• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2015년도 춘계 학술논문 발표대회
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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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• 제55권5호
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• pp.1616-1629
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• 2023

To quantify the conservatism of existing ASME strain-based evaluation methods for seismic loading, this paper presents very low cycle fatigue test data of elbows under various cyclic loading conditions and comparison of evaluation results with experimental failure cycles. For strain-based evaluation methods, the method presented in ASME BPVC CC N-900 and Sec. VIII are used. Predicted failure cycles are compared with experimental failure cycle to quantify the conservatism of evaluation methods. All methods give very conservative failure cycles. The CC N-900 method is the most conservative and prediction results are only ~0.5% of experimental data. For Sec. VIII method, the use of the option using code tensile properties gives ~3% of experimental data, and the use of the material-specific reduction of area can reduce conservatism but still gives ~15% of experimental data.

• 한국구조물진단유지관리공학회 논문집
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• 제15권1호
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• pp.180-188
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• 2011

최근 대구경철근에 적용할 수 있는 기계적 철근이음장치에 대한 연구가 진행되고 있다. 본 논문에서는 대구경 기계적 철근이음재에 대한 구조적 성능평가에 대하여 연구가 수행되었다. 원자력발전소에 대구경의 기계적 철근이음장치를 적용하기 위해서 2가지 형태의 철근이음장치에 대한 실험이 수행되었으며, 원자력발전소에 적용되는 11번과 14번 및 18번에 대한 대구경 철근 이음 장치가 조립되어 정적 및 동적실험이 수행되었다. 실험은 ASME SEC III DIV.2 CC-4330에 따라 이루어졌다.

• 한국전산구조공학회논문집
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• 제28권5호
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• pp.535-542
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• 2015

원자로 격납건물은 냉각재상실사고와 같이 내부의 과도한 압력이 유발되는 사고에 있어서도 방사성 물질이 외부로 누출되지 않도록 막는 최종의 방벽이다. 이러한 격납건물의 기능적 중요성에 기인하여, 건설 초기 구조건전성시험(SIT)을 수행한다. 이러한 SIT거동을 가장 실제와 가깝게 예측하기 위한 해석 연구를 수행하였다. 해당 연구의 결과는 2편의 논문으로 정리되었는데, 본 논문은 그 중 II편으로 I편의 해석모델 구성 시의 주요 고려사항의 분석 및 예비해석 결과를 반영한 상세 해석 모델의 구성 과정 및 해석 결과를 제시하고 있다. 특히 비부착식 텐던으로 시공된 구조물에서 덕트관에 의한 강성 저감효과 및 덕트관을 사이에 둔 텐던과 콘크리트간의 밀착 여부에 따른 영향을 해석 시 최대한 고려하고자 하였다. 이러한 과정을 통해 구축된 해석 모델에 따른 변위과 신고리 3호기 SIT 측정변위를 비교한 결과, ASME CC-6000 기준을 충분히 만족시키는 결과가 나타남을 확인하였다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2092-2099
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• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.