• Proceedings of the KWS Conference
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• 1996.05a
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• pp.34-37
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• 1996

The expanding method of tube to tubesheet joint part on neclear steam generators are classified into three classes of roller expanding, explosive expanding and hydraulic expanding. After the expanded Mock-Up specimen are made by the three expanding method. The general properties, microstructure/microvickers hardness, pull-out strength, hydraulic leak pressure, of tube to tubesheet joint part were inspected. and We evaluated the operation efficiency of expansion, reproduction of expanded joint about three expanding method. Through the overall evaluation of tube to tubesheet joint part, The hydraukic expanding and explosive expanding could be certificated more useful expanding method.

• Applied Chemistry for Engineering
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• v.9 no.2
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• pp.268-272
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• 1998

Water leak phenomena in the liquid sodium which is a coolant of liquid metal reactor, were investigated by carrying out sodium-water reaction experiment. It was confirmed that sodium and water react each other by the analysis of material composition of aspecimen at the end of experiment. When steam of $100kg/cm^2$ was passed through the leak path of the specimen for 4 hours, reaction products from sodium-water reaction were observed on the leak site. However, re-opening phenomena were not observed at this condition. It was interpretted that the reaction product precipitated on leak path and thermal transient caused self-plugging and re-openning phenomena, respectively.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.10
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• pp.1075-1081
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• 2011

A leak-detection method has been developed by measuring the pressure variation between the inner and outer heattransfer tubes of a double-wall tube steam generator. An experiment was carried out to measure the leak rate in the gap between two surfaces pressed with a hydraulic press in order to simulate the phenomena, and a correlation was determined for the leak rate in a micro gap. However, in the correlation, the gap width and friction coefficient were coupled with the surface roughness, which affects the two parameters. The two parameters were separated using a surface-contact model to develop a correlation for the friction coefficient. The correlation was compared with the existing correlations used for crack analysis. Although the applied ranges of Reynolds numbers were different, the developed correlation for Reynolds numbers of 0.1.0.35 showed similar tendencies to existing correlations used for higher Reynolds numbers.

• Journal of Mechanical Science and Technology
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• v.19 no.5
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.435-443
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• 1998

Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr$_2$O$_4$-TiO$_2$, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.77-87
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• 2000

It is reported that the toughness and safety margins of the SA106 Gr.C main steam line piping steel is reduced due to dynamic strain aging (DSA) at the reactor operating temperature for Leak-Before-Break (LBB) application. In this study, intercritical annealing in two-phase ($\alpha$+${\gamma}$）region was performed to investigate the possibility of improving the toughness and reducing DSA susceptibility. The manifestations of DSA were still observed in the tensile tests of the annealed specimens. However, the ductility loss caused by DSA was smaller than that in the as-received material. Furthermore, the intercritical annealing was able to increase the Charpy impact toughness by 1.5 times compared to as-received. With the heat treatment, we could obtain microstructural changes such as the cleaner retained ferrite, increased ferrite content and somewhat finer grain size. It is considered that the reduced DSA was induced by cleaner retained ferrite, which in turn resulted in higher impact toughness in addition to the general toughening due to finer grain sizes and increased ferrite content.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.299-305
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• 1991

Corrosive environment may be generated within steam generators from condenser cooling water in-leakage. Theoretical analysis of the accumulation of chloride as a sea water impurity is being carried out for the condenser cooling water used at YGN-1,2 nuclear power stations. Calculations have shown that highly concentrated chloride solution would be produced within the steam generators in the case of sea water in-leakage. Maximum allowable design condenser leak rate(0.5 gpm) leads chloride concentration of 2.3 ppm at steam generetor and 0.6 ppm at hotwell with the maximum blowdown rate and condensate purification. Concentration factor at steam generator is dependent only on both blowdoum rate and condensate purification efficiency as follows, Concentration Factor（equation omitted）(B$\neq$O) Blowdown and condensate purification are evaluated as the only effective measures to remove impurities from the secondary systems.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.10
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• pp.1097-1103
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• 2011

Liquid droplet impingement erosion (LDIE) is caused by the impact of high-velocity droplets entrained in steam or air on metal. The degradation caused by the LDIE has been experienced in steam turbine internals and high-velocity airplane components (particularly canopies). Recently, LDIE has also been observed in the pipelines of nuclear plants. LDIE among the pipelines occurs when two-phase steam experiences a high pressure drop (e.g., across an orifice in a line to the condenser). In 2011, a nuclear power plant in Korea experienced a steam leak caused by LDIE in a pipe through which a two-phase fluid was flowing. This paper describes a study on the design change of a pipe affected by LDIE in order to mitigate the damage. The design change has been reviewed in terms of fluid dynamics by using the FLUENT code.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.181-187
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• 2016

A thermal power plant has a heat exchanger tube to collect and convert the heat generated from the high temperature and pressure steam to energy, but the tubes are arranged in a complex manner. In the event that a leakage occurs in any of these tubes, the high-pressure steam leaks out and may cause the neighboring tubes to rupture. This leakage can finally stop power generation, and hence there is a dire need to establish a suitable technology capable of detecting tube leaks at an early stage even before it occurs. As shown in this paper, by applying acoustic emission (AE) technology in existing boiler tube leak detection equipment (BTLD), we developed a system that detects these leakages early enough and generates an alarm at an early stage to necessitate action; the developed system works better that the existing system used to detect fine leakages. We verified the usability of the system in a 560MW-class thermal power plant boiler by conducting leak tests by simulating leakages from a variety of hole sizes (ⵁ2, ⵁ5, ⵁ10 mm). Results show that while the existing fine leakage detection system does not detect fine leakages of ⵁ2 mm and ⵁ5 mm, the newly developed system could detect leakages early enough and generate an alarm at an early stage, and it is possible to increase the signal to more than 18 dB.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.2
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• pp.158-164
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• 2004

Steam generator tubes experience widespread degradations such as stress corrosion cracking, wear, tube rupture, denting, fatigue and so on. The resulting damages can cause tube bursting or leak of the primary water which contains radioactivity Therefore the allowable size of the damage is required to be determined on the maintenance purpose. The burst pressure of a tube with a T-type combination crack consisting of longitudinal and circumferential cracks is obtained experimentally and analytically. Fracture parameters such as stress intensity factor and crack opening angle are investigated. Also the burst pressure for a T-type combination crack is compared with that of a single longitudinal crack to develop a length-based criteria.