Abstract: According to the characteristics of pool-type sodium-cooled fast reactor, and with the fast reactor hydraulic model, thermal model and neutron kinetics model thoroughly classified and developed, a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool-type sodium-cooled fast reactor accident analysis. Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation. The calculation results are consistent with the test data and other fast reactor system analysis code results, and the correctness of the FASYS code is proved.