AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.
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ROBINSON, GRAHAM S. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems..
Computer software. Vers. 00. USDOE. 23 Jul. 1999.
Web.
ROBINSON, GRAHAM S. (1999, July 23). Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. (Version 00) [Computer software].
ROBINSON, GRAHAM S. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems..
Computer software. Version 00. July 23, 1999.
@misc{osti_1256869,
title = {Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems., Version 00},
author = {ROBINSON, GRAHAM S.},
abstractNote = {AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.},
doi = {},
url = {https://www.osti.gov/biblio/1256869},
year = {Fri Jul 23 00:00:00 EDT 1999},
month = {Fri Jul 23 00:00:00 EDT 1999},
note =
}
