Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES

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To submit a request, click below on the link of the version you wish to order. Rules for end-users are
available here.

Program name | Package id | Status | Status date |
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DIF3D 8.0/VARIANT8.0 | CCC-0649/002 | Arrived | 04-DEC-2001 |

Machines used:

Package ID | Orig. computer | Test computer |
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CCC-0649/002 | UNIX W.S. |

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3. DESCRIPTION OF PROGRAM OR FUNCTION

DIF3D solves multigroup diffusion theory eigenvalue, adjoint, fixed source and criticality (concentration search) problems in 1-, 2- and 3-space dimensions for orthogonal (rectangular or cylindrical), triangular and hexagonal geometries. Anisotropic diffusion coefficients are permitted. Flux and power density maps by mesh cell and regionwise balance integrals are provided. Although primarily designed for fast reactor problems, upscattering and internal black boundary conditions are also treated.

The DIF3D8.0/VARIANT8.0 release differs from the previous DIF3D7.0 release in that it includes a significantly expanded set of solution techniques using variational nodal methods. DIF3D's nodal option solves the multigroup steadystate neutron diffusion equation in two- and three-dimensional hexagonal and cartesian geometries and solves the transport equation in two-and three-dimensional cartesian geometries. Eigenvalue, adjoint, fixed source and criticality (concentration) search problems are permitted as are anisotropic diffusion coefficients. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, upscattering and for finite difference option only internal black boundary conditions are also treated.

VARIANT solves the multigroup steady-state neutron diffusion and transport equations in two- and three-dimensional Cartesian and hexagonal geometries using variational nodal methods. The transport approximations involve complete spherical harmonic expansions up to order P5. Eigenvalue, adjoint, fixed source, gamma heating, and criticality (concentration) search problems are permitted. Anisotropic scattering is treated, and although primarily designed for fast reactor problems, upscattering options are also included.

Related and Auxiliary Programs: DIF3D reads and writes the standard interface files specified by the Committee on Computer Code Coordination (CCCC). DIF3D is included in the REBUS-3 code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

DIF3D solves multigroup diffusion theory eigenvalue, adjoint, fixed source and criticality (concentration search) problems in 1-, 2- and 3-space dimensions for orthogonal (rectangular or cylindrical), triangular and hexagonal geometries. Anisotropic diffusion coefficients are permitted. Flux and power density maps by mesh cell and regionwise balance integrals are provided. Although primarily designed for fast reactor problems, upscattering and internal black boundary conditions are also treated.

The DIF3D8.0/VARIANT8.0 release differs from the previous DIF3D7.0 release in that it includes a significantly expanded set of solution techniques using variational nodal methods. DIF3D's nodal option solves the multigroup steadystate neutron diffusion equation in two- and three-dimensional hexagonal and cartesian geometries and solves the transport equation in two-and three-dimensional cartesian geometries. Eigenvalue, adjoint, fixed source and criticality (concentration) search problems are permitted as are anisotropic diffusion coefficients. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, upscattering and for finite difference option only internal black boundary conditions are also treated.

VARIANT solves the multigroup steady-state neutron diffusion and transport equations in two- and three-dimensional Cartesian and hexagonal geometries using variational nodal methods. The transport approximations involve complete spherical harmonic expansions up to order P5. Eigenvalue, adjoint, fixed source, gamma heating, and criticality (concentration) search problems are permitted. Anisotropic scattering is treated, and although primarily designed for fast reactor problems, upscattering options are also included.

Related and Auxiliary Programs: DIF3D reads and writes the standard interface files specified by the Committee on Computer Code Coordination (CCCC). DIF3D is included in the REBUS-3 code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

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4. METHOD OF SOLUTION

Optimized iteration methods for the solution of large-scale fast-reactor finite-difference diffusion theory calculations are used. The methods utilised include a variant of the Chebyshev acceleration technique applied to the outer fission source iterations and an optimised block successive overrelaxation method for the within-group iterations.

A nodal solution otpion intended for analysis of LMFBR designs in two- and three-dimensional hexagonal geometries in incorporated in the DIF3D package.

