• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.296-301
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• 1996

이 연구의 목적은 원주방향 균열을 가진 페라이틱 배관의 파괴거동을 실험적으로 평가하는데 있다. 한계하중방법, SC.TNP 방법, R6방법, 그리고 ASME Code방법과 같은 여러 파괴거동 평가 방법의 타당성이 PWR 운전조건(압력:15.5MPa, 온도:228$^{\circ}C$)하에서의 직경 16인치의 대규모 배관파괴실험을 통해 조사된다. 모사지진하중, 단일주파수 사인함수하중, 정하중과 같은 여러 가지 형태의 하중이 배관의 하중지지능력에 미치는 영향이 조사된다. 또한 엘보우부위와 직관부의 영향과 표면균열 및 관통균일의 영향 등도 함께 조사된다. 결과는 다음과 같다. (1) 표면균열을 가진 배관의 파괴거동은 한계하중방법과 SC.TNP 방법에 의해 잘 예측할 수 있다. 반면 관통균열의 경우는 한계하중방법에 의해 잘 예측된다. (2) 모사지진하중하에서는 단일주파수 사인함수하중이나 정하중 하에서 보다 하중지지능력이 크게 예측된다. (3) 엘보우부위와 직관부, 관통균열과 표면균열 사이에 파괴거동에 대한 큰 차이는 없다.

• Journal of the KSME
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• v.33 no.5
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• pp.414-420
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• 1993

원자력 발전소의 운전환경에 따른 노화손상은 발전소 운전 내력에 연류된다. 통상SCC는 40년 또는 40년 이하의 수명을 가정하여 이에 따른 설계 여유도나 안전 계수를 기초로 하여 설계된다. 그러나 설계여유도나 안전계수는 운전중 변경될 수도 있다. 원자력 발전소 기기의 변경될 수도 있다. 원자력 발전소 기기의 수명평가를 위해서는 초기설계치 및 잔여수명이나 실제 설계치의 변화 등을 평가하여야 한다. 주요 물성치 및 손상 매개변수의 실제 변경 상황은 수명기간 설정에 근간이 된다. SSC의 수명평가를 위해 재료의 노화손상은 철저히 파악되어야 한다. 또한 기계적 하중, 열적하중 등의 노화 촉진요인과 정상운전 및 시험도 노화손상에 영향을 미치므로 이들에 대해 상세 평가를 수행하여야 한다. 수명평가는 운전환경 및 노화촉진요인들과 밀접한 관련이 있으므로, SSC의 수명을 평가하기 위해서는 이러한 복잡한 변수들 사이의 연관관계를 구체적 으로 고려하여야 한다. 그러므로 SSC의 수명을 평가하기 위해서는 이러한 복잡한 변수들 사이의 연관관계를 구체적으로 고려하여야 한다. 그러므로 SSC의 설계, 제작, 설치, 시험, 운전상태 및 보수주기 등에 대한 정보파악이 선행되어야 한다.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1996.11a
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• pp.913-920
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• 1996

In this study, the counting method for multi-axial and multi-point load states was proposed. Using this counting method, the load spectrum is generated from the service load history which is measured for boom structure of excavator. Loading state for loading points of boom structure is described as a multi-dimensional state space. From this load spectrum, the stress spectrum was generated by FEM analysis using the superposition of the unit load. The cumulated damage at the severe damage point of In nm structure by the failure example is calculated by Palmgren-Miner's rule. As a result of this study, the fatigue life estimation using the multi-axial and multi-point load counting method is useful.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.5
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• pp.453-460
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• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.19 no.8
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• pp.44-54
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• 2018

Stress and fatigue of the distributor, an equipment of the high-pressure evaporator for the HRSG, were evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the results of the piping system analysis model, reaction forces of the tubes connected to the distributor were derived and used as the nozzle load applied to the detailed analysis model of the distributor afterward. Next, the detailed model to analyze the distributor was constructed, the distributor being statically analyzed for the design condition with the steam pressure and the nozzle load. As a result, the maximum stress occurred at the bore of the horizontal nozzle, and the primary membrane stress at the shell and nozzle was found to be less than the allowable. Next, for the transient operating conditions given for the distributor, thermal analysis was performed and the structural analysis was carried out with the steam pressure, nozzle load, and thermal load. Under the transient conditions, the maximum stress occurred at the vertical downcomer nozzle, and of which fatigue life was evaluated. As a result, the cumulative usage factor was less than the allowable and hence the distributor was found to be safe from fatigue failure.

