.. _io_settings:

======================================
Settings Specification -- settings.xml
======================================

All simulation parameters and miscellaneous options are specified in the
settings.xml file.

-------------------------------
``<atomic_relaxation>`` Element
-------------------------------

The ``<atomic_relaxation>`` element determines whether the atomic relaxation
cascade, the X-ray fluorescence photons and Auger electrons emitted when an
inner-shell vacancy is filled, is simulated following photoelectric and
incoherent (Compton) scattering interactions. Disabling this can speed up
photon transport calculations where the detailed secondary particle cascade is
not of interest.

  *Default*: true

---------------------
``<batches>`` Element
---------------------

The ``<batches>`` element indicates the total number of batches to execute,
where each batch corresponds to a tally realization. In a fixed source
calculation, each batch consists of a number of source particles. In an
eigenvalue calculation, each batch consists of one or many fission source
iterations (generations), where each generation itself consists of a number of
source neutrons.

  *Default*: None

-----------------------------
``<collision_track>`` Element
-----------------------------

The ``<collision_track>`` element indicates to track information about particle
collisions based on a set of criteria and store these events in a file named
``collision_track.h5``. This file records details such as the position of the
interaction, direction of the incoming particle, incident energy and deposited
energy, weight, time of the interaction, and the delayed neutron group (0 for
prompt neutrons). Additional information such as the cell ID, material ID,
universe ID, nuclide ZAID, particle type, and event MT number are also stored.
Users can specify one or more criterion to filter collisions. If no criteria are
specified, it defaults to tracking all collisions across the model.

.. warning::
    Storing all collisions can be very memory intensive. For more targeted
    tracking, users can employ a variety of parameters such as ``cell_ids``,
    ``reactions``, ``universe_ids``, ``material_ids``, ``nuclides``, and
    ``deposited_E_threshold`` to refine the selection of particle interactions
    to be banked.

This element can contain one or more of the following attributes or
sub-elements:

  :max_collisions:
    An integer indicating the maximum number of collisions to be banked per file.

    *Default*: 1000

  :max_collision_track_files:
    An integer indicating the number of collision_track files to be used.

    *Default*: 1

  :mcpl:
    An optional boolean to enable MCPL_-format instead of the native HDF5-based
    format. If activated, the output file name and type is changed to
    ``collision_track.mcpl``.

    *Default*: false

    .. _MCPL: https://mctools.github.io/mcpl/mcpl.pdf

  :cell_ids:
    A list of integers representing cell IDs to define specific cells in which
    collisions are to be banked.

    *Default*: None

  :universe_ids:
    A list of integers representing the universe IDs to define specific
    universes in which collisions are to be banked.

    *Default*: None

  :material_ids:
    A list of integers representing the material IDs to define specific
    materials in which collisions are to be banked.

    *Default*: None

  :nuclides:
    A list of strings representing the nuclide, to define specific
    define specific target nuclide collisions to be banked.

    .. note::
      Electron and positron collision-track events are not associated with
      a specific nuclide. If a ``nuclides`` entry is specified, these events
      are omitted.

    *Default*: None

  :reactions:
    A list of integers representing the ENDF-6 format MT numbers or strings
    (e.g. (n,fission)) to define specific reaction types to be banked.

    *Default*: None

  :deposited_E_threshold:
    A float defining the minimum deposited energy per collision (in eV) to
    trigger banking.

    *Default*: 0.0

----------------------------------
``<confidence_intervals>`` Element
----------------------------------

The ``<confidence_intervals>`` element has no attributes and has an accepted
value of "true" or "false". If set to "true", uncertainties on tally results
will be reported as the half-width of the 95% two-sided confidence interval. If
set to "false", uncertainties on tally results will be reported as the sample
standard deviation.

  *Default*: false

-------------------------------------
``<create_delayed_neutrons>`` Element
-------------------------------------

The ``<create_delayed_neutrons>`` element indicates whether delayed neutrons
are created in fission. If this element is set to "true", delayed neutrons
will be created in fission events; otherwise only prompt neutrons will be
created.

  *Default*: true

-------------------------------------
``<create_fission_neutrons>`` Element
-------------------------------------

The ``<create_fission_neutrons>`` element indicates whether fission neutrons
should be created or not.  If this element is set to "true", fission neutrons
will be created; otherwise the fission is treated as capture and no fission
neutron will be created. Note that this option is only applied to fixed source
calculation. For eigenvalue calculation, fission will always be treated as real
fission.

  *Default*: true

--------------------
``<cutoff>`` Element
--------------------

The ``<cutoff>`` element indicates three kinds of cutoffs. The first is the
weight cutoff used below which particles undergo Russian roulette. Surviving
particles are assigned a user-determined weight. Note that weight cutoffs and
Russian rouletting are not turned on by default. The second is the energy cutoff
which is used to kill particles under certain energy. The energy cutoff should
not be used unless you know particles under the energy are of no importance to
results you care. The third is the time cutoff used to kill particles whose time
exceeds a specific cutoff. Particles will be killed exactly at the specified
time.

  :weight:
    The weight below which particles undergo Russian roulette.

    *Default*: 0.25

  :weight_avg:
    The weight that is assigned to particles that are not killed after Russian
    roulette.

    *Default*: 1.0

  :survival_normalization:
    If this element is set to "true", this will enable the use of survival
    biasing source normalization, whereby the weight parameters, weight and
    weight_avg, are multiplied per history by the start weight of said history.

    *Default*: false

  :energy_neutron:
    The energy under which neutrons will be killed.

    *Default*: 0.0

  :energy_photon:
    The energy under which photons will be killed.

    *Default*: 1000.0

  :energy_electron:
    The energy under which electrons will be killed.

    *Default*: 0.0

  :energy_positron:
    The energy under which positrons will be killed.

    *Default*: 0.0

  :time_neutron
    The time above which neutrons will be killed.

    *Default*: Infinity

  :time_photon
    The time above which photons will be killed.

    *Default*: Infinity

  :time_electron
    The time above which electrons will be killed.

    *Default*: Infinity

  :time_positron
    The time above which positorns will be killed.

    *Default*: Infinity

----------------------------
``<delayed_photon_scaling>``
----------------------------

Determines whether to scale the fission photon yield to account for delayed
photon energy. The photon yields are scaled as (EGP + EGD)/EGP where EGP and EGD
are the prompt and delayed photon components of energy release, respectively,
from MF=1, MT=458 on an ENDF evaluation.

