• Smart Structures and Systems
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• 제23권3호
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• pp.227-242
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• 2019

Seismic isolation systems employ structural control that protect both buildings and vibration-sensitive contents from destructive effects of earthquakes. Structural control is divided into three main groups: passive, active, and semi-active. Among them, semi-active isolation systems, which can reduce floor displacements and accelerations concurrently, has gained importance in recent years since they don't require large power or pose stability problems like active ones. However, their seismic performance may vary depending on the variations that may be observed in the mechanical properties of semi-active devices and/or seismic isolators. Uncertainties relating to isolators can arise from variations in geometry, boundary conditions, material behavior, or temperature, or aging whereas those relating to semi-active control devices can be due to thermal changes, inefficiencies in calibrations, manufacturing errors, etc. For a more realistic evaluation of the seismic behavior of semi-active isolated buildings, such uncertainties must be taken into account. Here, the probabilistic behavior of semi-active isolated buildings under historical pulse-like near-fault earthquakes is evaluated in terms of their performance in preserving structural integrity and protecting vibration-sensitive contents considering aforementioned uncertainties via Monte-Carlo simulations of 3-story and 9-story semi-active isolated benchmark buildings. The results are presented in the form of fragility curves and probability of failure profiles.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3474-3490
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• 2021

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

• International Journal of Naval Architecture and Ocean Engineering
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• 제12권1호
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• pp.868-880
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• 2020

One of the main concerns in the structural integrity of offshore pipelines is mechanical damage from external loads. Pipelines are exposed to fatigue failure in welded joints due to geometric discontinuity. In addition, fatigue loads such as currents, waves, and platform motions may cause significant plastic deformation and fracture or leakage within a relatively low-cycle regime. The 2007 ASME Div. 2 Code adopts the master S―N curve for the fatigue evaluation of welded joints based on the mesh-insensitive structural stress. An extension to the master S―N curve was introduced to evaluate the low-cycle fatigue strength. This structural strain method uses the tensile properties of the material. However, the monotonic tensile properties have limitations in describing the material behavior above the elastic range because most engineering materials exhibit hardening or softening behavior under cyclic loads. The goal of this study is to extend the cyclic stress-strain behavior to the structural strain method. To this end, structural strain-based procedure was established while considering the cyclic stress-strain behavior and compared to the structural strain method with monotonic tensile properties. Finally, the improved prediction method was validated using fatigue test data from full-scale girth-welded pipes.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.30-35
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• 2022

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

• 한국기계가공학회지
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• 제21권6호
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• pp.81-88
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• 2022

Titanium alloy (Ti-6Al-4V) has excellent mechanical properties and high specific strength; therefore, it is widely used in aerospace, automobile, defense, engine parts, and bio fields. Particularly in the aerospace field, as it has a low specific gravity and rigidity, it is used for the purpose of increasing energy efficiency through weight reduction of parts, and most have a thin-walled structure. However, it is extremely difficult to machine thin-walled shapes owing to vibration and deformation. In the case of thin-walled structures, the cutting forces and vibrations rapidly increase depending on the cutting conditions, significantly affecting the surface integrity and tool life. In this study, machining experiments on thin-wall milling of a titanium alloy (Ti-6Al-4V) were conducted for each experimental condition with different axial depths of cut, radial depth of cut, and machining sequence. The machining characteristics were analyzed, and an effective machining method was derived by a comprehensive analysis of the machined surface conditions and cutting signals.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1874-1889
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• 2022

In this study, we evaluate the thermal-hydraulic performance and economics of Printed Circuit Heat Exchanger (PCHE) according to the channel types and associated shape variables for the design of recuperators with Sodium-cooled Fast Reactors (SFRs). To perform the evaluations with variables such as the Reynolds number, channel types, tube diameter, and shape variables, a code for the heat exchanger is developed and verified through a comparison with experimental results. Based on the code, the volume and pressure drop are calculated, and an economic assessment is conducted. The zigzag type, which has bending angle of 80° and a tube diameter of 1.9 mm, is the most economical channel type in a SFR using CO₂ as the working fluid. For a SFR using N₂, we recommend the airfoil type with vertical and horizontal numbers of 1.6 and 1.1, respectively. The airfoil type is superior when the mass flow rate is large because the operating cost changes significantly. When the mass flow rate is small, volume is a more important design parameter, therefore, the zigzag type is suitable. In addition, we conduct a sensitivity analysis based on the production cost of the PCHE to identify changes in optimal channel types.

