DRAGON Code

About DRAGON

What is the code DRAGON

The DRAGON code[1, 2, 3] results from a concerted rationalization effort to unify inside a single code various numerical techniques and calculation methods that can be used to solve the neutron transport equation. The major concern of the DRAGON development team was to ensure that new calculation techniques could be easily implemented in the code. Accordingly the lattice code DRAGON is divided into modules which are linked together via the GAN generalized driver[4, 5]. Information between the various modules is transfered through well defined and documented data structures.[6]

The main components of DRAGON are:

  • a module for resonance self-shielding calculations;
  • a module to analyze various geometries and to generate a tracking file for collision probability evaluation;
  • a module for multigroup collision probability integration;
  • a module to solve the multigroup neutron transport equation using the collision probability method;
  • a module to solve the multigroup neutron transport equation using the method of characteristics;
  • an isotopic depletion module;
  • an editing module.

The EXCELL module represents the most versatile geometry analysis technique available in DRAGON. It can process 2-D clusters geometry as well as cartesian and hexagonal lattices of cells in both two and three dimensions.[7, 8, 9, 10, 11, 12] It can also take into account various types of boundary conditions including total or partial reflection at the cell edge or translation and reflection symetries. In addition it is the only module which is compatible both with the collision probability technique and the method of characteristics.[13,14] Two approximate collision probability integration techniques are also available in DRAGON, namely, the JPM[15, 16, 17, 18, 19, 20] and the SYBIL[21, 22] modules which are often used when faster but more approximate solutions to the transport equation are required.

Both multigroup flux solution modules in DRAGON, namely the FLU and MOC modules can analyze fixed sources or eigenvalue problems including buckling search with different leakage models.[23] They use the power iteration method and are accelerated with a multigroup rebalancing and a variational acceleration technique.[24]

Finally, DRAGON can process various format of microscopic cross-section libraries including:

It can also generate GOXS et ISOTXS format macroscopic cross-section libraries compatibles with codes such as TRANSX-CTR or TRANSX-2.[25, 32]

Copyrights for DRAGON

The development of DRAGON is financially supported, directly or indirectly, by various organizations including École Polytechnique de Montréal, Hydro-Québec and the Hydro-Québec chair in nuclear engineering, the Natural Science and Engineering Research Council of Canada (NSERC), Atomic Energy of Canada limited (AECL) and the CANDU Owners Group (COG). The code DRAGON and its users guide are and will remain the property of École Polytechnique de Montréal. The PostScript utility module used in DRAGON is based on PSPLOT which is owned by Kevin E. Kohler at the Nova Southeastern University Oceanographic Center in Florida.[33]

One can copy DRAGON free of charge. École Polytechnique de Montréal assumes no responsability, implicit or explicit, for the use of DRAGON.