  *Default*: true

--------------------------------
``<electron_treatment>`` Element
--------------------------------

When photon transport is enabled, the ``<electron_treatment>`` element tells
OpenMC whether to deposit all energy from electrons locally (``led``) or create
secondary bremsstrahlung photons (``ttb``).

  *Default*: ttb

.. _energy_mode:

-------------------------
``<energy_mode>`` Element
-------------------------

The ``<energy_mode>`` element tells OpenMC if the run-mode should be
continuous-energy or multi-group.  Options for entry are: ``continuous-energy``
or ``multi-group``.

  *Default*: continuous-energy

--------------------------
``<entropy_mesh>`` Element
--------------------------

The ``<entropy_mesh>`` element indicates the ID of a mesh that is to be used for
calculating Shannon entropy. The mesh should cover all possible fissionable
materials in the problem and is specified using a :ref:`mesh_element`.

----------------------------
``<event_based>``
----------------------------

Determines whether to use event-based parallelism instead of the default
history-based parallelism.

  *Default*: false

--------------------------------
``<free_gas_threshold>`` Element
--------------------------------

The ``<free_gas_threshold>`` element specifies the energy multiplier, expressed
in units of :math:`kT`, that determines when the free gas scattering approach is
used for elastic scattering. Values must be positive.

  *Default*: 400.0

-----------------------------------
``<generations_per_batch>`` Element
-----------------------------------

The ``<generations_per_batch>`` element indicates the number of total fission
source iterations per batch for an eigenvalue calculation. This element is
ignored for all run modes other than "eigenvalue".

  *Default*: 1

------------------------------
``<ifp_n_generation>`` Element
------------------------------

The ``<ifp_n_generation>`` element indicates the number of generations to
consider for the Iterated Fission Probability method.

  *Default*: 10

----------------------
``<inactive>`` Element
----------------------

The ``<inactive>`` element indicates the number of inactive batches used in a
k-eigenvalue calculation. In general, the starting fission source iterations in
an eigenvalue calculation can not be used to contribute to tallies since the
fission source distribution and eigenvalue are generally not converged
immediately. This element is ignored for all run modes other than "eigenvalue".

  *Default*: 0

--------------------------
``<keff_trigger>`` Element
--------------------------

The ``<keff_trigger>`` element (ignored for all run modes other than
"eigenvalue".) specifies a precision trigger on the combined
:math:`k_{eff}`. The trigger is a convergence criterion on the uncertainty of
the estimated eigenvalue. It has the following attributes/sub-elements:

  :type:
    The type of precision trigger. Accepted options are "variance", "std_dev",
    and "rel_err".

    :variance:
      Variance of the batch mean :math:`\sigma^2`

    :std_dev:
      Standard deviation of the batch mean :math:`\sigma`

    :rel_err:
      Relative error of the batch mean :math:`\frac{\sigma}{\mu}`

    *Default*: None

  :threshold:
    The precision trigger's convergence criterion for the
    combined :math:`k_{eff}`.

    *Default*: None

  .. note:: See section on the :ref:`trigger` for more information.

---------------------------
``<log_grid_bins>`` Element
---------------------------

The ``<log_grid_bins>`` element indicates the number of bins to use for the
logarithmic-mapped energy grid. Using more bins will result in energy grid
searches over a smaller range at the expense of more memory. The default is
based on the recommended value in LA-UR-14-24530_.

  *Default*: 8000

  .. note:: This element is not used in the multi-group :ref:`energy_mode`.

.. _LA-UR-14-24530: https://mcnp.lanl.gov/pdf_files/TechReport_2014_LANL_LA-UR-14-24530_Brown.pdf

---------------------------
``<material_cell_offsets>``
---------------------------

By default, OpenMC will count the number of instances of each cell filled with a
material and generate "offset tables" that are used for cell instance tallies.
The ``<material_cell_offsets>`` element allows a user to override this default
setting and turn off the generation of offset tables, if desired, by setting it
to false.

  *Default*: true

--------------------------------
``<max_lost_particles>`` Element
--------------------------------

This element indicates the maximum number of lost particles.

  *Default*: 10

------------------------------------
``<rel_max_lost_particles>`` Element
------------------------------------


This element indicates the maximum number of lost particles, relative to the
total number of particles.

  *Default*: 1.0e-6

-------------------------------------
``<max_particles_in_flight>`` Element
-------------------------------------

This element indicates the number of particles to run in flight concurrently
when using event-based parallelism. A higher value uses more memory, but
may be more efficient computationally.

  *Default*: 100000

---------------------------------
``<max_particle_events>`` Element
---------------------------------

This element indicates the maximum number of events a particle can undergo.

  *Default*: 1000000

-----------------------
``<max_order>`` Element
-----------------------

The ``<max_order>`` element allows the user to set a maximum scattering order
to apply to every nuclide/material in the problem.  That is, if the data
library has :math:`P_3` data available, but ``<max_order>`` was set to ``1``,
then, OpenMC will only use up to the :math:`P_1` data.

  *Default*: Use the maximum order in the data library

  .. note:: This element is not used in the continuous-energy
    :ref:`energy_mode`.

--------------------------------
``<max_history_splits>`` Element
--------------------------------

The ``<max_history_splits>`` element indicates the number of times a particle
can split during a history.

  *Default*: 1000

-----------------------------
``<max_secondaries>`` Element
-----------------------------

The ``<max_secondaries>`` element indicates the maximum secondary bank size.

  *Default*: 10000

------------------------
``<max_tracks>`` Element
------------------------

The ``<max_tracks>`` element indicates the maximum number of tracks written to a
track file (per MPI process).

  *Default*: 1000

--------------------------------------
``<max_write_lost_particles>`` Element
--------------------------------------

This ``<max_write_lost_particles>`` element indicates the maximum number of
particle restart files (per MPI process) to write for lost particles.

  *Default*: None

.. _mesh_element:

------------------
``<mesh>`` Element
------------------

The ``<mesh>`` element describes a mesh that is used either for calculating
Shannon entropy, applying the uniform fission site method, or in tallies. For
Shannon entropy meshes, the mesh should cover all possible fissionable materials
in the problem. It has the following attributes/sub-elements:

  :id:
    A unique integer that is used to identify the mesh.

  :dimension:
    The number of mesh cells in the x, y, and z directions, respectively.

    *Default*: If this tag is not present, the number of mesh cells is
    automatically determined by the code.

