last modified: 01-JUL-1980 | catalog | categories | new | search |

NEA-0596 FOCUS.

FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

top ]
1. NAME OR DESIGNATION OF PROGRAM:  FOCUS.
top ]
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
FOCUS NEA-0596/01 Tested 01-JUL-1980

Machines used:

Package ID Orig. computer Test computer
NEA-0596/01 IBM 3033 IBM 3033
top ]
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.
top ]
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.
top ]
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.
top ]
6. TYPICAL RUNNING TIME

Strongly dependent on the complexity of the problem, the particular quantity to be calculated and the statistical accuracy desired. Running time may vary from about 0.5 minutes to several hours (on the IBM 370).
top ]
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.
top ]
8. RELATED AND AUXILIARY PROGRAMS

ADX (2) is a code which calculates  adjoint cross sections and energy distributions from ENDF/B. The program ETOF (3) composes a system data file for FOCUS with all cross section data needed to treat a given system.
top ]
9. STATUS
Package ID Status date Status
NEA-0596/01 01-JUL-1980 Tested at NEADB
top ]
10. REFERENCES:
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).
top ]
11. MACHINE REQUIREMENTS

Core storage dependent on the complexity of the problem. In general 256 to 320 K bytes will be sufficient. Simultaneous access to up to 4 files on disc or tape, depending on the options selected.  Clock.
top ]
12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0596/01 FORTRAN+ASSEMBLER
top ]
13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:  IBM 370 OS/VS2.
top ]
14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
top ]
15. NAME AND ESTABLISHMENT OF AUTHOR

          J. E. Hoogenboom
          Delft University of Technology
          c/o Interuniversity Reactor Institute
          Makelweg 15
          2629 JB Delft, Netherlands.
top ]
16. MATERIAL AVAILABLE
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
top ]
17. CATEGORIES
  • C. Static Design Studies
  • J. Gamma Heating and Shield Design

Keywords: ENDF/B, Monte Carlo method, anisotropic scattering, cross sections, inelastic scattering, neutron flux, neutron transport theory, reaction kinetics, shielding, three-dimensional.