Study on three-dimensional burnup calculation based on Monte Carlo method

To calculate burnup for the cores with fast neutron spectrum, authors have written the linking codes to process the input and output of MCNP code and ORIGEN2 code. The code package which includes MCNP, ORIGEN2, the linking codes and cross sections files can be used to calculate the fuel isotopes compositions and burnup reactivity loss for a fast reactor core with heterogeneous configuration. The pseudo fission products are adopted to calculate burnup reactivity loss. The burnup results of the first core of China Experimental Fast Reactor and the benchmark on MOX fuel fast reactor of OECD/NEA, which are calculated with the code package, are presented in this paper. (author)