• Proceedings of the KIEE Conference
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• 1999.11b
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• pp.334-336
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• 1999

Piping systems of the power generating stations have been generally protected by hydraulic and mechanical snubbers which can allow large displacements arising from temperature change while those can reduce or absorb stresses due to vibrations. However these snubbers require amounts of budget for maintenance or replacement because of the leakage, lubrication and finally short life cycle. Recently the snubbers consisted of wire rope have been proved to reduce vibrations of piping systems. While the wire rope snubbers are free of maintenance such as leakage and lubrication imported price are so high. Now it is necessary to design. manufacture and certificate these wire rope snubbers.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.29-33
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• 2008

As the measures of nuclear power plant utilities and manufacturers to reduce the defects of tube expansion region during manufacturing steam generators, many types of NDEs(Non-Destructive Examinations) are conducted to inspect the expansion region. The expansion region of tube is subject to degrade because of stress concentration induced by tube expansion, sludge pile and high temperature. So the inspections for tube expansion region have been reinforced. Liquid penetrant test, helium leak test, Bobbin profile test and hydraulic test are performed to confirm the integrity of tube expanded by hydraulic expansion method. Liquid penetrant test and helium leak test are used to inspect seal weld region on tubesheet end part. Bobbin Profile test is used to inspect fully the expanded region of steam generator tube. Hydraulic test finally verifies the integrity of seal weld region on tubesheet end part.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.47-53
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• 2010

In this research project, development of methodology to measure the thickness of pipes in the wide range using magnetic field. The magnetic field spreading in the sensor and the plate was modeled in the cases of the various thicknesses in plate. Based on the analysis, sensors were designed, manufactured and tested to optimize the specifications of the sensor. The sensor can be used in high temperature through calibration. And the uncertainty of the sensor was estimated.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2631-2636
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• 2008

The most components and piping of the secondary side of domestic nuclear power plants were manufactured carbon-steel and low-alloy steel. Flow accelerated corrosion leads to wall thinning (metal loss) of carbon steel components and piping exposed to the flowing water or wet steam of high temperature, pressure, and velocity. The feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which increases as operating time progress. Several nuclear power plants in Korea have also experienced wall thinning damage in the shell wall around the impingement baffle. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the experimental results based on down-scaled experimental facility. The experiments were performed based on several types of impingement baffle plates which are installed in low pressure feedwater heater.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Tribology and Lubricants
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• v.38 no.4
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• pp.121-127
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• 2022

This study presents the development of an integrated oil purification system consisting of moisture removal, oil flushing, and oil filtering devices. In this system, the oil flushing device is combined with a micro-bubble generator. Oil purification is necessary for ensuring the high performance of the lubricant through the efficient removal of contaminants and thus enables good maintenance of mechanical systems. The developed purification system removes moisture, varnish, and solid particles. Moreover, during oil purification, the oil flushing device separates foreign materials and contaminants remaining in the lubricating oil piping or mechanical systems. The microbubble generator, which is combined with the oil flushing device, can separate harmful contaminants, such as sludge, wear particles, and rust, from piping or lubrication systems through the cavitation effect. Moisture is removed using a double high-vacuum chamber, while sludge and varnish are removed via electro-absorption using a high-voltage generator. Additionally, the total maintenance cost of the system is reduced through the use of domestically fabricated cartridge filters composed of glass fiber and cellulose. The heater, which maintains the temperature of the lubricant at 60℃, can process 41,000 L of lubricant simultaneously. Multiple tests confirmed that the proposed integrated purification system exhibits good performance in oil flushing and removal of water and varnish.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.6-12
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• 2015

In this study, the performance test was carried out under various operation conditions targeting four representing types of lubricant which are mostly used in nuclear power plants, and mutually compared the characteristic and performance of lubricant. Especially, introducing the concept of the thread friction coefficient (hereinafter, TFC), which makes the friction relation between the stem nut and stem screw dimensionless. The test was performed to compare the lubricant performance for the four kinds of lubricant (Texaco, Alvania, Mobilux, MOVLL). In a test of the room temperature stem, the TFC of MOV Long Life shows the lowest value, next to Alvania EP2, Texaco EP2, and Mobilux EP0 in that order. And in a test of the high temperature stem, the TFC of Texaco EP2 shows the lowest, next to MOV Long Life, Alvania EP2, and Mobilux EP0 in that order. From the test result of the aging condition, three types of lubricant (MOV LL, Texaco EP2, Alvania EP2) show similar patterns up to 36 months, but in 60 months, the TFC of lubricant are increased rapidly.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• 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.