• Journal of the Korean Society for Nondestructive Testing
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• v.7 no.1
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• pp.51-53
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• 1987

• Proceedings of the KWS Conference
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• 1999.05a
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• pp.141-142
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• 1999

• Journal of the Korean Society of Safety
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• v.22 no.1 s.79
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• pp.13-18
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• 2007

Structural integrity assessment of thin-walled pipes and pipe items has become one of the major issues in the nuclear power plant. ASME Section XI Code Case N-597-2 provides a criterion for acceptance of the pipes. But the code case has several limitations for application and sometimes gives too conservative or non-conservative results. So it is necessary to understand fully the technical bases of the code case. In the code case N-597, the allowable local thicknesses of thinned straight pipes are given for three different cases. Because of the different technical base, each case gives different thickness values and sometimes gives contradictory values. In this paper attempts were made in order to propose a unified rule for the allowable local thickness and in order to remove or relax the restrictions on the application of the code case. For this purpose elastic stress analyses were made using the finite element method and the stress results were examined. Based on the obtained bending stress results, a very simple procedure was proposed to obtain the consistent allowable local thickness for the thinned straight pipes.

• Journal of the Korean Institute of Gas
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• v.18 no.6
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• pp.14-20
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• 2014

In this study, the stress analysis for the piping support design is performed as per the rules of the ASME Code, Section III, subsection NF-Component Support which provides a simplified method of design analysis for piping support. This method makes use of simple equations and conservative allowable stress limits for design and service loadings. For the base plate, code equation is satisfied within the allowable limits. Both anchor bolts and pipe strap are governed by the their interaction equations. The stresses resulting from various loadings and their combinations are within the allowable limits specified in the above mentioned ASME Code. Thus, it was proved that the structural integrity of the pump assembly was satisfactory.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.4
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• pp.313-320
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• 1998

To record and evaluate the flaws which were found during pre-service/in-service inspection performance of nuclear power plants in Korea, the center line beam method described in ASME code and 6 dB drop method stated in CSA code were used. The measured through wall dimensions and lengths by these methods were compared and analyzed in this report. With the measured data analysis, the ekact understanding and use of these methods improves the reliability of flaw sizing and assures the integrity of nuclear power plant components.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.114.2-114.2
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• 2005

• Journal of the Korean Institute of Gas
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• v.7 no.3 s.20
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• pp.58-64
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• 2003

This paper presents design safety analysis of pressure vessels. The gas pressure and thermal loads are applied to the pressure vessel simultaneously. In this study, ASME Sec. VIII Div. 2 code was accepted for the safety design of high-pressure vessel. And this result was analyzed using a coupled thermal-mechanical FEM analysis technique. The FEM computed result shows that ASME design code may not guarantee for combined loads of high gas pressure and thermal loads. And solid pressure vessel may be safe compared to other pressure vessels with supporting rings round the cylinder body.

• Nuclear industry
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• v.4 no.2 s.18
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• pp.49-55
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• 1984

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1997.04a
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• pp.114-120
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• 1997

원자력공사에 관련된 장치와 부속 기기를 설계 및 제작 또는 설치하려면 ASME에서 인증하는 자격이 필요하고, 자격을 취득하거나 갱신하기 위하여 시설, 품질보증 프로그램, 설계 및 제작능력등을 ASME 위원회의 실사를 받아야 한다. 이를 위한 일련의 과정 중에서 설계능력을 증명할 수 있는 설계해석 보고서를 제출하여야 하며 이를 위하여 현대중공업이 설계한 Chiler Surge Vessel의 설계조건 및 하중 및 하중조합에 대하여 Design Report를 작성하였으며, ASME Code의 요구조건을 모두 만족하는 것으로 평가되었다.

• Journal of the Korean Society for Nondestructive Testing
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• v.8 no.1
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• pp.6-11
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• 1988

The comparison and analysis of flaw detectability, flaw sizing methods and allowable criteria in accordance with the applicable codes, ASME and RCC-M Code, in ultrasonic examination of pressure vessel in nuclear power plants were studied. Accordding to the codes, calibration blocks were made. The artificial flaws such as disc and band types in test specimens were machined. They were detected and evaluated with etch code requirements and measured values were compared with the actual flaws.