• 한국안전학회지
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• 제11권3호
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• pp.3-9
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• 1996

Dynamic fracture characteristics of Polycarbonate WL-RDCB specimen were investigated. The dynamic crack propagation velocities in these specimens were measured by using both high speed camera system and silver paint grid method developed and justified in the INHA Fracture Mechanics Laboratory. The measured crack propagation velocities were fed into the INSAMCR code(a dynamic finite element code which has been developed in the INBA Fracture Mechanics Laboratory) to extract the dynamic stress intensity factors. It has been confirmed that both dynamic crack arrest toughness and the static crack arrest toughness depend on both the geometry and the dynamic crack propagation velocity of specimens. The maximum dynamic crack propagation velocity of Polycarbonate WL-RDCB specimen was found to be dependent on the material property, geometry and the type of loading. The dynamic cracks in these Polycarbonate WL-RDCB specimens seemed to propagate in a successive manner, involving distinguished 'propagation-arrest-propagation-arrest' steps on the microsecond time scale. It was also found that the relat-ionship between dynamic stress intensity factor and dynamic crack propagation velocities might be represented by the typical '$\Gamma$'shape.

• 대한기계학회논문집
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• 제14권1호
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• pp.112-118
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• 1990

본 연구에서는 ASTM-E24.01.06에서 제안하고 있는 실험방법을 응용하여 균열 정지 파괴인성값을 측정하였다.즉 쐐기와 분리형 부싱(wedge and split bushing)으 로 압축하중을 가함으로 균열선 웨지하중 시편[crack line wedge loaded specimen(CL- WL시편)]에 인장력을 발생시켜서 균열정지 응력확대계수( $K_{1a}$)를 결정하였다. 그리고 균열개시 응력확대계수가 균열정지 응력확대계수에 미치는 영향들을 여러가지 재료들에 대하여 체계적으로 검토하였다.다.

• 한국정밀공학회지
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• 제15권11호
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• pp.251-257
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• 1998

동적 파괴인성치 측정시스템과 동적 2차원 유한요소해석 프로그램을 개발하여 원자로압력용기에 사용하는 강(SA508 cl.3, SA516 gr.70)의 동적 파괴인성치와 동적 균열정지인성치를 평가하고 이에 대한 유용성을 확인하였으며, 이 시스템 을 이용하여 재료의 동적 파괴특성을 규명하였다. SA508 cl.3와 SA516 gr.70의 동적 균열전파속도(a)에 대응하는 동적 응력확대계수 (K(a))에 대한 실험식을 얻었으며, 동적 응력확대계수와 동적 균열전파속도와의 관계는 전형적인 "$\Gamma$" 형으로 나타남을 확인하였다.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.