• Proceedings of the KSME Conference
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• 2001.11a
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• pp.61-66
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• 2001

Fracture behaviors of pipes with local wall thinning is very important for the integrity of nuclear power plant. In pipes of energy plants, sometimes, the local wall thinning may result from severe drosion-corrosion damage. However, effect of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. From the tests, fracture behaviors and fracture strength of locally thinned pipe were manifested systematically. The observed failure modes were divided into four types; ovalization, ovalization+cracking, local buckling and local buckling+cracking. Also, maximum load was successfully evaluated.

• Journal of the KSME
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• v.33 no.5
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• pp.421-434
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• 1993

원자력안전기술원이 참여하고 있는 배관건전성 국제공동연구 프로그램인 IPIRG-2에 대하여 그 배경과 과제의 구성 및 공동연구방법을 간략하게 소개하였으며, 그 1단계 연구 프로그램인 IPIRG-1에서의 연구결과를 요약기술하였다. 규제기관과 원자력 산업계가 모두 관심을 가지고 있는 기술적인 주제들이 다루어지고 있는 이 국제공동연구는 원자력발전소의 건전성에 관한 매우 중요한 결과를 제공할 것으로 믿는다. 또, 여러 기관이 컨소시엄 형태로 공동연구를 수행함으로써 연구의 중복성을 피하고 연구비의 한계성을 극복하면서 각 기관이 보유한 고유의 기술들을 집 대성한다는 점에 대해서도 주목할 가치가 있다. 이 글에서는 IPIRG-2 프로그램의 전반적인 목 표와 연구수행체계만 간략하게 소개하였는데, 다음 기회에는 각 소과제에 대한 기술적인 내용에 대해서 기술하게 될 것이다.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.351-356
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• 2001

A potential loss of structural integrity due to aging of nuclear piping may have a significant effect on the safety of nuclear power plants. In particular, failures due to the erosion and corrosion defects are a major concern. As a result, there is a need to assess the remaining strength of pipe with erosion/corrosion defects. In this paper, a limit load solution for the eroded and corroded SA106 Grade B pipes subjected by internal pressure is developed. based in 3-D finite element analyses, considering a wide range of the shape of pipeline, flaw depth and axial flaw length parametrically.

• Plant Journal
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• v.12 no.4
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• pp.44-50
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• 2016

To model the compression system with more integrity, the pipeline dynamic model was applied to the compression system model. To combine the pipeline dynamic model and the compression system model, appropriate boundary conditions were selected on each end of connecting pipe with compressor, plenum and throttle valve. Simulation result illustrate the effect of pipeline dynamic model on the stability of compression system.

• Structural Engineering and Mechanics
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• v.59 no.5
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• pp.795-802
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• 2016

Gas pipelines are types of structures that are highly susceptible to corrosion. Sometimes, the pipes are subjected to a thinning of the wall thickness at the inside or outside wall due to erosion/corrosion. Therefore, it is important to evaluate the strength of the pipes undergoing corrosion to maintain the integrity of the piping systems. The main purpose of this study is to understand failure aspects caused by degradation of metal due to corrosion through. The ASME standard offers a relationship for the yielding pressure of the corroded pipes which was compared with the finite element results. The results demonstrate to obtain accurate results, the ASME relationship is unreliable. Moreover, pitting corrosion must be considered critical more than of other types.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.742-749
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• 2003

Failure of a pipeline due to local wall thinning is getting more attention in the nuclear power plant industry. Although guidelines such as ANSI/ASME B31G and ASME Code Case N597 are still useful fer assessing the integrity of a wall thinned pipeline, there are some limitations in these guidelines. For instance, these guidelines consider only pressure loading and thus neglect bending loading. However, most Pipelines in nuclear power plants are subjected to internal pressure and bending moment due to dead-weight loads and seismic loads. Therefore, an assessment procedure for locally wall thinned pipeline subjected to combined loading is needed. In this paper, three-dimensional finite element(FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Maximum moments based on true ultimate stress(${\alpha}$$\sub$u,t/) were obtained from FE results to predict the failure of the pipe. These results were compared with test results, which showed good agreement. Additional finite element analyses were performed to investigate the effect of key parameters, such as wall thinned depth, wall thinned angle and wall thinned length, on maximum moment. Also, the effect of internal pressure on maximum moment was investigated. Change of internal pressure did not show significant effect on the maximum moment.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.5
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• pp.453-460
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• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.8-14
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• 2013

Steam turbine blades and discs of nuclear power plants are one of the most highly stressed areas of turbine rotor, and periodic inspection of the blade roots is essential for monitoring integrity and preventing turbine failure. Ultrasonic technique is applied for volumetric inspection of blade root. However, the complexity of blade root geometry imposes challenges to inspection of blades and discs. Recently, phased array ultrasonic inspection technology is being applied to numerous power generation inspection applications including turbine rotor. The phased array ultrasonic technique requires customized inspection wedges which are generally necessary to generate effectively higher incident angle. But the usage of this wedge can cause access limitation for the lower stage blades of turbine because of the wedge front length. Therefore, the shear wave phased array probe which can generate high inspection angle without wedge is essentially necessary. In this study, feasibility study is conducted for the shear wave phased array ultrasonic probe application to blade and disc inspection. As results, the experimental results show that the shear wave phased array probe can detect the flaw and measure its size with reliable accuracy. Therefore if this shear wave phased array probe is applied to field inspection of blade and disc, more reliable inspection is expected for turbine having access limitation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.62-69
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• 2016

Current guidance considers that uniaxially loaded specimen with a deep crack is used for the determination of the ductile-to-brittle transition temperature. However, reactor pressure vessel is under biaxial loading in real and the existence of deep crack is not probable through periodic in-service-inspection. The elastic stress intensity factor and the elastic-plastic J-integral which were used for crack-tip stress field and fracture mechanics assessment parameters. The difference of the loading condition and crack geometry can significantly influence on these parameters. Thus, a constraint effect caused by differences between standard specimens and a real structure can over/underestimate the fracture toughness, and it affects the results of the structural integrity assessment, consequentially. The present paper investigates the constraint effects by evaluating the master curve $T_0$ reference temperature of PCVN (Pre-cracked Charpy V-Notch) and small scale cruciform specimens which was designed to simulate biaxial loading condition with shallow crack through the fracture toughness tests and 3-dimensional elastic-plastic finite element analyses. Based on the finite element analysis results, the fracture toughness values of a small scale cruciform specimen were estimated, and the geometry-dependent factors of the cruciform specimen considered in the present study were determined. Finally, the transferability of the test results of these specimens was discussed.