  :lower_left:
    The Cartesian coordinates of the lower-left corner of the mesh.

    *Default*: None

  :upper_right:
    The Cartesian coordinates of the upper-right corner of the mesh.

    *Default*: None

-----------------------
``<no_reduce>`` Element
-----------------------

The ``<no_reduce>`` element has no attributes and has an accepted value of
"true" or "false". If set to "true", all user-defined tallies and global tallies
will not be reduced across processors in a parallel calculation. This means that
the accumulate score in one batch on a single processor is considered as an
independent realization for the tally random variable. For a problem with large
tally data, this option can significantly improve the parallel efficiency.

  *Default*: false

--------------------
``<output>`` Element
--------------------

The ``<output>`` element determines what output files should be written to disk
during the run. The sub-elements are described below, where "true" will write
out the file and "false" will not.

  :summary:
    Writes out an HDF5 summary file describing all of the user input files that
    were read in.

    *Default*: true

  :tallies:
    Write out an ASCII file of tally results.

    *Default*: true

  .. note:: The tally results will always be written to a binary/HDF5 state
            point file.

  :path:
    Absolute or relative path where all output files should be written to. The
    specified path must exist or else OpenMC will abort.

    *Default*: Current working directory

-----------------------
``<particles>`` Element
-----------------------

This element indicates the number of neutrons to simulate per fission source
iteration when a k-eigenvalue calculation is performed or the number of
particles per batch for a fixed source simulation.

  *Default*: None

------------------------------
``<photon_transport>`` Element
------------------------------

The ``<photon_transport>`` element determines whether photon transport is
enabled. This element has no attributes or sub-elements and can be set to
either "false" or "true".

  *Default*: false

-----------------------
``<plot_seed>`` Element
-----------------------

The ``<plot_seed>`` element is used to set the seed for the pseudorandom number
generator during generation of colors in plots.

  *Default*: 1

.. _properties_file:

-----------------------------
``<properties_file>`` Element
-----------------------------

  The ``properties_file`` element has no attributes and contains the path to a
  properties HDF5 file to load cell temperatures/densities and material
  densities.

  *Default*: None

---------------------
``<ptables>`` Element
---------------------

The ``<ptables>`` element determines whether probability tables should be used
in the unresolved resonance range if available. This element has no attributes
or sub-elements and can be set to either "false" or "true".

  *Default*: true

  .. note:: This element is not used in the multi-group :ref:`energy_mode`.

------------------------
``<random_ray>`` Element
------------------------

The ``<random_ray>`` element enables random ray mode and contains a number of
settings relevant to the solver. Tips for selecting these parameters can be
found in the :ref:`random ray user guide <random_ray>`.

  :distance_inactive:
    The inactive ray length (dead zone length) in [cm].

    *Default*: None

  :distance_active:
    The active ray length in [cm].

    *Default*: None

  :ray_source:
    Specifies the starting ray distribution, and follows the format for
    :ref:`source_element`. It must be uniform in space and angle and cover the
    full domain. It does not represent a physical neutron or photon source -- it
    is only used to sample integrating ray starting locations and directions.

    *Default*: None

  :adjoint_source:
    Specifies an adjoint fixed source for adjoint transport simulations, and
    follows the format for :ref:`source_element`. The distributions which make
    up the adjoint source are subject to the same restrictions as forward
    fixed sources in Random Ray mode.

    *Default*: None

  :adjoint:
    Specifies whether to perform adjoint transport. The default is 'False',
    corresponding to forward transport.

    *Default*: None

  :volume_estimator:
    Specifies choice of volume estimator for the random ray solver. Options
    are 'naive', 'simulation_averaged', or 'hybrid'. The default is 'hybrid'.

    *Default*: None

  :volume_normalized_flux_tallies:
    Specifies whether to normalize flux tallies by volume (bool). The
    default is 'False'. When enabled, flux tallies will be reported in units
    of cm/cm^3. When disabled, flux tallies will be reported in units of cm
    (i.e., total distance traveled by neutrons in the spatial tally
    region).

    *Default*: None

  :sample_method:
    Specifies the method for sampling the starting ray distribution. This
    element can be set to "prng" or "halton".

    *Default*: prng

  :source_region_meshes:
    Relates meshes to spatial domains for subdividing source regions with each domain.

    :mesh:
      Contains an ``id`` attribute and one or more ``<domain>`` sub-elements.

      :id:
        The unique identifier for the mesh.

      :domain:
        Each domain element has an ``id`` attribute and a ``type`` attribute.

        :id:
          The unique identifier for the domain.

        :type:
          The type of the domain. Can be ``material``, ``cell``, or ``universe``.

  :diagonal_stabilization_rho:
    The rho factor for use with diagonal stabilization. This technique is
    applied when negative diagonal (in-group) elements are detected in
    the scattering matrix of input MGXS data, which is a common feature
    of transport corrected MGXS data.

    *Default*: 1.0

----------------------------------
``<resonance_scattering>`` Element
----------------------------------

The ``resonance_scattering`` element indicates to OpenMC that a method be used
to properly account for resonance elastic scattering (typically for nuclides
with Z > 40). This element can contain one or more of the following attributes
or sub-elements:

  :enable:
    Indicates whether a resonance elastic scattering method should be turned
    on. Accepts values of "true" or "false".

    *Default*: If the ``<resonance_scattering>`` element is present, "true".

  :method:

    Which resonance elastic scattering method is to be applied: "rvs" (relative
    velocity sampling) or "dbrc" (Doppler broadening rejection correction).
    Descriptions of each of these methods are documented here_.

    .. _here: https://doi.org/10.1016/j.anucene.2017.12.044

    *Default*: "rvs"

  :energy_min:
    The energy in eV above which the resonance elastic scattering method should
    be applied.

    *Default*: 0.01 eV

  :energy_max:
    The energy in eV below which the resonance elastic scattering method should
    be applied.

    *Default*: 1000.0 eV

  :nuclides:

    A list of nuclides to which the resonance elastic scattering method should
    be applied.

    *Default*: If ``<resonance_scattering>`` is present but the ``<nuclides>``
    sub-element is not given, the method is applied to all nuclides with 0 K
    elastic scattering data present.

  .. note:: If the ``resonance_scattering`` element is not given, the free gas,
            constant cross section scattering model, which has historically been
            used by Monte Carlo codes to sample target velocities, is used to
            treat the target motion of all nuclides.  If
            ``resonance_scattering`` is present, the constant cross section
            method is applied below ``energy_min`` and the target-at-rest
            (asymptotic) kernel is used above ``energy_max``.

