• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.04a
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• pp.301-308
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• 2004

The present study mainly focuses on the inelastic stress analysis of the 1/4 scale prestressed concrete containment vessel model(PCCV) under internal pressure and evaluates not only failure mode but also ultimate pressure capacity of the PCCV. Inelastic analysis is carried out 2D axisymmertic FE model and 3D FE model using four concrete material models which are Drucker-Prager Model, Chen-Chen Model, Damaged Plasticity Model and Menetrey-Willam Model. The uplift phenomenon of the basemat is considered in the 2D axisymmetric FE models. It is found from the 2D axisymmetric analysis results that both of Drucker-Prager model and Damaged Plasticity Model have a good performance and the uplift of the basemat is too small to influence on the global behavior of the PCCV. The FE analysis results on the ultimate pressure and failure mode have a good agreement with experimental results.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.10a
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• pp.312-319
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• 2004

This paper is mai y focused to develop new concrete material model such as ultimate failure surface in compression-compression region, hardening rule and cracking criteria which are basically used in the nonlinear finite element analysis of nuclear prestressed concrete containment building. From the Kepri's experimental results, failure surface of the concrete based on the elasto-plastic material model is modified and new cracking criteria is proposed. Nonlinear FE analysis program using a new material model is implemented to analysis plane concrete. Finally, numerical simulation to compare the performance of the new material model with experimental results is employed. The numerical results by the proposed model in this study agree very well with the experimental data.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.10
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• pp.1274-1284
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• 1999

The Regulatory Guide 1.82 recommends an analysis of hydraulic performance for sump of ECCS (Emergency Core Cooing System) when LOCA(Loss of Coolant Accident) occurs in a nuclear power plant. The present study deals with 3-dimensional, unsteady, turbulent and two-phase flow simulation to examine the behavior of mixture of reactor coolant and debris near the floor of containment building in conjunction with appropriate assumptions. The dispersed solid model has been adjusted to the interfacial momentum transfer between reactor coolant and debris. According to the results, the counterclockwiserecirculation zone had been formed in the region between sump and connection aisle about 376 second after LOCA occurs. The debris thickness accumulated on a sump screen periodically increases or decreases up to 2000 second, afterwards its peak decreases.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.493-498
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• 1996

A PC window-based computer code, CONPAS(CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree(CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree(LSET) for its detailed quantification. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, and sensitivity analysis, reporting aspects including tabling and graphic, and user-friend interface.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.2
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• pp.112-119
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• 2016

Many studies have been conducted on the performance of a passive autocatalytic recombiner (PAR), but not many have focused on the locations where the PAR is installed. During a severe accident in a nuclear reactor containment, a large amount of hydrogen gas can be produced and released into the containment, leading to hydrogen deflagration or a detonation. A PAR is a hydrogen mitigation method that is widely implemented in current and advanced light water reactors. Therefore, for this study, a PAR was installed at different locations in order to investigate the difference in hydrogen reduction rate. The results indicate that the hydrogen reduction rate of a PAR is proportional to the distance between the hydrogen induction location and the bottom wall.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2007.04a
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• pp.694-697
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• 2007

In this paper, a nonlinear finite element analysis program NUCAS, which has been developed for assessment of pressure capacity and failure mode for nuclear containment building is described. Degenerated shell element with assumed strain method and low-order solid element with enhanced assumed strain method is adapted to microscopic material and elasto-plastic material model, respectively. Finally, the performance of the developed program is tested and demonstrated with several examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.

• 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.

• Journal of the Korean Society for Marine Environment & Energy
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• v.1 no.1
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• pp.29-38
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• 1998

This paper summarizes the results of the investigations performed by the authors for a number of years on the performance of oil fences in currents and waves. The leakage inception velocity of currents, the wave-exciting motion of fences and the deformation of the fence skirt due to currents are investigated. The results show that improvement of the fence performance is necessary in order to utilize the oil fences as an effective containing device for spilt oils in ocean environments. As an approach toward improving the fence effectiveness, tandem fences in currents are investigated. It was found that an appropriate deployment of tandem fences of identical configuration could significantly improve the oil containment capability.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.825-837
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• 2017

Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

• Proceedings of the Korean Geotechical Society Conference
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• 2010.09c
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• pp.135-139
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• 2010

A construction technique to install the soil-bentonite (SB) cut-off wall for in-situ geoenvironmental containment by employing the trench cutting and re-mixing deep wall method is first presented in this paper. The laboratory test results on the hydraulic barrier performance of SB in relation to the chemical compatibility are then discussed. Hydraulic conductivity tests using flexible-wall permeameters as well as swell tests were conducted for SB specimens exposed to various types and concentrations of chemicals (calcium chloride, heavy fuel oil, ethanol, and/or seawater) in the permeant and/or in the pore water of original soil. For the SB specimens in which the pore water of original soil did not contain such chemicals and thus the sufficient bentonite hydration occurred, k values were not significantly increased even when permeated with the relatively aggressive chemical solutions such as 1.0 mol/L $CaCl_2$ or 50%-concentration ethanol solution. In contrast, the SB specimens containing $CaCl_2$ in the pore water had the higher k values. The excellent linear correlation between log k and swelling pressure implies that the swelling pressure can be a good indicator for the hydraulic barrier performance of the SB.