Numerical solutions of the neutron transport equation Public Deposited

http://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/fj2365253

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  • Because of the high cost of fuel for nuclear reactors, fuel cycle costs must be predicted accurately. This leads to the high cost computer computations and a search for cheaper yet still accurate methods. For this paper we have chosen to study an alternate method for calculating thermal neutron cell averaged cross sections. The method developed involves an expansion of the slab geometry transport equation by means of eigenfunctions. It utilizes moments matching boundary conditions similar to P[subscript]N methods. Theory and calculations are given in slab and cylindrical geometry. The method is extended to multigroup calculations and comparisons are made with standard calculational methods. Results indicate that this method can be used as an alternative to more standard methods such as integral transport and Monte Carlo theories. Accuracy of this method is quite good, better than P₃ theory in most cases. Computer calculation time is at least 50% faster than integral transport theory and, of course, only a small fraction of Monte Carlo theory.
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