• 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.

• 대한기계학회논문집A
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• 제31권2호
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• 대한기계학회논문집A
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• 제22권4호
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• pp.834-842
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• 1998

This paper examines the crack growth behavior of 7075-T651 and 5052-H32 aluminum alloy under high-low block loading condition. The cantilever beam type specimen with a chevron notch is used in this study. The crack growth and closure ae investigated by compliance method. The applied stress ratios are R=0.15, 0.0, -0.15 and R=-0.15, 0.0, 0.15. The crack growth rate was found to increase as the load amplitude increases. However,${\bigtriangleup}K_eff$ was almost independent on the stress ratio. The experimental constants of 7075-T651 and 5052-H32 in Paris law were c`=1-1.3${\times}{10^-7},m`=3~3.2 and c`=4~6{\times}{10^-9}, m`=4.3-4.8$, respectively. $K_op$ of 7075-T651 and 5052-H32 becomes smaller as the stress ratio decreases. It seems that the crack closure affects $K_op$.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.58-65
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• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1186-1191
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• 2003

The 40% of wall criterion, which is generally used for the plugging of steam generator tubes, may be applied only to a single crack. In the previous study, a total of 9 failure models were introduced to estimate the local failure of the ligament between cracks and the optimum coalescence model of multiple collinear cracks was determined among these models. It is, however, known that parallel axial cracks are more frequently detected during an in-service inspection than collinear axial cracks. The objective of this study is to determine the plastic collapse model which can be applied to the steam generator tube containing two parallel axial through-wall cracks. Nine previously proposed local failure models were selected as the candidates. Subsequently interaction effects between two adjacent cracks were evaluated to screen them. Plastic collapse tests for the plate with two parallel through-wall cracks and finite element analyses were performed for the determination of the optimum plastic collapse model. By comparing the test results with the prediction results obtained from the candidate models, a plastic zone contact model was selected as an optimum model.

• 한국정밀공학회지
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• 제22권10호
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• pp.137-143
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• 2005

The integrity of steam generator tubes in nuclear power plant should be maintained sufficiently during operation. For sake of this, complicated assessment procedures are required such as fracture mechanics analysis, etc. The integrity assessment of tubes has been performed by using conventional deterministic approaches while there are many uncertainties to carry out a rational evaluation. In this respect, probabilistic integrity assessment is considered as an alternative method for integrity assessment. The objectives of this study are to develop an integrity assessment system based on probabilistic fracture mechanics and to predict the failure probability of steam generator tubes containing an axial through-wall crack. The developed integrity assessment system consists of three evaluation modules, which apply first order reliability method, second order reliability method and Monte Carlo simulation method, respectively. The system has been applied to predict failure probability of steam generator tubes and the estimation results showed a promising applicability of the probabilistic integrity assessment system.

• 대한기계학회논문집A
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• 제25권7호
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• pp.1031-1038
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• 2001

In order to verify the analytical methods predicting failure behavior of cracked piping, full-scale pipe tests are crucial in nuclear power plant piping. For this reason, series of international test programs have been conducted. However, full-scale pipe tests require expensive testing equipment and long period of testing time. The objective of this paper is to develop a test system which can economically simulate the full-scale pipe test regarding the integrity evaluation. This system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system was developed for the integrity evaluation of nuclear piping based on the methodology of hardware-in-the-loop (HiL) simulation. Using this simulator, the piping integrity can be evaluated based on the elastic-plastic behavior of full-scale pipe, and the high cost full-scale pipe test may be replaced with this economical system.

• 한국안전학회지
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• 제23권5호
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• pp.61-66
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• 2008

It is important to calculate the exact crack opening area in the cracked pipe subjected to axial force and bending moment. Among many solutions for obtaining the crack opening displacement, Paris-Tada's expression, which is derived from energy method, is open used in fracture analysis for piping crack problems because of its simplicity. But Paris-Tada's equation has conservativeness when radius over thickness ratio(R/t) is ten or less, for it is based on the stress intensity factor solution having a compliance function derived from a simple shell theory. In this paper we derived a new expression using a different stress intensity factor solution which is able to consider the variation of compliance through wall thickness in a cracked pipe. Conservativeness of both equations was examined and compared to finite element analysis results. Conservativeness of the new equation is decreased when R/t > 10 and increased slightly when R/t < 10 compared with Paris-Tada's. But Both equations were highly conservative when R/t < 10 compared with finite element analysis results.

• 한국구조물진단유지관리공학회 논문집
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• 제20권1호
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• pp.112-117
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• 2016

최근 인공지반녹화가 주변 생활공간의 쾌적화, 에너지 절약효과, 도시 미관의 환경개선 등의 장점을 갖고 있어 콘크리트 위에 녹화 공간 조성이 증가하고 있다. 이에 따라 건축공사 표준 시방서, 건축물 녹화 설계기준 등에서는 방수층의 수밀안정성 확보를 위해 일정 품질 이상의 방근층 설치를 규정하고 있으나, 실제 우리나라의 많은 현장은 방수층 위에 누름콘크리트 타설 후 인공녹화를 시행하고 있는 실정이다. 이에 본 연구에서는 방근층을 시공하지 않은 현장을 대상으로 지표를 설정하였으며, 누름콘크리트의 균열에 따른 식물의 뿌리가 균열부로 침입 및 관통하여 누름콘크리트 하단부의 아스팔트 방수시트에 미치는 영향을 평가하였다.

• 대한기계학회논문집A
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• 제36권10호
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• pp.1115-1130
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• 2012

최근 원자로 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 균열로 인한 냉각수 누출사고가 발행하고 있다. 이러한 사고의 원인은 용접에 의한 인장잔류응력, 농축된 붕산수 및 응력부식에 민감한 재료로 인한 일차수응력부식균열(PWSCC : primary water stress corrosion cracking)인 것으로 판명되었다. PWSCC 평가는 원자로 건전성 평가의 주요 관심사로서 용접에 의해 발생되는 잔류응력을 정확하게 예측함으로써 가능하다. 본 연구에서는 유한요소해석을 이용하여 국내 원자로의 일반적인 J-groove 용접부의 해석절차를 소개하고, 용접해석 관련 변수의 민감도 해석을 통해 잔류응력 예측기법을 제시하고자 한다. 이를 위해 2 차원 및 3 차원 요한요소해석 방법을 바탕으로 변수 민감도 해석을 수행하였으며, 기존 연구결과와 비교를 통해 해석절차 및 방법의 유용성을 검정하였다.