• Journal of the wind engineering institute of Korea
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• v.22 no.4
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• pp.163-169
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• 2018

In this study, wind loads exerted on the offshore wind turbine rotor in parked condition were predicted with variations of wind speeds, yaw angles, azimuth angle, pitch angles, and power of the atmospheric boundary layer profile. The calculated wind loads using blade element theorem were compared with those of estimated aerodynamic loads for the simplified blade shape. Wind loads for an NREL's 5 MW scaled offshore wind turbine rotor were also compared with those of NREL's FAST results for more verification. All of the 6-component wind loads including forces and moments along the three axis were represented on a non-rotating coordinate system fixed at the apex of rotor hub. The calculated wind loads are applicable for the dynamic analysis of the wind turbine system, or obtaining the over-turning moment at the foundation of support structure for wind turbine system.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.33 no.2
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• pp.39-45
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• 2005

The aim of this research is to predict the thickness and loading noise of the round-tip shaped Hartzell propeller currently used in the general aviation aircraft. Before implementing the noise analysis, the pressure distribution on the propeller was obtained by using the free wake panel method and unsteady Bernoulli's equation. The noise signal at observer position can be obtained by using the FW-H equation. The noise prediction results for the propeller indicates that the thickness noise has s symmetric directivity pattern with respect to the tip path plane, while the noise due to loading shows higher noise directivity toward downstream than the upstream direction from the rotor plane. The loading noise is dominant rather than the thickness noise in normal operating condition.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.393-402
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• 1995

The structural analysis of the reactor coolant system mainly consist of too fields. The one is the static analysis considering the impact of pressure and temperature built up during normal operation. The other is the dynamic analysis to estimate the impact of postulated events such as the seismic loads or postulated branch line pipe breaks event. Since the most important goal of the RCS structural analysis is to prove the safety of the RCS during normal operation or postulated events, a widely proven theory having enough conservatism is adopted. The load occurring on the RCS during normal operation is considered as the basic design loading condition throughout whole plant life time. The most typical characteristic of the RCS during normal operation is the thermal expansion of the RCS caused by reactor coolant with high temperature and pressure. Therefore, the exact estimation on the thermal movement of the RCS is needed to get more clear understanding on the thermal movement behavior of the RCS. In this study, the general structural analysis concept and modeling method to evaluate the thermal movement of the RCS under the normal plant operation condition are presented. To discuss the validation of the suggested analysis, analysis results are compared with the measured data which ore referred from the standardized 1000 MWe PWR plant under construction.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.168.2-168.2
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.324-329
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• 1998

본 연구는 주증기배관으로 사용되고 있는 SA106. Gr.C의 모재와 용접계에 대해서 파괴인성에 미치는 온도와 하중속도의 영향을 살펴보기 위해서 다양한 온도와 하중속도에서 J-R 시험 및 인장시험을 수행하였다. 두 재료 모두 동적변형시효의 영향을 받고 있는 온도영역에서 약 40% 정도의 파괴인성 감소가 관찰되었으며, 하중속도에 따른 파괴인성 감소영역은 serration과 인장강도 증가 영역의 하중속도 의존성과 일치하였다. 원자력발전소 운전온도에서 모재와 용접재 모두 하중선변위속도가 4.0mm/min 일 때 파괴인성치의 최저를 보였으며, 하중속도가 증가함에 따라 증가하여 동적하중속도(800, 40mm/min)일 때 최대를 보였다. 모재와 용접재를 비교하면 용접재에서 serration이 뚜렸했고, 보다 넓은 온도영역에서 관찰되었다. 또한 인장강도의 증가가 보다 고온에서 형성되었다. 이러한 특성은 용접재가 모재에 비해 냉각률이 크고 미세한 결정입으로 이루어져있으며, 망간의 함량이 높기 때문으로 판단된다.