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    Fracture Mechanics Analysis of Boshehr Nuclear Power Plant (BNPP) RPV during LBLOCA Using ABAQUS Software

    , M.Sc. Thesis Sharif University of Technology Mirseifi, Miryousef (Author) ; Ghafari, Mohsen (Supervisor)
    Abstract
    In nuclear power plants (NPPs), the integrity of the Reactor Pressure Vessel (RPV) is one of the most essential issues in assessing the life of the NPPs. The RPV in a nuclear power plant cannot be replaced during plant life. Hence, maintaining the integrity of the RPV nuclear power plant is of great importance. In this study, the RPV integrity of Boshehr Nuclear Power Plant (BNPP) gainst crack formation and growth in the case of LBLOCA is evaluated and studied by the finite element method (FEM) in ABAQUS using the J-Integral method and also the weight function, which is a function of crack shape. According to the results, stress values increased significantly during the LBLOCA incident. The... 

    Flow Pattern Prediction in iPWR SMR with Natural Circulation by Coupling of RELAP and ANSYS CFX Code

    , M.Sc. Thesis Sharif University of Technology Emampour, Mohammad Hassan (Author) ; Ghafari, Mohsen (Supervisor)
    Abstract
    Selection of appropriate thermohydraulic tool for analyzing nuclear reactors is a trade of between accuracy and calculation run-time. Nuclear reactors analyses perform on two levels including system and sub-channel. In this regard or the system codes' one-dimensional approach is selected or the CFD codes for considering the three-dimensional and non-equilibrium phenomenon are employed. In this research a T-H tool for prediction of light-water cooled reactors core is developed. Although this code requires lower CPU and run-time in comparison with CFD codes, in contrast with system codes can report non-equilibrium and three-dimensional phenomenon. This code has a modular implementation based... 

    Simulation of Bushehr Nuclear Power Plant Hot Channel Using ANSYS CFX 18.0 Software

    , M.Sc. Thesis Sharif University of Technology Asadi, Ali Asghar (Author) ; Ghafari, Mohsen (Supervisor) ; Hosseini, Abolfazl (Co-Supervisor)
    Abstract
    In nuclear power plants, the fission reaction takes place inside the fuel rods, and the heat generated inside the fuel rods is transferred through the clad wall to the cooling fluid, which is single-phase, subcooled and in the turbulence zone. In the hot channel of the reactor core, where the axial and radial heat flux reaches its maximum state, the fluid leaves the single-phase state and some steam is formed in the area close to the clad wall.Therefore, in some cases, in the hot channel, the steam near the clad wall may reach a point that reduces the heat transfer coefficient, rapidly increases the wall temperature and thus destroys or melts the sheath surface. If this happens, it means...