• Magazine of the Korean Society of Agricultural Engineers
• /
• v.39 no.4
• /
• pp.106-113
• /
• 1997

Dynamic loading of structures often causes excursions of stresses well into the inelastic range, and the influence of the geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. A structure in a nuclear power plant is a structure of importance which puts emphasis on safety. A nuclear container is a pressure vessel subject to internal pressure and this structure is constructed by a reinforced concrete or a pre-stressed concrete. In this study, the material nonlinearity effect on the dynamic response is formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion is numerically solved by a central difference scheme. The constitutive relation of concrete is modeled according to a Drucker-Prager yield criterion in compression. The reinforcing bars are modeled by a smeared layer at the location of reinforcements, and the steel layer model under Von Mises yield criteria is adopted to represent an elastic-plastic behavior. To investigate the dynamic response of a nuclear reinforced concrete containment structure, the steel-ratios of 0, 3, 5 and 10 percent, are considered. The results obtained from the analysis of an example were summarized as follows 1. As the steel-ratio increases, the amplitude and the period of the vertical displacements in apex of dome decreased. The Dynamic Magnification Factor(DMF) was some larger than that of the structure without steel. However, the regular trend was not found in the values of DMF. 2. The dynamic response of the vertical displacement and the radial displacement in the dome-wall junction were shown that the period of displacement in initial step decreased with the steel-ratio increases. Especially, the effect of the steel on the dynamic response of radial displacement disapeared almost. The values of DMF were 1.94, 2.5, 2.62 and 2.66, and the values increased with the steel-ratio. 3. The characteristics of the dynamic response of radial displacement in the mid-wall were similar to that of dome-wall junction. The values of DMF were 1.91, 2.11, 2.13 and 2.18, and the values increased with the steel-ratio. 4. The amplitude and the period of the hoop-stresses in the dome, the dome-wall junction, and the mid-wall were shown the decreased trend with the steel-ratio. The values of DMF were some larger than those of the structure without steel. However, the regular trend was not found in the values of DMF.

• Journal of Navigation and Port Research
• /
• v.37 no.2
• /
• pp.129-135
• /
• 2013

Recently, waste water treatment system is developed in small and middle size to get more economic advantage. Freeze concentration system has high thermodynamic efficiency and low energy consumption, can re-use purified water and cold energy obtained from ice. This study was experimentally performed to investigate pollution containment in frozen layer by cooling wall temperature, air-bubble flow methods, initial ice-lining thickness of frozen layer in NaCl aqueous solution and the representative heavy metals, Pb and Cr aqueous solution. As the result, a decrease in the cooling wall temperature bring a higher growth rate of ice front and the more solute was involved in frozen layer. The method to inject directly air-bubble into ice-liquid interface through ring shape nozzle gave high purity of ice compared to indirect method. Ice lining in 5mm thickness resulted in frozen layer with higher purity than 1mm thickness.

• Nuclear Engineering and Technology
• /
• v.56 no.1
• /
• pp.78-91
• /
• 2024

Double steel plate concrete composite shear wall (SCSW) has been widely utilized in nuclear power plants and high-rise structures, and its shear connectors have a substantial impact on the seismic performance of SCSW. Therefore, in this study, the mechanical properties of SCSW with angle stiffening ribs as shear connections were parametrically examined for the reactor containment structure of nuclear power plants. The axial compression ratio of the SCSW, the spacing of the angle stiffening rib arrangement and the thickness of the angle stiffening rib steel plate were selected as the study parameters. Four finite element models were constructed by using the finite element program named ABAQUS to verify the experimental results of our team, and 13 finite element models were established to investigate the selected three parameters. Thus, the shear capacity, deformation capacity, ductility and energy dissipation capacity of SCSW were determined. The research results show that: compared with studs, using stiffened ribs as shear connectors can significantly enhance the mechanical properties of SCSW; When the axial compression ratio is 0.3-0.4, the seismic performance of SCSW can be maximized; with the lowering of stiffener gap, the shear bearing capacity is greatly enhanced, and when the gap is lowered to a specific distance, the shear bearing capacity has no major affect; in addition, increasing the thickness of stiffeners can significantly increase the shear capacity, ductility and energy dissipation capacity of SCSW. With the rise in the thickness of angle stiffening ribs, the improvement rate of each mechanical property index slows down. Finally, the shear bearing capacity calculation formula of SCSW with angle stiffening ribs as shear connectors is derived. The average error between the theoretical calculation formula and the finite element calculation results is 8% demonstrating that the theoretical formula is reliable. This study can provide reference for the design of SCSW.

