• 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.55 no.5
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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.

• Journal of the Korea institute for structural maintenance and inspection
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• v.15 no.1
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• pp.180-188
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• 2011

Recently a number of researches about mechanical splice have been studied to apply on a large diameter reinforcing steel bars of spliced sleeves. In this study the structural performance of large diameter reinforcing bars with spliced sleeves was evaluated. For the application of nuclear power plant structures, two different types of existing splices with #11, 14, 18 rebars were fabricated and static and dynamic test were performed on the basis of ASME SEC III DIV.2CC-4330.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.535-542
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. In this paper, numerical analyses are presented, which simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. A sophisticate structural analysis model is developed to simulate the structural behavior during the SIT properly based on various preliminary analysis results considering contact condition among structural elements. From the comparison of the analysis and test results based on the acceptance criteria of ASME CC-6000, it can be concluded that the construction quality of the containment has been well maintained and the acceptable performance of new design features has been verified.

• Nuclear Engineering and Technology
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• v.52 no.9
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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.