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Two-Dimensional, Two-Phase Simulation of the Fuel Channel in a Pressurized Water Reactor (PWR) – Case Study: Bushehr Nuclear Power Plant

Dashti, Mohammad | 2025

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 58278 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Ghaffari, Mohsen
  7. Abstract:
  8. In nuclear power plants, heat is generated through nuclear fission reactions within the fuel pellets and is transferred via the fuel cladding to the flowing coolant. At the inlet of the fuel channel, the coolant is in a single-phase, subcooled, and turbulent state. As the coolant flows along the channel, due to the concentration of heat generation at the center of the reactor core, the heat flux increases in a sinusoidal manner. This results in a rise in the coolant's temperature gradient in both radial and axial directions. Under normal operating conditions, the coolant temperature near the channel wall may approach the boiling point, leading to the formation of vapor bubbles on the wall. These bubbles can either enhance or reduce the heat transfer coefficient, thereby affecting the wall temperature. This phenomenon can lead to cracking and leakage of radioactive materials in the primary circuit. Additionally, due to the reduction in coolant density and the consequent increase in flow velocity, the control rods of the reactor may be damaged. Moreover, the decrease in moderator density may result in reactor power fluctuations and potential damage to the steam turbine or steam generator. Studying the variations of thermohydraulic parameters that may lead to such damage emphasizes the necessity of analyzing fluid flow in the reactor’s fuel channel in both radial and axial dimensions. Since bubble formation occurs on the channel walls, experimental analysis of this phenomenon under high-pressure conditions is not economically feasible. Therefore, two main approaches have been adopted in recent years: using refrigerant gases such as R12 and R134a to simulate the conditions of a pressurized light water reactor, and numerical simulation of the fuel channel. Several codes and research studies have been conducted in this area. These include the one-dimensional simulation of the Bushehr power plant fuel channel by Hajizadeh et al [1]. , the RELAP code, the ANSYS code, and the experimental work by Gagner [2]. In this context, the present study simulates the nuclear power plant fuel channel in two dimensions using the drift-flux equations proposed by Keldya et al [3]. and the Lahey [4] model. For validation purposes, two experiments selected by Keldya and colleagues [3] from Deborah’s data, as well as simulation results from Hajizadeh et al [1] , have been used. Finally, the Bushehr fuel channel is simulated and validated using the ANSYS code proposed by Asadi [5]. The results confirm that the developed program is capable of simulating two-phase flow. Furthermore, it is found that although the fluid temperature near the hot channel wall closely approaches the boiling point, no bubble formation occurs
  9. Keywords:
  10. Thermohydraulic Code Development ; Two Phase Flow ; Single Phase Flow ; Lahey Model ; Heat Flux ; Drift Flux Model ; Bushehr Nuclear Power Plant ; Pressurized Light Water Reactor

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