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Documentation
The computer code DRAGON contains a collection of models which can simulate the neutron 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 three such algorithms. The JPM option which solves the integral transport equation using the J± method, (interface current method applied to homogeneous blocks); the SYBIL option which solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D cartesian or hexagonal assemblies; and the EXCELL option which solves the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies.
The execution of DRAGON is controlled by the generalized GAN driver. It is modular and can be interfaced easily with other production codes.