Burnup dependent neutron flux and heat source distribution in sphere-pac mixed carbide fuels Public Deposited

http://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/pn89d9620

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  • Through a joint effort by the Swiss Federal Institute for Reactor Research and Oregon State University, the fuels modeling code SPECKLE is being developed to model sphere-pac mixed carbide nuclear fuel. Major parameters, such as fission gas release, fuel restructuring phenomena, and the temperature and porosity distributions in the fuel pin were modeled in the initial version SPECKLE-I. Additional modeling efforts to improve the thermal models and to characterize the mechanical aspects of sphere-pac fuel behavior provide the basis for subsequent versions of SPECKLE. A part of the improvements in modeling the thermal behavior has been to develop the capability to include changes due to burnup in the calculated heat source distribution. To achieve this, several routines were developed to provide a neutronics/heat source module for SPECKLE. This module includes a fitting routine (CUBFIT) to provide the option of altering the lethargy group structure of global reactor parameters from the reactors used for test pin irradiations. A neutronics routine (NEUTRON) based on transport theory provides average group fluxes in each of ten concentric rings in the fuel pin, and the burnup routine (BURNUP) provides new values for uranium and plutonium number densities across the pin. Also described are modifications to the SPECKLE-I heat source routine to incorporate burnup dependence as well as porosity changes in the heat source distribution in the pin.
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