USCD1234 DRAGON 3.05D.
DRAGON 3.05D, Reactor Cell Calculation System with Burnup
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RELATED OR AUXILIARY PROGRAMS, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| DRAGON 3.05D | USCD1234/01 | Tested | 18-MAY-2011 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| USCD1234/01 | Linux-based PC,UNIX W.S. | Linux-based PC |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
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 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.
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4. METHODS
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.
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.
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8. RELATED OR AUXILIARY PROGRAMS
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.
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.
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USCD1234/01, included references:
- Robert ROY and Alain HEBERT:The GAN Generalized Driver (March 2000) (IGE158.pdf)
- Robert ROY:
The CLE-2000 tool-box (December 1999) (IGE163.pdf)
- G. MARLEAU, A. HEBERT and R. ROY:
A User Guide for DRAGON 3.05D (Technical report IGE-174 Rev. 6D, March 2007)
- A. HEBERT, G. MARLEAU and R. ROY:
A Description of the Data Structures for DRAGON 3.05D
(Technical report IGE-232 Rev. 4D, March 2007)
- G. MARLEAU:
The Excell Geometries Numbering Scheme in Dragon
(Technical report IGE-233, December 1997)
- G. MARLEAU:
DRAGON Theory Manual - Part 1: Collision Probability Calculations
(Technical report IGE-236 Revision 1, October 2001)
- G. MARLEAU and E. VARIN:
An History Interface Between Dragon And Donjon:
The Hst: Module
(Technical report IGE-259 R1, December 2004)
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USCD1234/01
This package was tested at the NEA Databank on:- COMPUTER : Apple iMac i7 2800MHz
- OPERATING SYSTEM : Ubuntu Linux 9.10 Kernel 2.6
- COMPILER : fort77 Version 1.15
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USCD1234/01
COPYRIG~.TXT: text file containing the DRAGON copyrightREADME.TXT: the file you are currently reading
Guides: directory containing the IGE reports associated with DRAGON, GANLIB and
CLE-2000
Sources: source code for DRAGON and GANLIB and scripts for the dragon
installation
Tests: test cases and dragon running scripts
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- B. Spectrum Calculations, Generation of Group Constants and Cell Problems
- C. Static Design Studies
- K. Reactor Systems Analysis
Keywords: burnup, cell calculation, depletion, fuel management, neutron transport equation.