8. RELATED AND AUXILIARY PROGRAMS
The following programs are all part of the PREPRO2010 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
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
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).
SPECTRA: Convert model and general tabulation to linearized spectra (MF=5).
VIRGIN: Calculates uncollided flux and reactions due to transmission of a
monodirectional beam of neutrons through any thickness of material.