• 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.

• Computers and Concrete
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• v.13 no.1
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• pp.1-16
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• 2014

To evaluate the behavior of the advanced unbonded pre-stressed concrete containment vessel (UPCCV) for one typical China nuclear power plant under Japan's March 11 earthquake, five nonlinear time history analysis and a nonlinear static analysis of a 1:10 scale UPCCV structure have been carried out with MSC.MARC finite element program. Comparisons between the analytical and experimental results demonstrated that the developed finite element model can predict the earthquake behavior of the UPCCV with fair accuracy. The responses of the 1:10 scale UPCCV subjected to the 11 March 2011 Japan earthquakes recorded at the MYG003 station with the peak ground acceleration (PGA) of 781 gal and at the MYG013 station with the PGA of 982 gal were predicted by the dynamic analysis. Finally, a static analysis was performed to seek the ultimate load carrying capacity for the 1:10 scale UPCCV.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.512-527
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• 2004

This paper describes the finite element (FE) analysis results of a 1/4 scale model of a prestressed concrete containment vessel (PCCV) by considering the tension stiffening effect, which is a result of the bond effect between the concrete and the steel. The tension stiffening model is assumed to be an exponential form based on the relationship between the average stress and the average strain of the concrete. The objective of the present FE analysis is to evaluate the ultimate internal pressure capacity of the PCCV, as well as its failure mechanism, when the PCCV model is subjected to a monotonous internal pressure beyond is design pressure capacity. With the commercial code ABAQUS, the FE analysis used two concrete failure criteria: a 2-dimensional axi-symmetric model with modified Drucker-Prager failure criteria and a 3-dimensional model with a damaged plasticity mod디. The results of our FE analysis on the ultimate pressure capacity and failure modes of PCCV have a good agreement with the experimental data.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

• 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.

• Journal of the Korean Society of Safety
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• v.11 no.2
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• pp.79-88
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• 1996

A hypothetical accident in the containment vessel of liquid metal reactor could cause a pressure, temperature rise, and a strong aerosol release. The computer codes relating to the modelization of these accident make it necessary to use various input parameter, among which is the dynamic shape factor of aerosols produced. Combustion experiments of sodium spray fire carried out in a closed vessel, which was vertical cylinder made of 1.2m in diameter and 1.8m hight with a volume of 1.7$m^3$. The results of theoretical analysis presented here was compared to data obtained from experiments. The experimental results were summarized as follows. 1) The aerodynamic diameter and geometric diameter of aerosols are decreasing with increasing of injection pressure and injection temperature of sodium 2) The dynamic shape factor of aerosol is proportional to the aerodynamic diameter for a given particle. 3) The correspondence between the aerodynamic diameter and geometric diameter can be as $D_{ae}=0.70 D_{ge}$. 4) Peak pressure rose with increase in pressure and temperature of injection sodium, being more sensitive to the injection pressure than the injection temperature.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.177-185
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• 2022

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.

• 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.