the NUREG/CR-6361 benchmark guide, which provides specifications for single pin-cells, single assemblies, and the whole core classified depending on the nuclear properties and structural characteristics. McCARD code was verified by comparing its results with those of SCALE code for single pin-cell and single assembly benchmark problems. The difference in the multiplication factor obtained through the two codes did not exceed 90 pcm. The benchmark guide treats a total of 173 whole core experiments. The experiments are categorized as simple lattices, separator plates, reflecting walls, reflecting walls and separator plates, burnable absorber fuel rods, water holes, poison rods, and borated moderator. As a result of numerical simulation using McCARD, the mean value of the multiplication factors is 1.00223 and the standard deviation of the multiplication factors is 285 pcm. The difference between the multiplication factors and the experimental value is in the range of -665 pcm to + 1609 pcm. In addition, statistics of results for experiments categorized by reactor shape, additional structure, burnable poison, etc., are detailed in the main text.