2. 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, including the name and a file system location where the complete library can be found. In multi-group mode, this mgxs.xml file contains this same meta-data describing the nuclide or material, but also contains the group-wise nuclear data. This portion of the manual describes the format of the multi-group data library required to be used in the mgxs.xml file.

Similar to the other input file types, the multi-group library is provided in the XML format. This library must provide some meta-data about the library itself (such as the number of groups and the group structure, etc.) as well as the actual cross section data itself for each of the necessary nuclides or materials.

2.1. MGXS Library Specification

The multi-group library meta-data is contained within the groups, group_structure, and inverse_velocities elements. The actual multi-group data itself is contained within the xsdata element.

2.1.1. <groups> Element

The <groups> element has no attributes and simply provides the number of energy groups contained within the library.

Default: None, this must be provided.

2.1.2. <group_structure> Element

The <group_structure> element has no attributes and should be provided as a monotonically increasing list of bounding energies, in MeV, for a number of groups. To provide proper energy boundaries, the length of the data within the <group_structure> element should be one more than the number of groups in the problem. For example, a two-group problem could be specified as:

<group_structure> 0.0 0.625E-6 20.0 </group_structure>

Default: None, this must be provided.

2.1.3. <inverse_velocities> Element

The <inverse_velocities> element optionally indicates the average inverse velocity corresponding to each of the groups in the problem. This element should therefore be an array with a length which matches the number of groups set in the groups element.

Default: Should this be needed by the presence of an inverse-velocity score in the tallies.xml file and not provided in this element, OpenMC will simply convert the group mid-point energy to an inverse of the velocity and use this information for tallying.

2.1.4. <xsdata> Element

The <xsdata> element contains the nuclide or material-specific meta-data as well as the actual cross section data. The following are the attributes/sub-elements required to describe the meta-data:

name:

The name of the microscopic or macroscopic data set. An extension to the name must be provided (e.g., the .300K in UO2.300K). The name and extension together must be twelve or less characters in length. This extension must follow a period and be five characters or less in length. similar to the equivalent in the continuous-energy cross_sections.xml file, is used to denote variants of the particular nuclide or material of interest (i.e. the UO2 data in this example could have been generated at a temperature of 300K).

Default: None, this must be provided.
alias:

An alternative name to use for the microscopic or macroscopic data set.

Default: If no alias is provided, it will adopt the value of name.
kT:

The temperature times Boltzmann’s constant (in units of MeV) at which the data was generated.

Default: Room temperature, 2.53E-8 MeV
fissionable:

This element states whether or not the data in question is fissionable. Accepted values are “true” or “false”.

Default: None, this element must be provided.
representation:

This element provides the method used to generate and represent the multi-group cross sections. That is, whether they were generated with scalar flux weighting (or reduced to an equivalent representation) and thus are angle-independent, or if the data was generated with angular dependent fluxes and thus the data is angle-dependent. The options are either “isotropic” or “angle”.

Default: “isotropic”
num_azimuthal:

This element provides the number of equal width angular bins that the azimuthal angular domain is subdivided in the case of angle-dependent cross sections (i.e., “angle” is passed to the representation element). Note that these bins are equal in azimuthal angle widths, not equal in the cosine of the azimuthal angle widths.

Default: If representation is “angle”, this must be provided. This parameter is not used for other representation types.
num_polar:

This element provides the number of equal width angular bins that the polar angular domain is subdivided in the case of angle-dependent cross sections (i.e., “angle” is passed to the representation element). Note that these bins are equal in polar angle widths, not equal in the cosine of the polar angle widths.

Default: If representation is “angle”, this must be provided. This parameter is not used for other representation types.
scatt_type:

This element provides the representation of the angular distribution associated with each group-to-group transfer probability. The options are either “legendre”, “histogram”, or “tabular”. The “legendre” option means the angular distribution has been expanded via Legendre polynomials of the order provided in the “order” element. The “histogram” option means the angular distribution is provided in an equi-width histogram format with a number of bins as provided in the “order” element. This is useful when the angular distribution was obtained from a Monte Carlo tally and thus is natively in the histogram format. The “tabular” option means the angular distribution is provided in an equi-spaced point-wise representation.

Default: “legendre”
order:

This element provides either the Legendre order, number of bins, or number of points used to describe the angular distribution associated with each group-to-group transfer probability. The specific meaning of this bin depends upon the value of scatt_type as discussed above.

