VIM Continuous Energy Monte Carlo Transport Code

Conference · Sun Apr 30 00:00:00 EDT 1995

OSTI ID:459230

Blomquist, R. N. ^[1]

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1. Argonne National Laboratory (ANL), Argonne, IL (United States)

VIM is a continuous energy neutron and photon transport code. 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. VIM was developed primarily as a reactor criticality code. Its tally and edit features are very easy to use, and automatically provide fission, fission production, absorption, capture, elastic scattering, inelastic scattering, and (n,2n) reaction rates for each edit region, edit energy group, and isotope, as well as the corresponding macroscopic information, including group scalar fluxes. Microscopic and macroscopic cross sections, including microscopic P_N group-to-group cross sections are also easily produced.

• Research Organization: Argonne National Laboratory (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• OSTI ID: 459230
• Report Number(s): CONF-950420; TRN: 97:007131
• Resource Relation: Conference: International Conference on Mathematics and Computations, Reactor Physics, and Environmental Analyses, Portland, OR (United States), 30 Apr - 4 May 1995; Other Information: PBD: 1995; Related Information: Is Part Of Proceedings of the international conference on mathematics and computations, reactor physics, and environmental analyses. Volume 1 and 2; PB: 1629 p.
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICS AND COMPUTING
NEUTRON TRANSPORT
COMPUTERIZED SIMULATION
PHOTON TRANSPORT
REACTOR CORES
V CODES
MONTE CARLO METHOD
Nuclear Criticality Safety Program (NCSP)
Continuous Energy-Dependent ENDF Nuclear Data
Multigroup Cross Sections
Neutron Transport Criticality Mode
Fixed Source Mode
Fixed Source Mode Photon transport