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

Asadi, Ali Asghar | 2022

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 55357 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Ghafari, Mohsen; Hosseini, Abolfazl
  7. Abstract:
  8. 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 that the bubbles have come out of the nuclear boiling state and have reached a critical heat flux, which is defined in terms of safety and a parametric nuclear power plant as DNBR.Therefore, accurate prediction of the two-phase flow pattern and heat transfer inside the pressurized chamber is very important for the safety of the equipment. Due to the safety requirements of the nuclear power plant and its costs, it is not possible to perform such an experiment experimentally, so in this case, the code based on computational fluid dynamics, including CFX, is referred to.In this project, the model of Eulerian two liquids is combined and solved with the wall boiling model in CFX software. For validation, first experiments based on the parameters of volume fraction, fluid temperature, wall temperature, vapor velocity and experiments based on the determination of critical heat flux were performed and then, the simulation of the channel of a fuel assembly in nominal, hot and accident conditions was performed
  9. Keywords:
  10. Two Phase Flow ; Nuclear Power Plants ; Critical Heat Flux ; Nuclear Safety ; CFX Code ; Nuclear Power Plant Safety ; Heat Flux Partitioning

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