• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.130-139
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• 2023

This paper examines the quantification of uncertainty for welding residual stresses in dissimilar metal welds used in nuclear power plants. A mock-up of a dissimilar metal weld pipe, consisting of carbon and stainless steel pipes, was fabricated to measure the residual stress. A Round-Robin analysis was conducted by Korean institutions to assess the welding residual stress. The analysis was carried out in the second order, and the data obtained by each institution was evaluated based on the information provided. Using the Round-Robin results, the distribution of uncertainty in welding residual stresses among Korean institutions was evaluated. The quantification of uncertainty for Korean institutions was found to have a wider range compared to the distribution of welding residual stresses observed in overseas institutions. This study is considered useful in the establishment of comprehensive strategies for evaluating welding residual stress analysis methods used by domestic institutions.

• 대한기계학회논문집A
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• 제21권11호
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• pp.1786-1795
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• 1997

In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to modify two J-R curve prediction methods previously proposed by the authors and to propose an additional J-R curve prediction method for nuclear piping materials. In the first method which is based on the elastic-plastic finite element analysis, a blunting region handling procedure is added to the existing method. In the second method which is based on the empirical equation, a revised general equation is proposed to apply to both carbon steel and stainless steel. Finally, in the third method, both full stress-strain curve and finite element analysis results are used for J-R curve prediction. A good agreement between the predicted results based on the proposed methods and the experimental ones is obtained.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.54-60
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• 2017

When a sudden rupture occurs in high energy lines, ejection of inner fluid with high temperature and pressure causes blast wave as well as thrust forces on the ruptured pipe itself. The present study is to examine pipe whip behaviors and blast wave phenomena under postulated pipe break conditions. In this context, typical numerical models were generated by taking a MSL (Main Steam Line) piping, a steam generator and containment building. Subsequently, numerical analyses were carried out by changing break locations; one is pipe whip analyses to assess displacements and stresses of the broken pipe due to the thrust force. The other is blast wave analyses to evaluate the broken pipe due to the blast wave by considering the pipe whip. As a result, the stress value of the steam generator increased by about 7~21% and von Mises stress of steam generator outlet nozzle exceeded the yield strength of the material. In the displacement results, rapid movement of pipe occurred at 0.1 sec due to the blast wave, and the maximum displacement increased by about 2~9%.

• 한국항해항만학회지
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• 제45권4호
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• pp.212-223
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• 2021

본 논문에서는 해양구조물의 배관에 작용하는 환경조건과 구조물의 움직임에 따른 구조안전성 평가를 수행하였다. 배관에 작용하는 조건은 N₂ generator의 설계 조건을 분석하여 최고온도와 최저온도 조건을 적용하였다. 구조물의 움직임은 DNV 규칙에 따라 계산하여 적용하였다. 각각의 조건을 조합하고 열 하중, 운동 하중 그리고 배관지지대의 유무에 따라 총 26가지 하중 조합을 구성하였고 상용프로그램인 MSC Patran/Nastran을 이용하여 해석을 진행하였다. 열해석은 Steady-state 방법인 Sol 153, 열-구조 연성 해석은 Linear-static 방법인 Sol 101을 각각 적용하여 수행하였다. 해석 결과, Set 1과 Set 2에서는 배관 내의 온도가 낮을수록, Set 3에서는 온도가 높을수록, Set 4에서는 배관 내외부의 온도 차가 클수록 응력이 증가하는 경향이 있었다. 하지만, 온도 하중만 있는 조건과 운동 하중만 있는 조건에서의 응력의 합이 두 하중의 복합 하중 조건에서의 응력과 같은 값을 나타내지는 않았다. 즉, 운동 하중에 의한 영향은 운동의 방향, 배관의 배치나 지지대의 위치 등에 따라 달라진다는 것을 알 수 있다. 따라서, 설계 시점에서 배관에 작용하는 운동 하중의 크기와 방향, 배관의 배치 그리고 배관 지지대의 위치 등을 종합적으로 고려할 필요가 있다.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.24-31
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• 2018

The mechanical design procedure is studied for the PCHE(printed circuit heat exchanger) with electrochemical etched flow channels. The effective heat transfer plates of PCHE are assembled by diffusion bonding to make a module. PCHE is widely used for industrial applications due to its compactness, cost efficiency, and serviceability at high pressure and/or temperature conditions. The limitations and technical barriers of PCHE are investigated for application to nuclear components. Rules for design and fabrication of PCHE are specified in ASME Section VIII but not in ASME Section III of nuclear components. Therefore, the calculation procedure of key dimensions of PCHE is defined based on ASME section VIII. The effective heat transfer region of PCHE is defined by several key dimensions such as the flow channel radius, edge width, wall thickness, and ridge width. The mechanical design procedure of key dimensions was incorporated into a program for easy use in the PCHE design. The effect of assumptions used in the key dimension calculation on stress values is numerically investigated. A comparative analysis is done by comparing finite element analysis results for the semi-circular flow channels with the formula based sizing calculation assuming rectangular cross sections.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.1-11
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• 2015

Feedwater Heaters are important components in a nuclear power plant. As the age of heater increases, the maintenance cost required for continuous operation also increases. Most heaters have the carbon steel shells, tube support plates and flow baffles. The carbon steel is susceptible to flow-accelerated corrosion. This is especially true if the flow has a two-phase mixture of steam and condensate. The wall thinning around the wet steam entrance area of the shell is inevitable during some long term operation. The structural integrity of the feedwater heater shell affects the safe operation of the nuclear power plant. Therefore, it is needed for the thinned shell to be repaired. The maintenance method for preventing failure of the shell should be determined by investigating various factors including the stress distribution of thinned area. The stress analysis of the shell including the steam entrance region is studied in this paper. The results of thinned shell is compared with that of intact shell.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.34-41
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• 2010

This work is concerned with the analysis of stress corrosion cracking(SCC) behavior of Alloy 600 nozzle penetration mock-up according to a residual stress induced by a dissimilar metal welding(DMW) in a nuclear reactor pressure vessel. The effects of the dimension and materials of the nozzle penetration on the deformation and the residual stress induced by DMW were investigated using a finite element analysis(FEA). The inner diameter(ID) change of the nozzle by DMW and its dependance on the design variables, calculated by FEA, were well consistent with those measured from the mock-up. Accelerated SCC tests were performed for three mock-ups with different wall thicknesses in a highly acidic solution to investigate mainly the effect of the residual stress on the SCC behavior of Alloy 600 nozzle. From a destructive examination of the mock-up after the tests, the SCC behavior of the nozzle was fairly related with the residual stress induced by DMW : axial cracks were found in the ID surface of the nozzle within the J-weld region where the highest tensile hoop stress was predicted by FEA, while circumferential cracks were observed beyond both J-weld root and toe where the highest tensile axial stress was expected.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.83-88
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• 2010

Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.17-22
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• 2016

Alloy 600 widely used in nuclear power plant is susceptible to primary water stress corrosion cracking (PWSCC). It is important to prioritize the inspection of Alloy 600 components using PWSCC susceptibility index. Plant-specific model for the susceptibility index was reviewed. The normalized PWSCC susceptibility index to a reference value is suggested and applied. The result was found to be reasonable.