  .. note:: This element is not used in the multi-group :ref:`energy_mode`.

----------------------
``<run_mode>`` Element
----------------------

The ``<run_mode>`` element indicates which run mode should be used when OpenMC
is executed. This element has no attributes or sub-elements and can be set to
"eigenvalue", "fixed source", "plot", "volume", or "particle restart".

  *Default*: None

------------------
``<seed>`` Element
------------------

The ``seed`` element is used to set the seed used for the linear congruential
pseudo-random number generator.

  *Default*: 1

-----------------------------------
``<shared_secondary_bank>`` Element
-----------------------------------

  The ``shared_secondary_bank`` element indicates whether to use a shared
  secondary particle bank. When enabled, secondary particles are collected into
  a global bank, sorted for reproducibility, and load-balanced across MPI ranks
  between generations. If not specified, the shared secondary bank is enabled
  automatically for fixed-source simulations with weight windows active, and
  disabled otherwise.

.. _source_element:

--------------------
``<source>`` Element
--------------------

The ``source`` element gives information on an external source distribution to
be used either as the source for a fixed source calculation or the initial
source guess for criticality calculations. Multiple ``<source>`` elements may be
specified to define different source distributions. Each one takes the following
attributes/sub-elements:

  :strength:
    The strength of the source. If multiple sources are present, the source
    strength indicates the relative probability of choosing one source over the
    other.

    *Default*: 1.0

  :type:
    Indicator of source type. One of ``independent``, ``file``, ``compiled``, or
    ``mesh``. The type of the source will be determined by this attribute if it
    is present.

  :particle:
    The source particle type, specified as a PDG number or a string alias (e.g.,
    ``neutron``/``n``, ``photon``/``gamma``, ``electron``, ``positron``,
    ``proton``/``p``, ``deuteron``/``d``, ``triton``/``t``, ``alpha``, or GNDS
    nuclide names like ``Fe57``).

    *Default*: neutron

  :file:
    If this attribute is given, it indicates that the source type is ``file``,
    meaning particles are to be read from a binary source file whose path is
    given by the value of this element.

    *Default*: None

  :library:
    If this attribute is given, it indicates that the source type is
    ``compiled``, meaning that particles are instantiated from an externally
    compiled source function. This source can be completely customized as needed
    to define the source for your problem. The library has a few basic
    requirements:

    * It must contain a class that inherits from ``openmc::Source``;
    * The class must implement a function called ``sample()``;
    * There must be an ``openmc_create_source()`` function that creates the source
      as a unique pointer. This function can be used to pass parameters through to
      the source from the XML, if needed.

    More documentation on how to build sources can be found in
    :ref:`compiled_source`.

  :parameters:
    If this attribute is given, it indicated that the source type is
    ``compiled``. Its value provides the parameters to pass through to the class
    generated using the ``library`` parameter. More documentation on how to
    build parametrized sources can be found in
    :ref:`parameterized_compiled_source`.

  :space:
    An element specifying the spatial distribution of source sites. This element
    has the following attributes:

    :type:
      The type of spatial distribution. Valid options are "box", "fission",
      "point", "cartesian", "cylindrical", "spherical", "mesh", and "cloud".

      A "box" spatial distribution has coordinates sampled uniformly in a
      parallelepiped.

      A "fission" spatial distribution samples locations from a "box"
      distribution but only locations in fissionable materials are accepted.

      A "point" spatial distribution has coordinates specified by a triplet.

      A "cartesian" spatial distribution specifies independent distributions of
      x-, y-, and z-coordinates.

      A "cylindrical" spatial distribution specifies independent distributions
      of r-, phi-, and z-coordinates where phi is the azimuthal angle and the
      origin for the cylindrical coordinate system is specified by origin.

      A "spherical" spatial distribution specifies independent distributions of
      r-, cos_theta-, and phi-coordinates where cos_theta is the cosine of the
      angle with respect to the z-axis, phi is the azimuthal angle, and the
      sphere is centered on the coordinate (x0,y0,z0).

      A "mesh" spatial distribution samples source sites from a mesh element
      based on the relative strengths provided in the node. Source locations
      within an element are sampled isotropically. If no strengths are provided,
      the space within the mesh is uniformly sampled.

      A "cloud" spatial distribution samples source sites from a list of spatial
      positions provided in the node, based on the relative strengths provided
      in the node. If no strengths are provided, the positions are uniformly
      sampled.

      *Default*: None

    :parameters:
      For a "box" or "fission" spatial distribution, ``parameters`` should be
      given as six real numbers, the first three of which specify the lower-left
      corner of a parallelepiped and the last three of which specify the
      upper-right corner. Source sites are sampled uniformly through that
      parallelepiped.

      For a "point" spatial distribution, ``parameters`` should be given as
      three real numbers which specify the (x,y,z) location of an isotropic
      point source.

      For an "cartesian" distribution, no parameters are specified. Instead,
      the ``x``, ``y``, and ``z`` elements must be specified.

      For a "cylindrical" distribution, no parameters are specified. Instead,
      the ``r``, ``phi``, ``z``, and ``origin`` elements must be specified.
      Optionally, the ``r_dir`` and ``z_dir`` elements could be specified.

      For a "spherical" distribution, no parameters are specified. Instead,
      the ``r``, ``theta``, ``phi``, and ``origin`` elements must be specified.

      *Default*: None

    :x:
      For an "cartesian" distribution, this element specifies the distribution
      of x-coordinates. The necessary sub-elements/attributes are those of a
      univariate probability distribution (see the description in
      :ref:`univariate`).

    :y:
      For an "cartesian" distribution, this element specifies the distribution
      of y-coordinates. The necessary sub-elements/attributes are those of a
      univariate probability distribution (see the description in
      :ref:`univariate`).

    :z:
      For both "cartesian" and "cylindrical" distributions, this element
      specifies the distribution of z-coordinates. The necessary
      sub-elements/attributes are those of a univariate probability
      distribution (see the description in :ref:`univariate`).

    :r:
      For "cylindrical" and "spherical" distributions, this element specifies
      the distribution of r-coordinates (cylindrical radius and spherical
      radius, respectively). The necessary sub-elements/attributes are those
      of a univariate probability distribution (see the description in
      :ref:`univariate`).

