Fluence based neutron balance approach using spatial flux calculations




Bagdatlioglu, Cem

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This thesis describes the addition of spatially dependent power sharing to a methodology for calculating the input and output isotopics and burnup of nuclear reactors within a nuclear fuel cycle simulator. This methodology carries out neutron balance and depletion calculations by using pre-calculated fluence-based libraries. These libraries track the transmutation and neutron economy evolution of unit masses of isotopes available in input fuel. The current work generalizes the method to simulate reactors that contain more than one type of fuel in their core, for instance breeders with a driver-blanket configuration. To achieve this, spatial flux calculations are used to determine the fluence-dependent relative average flux inside macroscopic spatial regions. These fluxes are then used to determine the relative average power of macroscopic spatial regions as well as to more accurately calculate region-specific transmutation rates. The paper presents several cases where the fluence based approach alone would not have been sufficient to determine results.



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