NEA-1591 OMEGA.
OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS, CPU, FEATURES, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| OMEGA | NEA-1591/01 | Tested | 12-MAR-1999 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-1591/01 | IBM PC | PC Pentium II 400 |
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3. DESCRIPTION OF PROBLEM OR FUNCTION
OMEGA is a Monte Carlo code for the solution of the stationary neutron transport equation with k-eff as the eigenvalue. A three-dimensional geometry is permitted consisting of a very general arrangement of three basic shapes
(columns with circular, rectangular, or hexagonal cross section with a finite height and different material layers along their axes). The main restriction is that all the basic shapes must have parallel axes. Most real arrangements of fissile material inside and outside a reactor (e.g., in a fuel storage or transport container) can be described without approximation. The main field of application is the estimation of criticality safety. Many years of experience and comparison with reference cases have shown that the code together with the built-in cross section libraries gives reliable results. The following results can be calculated:
- the effective multiplication factor k-eff
- the flux distribution
- reaction rates
- spatially and energetically condensed cross sections for later use in a subsequent OMEGA run. A running job may be interrupted and continued later, possibly with an increased number of batches for an improved statistical accuracy. The geoetry as well as the k-eff results may be visualized. The use of the code is demonstrated by many illustrating examples.
OMEGA is a Monte Carlo code for the solution of the stationary neutron transport equation with k-eff as the eigenvalue. A three-dimensional geometry is permitted consisting of a very general arrangement of three basic shapes
(columns with circular, rectangular, or hexagonal cross section with a finite height and different material layers along their axes). The main restriction is that all the basic shapes must have parallel axes. Most real arrangements of fissile material inside and outside a reactor (e.g., in a fuel storage or transport container) can be described without approximation. The main field of application is the estimation of criticality safety. Many years of experience and comparison with reference cases have shown that the code together with the built-in cross section libraries gives reliable results. The following results can be calculated:
- the effective multiplication factor k-eff
- the flux distribution
- reaction rates
- spatially and energetically condensed cross sections for later use in a subsequent OMEGA run. A running job may be interrupted and continued later, possibly with an increased number of batches for an improved statistical accuracy. The geoetry as well as the k-eff results may be visualized. The use of the code is demonstrated by many illustrating examples.
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4. METHOD OF SOLUTION
The Monte Carlo method is used with neutrons starting from an initial source distribution. The histories of a generation (or batch) of neutrons are followed from collision to collision until the histories are terminated by capture, fission, or leakage. For the solution of the eigenvalue problem, the starting positions of the neutrons for a given generation are determined by the fission points of the preceding generation. The summation of the results starts only after some initial generations when the spatial part of the fission source has converged. At present the code uses the BNAB-78 subgroup library of the Obninsk data group (26 energy groups with subgroups in the resonance region) for epithermal and fast neutron energies. Spatially dependent resonance self-shielding is accounted for by the so-called subgroup method using the information given by the subgroup parameters of the library. A THERMOS type library is used for thermal energies where upscattering and molecular binding effects (e.g., for light water) are taken into account. The reliability of the data base for criticality problems was shown by many comparisons with reference cases.
The Monte Carlo method is used with neutrons starting from an initial source distribution. The histories of a generation (or batch) of neutrons are followed from collision to collision until the histories are terminated by capture, fission, or leakage. For the solution of the eigenvalue problem, the starting positions of the neutrons for a given generation are determined by the fission points of the preceding generation. The summation of the results starts only after some initial generations when the spatial part of the fission source has converged. At present the code uses the BNAB-78 subgroup library of the Obninsk data group (26 energy groups with subgroups in the resonance region) for epithermal and fast neutron energies. Spatially dependent resonance self-shielding is accounted for by the so-called subgroup method using the information given by the subgroup parameters of the library. A THERMOS type library is used for thermal energies where upscattering and molecular binding effects (e.g., for light water) are taken into account. The reliability of the data base for criticality problems was shown by many comparisons with reference cases.
