IAEA0932: VIRGIN2007, Calculates Uncollided Neutron Flux and Neutron Reactions from Transmission in ENDF Format

Program-name Package-ID Status VIRGIN-86.1 IAEA0932/01 Obsolete VIRGIN-86.1 IAEA0932/02 Obsolete VIRGIN-89.1 IAEA0932/03 Obsolete VIRGIN-92.1 IAEA0932/04 Obsolete VIRGIN-94.1 IAEA0932/05 Obsolete VIRGIN-96.1 IAEA0932/06 Obsolete VIRGIN2000 IAEA0932/07 Obsolete VIRGIN2002 IAEA0932/08 Obsolete VIRGIN2004 IAEA0932/09 Obsolete VIRGIN2007 IAEA0932/10 Tested

Machines used: Package-ID Orig.Computer Test Computer IAEA0932/10 Many Computers Linux-based PC,PC Windows

3. DESCRIPTION OF PROGRAM OR FUNCTION - VIRGIN calculates uncollided flux and reactions due to transmission of a monodirectional beam of neutrons through any thickness of material. In order to simulate an experimental measurement the results are given as integrals over energy tally groups (as opposed to pointwise in energy).

IAEA0932/10:

This version include the updates up to JANUARY 30, 2007.
Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations.
The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI.

MODIFICATIONS FROM PREVIOUS VERSIONS:
-------------------------------------
Virgin VERS. 2007-1 (JAN. 2007): checked against all ENDF/B-VII; increased incore page size from 60,000 to 240,000

4. METHOD OF SOLUTION - By taking the ratio of reactions to flux in each group an equivalent spatially dependent group averaged cross section is calculated.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM - The evaluated data must be in the ENDF/B format. However it must be linear-linear interpolable in energy-cross section between tabulated points. Since only cross sections (file 3) are used, this program will work on any version of ENDF/B.

6. TYPICAL RUNNING TIME - A few seconds.

7. UNUSUAL FEATURES OF THE PROGRAM - The tally group structure may be any set of monotonically increasing energy boundaries. There is no limit to the number of tally groups. By using the input parameters the user may specify any arbitrary tally group structure.

8. RELATED AND AUXILIARY PROGRAMS -
The following programs are all part of the PREPRO2002 package.
ACTIVATE: is designed to create file 10 activation cross sections by
          combining file 3 cross sections and file 9 multipliers
COMPLOT: Compares ENDF/B formatted data from two separate input
         files. Results are in graphical form.
CONVERT: Automatically converts a FORTRAN program for use on
         any one of a variety of: (1) computers; (2) compilers:
         (3) precisions; (4) installations; (5) standard or
         non-standard file names.
DICTIN:  Creates a reaction index for each material.
EVALPLOT:Plots data in the ENDF/B format.
FIXUP:   Reads evaluated data in the ENDF/B format; performs
         corrections and outputs the results in the ENDF/B format.
GROUPIE: calculates unshielded group averaged cross sections,
         Bondarenko self-shielded group averaged cross sections,
         and multiband parameters from data in the ENDF/B format.
LEGEND:  Calculates linearly interpolable tabulated angular
         distributions starting from data in the ENDF/B format.
LINEAR:  Converts cross sections in the ENDF/B format (File 3,
         23, and 27) to linearly interpolable form (in energy
         and cross section) and outputs the result in the ENDF/B
         format.
MERGER:  Selectively retrieves data by MAT/MF/MT or ZA/MF/MT from
         up to 10 ENDF/B data tapes and merges the data into a
         single MAT/MF/MT ordered output file.
MIXER:   Calculates the energy dependent cross sections for a
         composite mixture.
RECENT:  Reconstructs energy-dependent cross sections from a
         combination of resonance parameters and tabulated
         background cross sections in the ENDF/B format.
RELABEL: relabels a ENDF/B preprocessing program so that
         statement labels are in increasing order in increments of
         10 within each routine.
SIGMA-1: Doppler broadens evaluated cross sections in the
         linear-linear interpolation form of the ENDF/B format.
SIXPAK:  Checks all double-differential ENDF/B-VI format data
         (MF=6) and outputs equivalent uncorrelated data (MF=4,
         5, 12, 14, and 15)
VIRGIN:  Calculates uncollided flux and reactions due to
         transmission of a monodirectional beam of neutrons
         through any thickness of material.

9. STATUS
IAEA0932/01: 29-MAY-1989 Obsolete
IAEA0932/02: 29-MAY-1989 Obsolete
IAEA0932/03: 28-AUG-1992 Obsolete
IAEA0932/04: 22-JUL-1994 Obsolete
IAEA0932/05: 07-APR-1997 Obsolete
IAEA0932/06: 11-DEC-2000 Obsolete
IAEA0932/07: 24-JUL-2003 Obsolete
IAEA0932/08: 29-SEP-2006 Obsolete
IAEA0932/09: 10-DEC-2007 Obsolete
IAEA0932/10: 07-DEC-2007 Tested at NEADB

10. REFERENCES -
IAEA0932/10:
- IAEA Nuclear Data Section:
PREPRO2007, 2007 ENDF/B Pre-processing Codes (ENDF/B-VII Tested)
IAEA-NDS-39, Rev. 13, March 17, 2007

11. MACHINE REQUIREMENTS - The package is designed to be used with many machines. Make files for almost all kind of computers are included.

12. PROGRAMMING LANGUAGE(S) USED -
IAEA0932/10: FORTRAN-77

13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED - Several OS can be used.

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
Nuclear Data Section
IAEA
P.O. Box 100
A-1400 VIENNA
AUSTRIA

16. MATERIAL AVAILABLE -
IAEA0932/10:
General description of the PrePro2007 programs
Documentation
Compressed files for MAC, Readme file for MAC
Author's Make file for PCLinux, Author's LINUX executable
Author's PCWindows compile batch
Author's PCWindows executable
Fortran source files for different platfor
UNIX Makefiles for different platforms
Author's VMS executable, Author's compile command file

17. CATEGORIES -

- M. Data Management

Keywords: DATA PROCESSING, ENDF/B, NEUTRON CROSS SECTIONS, NEUTRON FLUX, TRANSMISSION