• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2015.05a
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• pp.9-10
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• 2015

Generally significant reinforcement is used in nuclear power plant structures and may cause potential problems when concrete is poured. In particular pouring concrete into structural member joint area is more difficult than other areas since the joint area is very congested due to the crossed bars and the embedded plates, The purpose of this study is to solve these problems by applying Gr.80(550MPa) shear bars to containment structures of nuclear power plant. In order to apply them to containment structures, it is necessary to change ASME-CC code (ASME Sec.III Div.2). The structural performance tests of wall & beam have been done to compare Gr.80(550Mpa) with Gr.60(420MPa) shear bars. The test results and code change proposal were presented to ASME-CC Committee last year and the discussion for code change will be expected to proceed in the near future.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.185-196
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• 1996

The purpose of this paper is to modify the ASME Code Z-Factor, which is used in the evaluation of circumferential surface crack in nuclear ferritic pipings. The ASME Code Z-Factor is a load multiplier to compensate plastic load with elasto-plastic load. The current ASME Code Z-Factor underestimates pipe maximum load. In this study, the original SC. TNP method is modified first because the original SC. TNP method has a problem that the maximum allowable load predicted from the original SC. TNP method is slightly higher than that measured from the experiment. Then the new Z-Factor is developed using the modified SC. TNP method. The desirability of both the modified SC. TNP method and the new Z-Factor is examined using the experimental results for the circumferential surface crack in pipings. The results show that (1) the modified SC. TNP method is good for predicting the circumferential surface crack behavior in pipings, and (2) the Z-Factor obtained from the modified SC. TNP method well predicts the behavior of circumferential surface crack in ferritic pipings.

• Journal of Welding and Joining
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• v.13 no.1
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• pp.51-61
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• 1995

품질보증 활동은 제품의 품질등급을 그 규격기준에 따라 규정하고, 발주자가 지정하는 품질의 제품을 실현하는 것이지만, 압력용기 등의 특정 제품에 대해서는 공적인 제3자가 기관에 의한 검사 인증제도가 요구되는 경우가 있다. 미국에서는 압력용기 등의 규격체계는 민간 Base의 ASME Code를 중심으로 하고, 운용에 있어서의 품질보증제도는 ASME Stamp에 의한 제작공장 인증과 적검사기관을 대행하는 보험회사와의 협조에 의한 개별제품인증이 행해진다. 소위 체제 지향의 품질 관리방식을 채택하고 있다. 미국 외의 구라파제국의 이와는 상대적으로 제품지향의 제품질관리를 하였다고 할 수가 있다. 그러나 87년에 제정되어 세계적으로 확산되고 있는 ISO-9000 시리즈는 미국식의 체제지향의 품질관리로서 양자간 접근방식이 대동소이하다. 그러므로 ASME Code의 품질관리에 대하여 그 본질과 구조 및 활동 내역 등을 고찰해 보는 것은 품질관리의 시대적 조류 파악에 도움이 될 것이다.

• Corrosion Science and Technology
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• v.11 no.6
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• pp.225-231
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• 2012

Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers).

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.6
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• pp.930-937
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• 2003

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori #1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts fur man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME Code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME Code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety fer helical-coil method which is used according to Code Case-N-496-1. From analysis results, we found .that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10％ by manufacture integrity criteria and 2／3 yield strength criteria on ASME Code. It was also confirmed that the helical-coil repair method would be safe.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.17 no.4
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• pp.150-159
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• 2018

The present research deals with a finite element analysis and fatigue evaluation of a steam separator of a high-pressure evaporator for the Heat Recovery Steam Generator (HRSG). The fatigue during the expected life of the HRSG was evaluated according to the ASME Boiler and Pressure Vessel Code Section VIII Division 2 (ASME Code). First, based on the eight transient operating conditions prescribed for the HRSG, temperature distribution of the steam separator was analyzed by a transient thermal analysis. Results of the thermal analysis were used as a thermal load for the structural analysis and used to determine the mean cycle temperature. Next, a structural analysis for the transient conditions was carried out with the thermal load, steam pressure, and nozzle load. The maximum stress location was found to be the riser nozzle bore, and hence fatigue was evaluated at that location, as per ASME Code. As a result, the cumulative usage factor was calculated as 0.00072 (much less than 1). In conclusion, the steam separator was found to be safe from fatigue failure during the expected life.

• Journal of the KSME
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• v.33 no.8
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• pp.746-751
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• 1993

발전로 계통설계와 ASME Code의 적용과의 상호연관성을 이해하기 위하여, 관련되는 내용으 로서, 계통설계의 개요, 원자력발전소의 기기 및 구조물의 등급분류 및 실례를 개괄적으로 살펴 보았다. 앞에서 본 바와 같이 계통설계 결과로서 부여되는 기기 및 구조물의 등급은 바로 적용 ASME Code 또는 기타 산업기술기준을 결정하는 근거가 되는데, 이것은 원자력 규제법규나 지침 등에 근거하고 있음을 알 수 있었다. 이와 같은 내용에 대한 정확한 이해는 각 기기나 구조물의 설계에 꼭 필요함은 물론, 원자력발전소의 인허가를 위시한 규제업무를 충분한 안전성을 확보 하면서, 합리적으로 수행하는데 매우 유용하리라고 판단된다. 뿐만 아니라 앞으로 국내에서 독 자적인 기술기준을 마련하여 적용하고자 할 때 이러한 사항들이 일관성 있고 균형있게 반영되 어야 할 것으로 생각된다.

• Journal of the Korean Society for Nondestructive Testing
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• v.4 no.2
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• pp.20-29
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• 1985

There are two basic concepts in industrial radiographic examination; one is a radio-graphic sensitivity, and the other is a acceptance criteria. The comparison of these main points are studied for KS Standard and ASME Boiler and Pressure Vessel Code. From the results of the experiment, higher radiographic sensitivity is required in KS Standard when the thickness of material to be examined is less than 20mm in single wall technique. The acceptance criteria for linear type indications are described on same concept in two standards, whereas the acceptance criteria for rounded indications of KS Standard which mainly depends upon the object thickness are more severe than those of ASME BPV Code.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.418-426
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• 1995

설계의 신뢰성은 응력해석을 통하여 확인될 수 있으며, 해석결과는 대상 부품의 구조적 건전성을 입증하는 근거가 된다. 본 보고서는 ANSYS의 피로해석 모듈을 이용한 CANDU 6핵연료채널의 응력해석 및 ASME Code에 따른 해석 절차 개발을 소개하였다. 응력해석은 ASME Code Section III NB-3200 의 $\ulcorner$Design by Analysis$\lrcorner$에 기초한 해석절차에 따라 수행하였으며, 체계적인 해석을 위해 자료 처리용 ANSYS 매크로 및 FORTRAN 프로그램을 개발하였다. 해석은 각 조건에 따라 기계적응력과 열응력해석으로 분리하여 수행한 후 조합되었으며, ANSYS 피로해석 모듈을 이용하여 선정된 절점들의 기계적응력과 열응력의 합에 대한 최대응력강도범위를 계산하였다. 응력해석 결과, CANDU 6 핵연료채널의 구조적 건전성이 입증되었으며, ANSYS를 이용한 ASME Code해석절차가 확립되어 CANDU 원자로 해석의 신뢰성을 크게 향상 시켰음은 물론 독자적인 수행을 위한 발판을 마련하였다.