Abstract: Background: The standardized computer analysis for licensing evaluation (SCALE) developed in the Oak Ridge National Laboratory (ORNL) of USA has been widely used in criticality safety, neutron physics, radiation shielding, and sensitivity and uncertainty analysis. However, the burnup calculation for single-fluid, two-zone molten salt reactor (MSR) has not been well dealt with in SCALE/TRITON due to the cell information card (Celldata) which is used in unit cell calculations to generate problem-dependent multigroup cross sections. Purpose:This study aims to develop and evaluate possible solutions to the problem above. Methods: Based on external program, three methods (i.e., homogeneous mixing method, equivalent volume method and average cross section method), are developed without any modification of the existing codes in SCALE6 and they are tested in a MSR with two-zone core. Test results are compared and analyzed. Results: Comparison of the three methods indicates that the results obtained by average cross section method are almost equal to those obtained by homogeneous mixing method and moreover they accord well with the results given in ORNL's work, whilst the equivalent volume method is not sufficient to describe the difference of unit cells in the core. Conclusion: The average cross section method with a relatively high computational efficiency and accuracy is recommended for burnup calculation in the MSR with two or more zones when using SCALE/TRITON.

Key words: MSR, Burnup, Average cross section, Enriched uranium