• Journal of the Computational Structural Engineering Institute of Korea
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• v.29 no.4
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• pp.347-354
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• 2016

Because typical gas turbine systems have frequent startup and shutdown operations, it is likely to cause cracks at the gas turbine casing and gas leakages at casing flanges due to thermal fatigue and embrittlement. Therefore, the evaluation of structural integrity and gas leakage at the gas turbine casings must be performed. In this paper, we have evaluated the structural integrity of the turbine casing and bolts under a normal operation in accordance with ASME B&PVC and evaluated the leakage at casing flanges by examination of contact pressure calculated using the finite element analysis. Finally, we propose a design flow including finite element modeling, the interpretation and evaluation methods for gas turbine casings. This may be utilized in the design and development of gas turbine casings.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.331-340
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• 2004

Deployment of an advanced on-line monitoring of the component integrity offers a prospect of improved performance, enhanced safety, and reduced overall cost for nuclear power plants. Ultrasonic guided waves have been known as one of the promising techniques that could be utilized for on-line monitoring. The present work is aimed at developing a new method for on-line monitoring of the pipes during the operation period of nuclear power plants. For this purpose, the steam generator (S/G) tube was selected as an object of tile experiment. Dispersion corves and the incident angles corresponding to the specific modes were calculated for the S/G tube. The modes of guided waves were identified by the time-frequency diagrams obtained by the short time Fourier transform. It was experimentally confirmed that there was no mode conversion when the ultrasonic guided waves passed over the curved region of the S/G tube. An optimum mode of guided wave for the S/G tube was suggested and verified by the experiment.

• Journal of Mechanical Science and Technology
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• v.14 no.1
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• pp.72-83
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• 2000

In recent years considerable progress has been made in the characterisation of surface finish in three dimensions, and in the development of protocols which can be used for international standardisation. Although the subject as it has currently developed has much further to go if the process of surface characterisation is to impact on manufacture, control and specification of the manufacturing process itself. Researchers in this important area are beginning to realise that if the subject is to have great impact on manufacturing industries, surface characterisation must be broadened to include measures of surface integrity of the component and in addition be related to the functional demands imposed on the surface. The functional demands being a requirement of the engineering situation in which the components are employed. If these three factors are considered simultaneously, surface characterisation, surface integrity and component function, then a new and important subject is born, the subject of the Engineered Surface. Part 1 of this paper attempts to draw together the elements which go together to create the subject, 'The Engineered Surface'. The paper presents a method by which this important subject can be developed to the benefit of manufacturing industries. The paper also discusses the importance of a co-ordinated approach to the subject and the way that information can be documented to eventually provide a useful atlas of controlling parameters which are essential for a range of material processing industries as they strive to meet the ever more stringent and cost effective requirements of the manufacture.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.7
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2219-2227
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• 2002

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.8 s.251
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• pp.993-1000
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• 2006

Operating experience of steam generators has shown that cracks of various morphology frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.8 s.179
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• pp.2015-2021
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• 2000

In order to evaluate the integrity of nuclear power plant components, the analysis based on fracture mechanics is crucial. For this purpose, finite element method is popularly used to obtain J-integral. However, it is time consuming to design the finite element model of a cracked structure. Also, the J-integral should be verified by alternative methods since it may differ depending on the calculation method. The objective of this paper is to develop a three-dimensional elastic-plastic J-integral analysis system which is named as EPAS program. The EPAS program consists of an automatic mesh generator for a through-wall crack and a surface crack, a solver based on ABAQUS program, and a J-integral calculation program which provides DI (Domain Integral) and EDI (Equivalent Domain Integral) based J-integral calculation. Using the EPAS program, an optimized finite element model for a cracked structure can be generated and corresponding J-integral can be obtained subsequently.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.12 s.255
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• pp.1618-1625
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• 2006

During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective steam generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement.