Abstract: Based on the linear nuclide chain method and the burn-up features of reactors, the depletion nuclide database is established and the densities of all the nuclides in the burnup system are calculated using the backtracking algorithm. The depletion code can solve the burn-up problem independently and have integrated output form, which makes it convenient to couple with the Monte Carlo program by the interface module to do the coupling transport-depletion calculation for reactor systems. The benchmark burn-up of the first core of China Experimental Fast Reactor(CEFR) is calculated with the transport-depletion code to verify the function correctness. The comparison with the results obtained already using other codes shows that the new depletion program can calculate the density of actinides with high fidelity, including fissionable nuclides and some minor actinides with puny yields. There is difference with the inventory of Pu241. This should be analyzed further to determinate whether the difference comes from the depletion program or from the related cross section data. The calculation results of fission products and reactivity tally with the designed results well.