Abstract: When a large commercial reprocessing plant has not been built, the use of burnup credit (BUC) technology can realize the intensive storage of spent fuel, which is helpful to alleviate the operating pressure of nuclear power plants. Compared with the traditional new fuel assumption, BUC technology involves a wide variety of nuclides and the numerical calculation of burnup is complex. The actual burnup is difficult to fully consider the actual depletion process of components in the reactor, Moreover, due to the approximation introduced by depletion chain decomposition, the randomness of nuclear reaction itself and the inherent error of numerical calculation method, it is difficult to obtain accurate nuclear fuel composition in burnup calculation. Therefore, it is necessary to quantify the uncertainty introduced in the depletion calculation reasonably and conservatively. In order to quantify the critical safety uncertainty transmitted from burnup calculation in the application of BUC, based on the parameter statistical method, 7 benchmark data of PWR chemical analysis published by ORNL and OCRWM, 56 groups of spent fuel PWR components, with enrichment coverage of 2.56%4.11% and burnup coverage of 11.547.3 GW·d/tU were selected. The experimental measured values of spent fuel samples used as the benchmark to compare with the calculated values of nuclides, the nuclide bias and bias uncertainty of burnup calculation were analyzed. Then, nuclide boundary method, Monte Carlo sampling and Latin hypercube sampling were used to modify nuclide. At the same time, the corrected nuclides were combined into different spent fuel assembly components. TRITON and CSAS25 modules were used with ENDF/BⅤ nuclear database for burnup calculation and criticality calculation respectively. The critical calculation results took the kinf uncertainty calculated by Monte Carlo sampling method as a reference. Subsequently, the effects of nuclide boundary method, Monte Carlo sampling and Latin hypercube sampling method on the critical calculation uncertainty were compared. The results show that, nuclides 235U, 238U, 239Pu, 240Pu, 241Pu and 242Pu can divide the burnup range of fuel assembly into three burnup subintervals, which include 1025 GW·d/tU, 2535 GW·d/tU and 3550 GW·d/tU. 234U, 238Pu and 241Am are divided into two fuel consumption subintervals, which include 1035 GW·d/tU and 3550 GW·d/tU. In the burnup subintervals, the nuclide concentration deviation of each spent fuel is distributed around the mean value by about 1δ. The nuclide deviation in the burnup sub interval is approximately constant. For the critical calculation, the kinf uncertainty calculation results of Latin hypercube sampling method and Monte Carlo sampling method are in good agreement. The Latin hypercube sampling method can consider the correlation between parameters, and the calculation results are more authentic, which can further improve the economy of the power plant.