• 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.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3130-3139
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• 2022

Direct containment heating (DCH) is one of the potential factors leading to early containment failure. DCH is closely related to safety analysis and containment performance evaluation of nuclear power plants. In this study, a DCH prediction program was developed to analyze the DCH loads of containment vessel. The phenomenological model of debris dispersal, metal oxidation reaction, debris-atmospheric heat transfer and hydrogen jet burn was established. Code assessment was performed by comparing with several separate effect tests and integral effect tests. The comparison between the predicted results and experimental data shows that the program can predict the key parameters such as peak pressure, temperature, and hydrogen production in containment well, and for most comparisons the relative errors can be maintained within 20%. Among them, the prediction uncertainty of hydrogen production is slightly larger. The analysis shows that the main sources of the error are the difference of time scale and the oxidation of cavity debris.

• International Journal of Naval Architecture and Ocean Engineering
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• v.8 no.6
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• pp.537-553
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• 2016

This paper addresses the safety of two-row tank design by performing the extensive sloshing model tests. Owing to the uncertainties entangled with the scale law transforming the measured impact pressure up to the full scale one, so called comparative approach was taken to derive the design sloshing load. The target design vessel was chosen as 230 K LNG-FPSO with tow-row tank arrangement and the reference vessel as 138 K conventional LNG carrier, which has past track record without any significant failure due to sloshing loads. Starting with the site-specific metocean data, ship motion analysis was carried out with 3D diffraction-radiation program, then the obtained ship motion data was used as 6DOF tank excitation for subsequent sloshing model test and analysis. The statistical analysis was carried out with obtained peak data and the long-term sloshing load was determined out of it. It was concluded that the normalized sloshing impact pressure on 230 K LNG-FPSO with two-row tank arrangement is higher than that of convectional LNG carrier, hence requires the use of reinforced cargo containment system for the sake of failure-free operation without filling limitation.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1650-1660
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• 2020

Comparing with the large containment, the gas can not flow freely within the local compartment due to the small volume of the compartment in case of serious accident, which affects the hydrogen flow distribution, and it will determines the location where high concentration occurs in compartment. In this paper, hydrogen distribution and possible hydrogen risk in the vessel under the different conditions are investigated. The results show that when the initial gas momentum is increased, the ability of gas enters into the upper region of the vessel will be strengthened, and the hydrogen volume fraction in the upper region of the vessel is higher. Comparing with horizontal source direction, when source direction is vertically towards upper space, hydrogen is more likely to accumulate in the upper region of the vessel. With the increasing of steam mass flow, the dilution effect of steam on the hydrogen volume fraction will be strengthened, while the pressure in the vessel is also increased. When steam flow is decreased, the hydrogen explosion risk is higher in the vessel. The experiment data can provide technical support for the validation of the CFD software and the mitigation of hydrogen risk in the containment compartment.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.485-497
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• 2024

Ring beam is the main anchorage zone of the tendons in the nuclear power prestressed concrete containment vessel (PCCV). Its safety is crucial and has a great influence on the overall performance of PCCV. In this paper, two half-scale ring beams were tested to investigate the mechanical performance of the anchorage zone in the PCCV under multidirectional pressure. The effect of working condition with different tension sequences was investigated. Additionally, a half axisymmetric plane model of the containment was established by the finite element simulation to further predict the experimental responses and propose the local reinforcement design in the anchorage zone of the ring beam. The results showed that the ultimate load of the specimens under both working conditions was greater than the nominal ultimate tensile force. The original reinforcement design could meet the bearing capacity requirements, but there was still room for optimization. The ring beam was generally under pressure in the anchorage area, while the splitting force appeared in the under-anchor area, and the spalling force appeared in the corner area of the tooth block, which could be targeted for local strengthening design.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.290-298
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• 1996

Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3493-3505
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• 2023

Reactor containment buildings serve as the last barrier to prevent radioactive leakage due to accidents and their safety is crucial in overpressurization conditions. Thus, the Regulatory Guide (RG) 1.216 has mentioned the global strain as one of failure criteria in the free-field for cylindrical prestressed concrete containment vessels (PCCV) subject to internal pressure. However, there is a limit that RG 1.216 shows the free-field without the specific locations of failure criteria and also the global strain corresponding to only azimuth 135° has been mentioned in NUREG/CR-6685, regardless of the elevations of the structure. Therefore, in order to reevaluate the failure criteria of the 1:4 scaled PCCV, the high fidelity simulation model based on the experimental test was significantly validated in this study, and it was interesting to find that the experimental and numerical result was very close to each other. In addition, for the consideration of the material uncertainties, the Latin hypercube method was used as a statistical approach. Consequently, it was revealed that the radial displacements of various azimuth area such as 120°, 135°, 150°, 180° and 210° at elevations 4680 mm and 6,200 mm can represent as the global deformation at the free-field, obtained from the statistical approach.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2006.04a
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• pp.639-646
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• 2006

A nonlinear finite element analysis is carried out to predict the ultimate internal pressure and failure mechanism of a 1/4 scale prestressed concrete containment vessel(PCCV) model using the commercial code ABAQUS. Therefore, this paper is mainly focused to compare the influence of concrete material model, tension stiffening parameter, uplift phenomenon and basemat. From the analysis results, nonlinear behavior of the PCCV showed a substantially different aspects in accordance with the nonlinear material model for the concrete as well as tension stiffening parameter. The boundary conditions beneath the basemat are considered to be a fixed condition and a nonlinear spring element to compare the influence of the uplift. The finite element analysis is considered with and without a basemat to find out the influence of the basemant itself. From the analysis results, the nonlinear behavior of the PCCV is entirely similar for the two cases.