• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.1-10
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• 2016

The number of nuclear power plants operating in Korea was 24 as of year 2015. Nine units out of 24 units have been operated for a period over 20 years. Kori unit 1 has been in operation for 40 years, and an extended operation for Wolsong unit 1 was decided in 2015. There has been reported some crackings in reactor internals in PWR have been reported in Europe, USA, Japan and Korea, and some of them were replaced with new one. Repair and replacement technologies for the reactor internals have been developing in order to meet the regulatory requirements for long term operation in Korea. The technologies will also be used for the exported nuclear units. It is required to review degradation history of the reactor internals worldwide as a part of the degradation management program development. Schematics of reactor internals designed and supplied by Westinghouse, Framatome and Combustion Engineering are described herein. Materials degradation history of reactor internals of PWR plants in USA, Japan and Europe is surveyed and summarized. Some events from Korean plants are also described. Aging management strategy for the internals is suggested.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Steel and Composite Structures
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• v.21 no.4
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• pp.703-753
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• 2016

The ratcheting behavior was studied experimentally for Z2CND18.12N elbow piping under cyclic bending and steady internal pressure. Dozens of cyclic plasticity models for structural ratcheting responses simulations were used in the paper. The four models, namely, Bilinear (BKH), Multilinear (MKIN/KINH), Chaboche (CH3), were already available in the ANSYS finite element package. Advanced cyclic plasticity models, such as, modified Chaboche (CH4), Ohno-Wang, modified Ohno-Wang, Abdel Karim-Ohno and modified Abdel Karim-Ohno, were implemented into ANSYS for simulating the experimental responses. Results from the experimental and simulation studies were presented in order to demonstrate the state of structural ratcheting response simulation by these models. None of the models evaluated perform satisfactorily in simulating circumferential strain ratcheting response. Further, improvement in cyclic plasticity modeling and incorporation of material and structural features, like time-dependent, temperature-dependent, non-proportional, dynamic strain aging, residual stresses and anisotropy of materials in the analysis would be essential for advancement of low-cycle fatigue simulations of structures.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.50-57
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• 2019

This paper investigates deformation characteristics of gas pipeline using the in-plane bending experiment and finite element analysis of a pipe bend. The effect of the bending angle and internal pressure on the deformation characteristics is analyzed. The pipe bend used in this study is API 5L X65 (out diameter: 20 inch) material with the thickness of 11.9 mm. The maximum load, displacement at maximum load, angle and local strain of 90° pipe bend are obtained from the in-plane bending experiment. Comparison between FE results and experimental data shows overall good agreements. In addition, the deformation characteristics of 22.5° and 45° pipe bend are calculated using the finite element analysis. As a result, the effect of the bend angle on the deformation characteristics is discussed.

• Journal of Korean Society of Coastal and Ocean Engineers
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• v.13 no.1
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• pp.35-45
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• 2001

The sea dike at Gun-Jang Industry Area had been constructed recently and both inside and outside areas of the dike show harmonic behaviour. To examine stability against piping, 2 dimensional seepage analysis was executed using finite element method. To investigate the harmonic motion of water level, unsteady and unsaturated flow analysis is needed, and specially harmonic motion in the both areas from the dike should be considered, Water level recorder was used to obtain tidal harmonic data, and sieve analysis has been earried out so that the distribution of grain diameter of the dike material is clearly informed. The calculated velocity was compared with critical velocity.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.24-31
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• 2018

In this paper, specimen tests with simulated large seismic conditions have been carried out to investigate damage characteristics such as structural deformation and crack initiation under seismic loading. The mechanical behavior of the specimens is predicted by numerical simulations and the strain-based damage evaluations are performed. Finite element analyses of the specimens under the simulated seismic loading at room and operating temperatures were carried out for low alloy steel and stainless steel materials. Peak strain amplitude, cumulative fatigue damage and cumulative strain limit damage are calculated considering the nature of cyclic loading. In all cases, the allowable damage criteria are exceeded at the time of observing cracks visually in the tests. Therefore, it is confirmed that the material behavior due to the large seismic loads can be predicted by the numerical method and the structural damage of the materials can be evaluated conservatively based on the strain criteria.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Journal of the Korean Society of Industry Convergence
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• v.25 no.5
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• pp.827-833
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• 2022

In this study, the prediction of the bending angle for the 350 A bellows-type expansion joints and the structural stability according to the load were determined. The stability of the 2km piping system was predicted by applying the allowable bending angle of the expansion pipe joint obtained from the analysis. The maximum bending angle was calculated through bending analysis of the bellows-type expansion joints, and the maximum bending angle by numerical calculation was about 1.8°, and the maximum bending angle of the bellows obtained by comparing the allowable strength of the material was about 0. 22°. This angle was very stable compared to the allowable bending angle (3°) of the expansion pipe joint regulation. By applying the maximum bending angle, the allowable maximum deflection of the 2 km pipe was about 3.8 m. When the seismic load was considered using regression analysis, the maximum deflection of the 2km pipe was about 142.3mm, and it was confirmed that the bellows-type expansion joints and the deflection were stable compared to the allowable maximum deflection of the pipe system. These research results are expected to present design and analysis guidelines for the construction of piping and the development of bellows systems, and to be used as basic data for systematic research.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.28-35
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• 2021

The flow accelerated corrosion (FAC) is one of significant aging and degradation mechanism and can affect structural integrity of CANDU feeder pipes. Pipe burst can occur under normal operation pressure (min. 10 MPa) if wall-thinning of the feeder pipe due to FAC is accumulated. Previous studies considered simple shapes of feeder pipe with local wall-thinning in order to conservatively assess structural integrity of wall-thinned feeder pipe. In this paper, a new FE model is developed, having an actual shape of the feeder pipe (double bent) as well as the actual wall-thinning shape and location based on the in-service inspection result. Then, the burst pressure assessment of the wall-thinned feeder pipe is performed using lower bound limit load analysis considering elastic-perfectly plastic material. In addition, an improved formulation to predict the burst pressure of the wall-thinned feeder pipe is presented and the safety margin is compared with an existing assessment method.