Jong Chull Jo

Pusan National University

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Myung Jo Jhung

Korea Institute of Nuclear Safety

Hho Jung Kim

Korea Institute of Nuclear Safety

F. J. Moody

Babcock & Wilcox (United States)

Woong Sik Kim

Korea Institute of Nuclear Safety

Jae Jun Jeong

Pusan National University

Dong Gu Kang

Korea Institute of Nuclear Safety

Seon Oh Yu

Korea Institute of Nuclear Safety

Byong Jo Yun

Pusan National University

Chang-Yong Choi

Jeonju University

Seok Ki Choi

Yonsei University

Cooperative Institutions

Korea Institute of Nuclear Safety

Korea Atomic Energy Research Institute

Korea Advanced Institute of Science and Technology

Pusan National University

Seoul National University

Hanyang University

Korea Institute of Energy Research

National University College

Sungkyunkwan University

Chung-Ang University

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Flow and Modal Analysis for the Investigation of PWR Safety Injection Line-Installed Thermal Sleeve Separation Failure Mechanism

Jong Chull Jo Myung Jo Jhung Hho Jung Kim

In conventional pressurized water reactors, a thermal sleeve (named simply ‘sleeve’ hereafter) is seated inside the nozzle part of each safety injection (SI) branch pipe to prevent and relieve potential excessive transient thermal stress in the nozzle wall when cold water is injected during the safety injection mode. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. This paper investigates the flow field in the pipe junction through a numerical simulation and vibration characteristics of thermal sleeves through a modal analysis to analyze the root cause of sleeve separation mechanism. By performing both flow simulation in the SI pipe junction and modal analysis of thermal sleeve, the fluid force and modal characteristics have been identified to be able to lead or contribute to separate thermal sleeves inside safety injection branch pipes in PWR plants from their original seating locations.

Pressurized water reactor

Transient (computer programming)

Safety valve

10.1115/pvp2005-71438

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

대한기계학회 춘추학술대회 (2004)

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

Flow boiling

Multiphase flow

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Numerical prediction of a flashing flow of saturated water at high pressure

Nuclear Engineering and Technology (2018)

Jong Chull Jo Jae Jun Jeong Byong Jo Yun F. J. Moody

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

10.1016/j.net.2018.06.002

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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

This study investigates the fluid-elastic instability characteristics of steam generator helical type tubes in operating nuclear power plants. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted by a general purpose computational fluid dynamics code employing the finite volume element modeling. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for helical type tubes with various conditions. Investigated are the effects of the helix angle, the number of supports and the status of the inner fluid on the modal, and fluid-elastic instability characteristics of the tubes, which are expressed in terms of the natural frequency, corresponding mode shape, and stability ratio. (authors)

Internal flow

Natural frequency

Elastic instability

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New features of the analytic function expansion nodal (AFEN) method and implementation in the hexagonal-Z KORDAX code

Transactions of the American Nuclear Society (2004)

Nam Zin Cho Jaejun Lee Do Sam Kim Chae Yong Yang Jong Chull Jo

Code (set theory)

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Dynamic characteristics of rectangular twisted beam with fins surrounded with fluid

Nuclear Engineering and Design (2006)

Myung Jo Jhung Jong Chull Jo

Natural frequency

Mode (computer interface)

10.1016/j.nucengdes.2006.03.021

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EQUIVALENT MATERIAL PROPERTIES OF PERFORATED PLATE WITH TRIANGULAR OR SQUARE PENETRATION PATTERN FOR DYNAMIC ANALYSIS

Nuclear Engineering and Technology (2006)

Myung Jo Jhung Jong Chull Jo

Square (algebra)

Penetration (warfare)

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Citations (28)

유체에 잠긴 다공 원통형 쉘의 자유진동해석

Jhung Myung Jo Jong Chull Jo

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Analytical Study of Two-Phase Flow Instability in a Natural Circulation PWR Auxiliary Feedwater System

Jong Chull Jo F. J. Moody Kyu Sik

A PWR incorporates a passive auxiliary feedwater system (PAFS), a closed natural circulation loop which is aligned to feed condensed water to its corresponding steam generator (SG). During its operation, saturated steam in the SG secondary side moves up due to buoyancy force and passes through a steam line, and then flows into a tube-tank type passive condensation heat exchanger where steam is condensed inside the tubes while the tube outer surfaces are cooled by the pool water. The condensate water is passively fed into the SG economizer by gravity. Because a natural circulation loop is susceptible to two-phase flow instability, it is requisite to confirm the system is designed adequately to avoid the potential challenges to its operational safety due to the instability. This paper presents an analytical approach for assessing if the PAFS has possible thermal and fluid mechanical characteristics which could lead to an undesirable unstable or oscillating condensate water level in the vertical pipe section. Both steady and unsteady analytical solutions for a simplified natural circulation loop model of the PAFS were derived in terms of the condensate water level and velocity in the vertical pipe section. From the solutions, the criteria for determining a potential for two-phase instability in the system were obtained.

Natural circulation

Boiler feedwater

Economizer

10.1115/pvp2013-97123

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Improved Features of the AFEN Method in Hexagonal-Z 3-D Geometry for Neutron Diffusion Calculation

Nam Zin Cho Jaejun Lee Do Sam Kim Chae Yong Yang Jong Chull Jo

10.11561/aesj.2005s.0.132.0

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Citations (1)