Computer Programs

NAME, COMPUTER, PROBLEM, SOLUTION, RESTRICTIONS, CPU, AUXILIARIES, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, AUTHOR, MATERIAL, CATEGORIES

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3. DESCRIPTION OF PROGRAM OR FUNCTION

COG is a modern, full-featured Monte Carlo radiation transport code that provides accurate answers to complex shielding, criticality, and activation problems.COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://cog.llnl.gov.

Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. A lattice feature simplifies the specification of regular arrays of parts. Parallel processing under MPI is supported for multi-CPU systems.

Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, pathlength stretching, point detectors, scattered direction biasing, and forced collisions.

Criticality – For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system.

Activation – COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source.

Coupled Problems – COG can solve coupled problems involving neutrons, photons, and electrons.

COG 11.1 Beta2 is an updated version of COG11 (RSICC C00777MNYCP00). New features in COG 11.1 Beta2 include:

A hybrid approach to detector score variance reduction in criticality problems;

Production and tracking of delayed fission gammas;

A treatment of nuclear resonance fluorescence;

Simulation of radiative decay;

A number of new data libraries.

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4. METHODS

COG is a high-resolution code for the Monte Carlo simulation of coupled neutron, proton, gamma-ray, and electron transport in arbitrary 3-D geometry. COG will transport neutrons with energies in the range of 10-5 eV to 150 MeV, protons with energies up to hundreds of GeV, and photons with energies in the range of 10 eV to 100 GeV. (COG’s energy ranges are limited by the available cross section sets and physics models). Via the EGS4 electron transport kernel, electrons in the range of 10 keV to a few thousand GeV can also be transported. The COG code is a significant upgrade from earlier Monte Carlo transport codes and has been written specifically to make it more versatile, accurate, and easier to use. COG has provisions for calculating deep penetration (shielding) problems, criticality problems, and neutron activation problems, and retains all of the standard capabilities found in other Monte Carlo transport codes. COG uses high-resolution pointwise cross-section databases and makes no compromises in the transport physics, so that the results of a COG run are limited only by the accuracy of the databases used. COG runs primarily on UNIX workstations – currently, UNIX workstations from Hewlett-Packard, Sun, Silicon Graphics, Compaq, and IBM are supported. There are also versions for the Power Macintosh and Linux/Intel PCs. COG can also be run concurrently over a network of processors or on an MPP to solve very large problems.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Size of problems may be limited by computer memory. COG will transport neutrons with energies in the range of 10-5 eV to 150 MeV, protons with energies up to hundreds of GeV, and photons with energies in the range of 10 eV to 100 GeV. (COG’s energy ranges are limited by the available cross section sets and physics models). Via the EGS4 electron transport kernel, electrons in the range of 10 keV to a few thousand GeV can also be transported.

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6. TYPICAL RUNNING TIME

Running time on a particular computer will vary widely, depending on geometric complexity, number of materials, and number of particles to be followed. For many problems, runs of a few minutes may give an approximate solution. For a statistically precise solution, hours of computation may be necessary. The parallel processing option is useful for shortening the run times of long problems.

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8. RELATED DATA LIBRARIES

Data libraries available for COG 11.1 Beta2:

Neutron Activation Library

ACTL92 LLNL's ACTL 1992

Photon Libraries

COGGXS defaults to EPDL97

EPDL89 LLNL's EPDL-1989

EPDL97 LLNL's EPDL-1997

Nuclear Resonance Fluorescence Library

COGNRF LLNL's Dr. James Hall data

Photonuclear Libraries

COGPNUC defaults to IAEAPNUC

IAEAPNUC IAEA

PN.ENDFB7R1 ENDF/B-VII.1

PN.MCNP.70u MCNP's .70u

Radiation Simulation Library

COGRS developed by LLNL's Dr. E. M. Lent

Delayed Fission Gamma Libraries developed by LLNL's Dr. E. M. Lent

DFG.ENDFB7R1 ENDF/B-VII.1

DFG.JEFF3.1.1 JEFF3.1.1

DFG.JENDL4 JENDL4

Neutron Libraries

ENDFB6R7 ENDF/B-VI.7

ENDFB6R8 ENDF/B-VI.8

ENDFB7R0 ENDF/B-VII.0

ENDFB7R0.BNL BNL's ENDF/B-VII.0

ENDFB7R1 ENDF/B-VII.1

ENDFB7R1.BNL BNL's ENDF/B-VII.1

ENDL2008 LLNL's ENDL-2008

ENDL90 LLNL's ENDL-1990

ENDL99 LNL's ENDL-1999

JEFF3.1 JEFF3.1

JEFF3.1.1 JEFF3.1.1

JEFF3.1.2 JEFF3.1.2

JENDL3.3 JENDL3.3

JENDL4 JENDL4

MCNP.50c MCNP's .50c

MCNP.51c MCNP's .51c

MCNP.55c MCNP's .55c

MCNP.66c MCNP's .66c

MCNP.70c MCNP's .70c

RED2002 Hybrid ENDFB/ENDL library by Dr. D. Cullen (2002)

Probability Table Libraries

PT.ENDFB7R0.BNL BNL's ENDF/B-VII.0

PT.ENDFB7R1.BNL BNL's ENDF/B-VII.1

PT.JEFF3.1 JEFF3.1

PT.JEFF3.1.1 JEFF3.1.1

PT.JEFF3.1.2 JEFF3.1.2

PT.MCNP.66c MCNP's .66c

PT.MCNP.70c MCNP's .70c

Thermal Libraries

T.ENDFB3R0 ENDF/B-III.0

T.ENDFB6R0 ENDF/B-VI.0

T.ENDFB6R2 ENDF/B-VI.2

T.ENDFB7R0 ENDF/B-VII.0

T.ENDFB7R0.BNL BNL's ENDF/B-VII.0

T.ENDFB7R0.LANL LANL's ENDF/B-VII.0

T.ENDFB7R1 ENDF/B-VII.1

T.ENDFB7R1.BNL BNL's ENDF/B-VII.1

T.JEF2.2 JEF2.2

T.JEFF3.0 JEFF3.0

T.JEFF3.1 JEFF3.1

T.JEFF3.1.1 JEFF3.1.1

T.JEFF3.1.2 JEFF3.1.2

Dosimetry Libraries

IRDF2002 IRDF-2002

IRDFF1.02 IRDFF Release 1.02

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10. REFERENCES

Background References:

R.J. Howerton, R.E. Dye, and S.T. Perkins, “Evaluated Nuclear Data Library,” Lawrence Livermore National Laboratory, Livermore, CA, UCRL-50400, Vol. 4, Rev. 1 (1981).

R. Kinsey, “Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF,” National Nuclear Data Center, Brookhaven National Laboratory, Upton NY, BNL-NCS-50496 (1979).

W. R. Nelson, H. Hirayama, and D. W. O. Rogers, “The EGS4 Code System,” SLAC-Report #265, Stanford University (December, 1985).

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Keywords: Monte Carlo method, activation, complex geometry, coupled, criticality, cross sections, electrons, gamma-ray, neutron, protons.