The computer code DRAGON contains a collection of models which can simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections which are supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The code DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a

source term. The current version of DRAGON contains many such algorithms. The SYBIL option which solves the integral transport equation using the collision probability method for simple one-dimensional (1–D) geometries (either plane, cylindrical or spherical) and the interface current method for 2–D Cartesian or hexagonal assemblies. The EXCELL option which solves the integral transport equation using the collision probability method for general 2–D geometries and for three-dimensional (3–D) assemblies. The MCCG option solves the integro-differential transport equation using the long characteristics method for general 2–D and 3–D geometries. The execution of DRAGON is controlled by the generalized GAN driver. It is modular and can be interfaced easily with other production codes.