Reaction extract

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The natural function of the FISPACT-II collapse is to reduce the full multi-group cross section data and a user-supplied incident particle spectrum into one-group effective cross sections to obtain reaction rates. For many reasons it may be desirable to have the full energy-dependent reaction rates and energy-dependent uncertainties available for visualisation and/or checking purposes.

A separate utility programme, extract_xs_endf allows the user to specify specific reactions and extract a variety of data on the energy-dependant reaction rates. It takes in the following arguments:

  1. fileroot name for the output, which will generate fileroot.out and fileroot.log
  2. projectile, given as a letter from the choices of: n, p, d, a, g
  3. energy group of the nuclear data, taken from the possible 1102, 709 or 162 group structures
  4. parent nuclide, which is the target nuclide for the reaction
  5. the mt number of the reaction, which must be drawn from the allowable values in the FISPACT-II mt list
  6. daughter nuclide, which is the product of the specific reaction - note that isomers are defined in the usual way, e.g. In116m
  7. [optional] the name of the files file (default files)

The files file must include a nuclide index file ind_nuc, incident particle spectrum fluxes file and the directory for the cross section data xs_endf, for example:

# Index of nuclides to be included
ind_nuc  /path/to/fispact/ENDFdata/TENDL2015data/tendl15_decay12_index
# Incident particle spectrum
fluxes   /some/data/directory/my_fluxes
# Library cross section data
xs_endf   /path/to/fispact/ENDFdata/TENDL2015data/tal2015-n/gxs-709

An example execution would be:

extract_xs_endf U238_capture n 709 U238 102 U239

which would generate U238_capture.out with an eight column output including:

  1. En-low = lower energy of the group
  2. En-high = higher energy of the group
  3. flux = group flux(i), as in the flux file
  4. flux-unc = 0.000000E+00 (this feature is turned off in distributed versions)
  5. gxs = group cross section, as in the TENDL file
  6. gxs_unc = mapped variance one sigma of the group cross section, in %
  7. greac-rate= flux(i)*gxs(i)/flux(1:709)
  8. cum-rate = incremental sum of the greaction-rates, in %