• 기계저널
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• 제33권8호
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• pp.708-716
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• 1993

기술기준의 개발은 원자로형의 공급국가에 따라 다양한 형태의 기술기준 적용으로 파생되는 국 산화 및 기술자립지연 등 문제를 해결하기 위하여 개발하고 있다. 일반기계기기는 전력생산과 직접 관련되면서도 ASME Sec. III의 적용을 받는 원자력기계기기와는 달리 구별되나 화력발전 용과는 사용조건이 다를 뿐 기능과 역할이 동일하다. 따라서 국내실정에 적합하면서 우수한 기 술성과 안전성을 확보할 수 있도록 일반기계기기의 기술기준을 개발하여 향후 국내 주도로 건 설되는 원전후속기 및 화력발전소에 점진적으로 활용함으로써 원전산업뿐만 아니라 화력발전산 업의 활성화에도 크게 기여해야 할 것으로 생각된다. 또한, 보일러, 터빈발전기 및 전기집진기등 금번에 기술기준개발 범위에서 제외된 기기의 기술기준도 개발하여 일반기계기기로 분류된 기 기의 실질적인 기술자립을 꾀하고 금번기술기준 제정 한 번으로 끝날 것이 아니라 기술기준의 유지관리를 위한 전담기구를 설립하여 기술기준을 지속적인 보완과 확대로 기술자립을 앞당 겨야 할 것으로 사료된다.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2004년도 추계학술발표대회 개요집
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• pp.174-176
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• 2004

The duplex stainless clad vessel with 38m & over thickness shall be performed to PWHT based on the ASME code. In this case, it is well-known that precipitators such as carbides and sigma($\sigma$) phase are formed at gram boundary between ferrite and austenite phase. Therefore, a weld test for simulating this situation has been planned and performed by 3309LMo71-1 for barrier layer and E2209Tl-1 for 2nd & over layer and then carried out to investigate the overlaid weld metal. Based on the test results, it could be concluded that PWHT should be carried out after the completion of 1st(barrier) layer and then 2 & over layer should be applied.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1151-1157
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• 2009

Nuclear power plants applying for the continued operation over design life are required to address the effects of reactor water environment in fatigue design requirement of the ASME Code. Reactor water environmental effects are generally evaluated by calculating fatigue correction factors on fatigue usage. This paper describes the application for pressurizer surge line of environmental fatigue correction factors and the strain rate impact in the application. From this paper, the environmental fatigue correction factors resulted from the assumption of a step change in temperature are especially compared with those calculated from the data measured during plant startup. As a conclusion of this paper, the design transient conditions applied to the fatigue design may be conservative in case of the environmental fatigue evaluation.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2001년도 추계학술대회(한국공작기계학회)
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• pp.275-279
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• 2001

There are two kinds of the design criteria of piping system in petrochemical plant design. The first is on static state evaluation by thermal growth and the other is on dynamic evaluation by piping vibration. In the static design evaluation, the ASME B31.3 code defines 7000 cycles of fatigue life in operating the piping system with design condition. However, the dynamic design evaluation in comparative with small displacements of high frequencies to static condition has not established clearly the method, yet. So, this study purposes to present the trial of a proposal of dynamic design criterion on the basis of static design method.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.30-40
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• 2016

The paper investigates the previous studies about welding residual stresses in nuclear components. First, various residual stress measurement methods are reviewed in applicability. Second a finite element welding residual stress analysis technique, which was developed from the viewpoint of FFS (Fitness-For-Service) assessment, is explained. Third, characteristics of the welding residual stresses on J-groove welds and butt welds were presented via investigating the previous studies. Last, engineering formulae for residual stresses in the FFS assessment codes such as R6 and API 579/ASME FFS-1 Code is summarized.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.107-114
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• 2016

Four cylindrically shaped IHXs(Intermediate Heat Exchangers) are installed in the PHTS(Primary Heat Transfer System) of the PGSFR(Prototype Gen IV Sodium cooled Fast Reactor). As for the IHX, the temperature difference of structure is inevitable result caused by heat transfer between primary coolant sodium and IHTS(Intermediate Heat Transport System) sodium. It is necessary to evaluate the high temperature structural integrity of IHXs which operate at the elevated temperature condition over the creep temperature. In this paper, the high temperature structural integrity of IHX under assumed loading conditions has been reviewed according to ASME code.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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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.

• 한국해양공학회지
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• 제9권1호
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• pp.83-91
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• 1995

Surface defects in structural members are apt to be origins of fatigue cracks growth, which may cause serious failure of whole structures. Most structure has a part where stress concentrates such as welded joints, corner parts, etc. And then, analysis on crack growth and penetration from these defects, therefore, is one of the most important subjects for the reliability of LBB design. The present paper has performed an experimental and analysis on the fatigue crack propagation by variety in crack length of surface cracked plate with stress concentration part. The crack growth behavior can be explained quantitatively by using Newman-Raju equation and the stress partitioning method proposed by ASME B&P Code Sec. XI. The stress concentration factor $K_t$ has affected on the crack growth. The crack growth after penetration depends upon the initial front side crack length.

• 한국해양공학회지
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• 제16권6호
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• pp.65-70
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• 2002

This paper presents a mechanical design of the deepwater pipeline wall thickness using the recent design rules. Characteristics and limitations of the new codes were identified through a case study design in the Gulf of Mexico. In addition to the ASME, API, and DVD codes, the code of federal regulations (CFR) was also utilized in the design. It was found that conservatism still exists within the collapse prediction for water depth greater than 1500m. Comparision of the results from DNV and API codes were presented.