### 2.1.32 - microscopic group constant calculation with set mdep

Posted:

**Mon Mar 01, 2021 9:04 am**The methodology, fourth input option and the output was changed.

Multiple set mdep cards per universe can be given. The outputs for different set mdep cards are given in XS_<uni>_<idx> tables, whre <idx> is the index of the set mdep card.

In the microscopic cross section output file, fourth and fifth columns are now the nuclide densities smeared to the homogenized volume with respective uncertainty. Sixth column and onwards are the cross sections with respective uncertainties previously present in column four and onwards. Values of column four times values of columns six, eight, ... are now equal to the macroscopic cross section of the nuclide in the homogenized volume (calculated for the selected materials).

Old parameter VR is now VOL, which should be equal to the volume of the homogenized universe.

The definition of fission yields for MT 18 reaction variants 181, 182, 183, ... is now different than earlier.

The microscopic cross section output file contains all fission yield tables and decay data.

See: https://serpent.vtt.fi/mediawiki/index. ... l#set_mdep.

Multiple set mdep cards per universe can be given. The outputs for different set mdep cards are given in XS_<uni>_<idx> tables, whre <idx> is the index of the set mdep card.

In the microscopic cross section output file, fourth and fifth columns are now the nuclide densities smeared to the homogenized volume with respective uncertainty. Sixth column and onwards are the cross sections with respective uncertainties previously present in column four and onwards. Values of column four times values of columns six, eight, ... are now equal to the macroscopic cross section of the nuclide in the homogenized volume (calculated for the selected materials).

Old parameter VR is now VOL, which should be equal to the volume of the homogenized universe.

The definition of fission yields for MT 18 reaction variants 181, 182, 183, ... is now different than earlier.

The microscopic cross section output file contains all fission yield tables and decay data.

See: https://serpent.vtt.fi/mediawiki/index. ... l#set_mdep.