• Nuclear Engineering and Technology
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• 제45권1호
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• pp.99-106
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• 2013

Thermal fatigue is a significant long-term degradation mechanism in nuclear power plants. In particular, as operating plants become older and life time extension activities are initiated, operators and regulators need screening criteria to exclude risks of thermal fatigue and methods to determine significant fatigue relevance. In general, the common thermal fatigue issues are well understood and controlled by plant instrumentation at fatigue susceptible locations. However, incidents indicate that certain piping system Tee connections are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentations. Therefore, in this study thermal fatigue evaluation of piping system Tee-connections is performed using the fluid-structure interaction (FSI) analysis. From the thermal hydraulic analysis, the temperature distributions are determined and their results are applied to the structural model of the piping system to determine the thermal stress. Using the rain-flow method the fatigue analysis is performed to generate fatigue usage factors. The procedure for improved load thermal fatigue assessment using FSI analysis shown in this study will supply valuable information for establishing a methodology on thermal fatigue.

• 대한기계학회논문집A
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• 제20권6호
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• pp.1872-1881
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• 1996

LBB(Leak-Before-Break) analysis is performed for the highest stress location of each different type of mateerials in the nuclear piping line. In most cases, the highest stress occurs in the pipe and nozzle interface location. i.e. terminal end. The current finite element analysis approach utilizes the symmetry condition both for locations near the nozzle and for locationa away from the nozzle to minimize the size of the finite element model and to make analysis simple when calculating the J-integral values at the crack tip. In other words, the nozzle is not included in the finite element model. However, in reality, the symmetric condition is not applicable for the pipe-nozzle interface location. Because the pipe-nozzle interface location is asymmetric due to different stiffenss of the pipe and nozzle(both material and dimensions). The simplified analysis approach for pipe-nozzle interface locaiton is too conservative for a smaller diameter piping. In tlhis paper, various analyses are performed for the range of materials and crack sizes to evaluate the nozzle effect for a LBB anlaysis. This paper presents methodology for developing the piping evaluaiton diagram at the pipe-nozzle interface location.

• 대한기계학회논문집A
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• 제32권8호
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• pp.668-677
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• 2008

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which consideded safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

• 한국가스학회지
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• 제17권2호
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• pp.50-58
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• 2013

공정플랜트에는 다양한 배관이 고압, 고온의 인화성, 폭발성 물질을 이송하고 있다. 잦은 설계 변경 및 증설 등으로 복잡한 형상으로 배관이 형성되어 있는 경우가 많으나 배관의 구조가 단순하여 실제 위험성에 비해 안전 관리가 부족한 경우가 많다. 본 연구에서는 국내 한 업체에서 부생가스를 연료로 사용하던 배관을 설계 변경하여 천연가스와 혼합하여 사용하도록 사례를 활용하여, 배관의 안전성을 평가 하는 방법을 예시하였다. 배관의 설계 변경 후 안전성을 ASME 기준에 따라 재평가하고, 배관의 주요 관리부위를 결정하는 방법을 제시하였다. 배관의 분기 및 루프 등이 다수 복잡하게 연결되어있는 가스혼합용 믹싱드럼 배관 시스템을 대상으로 해석하였다. 배관의 주요부위 응력 민감도를 이해하기 위해 배관의 지지대 구속조건 및 외부 온도를 변화시켜 가면서 이들의 영향을 연구하였다. 또한 부생가스가 포함하고 있는 수소가스에 의한 손상 가능성에 대해서도 논의하였다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.30-40
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• 2016

The paper investigates the previous studies about welding residual stresses in nuclear components. First, various residual stress measurement methods are reviewed in applicability. Second a finite element welding residual stress analysis technique, which was developed from the viewpoint of FFS (Fitness-For-Service) assessment, is explained. Third, characteristics of the welding residual stresses on J-groove welds and butt welds were presented via investigating the previous studies. Last, engineering formulae for residual stresses in the FFS assessment codes such as R6 and API 579/ASME FFS-1 Code is summarized.

