• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.1383-1384
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.725-726
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• 2005

• 한국가스학회지
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• 제12권1호
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• pp.36-41
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• 2008

본 논문에서는 완전밀폐식 LNG 저장탱크의 외부탱크 측벽면과 지붕 구조물에 대한 강도안전성을 유한요소법으로 해석하였다. 예응력 콘크리트 구조물로 건설된 외부탱크는 내부탱크의 붕괴로 인해 발생하는 LNG 유체정압과 유체동압, 그리고 태풍을 포함한 외부의 풍압하중을 받는다. FEM 해석결과에 의하면, 외부탱크의 측벽면과 지붕 구조물이 서로 연결되는 링빔 구조물은 저장탱크에 작용하는 대부분의 내 외부 하중을 담당하고 있다. 이러한 해석결과는 외부탱크의 설계 포인트를 링빔에 두고, 그 다음은 지붕구조물의 중심부에 대한 설계 안전성을 검토하는 것이다. 완전밀폐식 LNG 저장탱크 해석에서 사용한 해석모델은 LNG 누설에 의한 내부압력 및 태풍과 같은 외부압력이 결합된 복합하중에서도 안전한 강도안전성을 유지하고 있음을 알 수 있다.

• 한국전산구조공학회논문집
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• 제23권4호
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• pp.445-452
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• 2010

CANDU형 격납건물에 대하여 극한내압하중에 대한 확률론적 취약도 평가를 수행하였다. 격납건물 성능의 불확실성은 가동중 검사 결과를 통해 얻어진 재료 물성치 중앙값과 텐던 긴장력 중앙값을 적용하여 고려하였다. 격납건물은 개구부를 고려하여 3차원 유한요소로 모델링하였으며, 확률론적 취약도 평가를 위하여 대규모의 비선형 유한요소해석 모델을 적용하기에 적합한 효율적인 취약도 평가기법을 개발하였다. 월성 1호기 격납건물에 대한 물성치를 사용하였다. 개발된 새로운 취약도 평가기법을 도입하여 각각의 파괴모드에 대한 취약도 평가를 수행하였으며, 파괴모드 별, 신뢰도 수준별 취약도 곡선을 도출하였다. 벽체 중단부가 극한내압발생으로 인한 방사능물질 누출에 가장 취약한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제47권6호
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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.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2005년도 춘계 학술발표회 논문집
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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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• 제44권3호
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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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• 제42권2호
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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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• 제41권8호
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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.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.805-819
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• 2016

The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading.