• Proceedings of the Computational Structural Engineering Institute Conference
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• 2007.04a
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• pp.3-17
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• 2007

This paper provides an introduction to life-span simulation and numerical approach to support the performance design processes of reinforced concrete structures. An integrated computational system is proposed for life-span simulation of reinforced concrete. Conservation of moisture, carbon dioxide, oxygen, chloride, calcium and momentum is solved with hydration, carbonation, corrosion, ion dissolution. damage evolution and their thermodynamic/mechanical equilibrium. Coupled analysis of mass transport and damage mechanics associated with steel corrosion is presented for structural performance assessment of reinforced concrete. Multi-scale modeling of micro-pore formation and transport phenomena of moisture and ions are mutually linked for predicting the corrosion of reinforcement and volumetric changes. The interaction of crack propagation with corroded gel migration can also be simulated. Two finite element codes. multi-chemo physical simulation code (DuCOM) and nonlinear dynamic code of structural reinforced concrete (COM3) were combined together to form the integrated simulation system. This computational system was verified by the laboratory scale and large scale experiments of damaged reinforced concrete members under static loads, and has been applied to safety and serviceability assessment of existing structures. Based on the damage details predicted by the nonlinear finite element analytical system, the life-span-cost of RC structures including the original construction costs and the repairing costs for possible damage during the service life can be evaluated for design purpose.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.191-202
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• 2013

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Structural Engineering and Mechanics
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• v.84 no.6
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• pp.743-765
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• 2022

Near-surface excavations may cause the tilting and destruction of the adjacent superstructures in big cities. The stability of a huge excavation and its nearby superstructures was studied in this paper. Some test instruments monitored the deformation and loads at the designed location. Then the numerical models of the excavation were made in FLAC3D (a three-dimensional finite difference code) and Plaxis-3D (a three-dimensional finite element code). The effects of different supporting and reinforcement tools such as nails, piles, and shotcretes on the stability and bearing capacity of the foundation were analyzed through different numerical models. The numerically approximated results were compared with the corresponding in-field monitored results and reasonable compatibility was obtained. It was concluded that the displacement in excavation and the settlement of the nearby superstructure increases gradually as the depth of excavation rises. The effects of support and reinforcements were also observed and modeled in this study. The settlement of the structure gradually decreased as the supports were installed. These analyses showed that the pile significantly increased the bearing capacity and decreased the settlement of the superstructure. As a whole, the monitoring and numerical simulation results were in good consistency with one another in this practically important project.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• Communications of Mathematical Education
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• v.30 no.3
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• pp.281-294
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• 2016

This study deals with the content that was developed by the authors and the utilization of the 'Interactive Engineering Math Laboratory (IEmath Lab).' IEmath Lab provides online review lectures as well as a wide range of examples and exercises from the curriculum of engineering mathematics courses. The lectures come with pre-coded Python-based SageMath cells through which students can run and modify the code directly from this free laboratory. IEmath Lab is accessible via mobile devices so that the students can use it anywhere, anytime for maximum learning effectiveness and achievement. IEmath Lab would be an ideal tool for the effective learning and teaching of engineering mathematics, which combines theory and practice.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.397-416
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• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part II, based on the verified finite element (FE) models of aircrafts Airbus A320 and A380, as well as the NPP containment and auxiliary buildings in Part I of this paper, the whole collision process is reproduced numerically by adopting the coupled missile-target interaction approach with the finite element code LS-DYNA. The impact induced damage of NPP plant under four impact locations of containment (cylinder, air intake, conical roof and PCS water tank) and two impact locations of auxiliary buildings (exterior wall and roof of spent fuel pool room) are evaluated. Furthermore, by considering the inner structures in the containment and raft foundation of NPP, the structural vibration analyses are conducted under two impact locations (middle height of cylinder, main control room in the auxiliary buildings). It indicates that, within the discussed scenarios, NPP structures can withstand the impact of both two aircrafts, while the functionality of internal equipment on higher floors will be affected to some extent under impact induced vibrations, and A380 aircraft will cause more serious structural damage and vibrations than A320 aircraft. The present work can provide helpful references to assess the safety of the structures and inner equipment of NPP plant under commercial aircraft impact.

• Tunnel and Underground Space
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• v.30 no.3
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• pp.242-255
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• 2020

The numerical simulations of Heater Experiment-D (HE-D) at the Mont Terri rock laboratory in Switzerland were performed to investigate an applicability of FLAC3D to reproduce the coupled thermo-hydro-mechanical (THM) behaviour in Opalinus Clay, as part of the DECOVLEX-2015 project Task B. To investigate the reliability of numerical simulations of the coupled behaviour using FLAC3D code, the simulation results were compared with the observations from the in-situ experiment, such as temperature at 16 sensors, pore pressure at 6 sensors, and strain at 22 measurement points. An anisotropic heat conduction model, fluid flow model, and transversely isotropic elastic model in FLAC3D successfully represented the coupled thermo-hydraulic behaviour in terms of evolution for temperature and pore pressure, however, performance of the models for mechanical behavior is not satisfactory compared with the measured strain.

• Coupled systems mechanics
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• v.8 no.4
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• pp.339-350
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• 2019

A numerical simulation of the incompressible multiphase hydraulic jump flow was performed to compare the interface prediction through the use of the three RANS turbulence models: $k-{\varepsilon}$, $RNGk-{\varepsilon}$ and SST $k-{\omega}$. A three dimensional no submerged hydraulic jump and a two dimensional submerged hydraulic jump were modeled. Both the geometry and the mesh were created using the open source Gmsh code. The project's geometry consists of a rectangular channel with length and height differences between the two dimensional and three dimensional simulations. Uniform hexahedral cells were used for the mesh. Three refining meshes were constructed to allow to verify simulation convergence. The Volume of Fluid (abbr. VOF) method was used for treatment of the air-water surface. The turbulence models were evaluated in three distinct set up configurations to provide a greater accuracy in the flow representation. In the two-dimensional analysis of a submerged hydraulic jump simulation, the turbulence model RNG RNG $k-{\varepsilon}$ provided a better interface adjust with the experimental results than the model $k-{\varepsilon}$ and SST $k-{\omega}$. In the three-dimensional simulation of a no-submerged hydraulic jump the k-# showed better results than the SST $k-{\omega}$ and RNG $k-{\varepsilon}$ capturing the height and length of the ledge with a better fit with the experimental results.