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Added mgxs library portion of manual
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docs/source/usersguide/mgxs_library.rst
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docs/source/usersguide/mgxs_library.rst
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.. _usersguide_mgxs_library:
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========================================
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Multi-Group Cross Section Library Format
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========================================
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OpenMC can be run in continuous-energy mode or multi-group mode, provided the
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nuclear data is available. In continuous-energy mode, the
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``cross_sections.xml`` file contains necessary meta-data for each data set,
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including the name and a file system location where the complete library
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can be found. In multi-group mode, this ``cross_sections.xml`` file contains
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this same meta-data describing the nuclide or material, but also contains the
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group-wise nuclear data. This portion of the manual describes the format of
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the multi-group data library required to be used in the ``cross_sections.xml``
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file.
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Similar to the other input file types, the multi-group library is provided in
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the XML_ format. This library must provide some meta-data about the library
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itself (such as the number of groups and the group structure, etc.) as well as
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the actual cross section data itself for each of the necessary nuclides or
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materials.
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.. _XML: http://www.w3.org/XML/
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------------------------------------------------
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MGXS Library Specification -- cross_sections.xml
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------------------------------------------------
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The multi-group library meta-data is contained within the groups_,
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group_structure_, and inverse_velocities_ elements.
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The actual multi-group data itself is contained within the xsdata_ element.
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.. _groups:
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``<groups>`` Element
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----------------------------------
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The ``<groups>`` element has no attributes and simply provides the number of
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energy groups contained within the library.
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*Default*: None, this must be provided.
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.. _group_structure:
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``<group_structure>`` Element
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-----------------------------
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The ``<group_structure>`` element has no attributes and should be provided as a
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monotonically increasing list of bounding energies, in MeV, for a number of
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groups. To provide proper energy boundaries, the length of the data within the
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``<group_structure>`` element should be one more than the number of groups in
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the problem. For example, a two-group problem could be specified as:
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.. code-block:: xml
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<group_structure> 0.0 0.625E-6 20.0 </group_structure>
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*Default*: None, this must be provided.
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.. _inverse_velocities:
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``<inverse_velocities>`` Element
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--------------------------------
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The ``<inverse_velocities>`` element optionally indicates the average
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inverse velocity corresponding to each of the groups in the problem.
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This element should therefore be an array with a length which matches the
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number of groups set in the groups_ element.
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*Default*: Should this be needed by the presence of an ``inverse-velocity``
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score in the ``tallies.xml`` file and not provided in this element, OpenMC
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will simply convert the group mid-point energy to an inverse of the velocity
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and use this information for tallying.
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.. _xsdata:
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``<xsdata>`` Element
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--------------------
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The ``<xsdata>`` element contains the nuclide or material-specific meta-data as
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well as the actual cross section data. The following are the
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attributes/sub-elements required to describe the meta-data:
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:name:
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The name of the microscopic or macroscopic data set.
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*Default*: None, this must be provided.
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:alias:
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The number of total fission source iterations per batch.
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*Default*: If no alias is provided, it will adopt the value of ``name``.
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:kT:
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The temperature the data was generated at.
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*Default*: Room temperature, 2.53E-8 MeV
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:fissionable:
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This element states whether or not the data in question is fissionable.
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Accepted values are ``true`` or ``false``.
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*Default*: None, this element must be provided.
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:representation:
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This element provides the method used to generate and represent the
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multi-group cross sections. That is, whether they were generated with
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scalar flux weighting (or reduced to an equivalent representation)
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and thus are angle-independent, or if the data was generated with angular
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dependent fluxes and thus the data is angle-dependent. The options are
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either ``isotropic`` or ``angle``.
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*Default*: ``isotropic``
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:num_azimuthal:
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This element provides the number of equi-width bins that the azimuthal
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angular domain is subdivided in the case of angle-dependent cross sections
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(i.e., ``angle`` is passed to the ``representation`` element).
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*Default*: If ``representation`` is ``angle``, this must be provided. If
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not, this parameter is not used.
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:num_polar:
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This element provides the number of equi-width bins that the polar angular
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domain is subdivided in the case of angle-dependent cross sections
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(i.e., ``angle`` is passed to the ``representation`` element).
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*Default*: If ``representation`` is ``angle``, this must be provided. If
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not, this parameter is not used.
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:scatt_type:
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This element provides the representation of the angular distribution
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associated with each group-to-group transfer probability. The options are
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either ``legendre``, ``histogram``, or ``tabular``.
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The ``legendre`` option means the angular distribution has been
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expanded via Legendre polynomials of the order provided in the ``order``
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element.
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The ``histogram`` option means the angular distribution is provided in
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an equi-width histogram format with a number of bins as provided in the
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``order`` element. This is useful when the angular distribution was
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obtained from a Monte Carlo tally and thus is natively in the histogram
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format.
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The ``tabular`` option means the angular distribution is provided in an
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equi-spaced point-wise representation.
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*Default*: ``legendre``
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:order:
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This element provides either the Legendre order, number of bins, or number
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of points used to describe the angular distribution associated with each
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group-to-group transfer probability. The specific meaning of this bin
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depends upon the value of ``scatt_type`` as discussed above.
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*Default*: None, this element must be provided.
