NEA-0596 FOCUS.
FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| FOCUS | NEA-0596/01 | Tested | 01-JUL-1980 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-0596/01 | IBM 3033 | IBM 3033 |
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3. DESCRIPTION OF PROBLEM OR FUNCTION
FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space.
Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and eigenfunction problems in which the neutron source distribution is given by the (unknown) fundamental mode eigenfunction distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included.
FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space.
Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and eigenfunction problems in which the neutron source distribution is given by the (unknown) fundamental mode eigenfunction distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included.
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4. METHOD OF SOLUTION
A tranformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for eigenfunction problems.
Completely arbitrary problems can be handled by defining a neutron source and/or neutron detector in simple user-written subroutines.
Importance sampling devices such as splitting, Russian roulette and path length stretching depending on energy and space region are available.
A tranformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for eigenfunction problems.
Completely arbitrary problems can be handled by defining a neutron source and/or neutron detector in simple user-written subroutines.
Importance sampling devices such as splitting, Russian roulette and path length stretching depending on energy and space region are available.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
Due to array dimensions the number of different cross section media in a system is limited to 16. Each medium can contain at most 10 different nuclides. The total number of different nuclides in the system is limited to 100. At most 9 fissionable nuclides are allowed in the system. No limits apply to the cross section data or geometry description.
Due to array dimensions the number of different cross section media in a system is limited to 16. Each medium can contain at most 10 different nuclides. The total number of different nuclides in the system is limited to 100. At most 9 fissionable nuclides are allowed in the system. No limits apply to the cross section data or geometry description.
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7. UNUSUAL FEATURES OF THE PROGRAM
An unusual feature of FOCUS as an adjoint Monte Carlo code is its ability to treat eigenfunction problems. An equivalent treatment of a one-velocity thermal group is introduced. Due to a strong control of the sequence of random numbers per particle history differences in estimated quantities from two systems due to (small) differences in geometry or cross section can be calculated with relatively small standard deviation.
An unusual feature of FOCUS as an adjoint Monte Carlo code is its ability to treat eigenfunction problems. An equivalent treatment of a one-velocity thermal group is introduced. Due to a strong control of the sequence of random numbers per particle history differences in estimated quantities from two systems due to (small) differences in geometry or cross section can be calculated with relatively small standard deviation.
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NEA-0596/01, included references:
- J.E. Hoogenboom:"FOCUS - A Versatile Non-Multigroup Adjoint Monte Carlo Neutron
Transport Code"
IRI-131-77-06/ THD-H-RF-144 (1979).
- J.E. Hoogenboom:
"ETOF - A Program to Prepare a Cross Section Data Tape from the
ENDF/B File for the Adjoint Monte Carlo Code FOCUS"
IRI-131-77-05/THD-H-RF-146 (1979).
- J.E. Hoogenboom and P.F.A. de Leege:
"ADX - A Code to Calculate Adjoint Cross Sections from the ENDF/B
File"
IRI-131-77-04/THD-H-RF-145 (1979).
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NEA-0596/01
| File name | File description | Records |
|---|---|---|
| NEA0596_01.001 | FOCUS SOURCE (F4,EBCDIC) | 2197 |
| NEA0596_01.002 | ETOF SOURCE (F4,EBCDIC) | 2229 |
| NEA0596_01.003 | ADX SOURCE (F4,EBCDIC) | 1834 |
| NEA0596_01.004 | ENDF/B RETRIEVAL ROUTINES (F4) | 1729 |
| NEA0596_01.005 | ICLOCK ASSEMBLER (EBCDIC) | 92 |
| NEA0596_01.006 | DATUM ASSEMBLER (EBCDIC) | 77 |
| NEA0596_01.007 | XEDIT SOURCE (F4,EBCDIC) | 307 |
| NEA0596_01.008 | RIGEL SOURCE (F4,EBCDIC) | 2003 |
| NEA0596_01.009 | SPHERE GEOM MODULE (F4) | 178 |
| NEA0596_01.010 | SLAB GEOM MODULE (F4) | 254 |
| NEA0596_01.011 | CYLINDER GEOM MODULE (F4) | 440 |
| NEA0596_01.012 | GENERAL GEOM MODULE (F4) | 1127 |
| NEA0596_01.013 | AND/IAND ASSEMBLER | 19 |
| NEA0596_01.014 | LOC | 6 |
| NEA0596_01.015 | IAND/AND | 19 |
| NEA0596_01.016 | IOR/OR | 19 |
| NEA0596_01.017 | IXCOR/XCOR | 19 |
| NEA0596_01.018 | ICOMPL/COMPL | 18 |
| NEA0596_01.019 | RANDOM NUMBER MODULES ASSEMBLER | 485 |
| NEA0596_01.020 | RMIDSQ RANDOM NUMBER GEN. ASSEMBLER | 52 |
| NEA0596_01.021 | USER ROUTINES FOR 1. SAMPLE PROBLEM | 154 |
| NEA0596_01.022 | USER ROUTINES FOR 2. SAMPLE PROBLEM | 57 |
| NEA0596_01.023 | USER ROUTINES FOR 3. S.P. (NORMALISATION) | 108 |
| NEA0596_01.024 | ENDF/B FILE FOR SAMPLE PROBLEM | 127 |
| NEA0596_01.025 | ADX CARD INPUT | 8 |
| NEA0596_01.026 | RIGEL CARD INPUT | 9 |
| NEA0596_01.027 | ENDF/B WITH ADJOINT DATA FOR S.P. | 385 |
| NEA0596_01.028 | ETOF CARD INPUT FOR 1ST S.P. | 5 |
| NEA0596_01.029 | ETOF CARD INPUT FOR 2ND S.P. | 13 |
| NEA0596_01.030 | FOCUS CARD INPUT FOR 1ST S.P. | 30 |
| NEA0596_01.031 | FOCUS CARD INPUT FOR 2ND S.P. | 30 |
| NEA0596_01.032 | FOCUS CARD INPUT FOR 3RD S.P. (NORMAL.) | 30 |
| NEA0596_01.033 | ADX CARD OUTPUT LOGICAL UNIT 21 | 34 |
| NEA0596_01.034 | ADX CARD OUTPUT LOGICAL UNIT 22 | 35 |
| NEA0596_01.035 | ADX CARD OUTPUT LOGICAL UNIT 23 | 211 |
| NEA0596_01.036 | ADX PRINTED OUTPUT | 57 |
| NEA0596_01.037 | RIGEL PRINTED OUTPUT | 50 |
| NEA0596_01.038 | ETOF PRINTED OUTPUT 1ST S.P. | 37 |
| NEA0596_01.039 | ETOF PRINTED OUTPUT 2ND S.P. | 58 |
| NEA0596_01.040 | XEDIT PRINTED OUTPUT 1ST S.P. | 138 |
| NEA0596_01.041 | FOCUS PRINTED OUTPUT 1ST S.P. | 353 |
| NEA0596_01.042 | FOCUS PRINTED OUTPUT 2ND S.P. | 254 |
| NEA0596_01.043 | FOCUS PRINTED OUTPUT 3RD S.P. (NORMAL.) | 400 |
| NEA0596_01.044 | FOCUS JOB CONTROL & INFORMATION | 500 |
Keywords: ENDF/B, Monte Carlo method, anisotropic scattering, cross sections, inelastic scattering, neutron flux, neutron transport theory, reaction kinetics, shielding, three-dimensional.