    :r_dir:
      For "cylindrical" distributions, this element specifies the direction
      of the cylinder r-axis at phi=0. Defaults to (1.0, 0.0, 0.0).

    :theta:
      For a "spherical" distribution, this element specifies the distribution
      of theta-coordinates. The necessary sub-elements/attributes are those of a
      univariate probability distribution (see the description in
      :ref:`univariate`).

    :phi:
      For "cylindrical" and "spherical" distributions, this element specifies
      the distribution of phi-coordinates. The necessary
      sub-elements/attributes are those of a univariate probability
      distribution (see the description in :ref:`univariate`).

    :z_dir:
      For "cylindrical" distributions, this element specifies the direction
      of the cylinder z-axis. Defaults to (0.0, 0.0, 1.0).

    :origin:
      For "cylindrical and "spherical" distributions, this element specifies
      the coordinates for the origin of the coordinate system.

    :mesh_id:
      For "mesh" spatial distributions, this element specifies which mesh ID to
      use for the geometric description of the mesh.

    :coords:
      For "cloud" distributions, this element specifies a list of coordinates
      for each of the points in the cloud.

    :strengths:
      For "mesh" and "cloud" spatial distributions, this element specifies the
      relative source strength of each mesh element or each point in the cloud.

    :volume_normalized:
      For "mesh" spatial distributions, this optional boolean element specifies
      whether the vector of relative strengths should be multiplied by the mesh
      element volume. This is most common if the strengths represent a source
      per unit volume.

      *Default*: false

    :bias:
      For "mesh" and "cloud" spatial distributions, this optional element
      specifies floating point values corresponding to alternative probabilities
      for each value/component to use for biased sampling.

  :angle:
    An element specifying the angular distribution of source sites. This element
    has the following attributes:

    :type:
      The type of angular distribution. Valid options are "isotropic",
      "monodirectional", and "mu-phi". The angle of the particle emitted from a
      source site is isotropic if the "isotropic" option is given. The angle of
      the particle emitted from a source site is the direction specified in the
      ``reference_uvw`` element/attribute if "monodirectional" option is
      given. The "mu-phi" option produces directions with the cosine of the
      polar angle and the azimuthal angle explicitly specified.

      *Default*: isotropic

    :reference_uvw:
      The direction from which the polar angle is measured. Represented by the
      x-, y-, and z-components of a unit vector. For a monodirectional
      distribution, this defines the direction of all sampled particles.

    :mu:
      An element specifying the distribution of the cosine of the polar
      angle. Only relevant when the type is "mu-phi". The necessary
      sub-elements/attributes are those of a univariate probability distribution
      (see the description in :ref:`univariate`).

    :phi:
      An element specifying the distribution of the azimuthal angle. Only
      relevant when the type is "mu-phi". The necessary sub-elements/attributes
      are those of a univariate probability distribution (see the description in
      :ref:`univariate`).

    :bias:
      For "isotropic" angular distributions, this optional element specifies a
      "mu-phi" angular distribution used for biased sampling.

  :energy:
    An element specifying the energy distribution of source sites. The necessary
    sub-elements/attributes are those of a univariate probability distribution
    (see the description in :ref:`univariate`).

    *Default*: Watt spectrum with :math:`a` = 0.988 MeV and :math:`b` =
    2.249 MeV :sup:`-1`

  :write_initial:
    An element specifying whether to write out the initial source bank used at
    the beginning of the first batch. The output file is named
    "initial_source.h5"

    *Default*: false

  :mesh:
    For mesh sources, this indicates the ID of the corresponding mesh.

  :source:
    For mesh sources, this sub-element specifies the source for an individual
    mesh element and follows the format for :ref:`source_element`. The number of
    ``<source>`` sub-elements should correspond to the number of mesh elements.

  .. note:: Biased sampling can be applied to the spatial and energy distributions
            of a source by using the ``<bias>`` sub-element (see
            :ref:`univariate` for details on how to specify bias distributions).

  :constraints:
    This sub-element indicates the presence of constraints on sampled source
    sites (see :ref:`usersguide_source_constraints` for details). It may have
    the following sub-elements:

    :domain_ids:
      The unique IDs of domains for which source sites must be within.

      *Default*: None

    :domain_type:
      The type of each domain for source rejection ("cell", "material", or
      "universe").

      *Default*: None

    :fissionable:
      A boolean indicating whether source sites must be sampled within a
      material that is fissionable in order to be accepted.

    :time_bounds:
      A pair of times in [s] indicating the lower and upper bound for a time
      interval that source particles must be within.

    :energy_bounds:
      A pair of energies in [eV] indicating the lower and upper bound for an
      energy interval that source particles must be within.

    :rejection_strategy:
      Either "resample", indicating that source sites should be resampled when
      one is rejected, or "kill", indicating that a rejected source site is
      assigned zero weight.

.. _univariate:

Univariate Probability Distributions
++++++++++++++++++++++++++++++++++++

Various components of a source distribution involve probability distributions of
a single random variable, e.g. the distribution of the energy, the distribution
of the polar angle, and the distribution of x-coordinates. Each of these
components supports the same syntax with an element whose tag signifies the
variable and whose sub-elements/attributes are as follows:

:type:
  The type of the distribution. Valid options are "uniform", "discrete",
  "tabular", "maxwell", "watt", "mixture", and "decay_spectrum". The "uniform"
  option produces variates sampled from a uniform distribution over a finite
  interval. The "discrete" option produces random variates that can assume a
  finite number of values (i.e., a distribution characterized by a probability
  mass function). The "tabular" option produces random variates sampled from a
  tabulated distribution where the density function is either a histogram or
  linearly-interpolated between tabulated points. The "watt" option produces
  random variates is sampled from a Watt fission spectrum (only used for
  energies). The "maxwell" option produce variates sampled from a Maxwell
  fission spectrum (only used for energies). The "mixture" option produces
  samples from univariate sub-distributions with given probabilities. The
  "decay_spectrum" option produces photon energies sampled from decay photon
  spectra in a depletion chain (only used for energies).

  *Default*: None

:parameters:
  For a "uniform" distribution, ``parameters`` should be given as two real
  numbers :math:`a` and :math:`b` that define the interval :math:`[a,b]` over
  which random variates are sampled.

  For a "powerlaw" distribution, ``parameters`` should be given as three real
  numbers :math:`a` and :math:`b` that define the interval :math:`[a,b]` over
  which random variates are sampled and :math:`n` that defines the exponent of
  the probability distribution :math:`p(x)=c x^n`

  For a "discrete" or "tabular" distribution, ``parameters`` provides the
  :math:`(x,p)` pairs defining the discrete/tabular distribution. All :math:`x`
  points are given first followed by corresponding :math:`p` points.