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:tabular_legendre:
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This optional element is used to set how the Legendre scattering kernel, if
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provided via the ``scatt_type`` element above, is represented and thus used
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during the scattering process. Specifically, the options are to either
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convert the Legendre expansion to a tabular representation or leave it as
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a set of Legendre coefficients. Converting to a tabular representation will
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cost memory but is likely to decrease runtime compared to leaving as a
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set of Legendre coefficients. This element has the following
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attributes/sub-elements:
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:enable:
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This attribute/sub-element denotes whether or not the conversion to the
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tabular format should be performed or not. A value of ``true`` means
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the conversion should be performed, ``false`` means it should not.
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*Default*: ``True``
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:num_points:
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If the conversion is to take place the number of tabular points is
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required. This attribute/sub-element allows the user to set the desired
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number of points.
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*Default*: ``33``
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The following attributes/sub-elements are the actual cross section values to
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be used during the transport process.
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:total:
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This element requires the group-wise total cross section ordered by
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increasing group index (i.e., fast to thermal). If ``representation`` is
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``isotropic``, then the length of this list should equal the number of
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groups described in the ``groups`` element. If ``representation`` is
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``angle``, then the length of this list should equal the number of groups
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times the number of azimuthal angles times the number of polar angles,
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with the inner-dimension being groups, intermediate-dimension being
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azimuthal angles and outer-dimension being the polar angles.
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*Default*: If not provided, it will be determined by summing the
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absorption and scattering cross sections.
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:absorption:
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This element requires the group-wise absorption cross section ordered by
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increasing group index (i.e., fast to thermal). If ``representation`` is
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``isotropic``, then the length of this list should equal the number of
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groups described in the ``groups`` element. If ``representation`` is
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``angle``, then the length of this list should equal the number of groups
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times the number of azimuthal angles times the number of polar angles,
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with the inner-dimension being groups, intermediate-dimension being
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azimuthal angles and outer-dimension being the polar angles.
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*Default*: None, this must be provided.
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:scatter:
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This element requires the scattering moment matrices presented with the
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columns representing incoming group and rows representing the outgoing
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group. That is, down-scatter will be above the diagonal of the resultant
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matrix. This matrix is repeated for every Legendre order (in order of
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increasing orders) if ``scatt_type`` is ``legendre``; otherwise, this
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matrix is repeated for every bin of the histogram or tabular
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representation. Finally, if ``representation`` is ``angle``, the above
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is repeated for every azimuthal angle and every polar angle, in that
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order.
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*Default*: None, this must be provided.
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:multiplicity:
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This element provides the ratio of neutrons produced in scattering
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collisions to the neutrons which undergo scattering collisions; that is,
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the multiplicity provides the code with a scaling factor to account for
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neutrons being produced in (n,xn) reactions. This information is assumed
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isotropic and therefore does not need to be repeated for every Legendre
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moment or histogram/tabular bin. This matrix follows the same arrangement
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as described for the ``scatter`` element, with the exception of the
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data needed to provide the scattering type information.
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*Default*: Multiplicities of 1.0 are assumed (i.e., (n,xn) reactions are
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neglected).
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The following fission-specific data are only needed should ``fissionable``
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be ``true``.
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:fission:
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This element requires the group-wise fission cross section ordered by
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increasing group index (i.e., fast to thermal). If ``representation`` is
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``isotropic``, then the length of this list should equal the number of
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groups described in the ``groups`` element. If ``representation`` is
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``angle``, then the length of this list should equal the number of groups
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times the number of azimuthal angles times the number of polar angles,
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with the inner-dimension being groups, intermediate-dimension being
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azimuthal angles and outer-dimension being the polar angles.
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*Default*: None, this must be provided if the material is fissionable.
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:k_fission:
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This element requires the group-wise kappa-fission cross section ordered by
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increasing group index (i.e., fast to thermal). If ``representation`` is
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``isotropic``, then the length of this list should equal the number of
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groups described in the ``groups`` element. If ``representation`` is
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``angle``, then the length of this list should equal the number of groups
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times the number of azimuthal angles times the number of polar angles,
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with the inner-dimension being groups, intermediate-dimension being
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azimuthal angles and outer-dimension being the polar angles.
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*Default*: None, this is required only if ``kappa-fission`` tallies are
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requested and the material is fissionable.
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:chi:
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This element requires the group-wise fission spectra ordered by
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increasing group index (i.e., fast to thermal). This element should be
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used if making the common approximation that the fission spectra does
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not depend on incoming energy. If the user does not wish to make this
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approximation, then this should not be provided and this information
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included in the ``nu_fission`` element instead. If ``representation`` is
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``isotropic``, then the length of this list should equal the number of
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groups described in the ``groups`` element. If ``representation`` is
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``angle``, then the length of this list should equal the number of groups
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times the number of azimuthal angles times the number of polar angles,
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with the inner-dimension being groups, intermediate-dimension being
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azimuthal angles and outer-dimension being the polar angles.
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*Default*: None, either this element is provided or ``nu_fission`` is
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provided in fission matrix form, or the material is not fissionable.
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:nu_fission:
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This element provides either the group-wise fission production cross
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section vector (i.e., if ``chi`` is provided), or is the group-wise fission
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production matrix. If providing the vector, it should be ordered the same
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as the ``fission`` data. If providing the matrix, it should be ordered
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the same as the ``multiplicity`` matrix.
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*Default*: None, either this element must be provided if the material
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is fissionable.
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