• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 봄 학술발표회 논문집
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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.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2006년도 정기 학술대회 논문집
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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.

• Nuclear Engineering and Technology
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• 제28권6호
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• pp.551-562
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• 1996

This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe accidents in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3&4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3&4, and recommends additional instrument that m8y prove to be of vital importance in managing the accident.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1135-1142
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• 2009

The activities related to this paper are to investigate the influence of nodalisation on simulating atmospheric stratification in the THAI experiment TH13 (ISP-47) with the German containment code COCOSYS. This article focuses on different nodalisations of the vessel dome, where an atmospheric stratification occurred due to a high helium content. The volume of the dome was divided into several levels that were varied horizontally into different geometries. These geometries differ in the number of zones as well as in the existence of zones that enable the direct rise of an ascending steam plume into the vessel dome. Additionally, the vertical subdivision of the vessel dome was increased to simulate density gradients in a more detailed way. It was pointed out that the proper simulation of atmospheric stratifications and their dissolution depends on both a suitable horizontal as well as vertical nodalisation scheme. Besides, the treatment of fog droplets has an influence if their settlement is not simulated correctly. This report gives an overview of the gained experience and provides nodalisation requirements to simulate atmospheric stratifications and their proper dissolution.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.77-86
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• 2008

This paper describes the nonlinear analyses of a 1:4 scale model of a prestressed concrete containment vessel (PCCV) using an axisymmetric model and a three-dimensional model. These two models are refined by comparison of the analysis results and with testing results. This paper is especially focused on the analysis of behavior under pressure and the temperature effects revealed using an axisymmetric model. The temperature-dependent degradation properties of concrete and steel are considered. Both geometric and material nonlinearities, including thermal effects, are also addressed in the analyses. The Menetrey and Willam (1995) concrete constitutive model with non-associated flow potential is adopted for this study. This study includes the results of the predicted thermal and mechanical behaviors of the PCCV subject to high temperature loading and internal pressure at the same time. To find the effect of high temperature accident conditions on the ultimate capacity of the liner plate, reinforcement, prestressing tendon and concrete, two kinds of analyses are performed: one for pressure only and the other for pressure with temperature. The results from the test on pressurization, analysis for pressure only, and analyses considering pressure with temperatures are compared with one another. The analysis results show that the temperature directly affects the behavior of the liner plate, but has little impact on the ultimate pressure capacity of the PCCV.

• Journal of Radiation Protection and Research
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• 제28권4호
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• pp.291-298
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• 2003

The KN-12 transport cask has been designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U) package. The containment boundary of the cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the cask are satisfied by maintaining a maximum air reference leak rate of $2.7{\times}10^{-4}ref.cm^3s^{-1}$ or a helium leak rate of $3.3{\times}10^{-4}cm^3s^{-1}$ for normal transport conditions and for hypothetical accident conditions.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.797-802
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• 2014

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

• Nuclear Engineering and Technology
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• 제56권2호
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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.

• Journal of Mechanical Science and Technology
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• 제18권6호
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• pp.990-1001
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• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Nuclear Engineering and Technology
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• 제55권7호
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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.