• 한국소음진동공학회논문집
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• 제19권11호
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• pp.1151-1157
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• 2009

Ensuring both product quality and reducing material cost are important issue for the design of the piping system of an air conditioner outdoor unit. This paper describes a shape optimization that achieves mass reduction of an air conditioner piping system while satisfying two design constraints on resonance avoidance and the maximum stress in the pipes. In order to obtain optimized design results with various analysis fields considered simultaneously, an automated multidisciplinary analysis system was constructed using PIAnO v.2.4, a commercial process integration and design optimization(PIDO) tool. As the first step of the automated analysis system, a finite element model is automatically generated corresponding to the specified shape of the pipes using a morphing technique included in HyperMesh. Then, the performance indices representing various design requirements (e.g. natural frequency, maximum stress and pipe mass) are obtained from the finite element analyses using appropriate computer-aided engineering(CAE) tools. A sequential approximate optimization(SAO) method was employed to effectively obtain the optimum design. As a result, the pipe mass was reduced by 18 % compared with that of an initial design while all the constraints were satisfied.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.140-145
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• 2023

When performing stress analysis for a nozzle in nuclear power plants, the nozzle loads should be determined conservatively. Existing stress analysis report of 3-D nozzle loads in nuclear power plants often provide only load magnitude not the sign (direction). Since calculated stress distribution depends on load direction, determining critical load directions for conservative stress analysis is crucial. In this study, an efficient method for determining critical load directions in nozzle loads is proposed. In the proposed method, stresses are firstly calculated using elastic finite element (FE) analysis for the uni-axial load in each direction. Then stress distributions for the multi-axial loads are analytically calculated using the principle of superposition. The calculated stress values are verified by comparing with FE analysis results under multi-axial loading. By using this method, the complex task of determining conservative load directions can be simplified.

• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.9-16
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• 2011

Primary water stress corrosion cracking (PWSCC) has been found in the weld region of the nuclear power plant. Welding can produce tensile residual stress. Tensile residual stress contributes to the initiation and growth of PWSCC. It is important to estimate weld residual stress accurately to predict or prevent the initiation and growth of PWSCC. This paper shows the results of finite element analysis and measurement experiment for weld residual stress. For the study, four kinds of specimen were fabricated with the materials used in the nuclear power plant. Residual stresses were measured by four kinds of methods of hole drilling, x-ray diffraction, instrumented indentation and sectioning. Through the study, numerical analysis and measurement results were compared and the characteristics of each measurement technique were observed.

• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.8-18
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• 2012

In early constructed nuclear power plants, Ni-based Alloys 82/182 had been widely used for dissimilar metal welds (DMW) as a weld filler metal. However, Alloys 82/182 have been proven to be susceptible to primary water stress corrosion cracking (PWSCC) in the nuclear primary water environment. The formation of crack due to PWSCC is also influenced by weld residual stresses. Thus, the accurate estimation of weld residual stresses of DMW is crucial to investigate the possibility of PWSCC and instability behaviors of crack due to PWSCC. In this context, the present paper investigates weld residual stresses of nuclear reactor piping nozzles based on 2-D axi-symmetric finite element analyses based on layer-based approach using maximum molten bead temperature. In particular, the effect of analysis parameters, i.e., a thickness of weld layer, an initial molten bead temperature, convection heat transfer coefficient, and geometric constraints on predicted weld residual stresses was investigated.

• 한국압력기기공학회 논문집
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• 제12권2호
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• pp.50-57
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• 2016

The study presents structural integrity assessment of ductility degradation of a baffle former assembly by performing finite element analysis considering real loading conditions and stress triaxiality. Variations of fracture strain curves of type 304 austenitic stainless steel with stress triaxiality are derived based on the previous study results. Temperature distributions during normal operation such as heat-up, steady state, and cool-down are calculated via finite element temperature analysis considering gamma heating and heat convection with reactor coolant. Variations of stress and strain state during long operation period are also calculated by performing sequentially coupled temperature-stress analysis. Fracture strain is derived by using the fracture curve and the stress triaxility. Finally, variations of ductility degradation damage indicator with the fracture strain and the equivalent inelastic strain are investigated. It is found that maximum value of the ductility degradation damage index continuously increases and becomes 0.4877 at 40 EFPYs. Also, the maximum value occurs at top and middle inner parts of the baffle former assembly before and after 20 EFPYs, respectively.