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Featured researches published by Woong Sik Kim.


Journal of Pressure Vessel Technology-transactions of The Asme | 2009

Numerical Simulation of Subcooled Flow Boiling Heat Transfer in Helical Tubes

Jong Chull Jo; Woong Sik Kim; Chang-Yong Choi; Yong Kab Lee

This paper addresses the numerical simulation of two-phase flow heat transfer in the helically coiled tubes of an integral type pressurized water reactor steam generator under normal operation using a computational fluid dynamics code. The shell-side flow field where a single-phase fluid flows in the downward direction is also calculated in conjunction with the tube-side two-phase flow characteristics. For the calculation of tube-side two-phase flow, the inhomogeneous two-fluid model is used. Both the Rensselaer Polytechnic Institute wall boiling model and the bulk boiling model are implemented for the numerical simulations of boiling-induced two-phase flow in a vertical straight pipe and channel, and the computed results are compared with the available measured data. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. Both the internal and external turbulent flows are simulated using the standard k-e model. From the results of the present numerical simulation, it is shown that the bulk boiling model can be applied to the simulation of two-phase flow in the helically coiled steam generator tubes. In addition, the present simulation method is considered to be physically plausible in the light of discussions on the computed results.


Transactions of The Korean Society for Noise and Vibration Engineering | 2004

Fretting-wear Characteristics of Steam Generator Helical Tubes

Myung Jo Jhung; Jong Chull Jo; Woong Sik Kim; Hho Jung Kim; Tae Hyung Kim

This study investigates the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the helical type tubes with various conditions. The wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting-wear characteristics of the tube.


12th International Conference on Nuclear Engineering, Volume 2 | 2004

A Direction to Adopting Risk-Informed and Performance-Based Regulation (RIPBR) in Korea

Woong Sik Kim; Key Yong Sung; Chang Ju Lee; Young Hwan Choi; Yong Seog Choi; Hho Jung Kim

The USNRC has pursued the incorporation of new regulatory approach of risk-informed and performance-based regulation (RIPBR) into nuclear safety regulation, as an alternative to improve existing nuclear safety regulation of nuclear power plants, which is deterministic and prescriptive. It focuses on the use of risk insight from probabilistic safety assessment (PSA). Recently, it becomes necessary to find a way to improve regulatory efficiency and effectiveness in order to cover the increasing regulatory needs in Korea. Also, the utility has optimized design and operation of the plant using PSA insight and equipment performance information. According to the increase of the necessity for regulatory improvement using risk and performance information, KINS (Korea Institute of Nuclear Safety) is developing, as a part of mid- and long-term project of Nuclear R&D program, how to adopt the RIPBR in Korean nuclear regulation. This paper presents the interim result of the study that comprises the assessment of feasibility to adopt the RIPBR, and basic directions and principles for implementing RIPBR model. It is concluded that adopting RIPBR is essential for risk management of the plant, public acceptance to nuclear safety, improvement of existing regulation, minimization of unnecessary regulatory burden, and effective use of regulatory resources, etc. Three basic directions and several principles that are necessary to implement RIPBR model are identified from the study. The application of the directions and principles to the assessment of RIPBR model to be established in the near future is expected to result in making the adoption of new regulatory system more objective and consistent.Copyright


Nuclear Engineering and Design | 2005

Dynamic characteristics of steam generator U-tubes with defect

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Hho Jung Kim


Nuclear Engineering and Technology | 2004

Fluidelastic Instability Characteristics of Helical Steam Generator Tubes

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Young Hwan Choi; Hho Jung Kim


Nuclear Engineering and Technology | 2003

Fretting-Wear Characteristics of Steam Generator Tubes by Foreign Object

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Young Hwan Choi; Hho Jung Kim


대한기계학회 춘추학술대회 | 2004

Numerical Analysis of Boiling-Induced Multiphase Flow in a Helical Steam Generator Coiled Tube

Jong Chull Jo; Hho Jung Kim; Woong Sik Kim; Seon Oh Yu; Yong Kab Lee


Problems Involving Thermal Hydraulics, Liquid Sloshing, and Extreme Loads on Structures | 2004

Prediction of Foreign Object-Caused Fretting-Wear on Helical Tubes in the Single-Phase External Flow Field

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Hho Jung Kim


Archive | 2004

Identification and Resolution of Safety Issues for the Advanced Integral Type PWR

Woong Sik Kim; Jong Chull Jo; Young Gill Yune; Hho Jung Kim


Archive | 2004

Fluid-elastic Instability of Helical Tubes Subjected to Single-Phase External Flow and Two-Phase Internal Flow

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Hho Jung Kim

Collaboration


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Hho Jung Kim

Korea Institute of Nuclear Safety

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Jong Chull Jo

Korea Institute of Nuclear Safety

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Myung Jo Jhung

Korea Institute of Nuclear Safety

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Young Hwan Choi

Korea Institute of Nuclear Safety

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Chang Ju Lee

Korea Institute of Nuclear Safety

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Key Yong Sung

Korea Institute of Nuclear Safety

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Young Gill Yune

Korea Institute of Nuclear Safety

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Seon Oh Yu

Korea Institute of Nuclear Safety

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Yong Seog Choi

Korea Institute of Nuclear Safety

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