  For a "decay_spectrum" distribution, ``parameters`` gives the atom densities
  in [atom/b-cm] for the nuclides listed in the ``nuclides`` element, in the
  same order.

  For a "watt" distribution, ``parameters`` should be given as two real numbers
  :math:`a` and :math:`b` that parameterize the distribution :math:`p(x) dx = c
  e^{-x/a} \sinh \sqrt{b \, x} dx`.

  For a "maxwell" distribution, ``parameters`` should be given as one real
  number :math:`a` that parameterizes the distribution :math:`p(x) dx = c x
  e^{-x/a} dx`.

  .. note:: The above format should be used even when using the multi-group
            :ref:`energy_mode`.

:interpolation:
  For a "tabular" distribution, ``interpolation`` can be set to "histogram" or
  "linear-linear" thereby specifying how tabular points are to be interpolated.

  *Default*: histogram

:pair:
  For a "mixture" distribution, this element provides a distribution and its
  corresponding probability.

  :probability:
    An attribute or ``pair`` that provides the probability of a univariate
    distribution within a "mixture" distribution.

  :dist:
    This sub-element of a ``pair`` element provides information on the
    corresponding univariate distribution.

:volume:
  For a "decay_spectrum" distribution, this attribute specifies the source
  region volume in cm\ :sup:`3`. It is used together with atom densities to
  determine the absolute photon emission rate. When a source uses a
  "decay_spectrum" energy distribution, the source strength is set from this
  emission rate.

:nuclides:
  For a "decay_spectrum" distribution, this element specifies a
  whitespace-separated list of nuclide names contributing to the decay photon
  source. The atom densities for these nuclides are given by the ``parameters``
  element in the same order. Nuclides are resolved against the depletion chain,
  and nuclides without decay photon spectra do not contribute to the
  distribution.

:bias:
  This optional element specifies a biased distribution for importance sampling.
  For continuous distributions, the ``bias`` element should contain another
  univariate distribution with the same support (interval) as the parent
  distribution. For discrete distributions, the ``bias`` element should contain
  floating point values corresponding to alternative probabilities for each
  value/component to be used for biased sampling.

  *Default*: None

---------------------------------------
``<source_rejection_fraction>`` Element
---------------------------------------

The ``<source_rejection_fraction>`` element specifies the minimum fraction of
external source sites that must be accepted when applying rejection sampling
based on constraints.

   *Default*: 0.05

--------------------------
``<source_point>`` Element
--------------------------

The ``<source_point>`` element indicates at what batches the source bank
should be written. The source bank can be either written out within a state
point file or separately in a source point file. This element has the following
attributes/sub-elements:

  :batches:
    A list of integers separated by spaces indicating at what batches a state
    point file should be written. It should be noted that if the ``separate``
    attribute is not set to "true", this list must be a subset of state point
    batches.

    *Default*: Last batch only

  :separate:
    If this element is set to "true", a separate binary source point file will
    be written. Otherwise, the source sites will be written in the state point
    directly.

    *Default*: false

  :write:
    If this element is set to "false", source sites are not written
    to the state point or source point file. This can substantially reduce the
    size of state points if large numbers of particles per batch are used.

    *Default*: true

  :overwrite_latest:
    If this element is set to "true", a source point file containing
    the source bank will be written out to a separate file named
    ``source.binary`` or ``source.h5`` depending on if HDF5 is enabled.
    This file will be overwritten at every single batch so that the latest
    source bank will be available. It should be noted that a user can set both
    this element to "true" and specify batches to write a permanent source bank.

    *Default*: false

  :mcpl:
    If this element is set to "true", the source point file containing the
    source bank will be written as an MCPL_ file name ``source.mcpl`` instead of
    an HDF5 file. This option is only applicable if the ``<separate>`` element
    is set to true.

    *Default*: false

-------------------------
``<state_point>`` Element
-------------------------

The ``<state_point>`` element indicates at what batches a state point file
should be written. A state point file can be used to restart a run or to get
tally results at any batch. The default behavior when using this tag is to
write out the source bank in the state_point file. This behavior can be
customized by using the ``<source_point>`` element. This element has the
following attributes/sub-elements:

  :batches:
    A list of integers separated by spaces indicating at what batches a state
    point file should be written.

    *Default*: Last batch only

--------------------
``<stride>`` Element
--------------------

The ``stride`` element is used to specify how many random numbers are allocated
for each source particle history.

  *Default*: 152,917

------------------------------
``<surf_source_read>`` Element
------------------------------

The ``<surf_source_read>`` element specifies a surface source file for OpenMC to
read source bank for initializing histories. This element has the following
attributes/sub-elements:

  :path:
    Absolute or relative path to a surface source file to read in source bank.

    *Default*: ``surface_source.h5`` in current working directory

-------------------------------
``<surf_source_write>`` Element
-------------------------------

The ``<surf_source_write>`` element triggers OpenMC to bank particles crossing
certain surfaces and write out the source bank in a separate file called
``surface_source.h5``. One or multiple surface IDs and one cell ID can be used
to select the surfaces of interest. If no surface IDs are declared, every surface
of the model is eligible to bank particles. In that case, a cell ID (using
either the ``cell``, ``cellfrom`` or ``cellto`` attributes) can be used to select
every surface of a specific cell. This element has the following
attributes/sub-elements:

  :surface_ids:
    A list of integers separated by spaces indicating the unique IDs of surfaces
    for which crossing particles will be banked.

    *Default*: None

  :max_particles:
    An integer indicating the maximum number of particles to be banked on
    specified surfaces per processor. The size of source bank in
    ``surface_source.h5`` is limited to this value times the number of
    processors.

    *Default*: None

  :max_source_files:
    An integer value indicating the number of surface source files to be written
    containing the maximum number of particles each. The surface source bank
    will be cleared in simulation memory each time a surface source file is
    written. By default a ``surface_source.h5`` file will be created when the
    maximum number of saved particles is reached.

    *Default*: 1

  :mcpl:
    An optional boolean which indicates if the banked particles should be
    written to a file in the MCPL_-format instead of the native HDF5-based
    format. If activated the output file name is changed to
    ``surface_source.mcpl``.

