• Proceedings of the Korea Concrete Institute Conference
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• 1998.04a
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• pp.193-198
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• 1998

In this study, the change of concrete temperature, strain and thermal stress were measured by using the embedded type concrete gauges in tendon gallery wall of containment building. A finite element analysis was performed to clarify the thermal behavior of concrete. The analytic and test results were investigated to improve the validity of analytic method. According to the test results, concrete temperature, strain and thermal stress were strongly affected by measuring point and environment condition of member. And the thermal stress was developed in the member which was not demoulded at early ages. This is caused by the change of internal temperature and restrained condition. A finite element effectively interpreted the test results by estimating the concrete properties and the site condition.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1383-1384
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.725-726
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• 2005

• Journal of the Korean Institute of Gas
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• v.12 no.1
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• pp.36-41
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• 2008

Using the finite element analysis, this paper presents the strength safety of a side wall of an outer tank and a roof structures in a full containment LNG storage tank system. The outer tank structure in which is constructed with a prestressed concrete is forced by internal hydrostatic and hydrodynamic pressures of a leaked LNG and an external wind pressure including a typhoon one. The FEM computed results show that the ring beam between a side wall of an outer tank and a roof structure supports most of the internal and the external loads. This means that the design point of the outer tank system is a ring beam structure and the other one is a center part of the roof structure. In this FE analysis model of a full containment LNG tank system, the outer tank and the roof structures are safe for the given combined loads such as an internal leaked LNG pressure and an external typhoon pressure.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.23 no.4
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• pp.445-452
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• 2010

In this paper an assessment of the internal pressure fragility of the CANDU type containment buildings is performed. The uncertainties of the performance of the containment buildings, material properties and tendon characteristics are referred from the in-service reports of Wolsung Unit 1. The containment buildings are modeled as a three-dimensional finite elements with considering the major opening and penetrations. A new method to evaluate the probabilistic fragility of the massive structural system is developed. The fragility curves of the target containment building are presented with repect to the failure modes and reliability levels. The center of wall is reveled as the most weak structural component of the containment building in the sense of the rupture and catastrophic rupture failure modes.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.756-765
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• 2015

Background: Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. Methods: The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. Results: The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ~40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. Conclusion: The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.151-158
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• 2005

For the safety analysis of large structures such as nuclear containment buildings, we conventionally prefer to use analytical approach using finite element method rather than empirical test. Therefor, this paper is mainly focused to develop low-order solid finite element model with the elasto-plastic material model for the safety analysis of nuclear containment building. Drucker-Prager failure criteria in uncracked concrete and maximum tensile stress criteria in cracked concrete are used to model the constitutive behavior of concrete. The concrete material model takes into account the aspects of tensile strain, compression strength reduction of concrete and shear transfer to improve the accuracy of the finite element analysis. Finally, numerical simulation to compare the performance of the developed model with experimental results is employed. The numerical results in this study agree very well with the experimental data.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.211-218
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• 2010

An analysis model on debris transport in the containment floor of pressurized water reactors is developed in which the flow field is calculated by Eulerian conservation equations of mass and momentum and the debris particles are traced by Lagrange equations of motion using the pre-determined flow field data. For the flow field calculation, two-dimensional Shallow Water Equations derived from Navier Stokes equations are solved using the Finite Volume Method, and the Harten-Lax-van Leer scheme is used for accuracy to capture the dry-to-wet interface. For the debris tracing, a simplified two-dimensional Lagrangian particle tracking model including drag force is developed. Advanced schemes to find the positions of particles over the containment floor and to determine the position of particles reflected from the solid wall are implemented. The present model is applied to calculate the transport fraction to the Hold-up Volume Tank in Advanced Power Reactors 1400. By the present model, the debris transport fraction is predicted, and the effect of particle density and particle size on transport is investigated.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1045-1052
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• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.