Analysis of Nuclide Composition of Spent Fuel Waste Cladding Residue

In order to analyze the nuclide content of spent fuel waste cladding residue, a multilayer calculation model was established for M310 nuclear power plant and fuel assembly. The variation of ²⁴⁴Cm content, total Pu content and ²⁴⁴Cm/Pu ratio with burnup and cooling time was calculated by SCALE code. The results show that the variation of ²⁴⁴Cm content, total Pu content and ²⁴⁴Cm/Pu ratio with burnup and cooling time can be fitted by third-order polynomial. This study provides data support for non-destructive analysis method of waste cladding residue.