    *Default*: false

    .. _MCPL: https://mctools.github.io/mcpl/mcpl.pdf

  :cell:
    An integer representing the cell ID used to determine if particles crossing
    identified surfaces are to be banked. Particles coming from or going to this
    declared cell will be banked if they cross the identified surfaces.

    *Default*: None

  :cellfrom:
    An integer representing the cell ID used to determine if particles crossing
    identified surfaces are to be banked. Particles coming from this declared
    cell will be banked if they cross the identified surfaces.

    *Default*: None

  :cellto:
    An integer representing the cell ID used to determine if particles crossing
    identified surfaces are to be banked. Particles going to this declared cell
    will be banked if they cross the identified surfaces.

    *Default*: None

.. note:: The ``cell``, ``cellfrom`` and ``cellto`` attributes cannot be
          used simultaneously.

.. note:: Surfaces with boundary conditions that are not "transmission" or "vacuum"
          are not eligible to store any particles when using ``cell``, ``cellfrom``
          or ``cellto`` attributes. It is recommended to use surface IDs instead.

------------------------------------
``<surface_grazing_cutoff>`` Element
------------------------------------

The ``<surface_grazing_cutoff>`` element specifies the surface flux cosine cutoff.

  *Default*: 0.001

-----------------------------------
``<surface_grazing_ratio>`` Element
-----------------------------------

The ``<surface_grazing_ratio>`` element specifies the surface flux cosine
substitution ratio.

  *Default*: 0.5

------------------------------
``<survival_biasing>`` Element
------------------------------

The ``<survival_biasing>`` element has no attributes and has an accepted value
of "true" or "false". If set to "true", this option will enable the use of
survival biasing, otherwise known as implicit capture or absorption.

  *Default*: false

.. _tabular_legendre:

---------------------------------
``<tabular_legendre>`` Element
---------------------------------

The optional ``<tabular_legendre>`` element specifies how the multi-group
Legendre scattering kernel is represented if encountered in a multi-group
problem.  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 of a
    Legendre scattering expansion to the tabular format should be performed or
    not. A value of “true” means the conversion should be performed, “false”
    means it will 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

  .. note:: This element is only used in the multi-group :ref:`energy_mode`.

.. _temperature_default:

---------------------------------
``<temperature_default>`` Element
---------------------------------

The ``<temperature_default>`` element specifies a default temperature in Kelvin
that is to be applied to cells in the absence of an explicit cell temperature or
a material default temperature.

  *Default*: 293.6 K

.. _temperature_method:

--------------------------------
``<temperature_method>`` Element
--------------------------------

The ``<temperature_method>`` element has an accepted value of "nearest" or
"interpolation". A value of "nearest" indicates that for each
cell, the nearest temperature at which cross sections are given is to be
applied, within a given tolerance (see :ref:`temperature_tolerance`). A value of
"interpolation" indicates that cross sections are to be linear-linear
interpolated between temperatures at which nuclear data are present (see
:ref:`temperature_treatment`). With the "interpolation" method, temperatures
outside of the bounds of the nuclear data may be accepted, provided they still
fall within the tolerance (see :ref:`temperature_tolerance`).

  *Default*: "nearest"

.. _temperature_multipole:

-----------------------------------
``<temperature_multipole>`` Element
-----------------------------------

The ``<temperature_multipole>`` element toggles the windowed multipole
capability on or off. If this element is set to "True" and the relevant data is
available, OpenMC will use the windowed multipole method to evaluate and Doppler
broaden cross sections in the resolved resonance range.  This override other
methods like "nearest" and "interpolation" in the resolved resonance range.

  *Default*: False

-------------------------------
``<temperature_range>`` Element
-------------------------------

The ``<temperature_range>`` element specifies a minimum and maximum temperature
in Kelvin above and below which cross sections should be loaded for all nuclides
and thermal scattering tables. This can be used for multi-physics simulations
where the temperatures might change from one iteration to the next.

  *Default*: None

.. _temperature_tolerance:

-----------------------------------
``<temperature_tolerance>`` Element
-----------------------------------

The ``<temperature_tolerance>`` element specifies a tolerance in Kelvin that is
to be applied when the "nearest" temperature method is used. For example, if a
cell temperature is 340 K and the tolerance is 15 K, then the closest
temperature in the range of 325 K to 355 K will be used to evaluate cross
sections. If the ``<temperature_method>`` is "interpolation", the tolerance
specified applies to cell temperatures outside of the data bounds. For example,
if a cell is specified at 695K, a tolerance of 15K and data is only available
at 700K and 1000K, the cell's cross sections will be evaluated at 700K, since
the desired temperature of 695K is within the tolerance of the actual data
despite not being bounded on both sides.

  *Default*: 10 K

.. _trace:

-------------------
``<trace>`` Element
-------------------

The ``<trace>`` element can be used to print out detailed information about a
single particle during a simulation. This element should be followed by three
integers: the batch number, generation number, and particle number.

  *Default*: None

.. _track:

-------------------
``<track>`` Element
-------------------

The ``<track>`` element specifies particles for which OpenMC will output binary
files describing particle position at every step of its transport. This element
should be followed by triplets of integers.  Each triplet describes one
particle. The integers in each triplet specify the batch number, generation
number, and particle number, respectively.

  *Default*: None

.. _trigger:

-------------------------
``<trigger>`` Element
-------------------------

OpenMC includes tally precision triggers which allow the user to define
uncertainty thresholds on :math:`k_{eff}` in the ``<keff_trigger>`` subelement
of ``settings.xml``, and/or tallies in ``tallies.xml``. When using triggers,
OpenMC will run until it completes as many batches as defined by ``<batches>``.
At this point, the uncertainties on all tallied values are computed and compared
with their corresponding trigger thresholds. If any triggers have not been met,
OpenMC will continue until either all trigger thresholds have been satisfied or
``<max_batches>`` has been reached.

The ``<trigger>`` element provides an active "toggle switch" for tally
precision trigger(s), the maximum number of batches and the batch interval. It
has the following attributes/sub-elements:

  :active:
    This determines whether or not to use trigger(s). Trigger(s) are used when
    this tag is set to "true".

  :max_batches:
    This describes the maximum number of batches allowed when using trigger(s).

    .. note:: When max_batches is set, the number of ``batches`` shown in the
              ``<batches>`` element represents minimum number of batches to
              simulate when using the trigger(s).

  :batch_interval:
    This tag describes the number of  batches in between convergence checks.
    OpenMC will check if the trigger has been reached at each batch defined
    by ``batch_interval`` after the minimum number of batches is reached.

