Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, RELATED OR AUXILIARY PROGRAMS, STATUS, REFERENCES, HARDWARE REQUIREMENTS, LANGUAGE, SOFTWARE REQUIREMENTS, 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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DIF3D10.0 | CCC-0784/01 | Arrived | 14-NOV-2011 |

Machines used:

Package ID | Orig. computer | Test computer |
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CCC-0784/01 | Linux-based PC,UNIX W.S. |

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3. DESCRIPTION OF PROGRAM OR FUNCTION

The DIF3D10.0 release revises the significantly expanded set of solution techniques using variational nodal methods introduced with DIF3D8.0/VARIANT8.0 release (distributed as CCC-0649). The nodal option of DIF3D solves the multigroup steady-state neutron diffusion equation in two- and three-dimensional hexagonal and Cartesian geometries and solves the transport equation in two-and three-dimensional Cartesian geometries. Eigen value, 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 P99. 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 (distributed by RSICC as CCC-653) code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

The DIF3D10.0 release revises the significantly expanded set of solution techniques using variational nodal methods introduced with DIF3D8.0/VARIANT8.0 release (distributed as CCC-0649). The nodal option of DIF3D solves the multigroup steady-state neutron diffusion equation in two- and three-dimensional hexagonal and Cartesian geometries and solves the transport equation in two-and three-dimensional Cartesian geometries. Eigen value, 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 P99. 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 (distributed by RSICC as CCC-653) code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

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4. METHODS

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.

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 99th order are allowed. Partial current expansions up to 9th order are allowed. Angular expansions of up to P99 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 limit to performing the calculation. Recent improvements have mitigated this problem significantly, but large energy group calculations (>100 groups) are still limited.

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 99th order are allowed. Partial current expansions up to 9th order are allowed. Angular expansions of up to P99 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 limit to performing the calculation. Recent improvements have mitigated this problem significantly, but large energy group calculations (>100 groups) are still limited.

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8. RELATED OR 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 (distributed as CCC-0653) code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

DIF3D reads and writes the standard interface files specified by the Committee on Computer Code Coordination (CCCC). DIF3D is included in the REBUS-3 (distributed as CCC-0653) code package and can thus be used to provide the neutronics solutions required in REBUS-3 depletion calculations.

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CCC-0784/01, included references:

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

Diffusion Theory Problems, ANL-82-64, Argonne National Laboratory, Argonne, IL

(1984).

- R. D. Lawrence:

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

Theory Calculations in Hexagonal Geometry, ANL-83-1, Argonne National

Laboratory, Argonne, IL (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, Argonne National Laboratory,

Argonne, IL (October 1995).

- C. H. Adams, et al.:

The Utility Subroutine Package Used by Applied Physics Division Export Codes,

ANL-83-3, Argonne National Laboratory, Argonne, IL (May 1992).

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Package ID | Computer language |
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CCC-0784/01 | C-LANGUAGE, FORTRAN-90 |

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13. SOFTWARE REQUIREMENTS

No special requirements are made on the operating system. The included installation procedure requires Fortran 90 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 and Linux workstations. Although developed on the Cray and IBM 30xx, the current version is tailored to Sun, Linux and MacOSX platforms.

No special requirements are made on the operating system. The included installation procedure requires Fortran 90 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 and Linux workstations. Although developed on the Cray and IBM 30xx, the current version is tailored to Sun, Linux and MacOSX platforms.

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CCC-0784/01

source codesample problem input and output

code dependent BCD and binary card image file descriptions

scripts

a README file

an internal memorandum describing the revised Variant formulations (PDF)

an internal technical memorandum for the HMG4C homogenization software (PDF)

Keywords: CCCC, complex geometry, criticality, diffusion, finite difference method, multigroup, neutron diffusion equation, nodal method, reactor physics, transport theory.