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Pin Power Reconstruction Method by Nodal Core Calculation Results

Kefalati, Mohadeseh | 2021

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 54518 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Vosoughi, Naser; Ghaffari, Mohsen
  7. Abstract:
  8. The widespread use of nuclear energy leads to obtain detailed information, such as neutron flux distribution (power) which is very effective in designing and evaluating the reactor safety. The neutron flux (power) reconstruction method uses the homogeneous flux distribution and the heterogeneous form function in a fuel assembly to calculate the heterogeneous power in the fuel rods. Therefore, this method has been widely developed in the last two decades. This study investigates to calculate two-dimensional and two-group neutron flux (power) in the fuel rod for both quadrilateral and hexagonal geometry related to core results by using nodal method. To achieve a more complete program and join programs to the nodal computational programs developed in the college, we developed data and program called PPR-S&H and used the benefits of the nodal method to reconstruct the distribution of power in the fuel rods in the core. To this aim, it was necessary to acquire boundary conditions in each fuel assembly, homogeneous flux distribution and the form function of each fuel rod. The boundary conditions were predetermined using the Nodal method. The reconstructed neutron flux (power) was simulated by using these conditions, finding the required parameters, and a polynomial equation. Then, the heterogeneous form function was calculated by using the homogeneous flux of each fuel rod. After calculations, PPR-S & H software was developed and standardized for BIBLIS-2D and Bushehr VVER-1000 hexagonal reactors. To benchmark the calculations, the Biblis-2D and the VVER-1000 Bushehr reactors were selected because the boundary conditions were performed by using the nodal method. In the Biblis-2D reactor, only the homogeneous flux and the relative power error of fuel assemblies are calculated; The maximum relative power error was 3.53%. the complete simulation of VVER-1000 Bushehr reactor shows that the maximum relative power error of fuel assembly, average relative power error of core, and maximum relative power error of the fuel rod in each fuel assembly in the core were 10.61%, 5.14%, and 10.88%, respectively
  9. Keywords:
  10. Neutron Flux ; Rectangular Geometries ; Bushehr Nuclear Power Plant ; Form Function ; Maximum Relative Power ; Reconstructed Neutron Flux ; Hexagonal Geometry ; Homogeneous Flux Distribution

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