• Journal of Mechanical Science and Technology
• /
• v.18 no.6
• /
• pp.990-1001
• /
• 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.

• International Journal of Concrete Structures and Materials
• /
• v.10 no.1
• /
• pp.47-60
• /
• 2016

The objective of this study was to evaluate the capability of different strength-maturity models to account for the effect of the hydration heat on the in-place strength development of high-strength concrete specifically developed for nuclear facility structures under various ambient curing temperatures. To simulate the primary containment-vessel of a nuclear reactor, three 1200-mm-thick wall specimens were prepared and stored under isothermal conditions of approximately $5^{\circ}C$ (cold temperature), $20^{\circ}C$ (reference temperature), and $35^{\circ}C$ (hot temperature). The in situ compressive strengths of the mock-up walls were measured using cores drilled from the walls and compared with strengths estimated from various strength-maturity models considering the internal temperature rise owing to the hydration heat. The test results showed the initial apparent activation energies at the hardening phase were approximately 2 times higher than the apparent activation energies until the final setting. The differences between core strengths and field-cured cylinder strengths became more notable at early ages and with the decrease in the ambient curing temperature. The strength-maturity model proposed by Yang provides better reliability in estimating in situ strength of concrete than that of Kim et al. and Pinto and Schindler.

• 한국전산유체공학회:학술대회논문집
• /
• 2008.03a
• /
• pp.308-315
• /
• 2008

극저온 유체를 운반하는 선박의 화물창은 보통 1차 방벽과 2차 방벽으로 구성되어 있다. 1차 방벽에 소량의 극저온 유체의 누출이 생기더라도, 액밀이 되는 2차 방벽에서 추가적인 극저온 유체의 누출을 방지할 수 있기 때문이다. 그러나 2차 방벽에 추가적인 손상이 생길 경우 유리솜으로 만들어진 Flat Joint를 거쳐 선체내벽까지 극저온 유체에 노출될 가능성이 있게 된다. 본 연구의 관심사는 2차 방벽의 손상 정도에 따라, 그리고 누출되는 극저온 유체의 양에 따라, 내측선체에 얼마나 영향이 미치는가를 알아보는 것이다. 이를 위해 본 연구에서는 극저온 유체를 운반하는 선박의 화물창을 구성하는 2차 방벽에 구멍을 뚫어 Flat Joint 사이로 극저온 유체가 흘러 들어가도록 격자를 생성한 후, CFD 소프트웨어를 이용하여 극저온 유체의 누출에 대한 계산을 수행 하였으며 이를 실험 결과와 비교 분석 하였다. 실험과 계산 결과를 살펴보면, 극저온 유체량에 따라 내측 선체에의 피해를 최소화 할 수 있음을 확인하였다.

