The computer code DRAGON contains a collection of models which can simulate a nuclear reactor fuel assembly neutronic, photonic, electronic and positronic behaviours. 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; isotopic depletion calculations and finally photons, electrons and positrons transport capabilities.
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, NXT and SALT options 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 CLE-2000 supervisor. It is modular and can be interfaced easily with other production codes.