The neutron diffusion and transport equations are solved using a variational nodal method with one mesh cell (node) per hexagonal assembly (Cartesian geometry node sizes are specified by the user). The nodal equations are derived from a functional incorporating nodal balance, and reflective and vacuum boundary conditions through Lagrange multipliers. Expansion of the functional in orthogonal spatial and angular (spherical harmonics) polynomials leads to a set of response matrix equations relating partial current moments to flux and source moments. The equations are solved by fission source iteration in conjunction with a coarse mesh rebalance acceleration scheme. The inner iterations are accelerated by a partitioned matrix scheme equivalent to a synthetic diffusion acceleration method.

Optimized iteration methods for the solution of large-scale fast-reactor finite-difference diffusion theory calculations are used. The methods utilised include a variant of the Chebyshev acceleration technique applied to the outer fission source iterations and an optimised block successive overrelaxation method for the within-group iterations.

A nodal solution otpion intended for analysis of LMFBR designs in two- and three-dimensional hexagonal geometries in incorporated in the DIF3D package.

The neutron diffusion and transport equations are solved using a variational nodal method with one mesh cell (node) per hexagonal assembly (Cartesian geometry node sizes are specified by the user). The nodal equations are derived from a functional incorporating nodal balance, and reflective and vacuum boundary conditions through Lagrange multipliers. Expansion of the functional in orthogonal spatial and angular (spherical harmonics) polynomials leads to a set of response matrix equations relating partial current moments to flux and source moments. The equations are solved by fission source iteration in conjunction with a coarse mesh rebalance acceleration scheme. The inner iterations are accelerated by a partitioned matrix scheme equivalent to a synthetic diffusion acceleration method.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Problem dimensions are all variable. Enough memory must be allocated to contain all the information for at least one energy group. Flux and source expansions of up to sixth order are allowed. Partial current expansions up to second order are allowed. Angular and scattering expansions of up to P5 are allowed. The typical limiting factor for a problem lies in the storage of response matrices for problems involving large numbers of unique node types. For highly heterogeneous problems involving thousands of different node types, calculation and storage of response matrices represent the primary computational cost.

Problem dimensions are all variable. Enough memory must be allocated to contain all the information for at least one energy group. Flux and source expansions of up to sixth order are allowed. Partial current expansions up to second order are allowed. Angular and scattering expansions of up to P5 are allowed. The typical limiting factor for a problem lies in the storage of response matrices for problems involving large numbers of unique node types. For highly heterogeneous problems involving thousands of different node types, calculation and storage of response matrices represent the primary computational cost.

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6. TYPICAL RUNNING TIME

A three-dimensional nodal calculation for a small LMR with 60 degree planar symmetry, 9 energy groups, 14 axial mesh planes and 16 rings of hexagons required 22 CPU seconds on a Sun SPARCstation 20 (61 seconds on a SPARCStation 5, 18 seconds on an IBM RS/6000), to perform 14 outer iterations with 28 inners/outer and a convergence criteria of 10-6. All of the test cases completed in less than 5 minutes on a IBM RS/6000 Model 270 and on a UltraSparc 60.

A three-dimensional nodal calculation for a small LMR with 60 degree planar symmetry, 9 energy groups, 14 axial mesh planes and 16 rings of hexagons required 22 CPU seconds on a Sun SPARCstation 20 (61 seconds on a SPARCStation 5, 18 seconds on an IBM RS/6000), to perform 14 outer iterations with 28 inners/outer and a convergence criteria of 10-6. All of the test cases completed in less than 5 minutes on a IBM RS/6000 Model 270 and on a UltraSparc 60.