• 한국전산유체공학회:학술대회논문집
• /
• 2008.10a
• /
• pp.308-315
• /
• 2008

극저온 유체를 운반하는 선박의 화물창은 보통 1차 방벽과 2차 방벽으로 구성되어 있다. 1차 방벽에 소량의 극저온 유체의 누출이 생기더라도, 액밀이 되는 2차 방벽에서 추가적인 극저온 유체의 누출을 방지할 수 있기 때문이다. 그러나 2차 방벽에 추가적인 손상이 생길 경우 유리솜으로 만들어진 Flat Joint를 거쳐 선체내벽까지 극저온 유체에 노출될 가능성이 있게 된다. 본 연구의 관심사는 2차 방벽의 손상 정도에 따라, 그리고 누출되는 극저온 유체의 양에 따라, 내측선체에 얼마나 영향이 미치는가를 알아보는 것이다. 이를 위해 본 연구에서는 극저온 유체를 운반하는 선박의 화물창을 구성하는 2차 방벽에 구멍을 뚫어 Flat Joint 사이로 극저온 유체가 흘러 들어가도록 격자를 생성한 후, CFD 소프트웨어를 이용하여 극저온 유체의 누출에 대한 계산을 수행 하였으며 이를 실험 결과와 비교 분석 하였다. 실험과 계산 결과를 살펴보면, 극저온 유체량에 따라 내측 선체에의 피해를 최소화 할 수 있음을 확인하였다.

• Journal of the Society of Naval Architects of Korea
• /
• v.54 no.5
• /
• pp.361-367
• /
• 2017

A membrane type LNG cargo tank is equipped with a pump tower and a liquid dome for loading and unloading of LNG. However, the membrane running continuously on the tank wall to prevent leakage of LNG is interrupted by the liquid dome, hence care should be taken in the design of liquid dome and its substructures. In case of GTT NO96 membrane type cargo containment system, chair structure is arranged along the periphery of the liquid dome targeting to support the membrane which is exposed to the both hull girder and thermal load. This paper proposes a new and simple chair structure, which outperforms traditional design from productivity point of view maintaining same level of structural safety. Strength assessment on the new design was performed to guarantee the structural safety of the new design, which includes strength, fatigue and crack propagation analysis.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 1998.10a
• /
• pp.3-8
• /
• 1998

The present work proposes the potential fuzzy framework, based on fuzzy set theory, for supporting decision-making problems, especially, selection problems of a best design in the area of nuclear energy system. The framework proposed is composed of the hierarchical structure module, the assignment module, the fuzzification module, and the defuzzification module. In the structure module, the relationship among decision objectives, decision criteria, decision sub-criteria, and decision alternatives is hierarchically structured. In the assignment module, linguistic or rank scoring approach can be used to assign subjective and/or vague values to the decision analyst's judgment on decision variables. In the fuzzification module, fuzzy numbers are assigned to these values of decision variables. Using fuzzy arithmetic operations, for each alternative, fuzzy preference index as a fuzzy synthesis measure is obtained. In the defuzzification module, using one of methods ranking fuzzy numbers, these indices are defuzzified to overall utility values as a cardinality measure determining final scores. According these values, alternatives of interest are ranked and an optimal alternative is chosen. To illustrate the applicability of the framework proposed to selection problem, as a case example, the best option choice of four design options under five decision criteria for primary containment wall thickening around large penetrations in an advanced nuclear energy system is studied.

• Structural Engineering and Mechanics
• /
• v.60 no.2
• /
• pp.337-350
• /
• 2016

Even though extensive researches have been performed for steam explosion due to their complex mechanisms and inherent uncertainties, establishment of severe accident management guidelines and strategies is one of state-of-the arts in nuclear industry. The goal of this research is primarily to examine effects of vessel failure modes and locations on nuclear facilities under typical steam explosion conditions. Both discrete and integrated models were employed from the viewpoint of structural integrity assessment of steel components and evaluation of the cracking and crushing in reinforced concrete structures. Thereafter, comparison of systematic analysis results was performed; despite the vessel failure modes were dominant, resulting maximum stresses at the all steel components were sufficiently lower than the corresponding yield strengths. Two failure criteria for the reinforced concrete structures such as the limiting failure ratio of concrete and the limiting strains for rebar and liner plate were satisfied under steam explosion conditions. Moreover, stresses of steel components and reinforced concrete structures were reduced with maximum difference of 12% when the integrated model was adopted comparing to those of discrete models.