A new method is developed to analyze CANDLE (Constant Axial shape of Neutron flux, nuclide densities andpower shape During Life of Energy producing reactor) burnup, where the microscopic group cross-sections are evaluatedat every space mesh by TLLI (Table Look-up and Linear Interpolation) method, and used to analyze a fast reactorwith natural uranium as a fresh fuel. The results are compared with the conventional method, where only one setof the microscopic group cross-sections is employed, to investigate the effects of space-dependency of the microscopicgroup cross-sections and feasibility of the old method.The differences of the effective neutron multiplication factor, burning region moving speed, spent fuel burnup andspatial distributions of nuclide densities, neutron fluence and power density may be considerable from the reactor designerpoint. However, they are small enough when we study about the characteristics of CANDLE burnup for differentdesigns.
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