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Adam Nelson 2016-10-10 16:10:39 -04:00
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8
docs/source/io_formats/mgxs_library.rst
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@ -6,7 +6,7 @@ Multi-Group Cross Section Library Format
OpenMC can be run in continuous-energy mode or multi-group mode, provided the
nuclear data is available. In continuous-energy mode, the
``cross_sections.xml`` file contains necessary meta-data for each data set,
``cross_sections.xml`` file contains necessary meta-data for each dataset,
including the name and a file system location where the complete library
can be found. In multi-group mode, the multi-group meta-data and the
nuclear data itself is contained within an ``mgxs.h5``. This portion of
@ -89,19 +89,19 @@ Temperature-dependent data, provided for temperature <TTT>K.
cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups, azimuthal, polar]. This is only required if the data set
[groups, azimuthal, polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **kappa-fission** (*double[]* or *double[][][]*) -- Kappa-Fission
(energy-release from fission) cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups, azimuthal, polar]. This is only required if the data set
[groups, azimuthal, polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **chi** (*double[]* or *double[][][]*) -- Fission neutron energy
spectra.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups, azimuthal, polar]. This is only required if the data set
[groups, azimuthal, polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **nu-fission** (*double[]* to *double[][][][]*) -- Nu-Fission
cross section.