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CCC-0519 AUS.

AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

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1. NAME OR DESIGNATION OF PROGRAM:  AUS98.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
AUS87 CCC-0519/01 Tested 29-JAN-1990
AUS98 CCC-0519/02 Arrived 23-MAR-2001

Machines used:

Package ID Orig. computer Test computer
CCC-0519/01 IBM 360 series IBM 360 series
CCC-0519/02 UNIX W.S.
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3. DESCRIPTION OF PROGRAM OR FUNCTION

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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4. METHOD OF SOLUTION

AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream.  A simple path is simply a sequence of modules, but the path is actually preprocessed and compiled using the Fortran 77 compiler.  This provides for complex module linking if required.  Some of the modules included in AUS98 are:

MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation.

ANAUSN One-dimensional discrete ordinates calculation.

ICPP Isotropic collision probability calculation in one dimension and for rod clusters.

POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics.

AUSIDD One-dimensional diffusion calculation.

EDITAR Reaction-rate editing and group collapsing following a transport calculation.

CHAR Lattice and global burnup calculation.

MICBURN Control of global burnup calculation and fuel management using microscopic nuclide cross sections.

BURNMAC Global burnup calculation and fuel management using macroscopic material cross sections.

AUSED Cross-section editing and maintenance.

ORNL Forming cross sections for standard transport codes such as DORT and KENO.

MERGEL Merging of cross sections files.

PLOTXS Interactive plotting of cross section files.

AUSPLOT Interactive plotting of fluxes and reaction rates following a transport calculation.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM:  None noted.
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6. TYPICAL RUNNING TIME

Running times on a Silicon Graphics Power Challenge vary from one second for a simple cell calculation to one or two minutes for a three dimensional diffusion equation.  The longest running test case which includes six three dimensional diffusion calculation took 2.2 minutes.
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8. RELATED AND AUXILIARY PROGRAMS

Auxiliary Programs : BCDBINXS : Changer of cross section library from                      ASCII to binary.
                     BCDBINFP : Changer of fission product decay                      library from ASCII to binary.

Related Data Libraries : AUS98 includes a cross-section library with                          200 neutron and 37 photon groups and a                          fission product decay library with 869                          fission products.  Both libraries are based                          on ENDF/B-VI.
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9. STATUS
Package ID Status date Status
CCC-0519/01 29-JAN-1990 Screened
CCC-0519/02 23-MAR-2001 Masterfiled Arrived
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10. REFERENCES
CCC-0519/01, included references:
- G. S. Robinson :
  Notes on Implementing the AUS87 Code Package
  Infomal Notes (1987)
- G. S. Robinson :
  A Guide to the AUS Modular Neutronics Codes System
  AAEC/E645 (April 1987)
- G. S. Robinson :
  EDITAR : A Module for Reaction Rate Editing and  Cross-Section
  Averaging Within the AUS Neutronics Code System
  AAEC/E621 (March 1986)
- G. S. Robinson :
  ICPP : A Collection Probability Module for the AUS Neutronics
  Code System
  AAEC/E620 (October 1985)
- G. S. Robinson :
  CHAR and BURMAC : Burnup Modules of the AUS Neutronics Code System
  AAEC/E624 (March 1986)
- G. S. Robinson :
  MIRANDA : Module Based on Multiregion Resonance Theory for
  Generating Cross Sections within the AUS Neutronics Code System
  AAEDC/E626 (December 1985)
- G. S. Robinson :
  Extension of the AUS Reactor Neutronics System for Application to
  Fusion Blanket Neutronics
  AAEC/E583 (March 1984)
- B. E. Clancy :
  ANAUSN : A One-Dimensional Multigroup Transport Theory Module for
  the AUS Reactor Neutronics System
  AAEC/E539 (May 1982)
- B. V. Harrington :
  AUS Module AUSED : An Editing Program for AUS Cross-Section Data
  Pools
  AAEC/E389 (May 1976)
CCC-0519/02, included references:
- G.S.Robinson and B.V. Harrington
AUS98 - The 1998 Version of the AUS Modular Neutronics Code System
ANSTO/E734 (July 1998)
- B.V. Harrington, J.P. Pollard and J.M. Barry
POW3D - Neutron diffusion Module of the AUS System, A User's Manual
ANSTO/E726 (November 1996)
- G.S. Robinson
Generation and Validation of a Cross Section Library Based on ENDF/B-VI for the
AUS Neutronics Code System
ANSTO/E712 (December 1993)
- G.S. Robinson
AUS Module MIRANDA - A Data Preparation Code Based on Multiregion Resonance
Theory.
ANSTO/E410 (July 1977)
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11. MACHINE REQUIREMENTS

