• 대한조선학회논문집
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• 제30권3호
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• pp.75-89
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• 1993

본 논문은 기본적으로 미국 기계학회의 성능시험 코드에 준하여 운동 모형시험에 불확실성 해석을 적용하였다. 총 불확실성은 시험준비 과정, 계기의 검증, 데이타 취득 과정 및 분석과정에서 발생하는 오차요인들과 오차 전파 방정식에 의해 구하여진 민감도와 결합하여 구해진다. 저항추진 시험과 같은 정적시험과 운동시험과 같은 동적시험의 차이들이 불확실성 해석의 모든 과정동안 분명하게 구별되며 또한 비대칭 고정오차항목이 고려되었다. 본 불확실성 해석이 수행된 운동 응답 항목은 무차원화된 주파수의 함수로 나타낸 상하동요와 종동요의 RAO이다. 운동시험에 대한 불확실성 해석은 측정 정도를 향상시키고 오차들을 정량화하는데 유용할 뿐만아니라 여러가지 운동 해석 프로그램에 대한 검증의 기준으로 제시될 수 있다.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.71-76
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• 2019

Actual stress distributions in the nozzle of a pressure vessel may not be in plane strain condition, implying that the crack-tip constraint condition may be relaxed in the nozzle. In this paper, a methodology for evaluating the fracture toughness of the ASME Code is presented considering the relaxation of the constraint effect in the nozzle of the reactor pressure vessel. The crack-tip constraint effect is quantified by the T-stress. The equation, which represent the relation between the fracture toughness in the lower constraint condition and the plane strain fracture toughness, is derived using the T-stress. This equation is similar to the method for evaluating the fracture toughness of the Master Curve for low constraint conditions. As a result of evaluating the fracture toughness considering the constraint effect in the reactor inlet, outlet and direct injection nozzles using the proposed equation, it was confirmed that the fracture toughness in the nozzles is higher than the plane strain fracture toughness. Applying the proposed evaluation methodology, it is possible to reflect the relaxation of the constraint effect in the nozzles of the reactor pressure vessel, therefore, the safe operation area on the pressure-temperature limit curve can be prevented from being excessively limited.

• 대한금속재료학회지
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• 제48권3호
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• pp.194-202
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• 2010

Application of a stronger and more durable material for reactor pressure vessels (RPVs) might be an effective way to insure the integrity and increase the efficiency of nuclear power plants. A series of research projects to apply the SA508 Gr.4 steel in ASME code to RPVs are in progress because of its excellent strength and durability compared to commercial RPV steel (SA508 Gr.3 steel). In this study, the microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure that has coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as $M_{23}C_6$ and $M_7C_3$ due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. In addition, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect, and the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

• 대한금속재료학회지
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• 제50권4호
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• pp.271-279
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• 2012

9-12% Cr steels have been used in thermal power plants which repeat start and stop operations. Major factors of fatigue life are temperature, frequency, stress ratio, holding time, microstructure, and environment. Normally, fatigue life decreases at high temperature, low frequency, high stress ratio, and long holding time conditions. A Mod.9Cr-1Mo steel, called G91, was developed at ORNL (Oak Ridge National Laboratory, USA) and was adopted as a high-temperature structural material in the ASME Code in 2004. However, its low-cycle fatigue and fatigue crack growth characteristics have been rarely studied. In this work, we have investigated the low-cycle fatigue crack growth behaviors of G91 steel under various test conditions in terms of temperature and stress ratio. As temperature and stress ratio increase, the crack growth rate becomes faster and striation distance also increases. On the other hand, the number of branch cracks decreases.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.36-48
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• 2021

As the material aging of nuclear power plants has been progressing in domestic and overseas, crack growth becomes one of the most important issues. In this respect, the crack growth assessment has been considered an essential part of structural integrity. The crack growth assessment for nuclear power plants has been generally performed based on ASME B&PV Code, Sec. XI but the idealization of crack shape and the conservative solutions of stress intensity factor (SIF) are used. Although finite element analysis (FEA) based on iterative crack growth analysis is considered as an alternative method to simulate crack growth, there are yet no guidelines to model the crack-tip spider-web mesh for such analysis. In this study, effects of various meshing factors on FE SIF calculation are systematically examined. Based on FEA results, proper criteria for spider-web mesh in crack-tip are suggested. The validation of SIF calculation method through mapping initial stress field is investigated to consider initial residual stress on crack growth. The iterative crack-tip modeling program to simulate crack growth is developed using the proposed criteria for spider-web mesh design. The SIF results from the developed program are validated by comparing with those from technical reports of other institutes.