|Program name||Package id||Status||Status date|
|Package ID||Orig. computer||Test computer|
|USCD1234/01||Linux-based PC,UNIX W.S.||Linux-based PC|
More information on DRAGON3 is available from the developer's home page at https://www.polymtl.ca/phys/en/dragon-download
The computer code DRAGON contains a collection of models that 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 supplied by means of standard libraries; resonance selfshielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that 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 flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that 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; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries.
The execution of DRAGON is managed via the GAN generalized driver. The code is modular and can be interfaced easily with other production codes the finite reactor code DONJON.
DRAGON can access directly microscopic cross-section libraries defined according to the following standard formats: DRAGLIB, MATXS, WIMS-D4 and WIMS-AECL. It has the capability of exchanging macroscopic cross-section libraries with a code such as TRANSX-CTR or TRANSX-2 by the use of GOXS and ISOTXS format files. The macroscopic cross section can also be read in DRAGON via the input data stream.
|Package ID||Status date||Status|
|USCD1234/01||18-MAY-2011||Tested at NEADB|
|Package ID||Computer language|
Keywords: burnup, cell calculation, depletion, fuel management, neutron transport equation.