CCC-0658 VIM3.6.
VIM4.0, Stead-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, FEATURES, AUXILIARIES, STATUS, REFERENCES, HARDWARE REQUIREMENTS, LANGUAGE, SOFTWARE REQUIREMENTS, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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3. DESCRIPTION OF PROGRAM OR FUNCTION
VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections. Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication).
Photon transport is simulated in the fixed source mode. The geometry options are infinite medium, combinatorial geometry, and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates. Boundary conditions include vacuum, specular and white reflection, and periodic boundaries for reactor cell calculations.
The VIM 4.0 distribution includes data from ENDF/B-IV, ENDF/B-V, ENDF/B-VI and JEF2.2. Binary sequential data libraries for use with the code system on IBM or Sun workstations are included. ASCII data libraries and a convenient means to convert them to binary on a target machine are included for users on other systems. In addition to be included in the RSICC distribution files, the VIM User Guide is available on the developer's web site http://www.ra.anl.gov/vimguide/.
VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections. Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication).
Photon transport is simulated in the fixed source mode. The geometry options are infinite medium, combinatorial geometry, and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates. Boundary conditions include vacuum, specular and white reflection, and periodic boundaries for reactor cell calculations.
The VIM 4.0 distribution includes data from ENDF/B-IV, ENDF/B-V, ENDF/B-VI and JEF2.2. Binary sequential data libraries for use with the code system on IBM or Sun workstations are included. ASCII data libraries and a convenient means to convert them to binary on a target machine are included for users on other systems. In addition to be included in the RSICC distribution files, the VIM User Guide is available on the developer's web site http://www.ra.anl.gov/vimguide/.
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4. METHODS
VIM uses standard Monte Carlo methods for particle tracking with several optional variance-reduction techniques. These include splitting/Russian roulette, non-terminating absorption with nonanalog weight cutoff energy. The keff is determined by the optimum linear combinations of two of the three eigenvalue estimates - analog, collision, and track length. Resonance and smooth cross sections are specified pointwise with linear-linear interpolation, frequently with many thousands of energy points. Unresolved resonances are described by the probability table method, which allows the statistical nature of the evaluated resonance cross sections to be incorporated naturally into self-shielding. Neutron interactions are elastic, inelastic and thermal scattering, (n,2n), fission, and capture, which includes (n,gamma), (n,p), (n,alpha), etc. Photon interaction data for pair production, coherent and incoherent scattering, and photoelectric events are taken from MCPLIB. Trajectories and scattering are continuous in direction, and anisotropic elastic and discrete level inelastic neutron scattering are described with probability tables derived from ENDF/B data. VIM has an automatic restart capability to permit user-directed statistical convergence.
In eigenvalue calculations, the beginning source sites are from a random (flat) guess, or can be provided via ASCII input, or from a previous calculation. The starting neutrons for each subsequent generation are randomly selected from the potential fission sites in the previous generation.
Track-length or collision estimates of reaction rates are automatically tallied by energy group and edit region to facilitate comparison to other calculations. Groupwise edits include isotopic and macroscopic reaction rates and cross sections, group-to-group scattering cross sections, net currents, and scalar fluxes. Particle pseudo-collisions are used to estimate microscopic group-to-group (n,2n), inelastic, and PN elastic scattering. The serial correlation of eigenvalue estimates is computed to detect underestimated errors.
VIM uses standard Monte Carlo methods for particle tracking with several optional variance-reduction techniques. These include splitting/Russian roulette, non-terminating absorption with nonanalog weight cutoff energy. The keff is determined by the optimum linear combinations of two of the three eigenvalue estimates - analog, collision, and track length. Resonance and smooth cross sections are specified pointwise with linear-linear interpolation, frequently with many thousands of energy points. Unresolved resonances are described by the probability table method, which allows the statistical nature of the evaluated resonance cross sections to be incorporated naturally into self-shielding. Neutron interactions are elastic, inelastic and thermal scattering, (n,2n), fission, and capture, which includes (n,gamma), (n,p), (n,alpha), etc. Photon interaction data for pair production, coherent and incoherent scattering, and photoelectric events are taken from MCPLIB. Trajectories and scattering are continuous in direction, and anisotropic elastic and discrete level inelastic neutron scattering are described with probability tables derived from ENDF/B data. VIM has an automatic restart capability to permit user-directed statistical convergence.
In eigenvalue calculations, the beginning source sites are from a random (flat) guess, or can be provided via ASCII input, or from a previous calculation. The starting neutrons for each subsequent generation are randomly selected from the potential fission sites in the previous generation.
Track-length or collision estimates of reaction rates are automatically tallied by energy group and edit region to facilitate comparison to other calculations. Groupwise edits include isotopic and macroscopic reaction rates and cross sections, group-to-group scattering cross sections, net currents, and scalar fluxes. Particle pseudo-collisions are used to estimate microscopic group-to-group (n,2n), inelastic, and PN elastic scattering. The serial correlation of eigenvalue estimates is computed to detect underestimated errors.
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6. TYPICAL RUNNING TIME
Varies widely, depending on geometric complexity, the number of isotopes, application of absorption weighting and splitting, overall scattering ratio, and desired statistics. A 6000-zone calculation of the LTR-IIa reactor keff to a one standard precision of 0.3% requires approximately 10 minutes on a Sun Sparc 20.
Varies widely, depending on geometric complexity, the number of isotopes, application of absorption weighting and splitting, overall scattering ratio, and desired statistics. A 6000-zone calculation of the LTR-IIa reactor keff to a one standard precision of 0.3% requires approximately 10 minutes on a Sun Sparc 20.
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13. SOFTWARE REQUIREMENTS
At RSICC the system was successfully tested on a Sun Solaris 2.6 UltraSparc 60 using F77 5.0 and C/C++ 5.0 and on an IBM RS/6000 Model 270 running AIX 4.3.3, XL Fortran 7.1 and XL C 4.4 XSEDIT calls DISSPLA to plot cross section data, although this coding can be easily removed.
At RSICC the system was successfully tested on a Sun Solaris 2.6 UltraSparc 60 using F77 5.0 and C/C++ 5.0 and on an IBM RS/6000 Model 270 running AIX 4.3.3, XL Fortran 7.1 and XL C 4.4 XSEDIT calls DISSPLA to plot cross section data, although this coding can be easily removed.
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Keywords: ENDF/B, Monte Carlo method, combinatorial geometry, criticality, gamma radiation, hexagonal lattices, multigroup theory, neutron, neutron transport equation, photon transport, rectangular, steady-state conditions, three-dimensional.