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NEA-1591/01
The batch file sample.bat executes the ten examples inp1,...,inp10 (see report VKTA-36/May 1996).A single case (say inp8) may be started by omega inp8.
test 1 : 0.94 min.
test 2 : 0.74 min.
test 3 : 0.43 min.
test 4 : 0.80 min.
test 5 : 1.52 min.
test 6 : 1.74 min.
test 7 : 0.63 min.
test 8 : 2.12 min.
test 9 : 1.70 min.
test 10 :1.73 min.
The batch file all16.bat executes the 16 examples cas01,...cas16 (see report VKTA-54/April 1998).
cas 01 : 0.66 min.
cas 02 : 19.56 min.
cas 03 : 3.14 min.
cas 04 : 2.04 min.
cas 05 : 5.59 min.
cas 06 : 6.72 min.
cas 07 : 3.17 min.
cas 08 : 53.89 min.
cas 09 : Stopped on error (SQRT argument negative in module transp)
cas 10 : 58.33 min.
cas 11 : 4.10 min.
cas 12 : 3.80 min.
cas 13 : 3.75 min.
cas 14 : 3.50 min.
cas 15 : 3.70 min.
cas 16 : Program stops without any error message.
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NEA-1591/01, included references:
- E. Seifert:A PC Version of the Monte Carlo Criticality Code OMEGA.
VKTA-36 (May 1996)
- E. Seifert:
MCNP and OMEGA Criticality Calculations.
VKTA-54 (April 1998)
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NEA-1591/01
File storage is not more than 10 MByte disk memory.[ top ]
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NEA-1591/01
ABSTRACT.TXT Program abstractEXAMPLES/ACOMMENT.TXT Directory description
EXAMPLES/EXAMPLE7.SET
EXAMPLES/INP* 10 input files for samples
EXAMPLE/INP*.CON CON files for 10 samples
EXAMPLE/INP*.OUT Output for 10 samples
EXAMPLES/OMEGA.BAT Batch file
EXAMPLES/SAMPLES.BAT Batch file for executing 10 samples
SRC/COMPIL1.BAT Batch file for OMEGA compilation
SRC/COMPIL2.BAT Batch file for FLIX compilation
SRC/FASTLIB Fast library (binary form)
SRC/FLIX.BAT Batch file
SRC/FLIX.EXE Executable
SRC/FLIX.FOR FLIX source
SRC/LIBRARY.BAT Batch file for library conversion
SRC/OMEGA.BAT Batch file
SRC/OMEGA.EXE Executable
SRC/OMEGA.FOR OMEGA source
SRC/SL25CARD Fast library (ASCII form)
SRC/SLCONV.EXE Executable
SRC/SLCONV.FOR Source of fast library conversion (ASCII to binary)
SRC/SLPROT
SRC/TERMLIB Thermal library (binary form)
SRC/TL21CARD Thermal library (ASCII form)
SRC/TLCONV.EXE Executable
SRC/TLCONV.FOR Source of thermal library conversion (ASCII to binary)
SRC/TLPROT
TEST1/ACOMMENT.TXT Directory description
TEST1/X10.SET
TEST1/TESTJOB1.BAT Batch file for executing the 1th test job
TEST1/X1
TEST1/X2
TEST1/X3
TEST1/X1.OUT
TEST1/X2.OUT
TEST1/X3.OUT
TEST1/X10.OUT
TEST1/X20.OUT
TEST1/X30.OUT
TEST1/X1.SET
TEST1/TEST1.DIF Dif file
TEST2/ACOMMENT.TXT Directory description
TEST2/EXAMPLE7.SET
TEST2/INP* 10 input files for samples
TEST2/INP*.OUT 10 output files for samples
TEST2/INP7FLIX
TEST2/OUT* 10 output files
TEST2/TEST2.BAT Batch file for executing the 2th test job
TEST2/TEST2.DIF Dif file
VKTA54/CAS* 16 input files
VKTA54/ACOMMENT.TXT Directory description
WORK/OMEGA.BAT Batch file for starting OMEGA
WORK/FLIX.BAT Batch file for starting FLIX
PKUNZIP.EXE Unzipping file
README.TXT Information file
Keywords: Monte Carlo method, criticality, flux distribution, neutron transport equation, reaction rates, three-dimensional.