References

[1] G. Marleau, R. Roy and A. Hébert, DRAGON: A Collision Probability Transport Code for Cell and Supercell Calculations, Report IGE-157, Institut de génie nucléaire, École Polytechnique de Montréal, Montréal, Québec (1994).
[2] G. Marleau, A. Hébert and R. Roy, ``New Computational Methods Used in the Lattice Code DRAGON'', Topical Meeting on Advances in Reactor Physics, pp 1.177-1.188, Charleston, South Carolina, March 8-11 1992.
[3] G. Marleau, A. Hébert and R. Roy, A User Guide for DRAGON. Version DRAGON_000331 Release 3.04, Report IGE-174 Rev. 5, Institut de génie nucléaire, École Polytechnique de Montréal, Montréal, Québec (2000).
[4] R. Roy and A. Hébert, The GAN Generalized Driver, Report IGE-158, Institut de génie nucléaire, École Polytechnique de Montréal, Montréal, Québec (2000).
[5] R. Roy, The CLE-2000 Tool-Box, Report IGE-163, Institut de génie nucléaire, École Polytechnique de Montréal, Montréal, Québec (1999).
[6] A. Hébert, G. Marleau and R. Roy, A Description of the DRAGON Data Structures. Version DRAGON_000331 Release 3.04, Report IGE-232 Rev. 3, Institut de génie nucléaire, École Polytechnique de Montréal, Montréal, Québec (2000).
[7] R. Roy, A. Hébert and G. Marleau, ``A Transport Method for Treating Three-Dimensional Lattices of Heterogeneous Cells'', Nucl. Sci. Eng., 101, 217-225 (1989).
[8] R. Roy, G. Marleau, J. Tajmouati and D. Rozon, ``Modeling of CANDU Reactivity Control Devices with the Lattice Code DRAGON'', Ann. nucl. Energy, 21, 115-132 (1994).
[9] R. Roy, ``Anisotropic Scattering for Integral Transport Codes. Part 1. Slab Assemblies'', Ann. nucl. Energy, 17, 379-388 (1990).
[10] R. Roy, ``Anisotropic Scattering for Integral Transport Codes. Part 2. Cyclic Tracking and its Application to XY Lattices'', Ann. nucl. Energy, 18, 511-524 (1991).
[11] R. Roy, A. Hébert and G. Marleau, ``A Cyclic Tracking Procedure for Collision Probability Calculations in 2-D Lattices'', International Topical Meeting on Advances in Mathematica, Computation and Reactor Physics, pp 2.2.4.1-2.2.4.14, Pittsburgh, Pennsylvania, April 28 - May 2, 1991.
[12] G. Marleau and R. Roy, ``Use of Specular Boundary Conditions for CANDU Cell Analysis'', Fourth International Conference on Simulation Methods in Nuclear Engineering, pp 5B.3.1-5B.3.13, Montréal, Québec, June 2-4, 1993.
[13] R. Roy, ``The Cyclic Characteristics Method'', International Conference on the Physics of Nuclear Science and Technology, Long Island, New York, October 5-8, 1998.
[14] R. Roy, ``The Cyclic Characteristics Method with Anisotropic Scattering'', M& C'99 Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, Madrid, Spain, September 27-30, 1999.
[15] F.E. Driggers, A Method for Calculating Neutron Absorption and Flux Spectra at Epithermal Energies, Report AECL-1996, Atomic Energy of Canada Limited, Chalk River, Ontario (1964).
[16] G. Marleau, M.L. Vergain, A. Hébert and R. Roy, ``Computation of the DP1 Collision Probabilities for Spherical and Cylindrical Geometries'', Ann. nucl. Energy, 17, 119-134 (1990).
[17] G. Marleau and A. Hébert, ``An Integral Transport Method for Treating CANDU and GCR Clusters'', Progress in Nuclear Energy, 18, 197-205 (1986).
[18] G. Marleau, R. Roy and A. Hébert, ``Analytic Reductions for Transmission and Leakage Probabilities in Finite Tubes and Hexahedra'', Nucl. Sci. Eng., 104, 209-216 (1990).
[19] M. Ouisloumen, G. Marleau, A. Hébert and R. Roy, ``Computation of Collision Probabilities for Mixed Hexagonal-Cylindrical Geometries Using the DP1 Approximation to the J± Technique'', International Topical Meeting on Advances in Mathematica, Computation and Reactor Physics, pp 2.2.1.1-2.2.1.12, Pittsburgh, Pennsylvania, April 28 - May 2, 1991.
[20] G. Marleau and A. Hébert, ``Analysis of Cluster Geometries Using the DP1 Approximation of the J± Technique'', Nucl. Sci. Eng., 111, 257-270 (1992).
[21] A. Hoffman et al., APOLLO: Code Multigroupe de résolution de l'équation du transport pour les neutrons thermiques et rapides, Note CEA-N-1610, Commisariat à l'énergie Atomique, France (1973).
[22] A. Hébert, Développement de la méthode SPH: Homogénéisation de cellules dans un réseau non uniforme et calcul des paramètres de réflecteur, Note CEA-N-2209, Commisariat à l'énergie Atomique, France (1981).
[23] G. Marleau and A. Hébert, ``Introduction of an Improved Critical Buckling Search in WIMS'', 1986 Simulation Symposium on Reactor Dynamics and Plant Control, Hamilton, Ontario, April 21-22, 1986.
[24] G. Marleau and A. Hébert, ``Solving the Multigroup Transport Equation Using the Power Iteration Method'', 1985 Simulation Symposium on Reactor Dynamics and Plant Control, Kingston, Ontario, April 22-23, 1985.
[25] R.E. Macfarlane, TRANSX-CTR: A code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis, Report LA-9863-MS, Los Alamos Scientific Laboratory, Los Alamos, New Mexico (1984).
[26] MATXS7A - 69 Neutron Group Cross Section Library in MATXS, DLC-117, RSIC Data Library Collection, Oak Ridge National Laboratory (1985).
[27] P. Vontobel and S. Pelloni, ``New JEF/EFF Based MATXS-Formatted Nuclear Data Libraries'', Nucl. Sci. Eng., 101, 298 (1989).
[28] J.R.Askew, F.J. Fayers and P.B. Kemshell, ``A General Description of the Lattice Code WIMS'', J. Brit. Nucl. Energy Soc., 5, 564, (1966).
[29] C.J. Taubman, The WIMS 69-Group Library Tape 166259, Report AEEW-M1324, U.K. Atomic Energy Authority, Winfrith (1975).
[30] J.J. Kim, J.T. Lee and H.R. Kim, ``Generation and Benchmarking of a 69 Group Cross Section Library for Thermal Reactor Applications'', J. Korean Nucl. Soc., 21, 245 (1989).
[31] J.V. Donnelly, WIMS-CRNL, A User's Manual for the Chalk River Version of WIMS, Report AECL-8955, Atomic Energy of Canada Limited, Chalk River, Ontario (1986).
[32] R.E. Macfarlane, TRANSX-2: A Code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes, Report LA-12312-MS, Los Alamos Scientific Laboratory, Los Alamos, New Mexico (1992).
[33] K.E. Kohler, PostScript for Technical Drawings PSPLOT: A FORTRAN-Callable PostScript Plotting Library User's Manual, Technical Report Nova Southeastern University, Oceanographic Center, 8000 North Ocean Drive, Dania, Florida; One can get a feel for the flavor of PSPLOT at http://www.nova.edu/ocean/ while access to the full psplot library is via anonymous ftp: whitetip.ocean.nova.edu (137.52.16.19) in the directory psplot.

 

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