• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

• Journal of computational fluids engineering
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• v.10 no.3 s.30
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.70-77
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• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.413-423
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• 1993

This paper reports an experimental study of direct containment heating (DCH) which would occur if the primary system pressure is still high at the time of vessel breach during a light water reactor core melt accident. The experiments were conducted in 1/30-scale cavity models of Kori unit 1 and 2 and Young Kwang unit 3 and 4 nuclear power plants. One 1/20-scale model of the Kori plant was also used to investigate the scaling effect. The primary variables in the experiments were initial vessel pressure, vessel breach size and cavity geometry. It is observed that higher initial pressure and larger breach size enhance the melt dispersal fraction. Also, the cavity geometry appears to affect the dispersal rate greatly. A simple correlation of melt dispersal fraction is proposed in terms of nondimensional effective period. This correlation shows good agreement with the present experimental data, the KAIST data and the BNL data.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.611-623
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• 2024

The ISP-47 TOSQAN experiment was analyzed with containmentFOAM which is an open-source CFD code based on OpenFOAM. The containment phenomena taking place during the experiment are gas mixing, stratification and wall condensation in a mixture composed of steam and non-condensable gas. The k-ω SST turbulence model was adopted with buoyancy turbulence models. The wall condensation model used is based on the diffusion layer approach. We have simulated the full TOSQAN experiment which had a duration 20000 s. Sensitivity studies were conducted for the buoyancy turbulence models with SGDH and GGDH and there were not significant differences. All the main features of the experiments namely pressure history, temperature, velocity and gas species evolution were well predicted by containemntFOAM. The simulation results confirmed the formation of two large flow stream circulations and a mixing zone resulting by the combined effects of the condensation flow and natural convection flow. It was found that the natural convection in lower region of the vessel devotes to maintain two large circulations and to be varied the height of the mixing zone as result of sensitivity analysis of non-condensing wall temperature. The computational results obtained with the 2D mesh grid approach were comparable to the experimental results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.3
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• pp.1-6
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• 2012

The containment Integrated Leakage Rate Testing(ILRT) of nuclear power plants in Korea is performed in accordance with NSSC(Nuclear Safety and Security Commission) code 2012-16 and ANSI/ANS 56.8-1994. Nuclear power plants in Korea and the United States are to apply same test criteria, ANSI/ANS 56.8-1994, except type A testing time. NPPs in Korea apply 24 hours according to NSSC code 2012-16, but NPPs in United States apply 8 hours according to 10CFR50 App. J for type A test. So, there are many difficulties in order to perform ILRT in Korea. In this study, I review the impact on the ILRT results and the effect of ILRT due to type A testing time. The future, we will continue study to enhance the test reliability and improve these problems.

• Computers and Concrete
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• v.14 no.2
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• pp.127-143
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• 2014

LOCA (Loss Of Coolant Accident) is one of the most important utmost accidents for Prestressed Concrete Containment Vessel (PCCV) due to its coupled effect of high temperature and inner pressure. In this paper, heat conduction analysis is used to obtain the LOCA temperature distribution of PCCV. Then the elastic internal force of PCCV under LOCA temperature is analyzed by using both simplified theoretical method and FEM (finite element methods) method. Considering the coupled effect of LOCA temperature, a nonlinear elasto-plasitic analysis is conducted for PCCV under utmost internal pressure considering three failure criteria. Results show that the LOCA temperature distribution is strongly nonlinear along the shell thickness at the early time; the moment result of simplified analysis is well coincident with the one of numerical analysis at weak constraint area; while in the strong constrained area, the value of moments and membrane forces fluctuate dramatically; the simplified and numerical analysis both show that the maximum moment occurs at 6hrs after LOCA.; the strain of PCCV under LOCA temperature is larger than the one of no temperature under elasto-plastic analysis; the LOCA temperature of 6hrs has the greatest influence on the ultimate bearing capacity with 8.43% decrease for failure criteria 1 and 2.65% decrease for failure criteria 3.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.