AUS98 is operable on SGI workstations under IRIX 6.4, Sun workstations under sunOS5.6, and should be operable on any UNIX workstation.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
CCC-0519/01 FORTRAN-77
CCC-0519/02 FORTRAN-77, C-LANGUAGE
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED

Compilation of AUS98 requires a fortran 77 compiler and an ANSI C standard compiler.  Execution also requires a Fortran 77 compiler.  The AUSPLOT module, which is a useful but not essential part of AUS98, requires a Tcl interpreter with Tk such as WISHX and the GNUPLOT interactive plotting program.
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by: Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.

Developed by:   Australian Nuclear Science and Technology
                Organisation
                Lucas Heights, Australia
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16. MATERIAL AVAILABLE
CCC-0519/02
C519tar1.gz: source code, script, data, test case input & output
C519tar1.lis list of files
readme.cd readme file
readme.rsi readme file
CCC-0519/01
File name File description Records
CCC0519_01.001 SVC Source (OS Supervisor call) 91
CCC0519_01.002 FORTLIB (Subroutine Library) 12854
CCC0519_01.003 AELINK1 Source (Aelink Program) 496
CCC0519_01.004 FORTCL (Compile and Load Module) 249
CCC0519_01.005 AUSED source (Module) 3508
CCC0519_01.006 ANAUSN Source (Module) 5489
CCC0519_01.007 AUSIDD Source (Module) 868
CCC0519_01.008 AUSYS Source (AUS Supervisor) 1107
CCC0519_01.009 BURNMAC Source (Module) 1216
CCC0519_01.010 CHAR Source (Char) 2400
CCC0519_01.011 EDITAR Source (Module) 5554
CCC0519_01.012 EXPAND Source (Module) 496
CCC0519_01.013 ICPP Source (Module) 9455
CCC0519_01.014 JOINER Source (Module) 137
CCC0519_01.015 MERGEL Source (Module) 210
CCC0519_01.016 MIRANDA Source (Module) 10156
CCC0519_01.017 ORNL Source (Module) 211
CCC0519_01.018 AUSEDCOM 69
CCC0519_01.019 CATPROC 123
CCC0519_01.020 LINKLIB Source (Load Standard Ausys Input) 76
CCC0519_01.021 RUN1 (Input for Test Case 1) 45
CCC0519_01.022 RUN2 (Input for Test Case 2) 76
CCC0519_01.023 RUN3 (Input for Test Case 3) 27
CCC0519_01.024 RUN4 (Input for Test Case 4) 52
CCC0519_01.025 RUN5 (Input for Test Case 5) 33
CCC0519_01.026 OUT1 (Output from Test Case 1) 609
CCC0519_01.027 OUT2 (Output from Test Case 2) 1717
CCC0519_01.028 OUT3 (Output from Test Case 3) 130
CCC0519_01.029 OUT4 (Output from Test Case 4) 1260
CCC0519_01.030 OUT5 (Output from Test Case 5) 448
CCC0519_01.031 ENDFB (128-Group Library) 445
CCC0519_01.032 ENDF200G (200+37)-Group Library 1450
CCC0519_01.033 Information file 108
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems
  • C. Static Design Studies
  • D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics
  • K. Reactor Systems Analysis
  • M. Data Management
  • X. Magnetic Fusion Research

Keywords: CTR, cell calculation, diffusion, multigroup, neutron.