• 한국전산구조공학회논문집
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• 제13권3호
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• pp.295-304
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• 2000

본 연구의 목적은 지진하중에 대한 응력 해석을 수행하여 ASME, Class 3 설계요건에 따라 스트레이너의 구조건진성을 평가하는 것이다. 스트레이너에 대한 설계요건이 ASME 코드 내에 명백하게 규정되어 있지 않기 때문에 본체는 밸브 설계요건인 ND-3500을 적용하고, 양쪽 플랜지 연결부는 배관 설계요건 중 ND-3658.3을 적용하였으며, 하단의 덮개 플랜지는 Appendix XI에 따라 설계 및 해석을 수행하였다. 본 연구에서는 T형 스트레이너를 쉘로 모델링하여 유한요소법을 사용하여 지진하중에 의해 스트레이너가 응답하는 모드 형상 및 고유진동수를 계산하여 충분히 강건한 구조물임을 입증한 후 정적 해석을 수행하여 주관과 분기 관이 접합하는 연결부위와 같은 위험단면에서의 막응력과 굽힘 응력을 구하였다. 각 하중조합에 대해 코드에서 규정하고있는 허용 값과 비교한 결과 스트레이너는 지진하중이 작용하는 경우 구조적 건전성을 유지하고 있음을 확인하였다. 아울러 인접 배관을 연결해주는 플랜지 연결부의 응력을 규정에 따라 구한 후 설계요건에 의한 허용 값과 비교하여 건전성을 만족함을 알 수 있었다.

• 한국전산구조공학회논문집
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• 제29권1호
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• pp.1-8
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• 2016

본 논문에서는 원심식 대형 압축기 구동용 모터 베이스 프레임의 구조해석 및 콘크리트 타설에 따른 구조안전성 평가를 수행하였다. 먼저 모터 베이스 프레임에 적용되는 네 가지 하중조건에 따른 구조해석을 진행하고 최대 비틀림 에너지 이론 및 Mohr-Coulomb 이론을 통하여 구조안전성을 평가하였다. 구조해석 결과에서 취약한 구조안전성을 나타낸 연결부 등의 불연속적인 부분에서 발생하는 국부응력에 대하여 ASME VIII Div. 2에 따른 구조안전성 평가를 적용함으로써 좀 더 합리적으로 구조안전성 평가를 수행할 수 있었다. 또한, 모터 베이스 프레임 내부에 콘크리트 타설 및 미타설에 따른 구조해석 및 ASME 구조안전성 평가를 통하여 모터 베이스 프레임의 구조안전성을 정량적으로 비교하여 콘크리트 타설로 인한 구조 안전성의 향상을 확인하였다.

• 열병합발전
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• 통권35호
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• pp.4-11
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• 2003

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.70-77
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• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• 한국안전학회지
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• 제21권6호
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• pp.20-25
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• 2006

In the design of class 1 components to apply ASME code section III NB, a fatigue is considered as one of the important failure mechanisms. Fatigue analysis procedure and standard fatigue design curve(S-N curve) is suggested in ASME code, which had to be performed to meet the integrity of components at the design step. As the plant life extension for operating power plants and the long-lived plant design, however, are being progressed, the fact which the existing ASME fatigue design curve can not consider fatigue effects sufficiently comes to the fore. To find the technical solution for these problems, a number of researches and discussion are continued up to now. In this study, the detailed fatigue analyses using the 3 dimensional modeling for the fatigue-weakened components were performed to develop the optimized fatigue analysis procedure and their results are compared with other reference solutions.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• 한국유체기계학회 논문집
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• 제8권3호
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• pp.16-25
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• 2005

The purpose of this study is localization of safety relief valves for Nuclear Service. The safety relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. We developed design technology used FEM ' CFM about safety relief valve for Nuclear Service according to ASME (or KEPIC) Code and KHNP's Technical Specification. To prove validity of a design technology, actually, we manufactured and inspected and tested the sample products designed according to a developed technology. The capacity qualification test was achieved according to requirement of ASME(or KEPIC) Code by NBBI and the functional qualification test was achieved according to ASME QME-1 for operating condition in technical specification of KHNP by NLI. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.310-319
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• 1997

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary(B$_1$), Secondary(C$_1$) and Peak(K$_1$) stress indices for pressure, the Primary (B$_2$), Secondary(C$_2$) and Peak(K$_2$) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage.

• Journal of Welding and Joining
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• 제32권1호
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• pp.71-78
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• 2014

Although the fatigue design curve of ASME Code has enough margin with respect to alternating stress and cycles, the welding residual stress(WRS) should be included in fatigue analysis. In this paper, WRS distribution in a nozzle with dissimilar metal weldment was obtained by finite element analysis and was added in fatigue analysis. The fatigue analysis was performed by following the ASME Code including thermal and stress analysis applying with postulated 30 transient conditions. The calculated results of a cumulative fatigue usage factors(CUF) were compared for the case of the models with or without WRS effects. The results showed that the CUF at weldment and heat affected zone was affected by the WRS.