• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.724-733
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• 2003

Fatigue design rules for welds in the ASME Boiler and Pressure Vessels Code are based on the use of Fatigue Strength Reduction Factors(FSRF) against a code specified fatigue design curve generated from smooth base metal specimens without the presence of welds. Similarly, stress intensification factors that are used in the ASME B3l.1 Piping Code are based on component S-N curves with a reference fatigue strength based on straight pipe girth welds. But the determination of either the FSRF or stress intensification factor requires extensive fatigue testing to take into account the stress concentration effects associated with various types of component geometry, weld configuration and loading conditions. As the fatigue behavior of welded joints is being better understood, it has been generally accepted that the difference in fatigue lives from one type of weld to another is dominated by the difference in stress concentration. However, general finite element procedures are currently not available for effective determination of such stress concentration effects. In this paper, a mesh-insensitive structural stress method is used to re-evaluate the S-N test data, and then more effective method is proposed for pressure vessel and piping fatigue design.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.107-114
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• 2016

Four cylindrically shaped IHXs(Intermediate Heat Exchangers) are installed in the PHTS(Primary Heat Transfer System) of the PGSFR(Prototype Gen IV Sodium cooled Fast Reactor). As for the IHX, the temperature difference of structure is inevitable result caused by heat transfer between primary coolant sodium and IHTS(Intermediate Heat Transport System) sodium. It is necessary to evaluate the high temperature structural integrity of IHXs which operate at the elevated temperature condition over the creep temperature. In this paper, the high temperature structural integrity of IHX under assumed loading conditions has been reviewed according to ASME code.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.35-39
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• 2009

In this study, Event Type Analysis System (ETAS) is developed for the accident evaluation in nuclear power plant. The ETAS system can be used in supporting regulator and/or operator under event situation in nuclear power plants. The ETAS system can categorize the all transient events to 3 categories such as Down-2000, Down-2173, and Slow Fluctuation. We develop the program structure for ETAS system and web-based ETAS system. The ETAS system will be used as sub module of Knowledge-Based Event Evaluation Network (K-EvENT) which is developing for the against the accident in nuclear power plants.

• KIEAE Journal
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• v.2 no.2
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• pp.33-38
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• 2002

There are a number of practical uses for devices which are capable of piping large amounts of light: illumination of areas where there would be maintenance, safety, or security problems with electronic light sources; piping sunlight into indoor areas for illumination; and the conversion of high luminous efficacy, good color quality, high intensity discharge lamps into more acceptable linear of area sources of light. Prism light guides are hollow structures which pipe light by means of total internal reflectin(TIR). These devices are unique in their ability to combine the efficiency of TIR with the relatively low cost of hollow structure. An important application stems from their ability to transform a point source of light such as an incandescent or discharge lamp into a linear or area source of light or illumination. We report the development of an economical, flexible prismatic film for fabricating the light guide wall. This guide geometries and films are currently under development.

• Transactions of the Korean Society of Machine Tool Engineers
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• v.17 no.1
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• pp.150-157
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• 2008

Bellows are commonly used in piping systems to absorb expansion and contraction in order to reduce stress. Unlike most piping components, bellows consist of a thin-walled shell of revolution with a corrugated meridian, in order to provide the flexibility needed to absorb mechanical movements. It is a composite shell structure consisting of at least one toroidal shell, an annular plate or conical shell. It is difficult to analyze the behavior of bellows because of its complex geometry. Simplified formulas for variable mechanical behaviors of bellows are provided by a standard called EJMA. An automatic design software for bellows is programming by using VBA(Visual Basic for Application) based on EJMA. Bellows engineers can effectively make a decision for bellows geometries because this software provides graphically design results in its post-processor. Bellows design software is expected to give quite a good guidance to practical design. The characteristics of bellows are also investigated through the automatic design process in bellows design software.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.1-8
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• 2010

Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.37 no.5
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• pp.787-799
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• 2017

Although the long-term leakage between the box culvert and the soil contact surface is one of the main causes of the failure in the embankment of the reservoir, there is a little studies on this matter. If a leakage occurs by the structure such as the buried box culvert of reservoir embankment is partially damaged, it is difficult to observe and there is a possibility of damage caused by piping. For these reasons, more research is necessary. In this study, the embankment type of the reservoir is divided into the core type and the homogeneous type when the damaged box culvert passing through the embankment of the reservoir is leaked due to the differential behavior of materials like differential settlement. In view of the condition, the seepage analysis of 2D was performed according to the water level change. The result of the study shows that the possibility of piping increases at the upper part rather than the bottom part of the box culvert when the leakage occurs to the box culvert passing through the embankment of the reservoir. Particularly, it is considered that the presence of the core helps to maintain the seepage stability of the embankment in case where the leakage occurs at the downstream side of the embankment. Also, if there is a drastic decrease on the internal pore water pressure in the embankment of reservoir, it is necessary to consider the possibility of piping.

• Journal of the Earthquake Engineering Society of Korea
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• v.15 no.3
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• pp.37-43
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• 2011

Reliability of piping systems is essential to the safety of any important industrial facilities. During an earthquake, damage to the piping system can occur. It can also cause considerable economic losses and the loss of life following earthquakes. Traditionally, the study of the secondary system was less important than primary structure system, however it has recently been emerging as a key issue for the effective maintenance of the structural system and to help reduce nonstructural earthquake damage. The primary objectives of this study are to evaluate seismic design requirements and the seismic performance of gas and fire protection piping systems installed in reinforced concrete (RC) buildings. In order to characterize the seismic behavior of the existing piping system in an official building, 10 simulated earthquakes and 9 recorded real earthquakes were applied to ground level and the building system by the newmark average acceleration time history method. The results developed by this research can be used for the improvement of new seismic code/regulatory guidelines of secondary systems as well as the improvement of seismic retrofitting or the strengthening of the current piping system.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.