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10. REFERENCES

- R.D. Lawrence:

Progress in Nodal Methods for the Solution of the Neutron

Diffusion and Transport Equations

Prog. Nucl. Energy, 17, 3, 271 (1986)

- P.J. Finck and K.L. Derstine:

The Application of Nodal Equivalence Theory to Hexagonal Geometry

Lattices

Proceedings of the International Topical Meeting Advances in

Mathematics, Computations and Reactor Physics, Pittsburgh, Pa.,

Vol. 4, pp 16.1 4-1 (1991)

- D. O'Dell:

Standard Interface Files and Procedures for Reactor Physics Codes

Version IV

UC-32, Los Alamos Scientific Laboratory (September 1977)

- B.J. Toppel:

A Users Guide for the REBUS-3 Fuel Cycle Analysis Capability

Argonne National Laboratory, ANL-83-2 (1983)

- R.D. Lawrence:

Progress in Nodal Methods for the Solution of the Neutron

Diffusion and Transport Equations

Prog. Nucl. Energy, 17, 3, 271 (1986)

- P.J. Finck and K.L. Derstine:

The Application of Nodal Equivalence Theory to Hexagonal Geometry

Lattices

Proceedings of the International Topical Meeting Advances in

Mathematics, Computations and Reactor Physics, Pittsburgh, Pa.,

Vol. 4, pp 16.1 4-1 (1991)

- D. O'Dell:

Standard Interface Files and Procedures for Reactor Physics Codes

Version IV

UC-32, Los Alamos Scientific Laboratory (September 1977)

- B.J. Toppel:

A Users Guide for the REBUS-3 Fuel Cycle Analysis Capability

Argonne National Laboratory, ANL-83-2 (1983)

CCC-0649/002, included references:

- K. L. Derstine:DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite-Difference

Diffusion Theory Problems," ANL-82-64 (1984).

- R. D. Lawrence:

The DIF3D Nodal Neutronics Option for Two- and Three-Dimensional

Diffusion Theory Calculations in Hexagonal Geometry," ANL-83-1 (1983).

- G. Palmiotti, E. E. Lewis, and C. B. Carrico:

VARIANT: VARIational Anisotropic Nodal Transport for Multidimensional

Cartesian and Hexagonal Geometry Calculation," ANL-95/40 (October 1995).

- C. H. Adams, et.al.:

The Utility Subroutine Package Used by Applied Physics

Division Export Codes," ANL-83-3 (May 1992).

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11. MACHINE REQUIREMENTS

The modular version of the code is in production use at Argonne National Laboratory on Unix Workstations Sun SPARCStation. External data storage must be available for approximately 40 scratch and interface files. Fourteen of these files are random access scratch files (grouped into 6 file groups), and the remainder are sequential access files with formatted or unformatted record types.

The modular version of the code is in production use at Argonne National Laboratory on Unix Workstations Sun SPARCStation. External data storage must be available for approximately 40 scratch and interface files. Fourteen of these files are random access scratch files (grouped into 6 file groups), and the remainder are sequential access files with formatted or unformatted record types.

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Package ID | Computer language |
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CCC-0649/002 | FORTRAN-77, C-LANGUAGE |

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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED

No special requirements are made on the operating system (SOLARIS 2.6 for SPARCStations and AIX 4.3.3 on the IBM). The included installation procedure requires Fortran 77 and C compilers. With modifications the program can be executed entirely in FORTRAN. Optional dynamic memory allocation and timing routines supplied from host machine libraries or code in "C" may be used on Unix workstations. Although developed on the Cray and IBM 30xx, the current version is tailored to Sun SparcStations and IBM AIX RS/6000.

No special requirements are made on the operating system (SOLARIS 2.6 for SPARCStations and AIX 4.3.3 on the IBM). The included installation procedure requires Fortran 77 and C compilers. With modifications the program can be executed entirely in FORTRAN. Optional dynamic memory allocation and timing routines supplied from host machine libraries or code in "C" may be used on Unix workstations. Although developed on the Cray and IBM 30xx, the current version is tailored to Sun SparcStations and IBM AIX RS/6000.

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CCC-0649/002

649mfmws.txt Content of packageC649.pdf documentation in electronic format

C649tar1.gz Compressed tar file

Readme.rsi RSICC readme file

Keywords: anisotropic scattering, criticality searches, eigenvalues, fast reactors, hexagonal, multigroup, steady-state conditions, three-dimensional, two-dimensional, x-y-z.