Default: None, this element must be provided.
tabular_legendre:
 

This optional element is used to set how the Legendre scattering kernel, if provided via the scatt_type element above, is represented and thus used during the scattering process. Specifically, the options are to either convert the Legendre expansion to a tabular representation or leave it as a set of Legendre coefficients. Converting to a tabular representation will cost memory but can allow for a decrease in runtime compared to leaving as a set of Legendre coefficients. This element has the following attributes/sub-elements:

enable:

This attribute/sub-element denotes whether or not the conversion to the tabular format should be performed or not. A value of “true” means the conversion should be performed, “false” means it should not.

Default: “true”
num_points:

If the conversion is to take place the number of tabular points is required. This attribute/sub-element allows the user to set the desired number of points.

Default: 33

The following attributes/sub-elements are the cross section values to be used during the transport process.

total:

This element requires the group-wise total cross section ordered by increasing group index (i.e., fast to thermal). If representation is “isotropic”, then the length of this list should equal the number of groups described in the groups element. If representation is “angle”, then the length of this list should equal the number of groups times the number of azimuthal angles times the number of polar angles, with the inner-dimension being groups, intermediate-dimension being azimuthal angles and outer-dimension being the polar angles.

Default: If not provided, it will be determined by summing the absorption and scattering cross sections.
absorption:

This element requires the group-wise absorption cross section ordered by increasing group index (i.e., fast to thermal). If representation is “isotropic”, then the length of this list should equal the number of groups described in the groups element. If representation is “angle”, then the length of this list should equal the number of groups times the number of azimuthal angles times the number of polar angles, with the inner-dimension being groups, intermediate-dimension being azimuthal angles and outer-dimension being the polar angles.

Default: None, this must be provided.
scatter:

This element requires the scattering moment matrices presented with the columns representing incoming group and rows representing the outgoing group. That is, down-scatter will be above the diagonal of the resultant matrix. This matrix is repeated for every Legendre order (in order of increasing orders) if scatt_type is “legendre”; otherwise, this matrix is repeated for every bin of the histogram or tabular representation. Finally, if representation is “angle”, the above is repeated for every azimuthal angle and every polar angle, in that order.

Default: None, this must be provided.
multiplicity:

This element provides the ratio of neutrons produced in scattering collisions to the neutrons which undergo scattering collisions; that is, the multiplicity provides the code with a scaling factor to account for neutrons being produced in (n,xn) reactions. This information is assumed isotropic and therefore does not need to be repeated for every Legendre moment or histogram/tabular bin. This matrix follows the same arrangement as described for the scatter element, with the exception of the data needed to provide the scattering type information.

Default: Multiplicities of 1.0 are assumed (i.e., (n,xn) reactions are neglected).

The following fission-specific data are only needed should fissionable be “true”.

fission:

This element requires the group-wise fission cross section ordered by increasing group index (i.e., fast to thermal). If representation is “isotropic”, then the length of this list should equal the number of groups described in the groups element. If representation is “angle”, then the length of this list should equal the number of groups times the number of azimuthal angles times the number of polar angles, with the inner-dimension being groups, intermediate-dimension being azimuthal angles and outer-dimension being the polar angles.

Default: None, this is required only if fission tallies are requested and the material is fissionable.
kappa_fission:

This element requires the group-wise kappa-fission cross section ordered by increasing group index (i.e., fast to thermal). If representation is “isotropic”, then the length of this list should equal the number of groups described in the groups element. If representation is “angle”, then the length of this list should equal the number of groups times the number of azimuthal angles times the number of polar angles, with the inner-dimension being groups, intermediate-dimension being azimuthal angles and outer-dimension being the polar angles.

Default: None, this is required only if kappa_fission tallies are requested and the material is fissionable.
chi:

This element requires the group-wise fission spectra ordered by increasing group index (i.e., fast to thermal). This element should be used if making the common approximation that the fission spectra does not depend on incoming energy. If the user does not wish to make this approximation, then this should not be provided and this information included in the nu_fission element instead. If representation is “isotropic”, then the length of this list should equal the number of groups described in the groups element. If representation is “angle”, then the length of this list should equal the number of groups times the number of azimuthal angles times the number of polar angles, with the inner-dimension being groups, intermediate-dimension being azimuthal angles and outer-dimension being the polar angles.

Default: None, either this element is provided or nu_fission is provided in fission matrix form, or the material is not fissionable.
nu_fission:

This element provides either the group-wise fission production cross section vector (i.e., if chi is provided), or is the group-wise fission production matrix. If providing the vector, it should be ordered the same as the fission data. If providing the matrix, it should be ordered the same as the multiplicity matrix.

Default: None, either this element must be provided if the material is fissionable.