• Corrosion Science and Technology
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• 제3권5호
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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.

• 비파괴검사학회지
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• 제24권4호
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• pp.331-340
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• 2004

기기 건전성의 온라인 검사 및 감시 기술은 기존의 원전의 가동 중 검사의 효율적 수행에 필요한 정보를 제공한다. 유도 초음파는 온라인 검사와 감시에 활용 가능성이 있는 기술 중의 하나로 알려져 있다. 본 연구에서는 온라인 검사 겐 감시 기술 개발의 일환으로 유도초음파를 이용하여 파이프의 건전성을 검사/감시하는 기술을 개발하고자 한다. 이를 위해 증기 발생기 세관을 실험 대상으로 하여 분산 선도 덴 특정 모드에 대응하는 초음파 입사각을 계산하였다. Short time Fourier transform을 이용한 시긴-주파수 분석을 통하여 유도초음파 모드를 확인하였으며, 유도 초음파가 증기발생기 세관의 곡관 부분을 통과할 때 모드 변환이 발생하지 않는 것을 실험적으로 확인하였다. 유도 초음파를 이용한 증기발생기 세관의 최적 검사 모드를 제안하고 실험에 의하여 이를 확인하였다.

• Journal of Mechanical Science and Technology
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• 제14권1호
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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.

• 대한기계학회논문집A
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• 제21권7호
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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.

• 대한기계학회논문집A
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• 제26권11호
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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.

• 대한기계학회논문집A
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• 제30권8호
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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.

• 대한기계학회논문집A
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• 제24권8호
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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.

• 대한기계학회논문집A
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• 제34권7호
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• pp.843-850
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• 2010

지지격자체는 경수로 핵연료집합체의 가장 중요한 핵심 구조부품이다. 지지격자체 설계시의 고려사항은 원자로 운전중에 연료봉의 지지건전성을 유지하도록 하는 것이다. 본 연구에서는 연료봉이 유동기인진동에 의해서 진동할 때 연료봉과 연료봉 지지부 사이에서 상대변위를 완화해 줌으로서 연료봉의 프레팅 마모손상 가능성을 감소시킬 수 있는 이동 가능한 연료봉 지지부로 구성된 새로운 지지격자체 형상을 제안하였다. 아울러 제안된 이동 가능 지지부의 연료봉 지지특성을 유한요소해석을 통해 분석하였다.

• 대한기계학회논문집A
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• 제30권12호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.281-286
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• 2008

Wall thinning defect due to flow accelerated corrosion is one of major aging phenomena in most power plant industries, and it results in reducing load carrying capacity of the piping systems. A failure testing system was set up for real scale elbows containing various simulated wall thinning defects, and monotonic in-plane bending tests were performed under internal pressure to find out the failure behavior of thinned elbows. Various finite element models were generated and analysed to figure out and simulate the behavior for other thinning shapes and loading conditions. This paper presents the decreasing trends of load carrying capacity according to the thinning dimensions which were revealed from the investigation of finite element analysis results. A mechanical integrity evaluation model for thinned elbows was proposed, also. This model can be used to calculate the TES plastic load of thinned elbows for general internal pressure, thinning location, and in-plane bending direction.