• 한국초전도ㆍ저온공학회논문지
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• 제25권1호
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• pp.16-20
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• 2023

The International Electrotechnical Commission (IEC) 61788-26:2020 provides guidelines for measuring the critical current of Rare-earth barium copper oxide (REBCO) tapes using two methods: linear ramp and step-hold methods. The critical current measurement criterion, 1 or 0.1 μV/cm of electric field from IEC 61788-26 has been normally applied to REBCO coils or magnets. No-insulation (NI) winding technique has many advantages in aspects of electrical and thermal stability and mechanical integrity. However, the leak current from the NI REBCO coil can cause distortion in critical current measurement due to the characteristic resistance which causes the radial current flow paths. In this paper, we simulated the NI REBCO coil by applying both linear ramp and step-hold methods based on a simplified equivalent circuit model. Using the circuit analysis, we analyzed and evaluated both methods. By using the equivalent circuit model, we can evaluate the critical current of the NI REBCO coil, resulting in an estimation error within 0.1%. We also evaluate the accuracy of critical current measurement using both the linear ramp and step-hold methods. The accuracy of the linear ramp method is influenced by the inductive voltage, whereas the accuracy of the step-hold method depends on the duration of the hold-time. An adequate hold time, typically 5 to 10 times the time constant (τ), makes the step-hold method more accurate than the linear ramp method.

• Structural Engineering and Mechanics
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• 제86권5호
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• pp.663-671
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• 2023

Freeze-thaw cycles induce strength loss at the frozen soil-concrete interface and deterioration of bonding, which causes construction engineering problems. To clarify the deterioration characteristics of the interface under the freeze-thaw cycle, a frozen soil-concrete sample was used as the research object, an interface scanning electron microscope test under the freeze-thaw cycle was carried out to identify the micro index information, and an interface shear test was carried out to explore the loss law of interface shear strength under the freeze-thaw cycle. The results showed that the integrity of the interface was destroyed, and the pore number and pore size of the interface increased significantly with the number of freeze-thaw cycles. The connection form gradually deteriorates from surface-to-surface contact to point-to-surface contact and point-to-point contact, and the interfacial shear strength decreases the most at 0-3 freeze-thaw cycles, with small decreases from to 3-8 cycles. After 12 freeze-thaw cycles, the interfacial shear strength tends to be stable, and shear the failure occurs internally in the soil.

• Coupled systems mechanics
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• 제12권5호
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• pp.429-444
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• 2023

The comprehension and structural modeling of masonry constructions is fundamental to safeguard the integrity of built cultural assets and intervene through adequate actions, especially in earthquake-prone regions. Despite the availability of several modeling strategies and modern computing power, modeling masonry remains a great challenge because of still demanding computational efforts, constraints in performing destructive or semi-destructive in-situ tests, and material uncertainties. This paper investigates the shear behavior of masonry walls by applying a plane-stress FE continuum model with the Modified Masonry-like Material (MMLM). Epistemic uncertainty affecting input parameters of the MMLM is considered in a probabilistic framework. After appointing a suitable probability density function to input quantities according to prior engineering knowledge, uncertainties are propagated to outputs relying on gPCE-based surrogate models to considerably speed up the forward problem-solving. The sensitivity of the response to input parameters is evaluated through the computation of Sobol' indices pointing out the parameters more worthy to be further investigated, when dealing with the seismic assessment of masonry buildings. Finally, masonry mechanical properties are calibrated in a probabilistic setting with the Bayesian approach to the inverse problem based on the available measurements obtained from the experimental load-displacement curves provided by shear compression in-situ tests.