    .. note:: If this tag is not present, the ``batch_interval`` is predicted
              dynamically by OpenMC for each convergence check. The predictive
              model assumes no correlation between fission sources
              distributions from batch-to-batch. This assumption is reasonable
              for fixed source and small criticality calculations, but is very
              optimistic for highly coupled full-core reactor problems.

-------------------------------------
``<uniform_source_sampling>`` Element
-------------------------------------

The ``<uniform_source_sampling>`` element indicates whether to sample among
multiple sources uniformly, applying their strengths as weights to sampled
particles.

  *Default*: False

------------------------
``<ufs_mesh>`` Element
------------------------

The ``<ufs_mesh>`` element indicates the ID of a mesh that is used for
re-weighting source sites at every generation based on the uniform fission site
methodology described in Kelly et al., "MC21 Analysis of the Nuclear Energy
Agency Monte Carlo Performance Benchmark Problem," Proceedings of *Physor 2012*,
Knoxville, TN (2012). The mesh should cover all possible fissionable materials
in the problem and is specified using a :ref:`mesh_element`.

-------------------------------
``<use_decay_photons>`` Element
-------------------------------

The ``<use_decay_photons>`` element indicates whether to produce decay photons
from neutron reactions instead of prompt photons. This is used in conjunction
with the direct 1-step method for shutdown dose rate calculations.

  *Default*: False

.. _verbosity:

-----------------------
``<verbosity>`` Element
-----------------------

The ``<verbosity>`` element tells the code how much information to display to
the standard output. A higher verbosity corresponds to more information being
displayed. The text of this element should be an integer between between 1
and 10. The verbosity levels are defined as follows:

  :1: don't display any output
  :2: only show OpenMC logo
  :3: all of the above + headers
  :4: all of the above + results
  :5: all of the above + file I/O
  :6: all of the above + timing statistics and initialization messages
  :7: all of the above + :math:`k` by generation
  :9: all of the above + indicate when each particle starts
  :10: all of the above + event information

  *Default*: 7

-------------------------
``<volume_calc>`` Element
-------------------------

The ``<volume_calc>`` element indicates that a stochastic volume calculation
should be run at the beginning of the simulation. This element has the following
sub-elements/attributes:

  :domain_type:
    The type of each domain for the volume calculation ("cell", "material", or
    "universe").

    *Default*: None

  :domain_ids:
    The unique IDs of domains for which the volume should be estimated.

    *Default*: None

  :samples:
    The number of samples used to estimate volumes.

    *Default*: None

  :lower_left:
    The lower-left Cartesian coordinates of a bounding box that is used to
    sample points within.

    *Default*: None

  :upper_right:
    The upper-right Cartesian coordinates of a bounding box that is used to
    sample points within.

    *Default*: None

  :threshold:
    Presence of a ``<threshold>`` sub-element indicates that the volume
    calculation will be halted based on a threshold on the error. It has the
    following sub-elements/attributes:

    :type:
      The type of the trigger. Accepted options are "variance", "std_dev",
      and "rel_err".

      :variance:
        Variance of the mean, :math:`\sigma^2`

      :std_dev:
        Standard deviation of the mean, :math:`\sigma`

      :rel_err:
        Relative error of the mean, :math:`\frac{\sigma}{\mu}`

      *Default*: None

    :threshold:
      The trigger's convergence criterion for the given type.

      *Default*: None

----------------------------
``<weight_windows>`` Element
----------------------------

The ``<weight_windows>`` element specifies all necessary parameters for
mesh-based weight windows. This element has the following
sub-elements/attributes:

  :id:
    A unique integer that is used to identify the weight windows

  :mesh:
    ID of a mesh that is to be used for weight windows

    *Default*: None

  :particle_type:
    The particle that the weight windows will apply to, specified as a PDG
    code or string (e.g., ``neutron``).

    *Default*: 'neutron'

  :energy_bins:
    Monotonically increasing list of bounding energies in [eV] to be used for
    weight windows

    *Default*: None

  :lower_ww_bounds:
    Lower weight window bound for each (energy bin, mesh bin) combination.

    *Default*: None

  :upper_ww_bounds:
    Upper weight window bound for each (energy bin, mesh bin) combination.

    *Default*: None

  :survival:
    The ratio of survival weight and lower weight window bound.

    *Default*: 3.0

  :max_lower_bound_ratio:
    Maximum allowed ratio of a particle's weight to the weight window's lower
    bound. A factor will be applied to raise the weight window to be lower than
    the particle's weight by a factor of max_lower_bound_ratio during transport
    if exceeded.

  :max_split:
    Maximum allowable number of particles when splitting

    *Default*: 10

  :weight_cutoff:
    Threshold below which particles will be terminated

    *Default*: :math:`10^{-38}`

--------------------------------------
``<weight_window_generator>`` Element
--------------------------------------

The ``<weight_window_generator>`` element provides information for creating a set of
mesh-based weight windows.

  :mesh:
    ID of a mesh that is to be used for the weight windows spatial bins

    *Default*: None

  :energy_bounds:
    The weight window energy bounds. If not present, the max/min energy of the
    cross section data is applied as a single energy bin.

    *Default*: None

  :particle_type:
    The particle that the weight windows will apply to, specified as a PDG
    code or string (e.g., ``neutron``).

    *Default*: neutron

  :max_realizations:
    The number of tally realizations after which the weight windows will stop updating.

    *Default*: 1

  :update_interval:
    The number of tally realizations between weight window updates.

    *Default*: 1

  :on_the_fly:
    Controls whether or not the tally results are reset after a weight window update.

    *Default*: true

  :method:
    Method used to update weight window values (one of 'magic' or 'fw_cadis')

    *Default*: magic

  :update_parameters:
    Method-specific update parameters used when generating/updating weight windows.

    For MAGIC:

      :value:
        The type of tally value to use when creating weight windows (one of 'mean' or 'rel_err')

        *Default*: 'mean'

      :threshold:
        The relative error threshold above which tally results will be ignored.

        *Default*: 1.0

      :ratio:
        The ratio of the lower to upper weight window bounds.

        *Default*: 5.0

    For FW-CADIS:

      :targets:
        A sequence of IDs corresponding to the tallies which cover phase
        space regions of interest for local variance reduction.

        *Default*: None

---------------------------------------
``<weight_window_checkpoints>`` Element
---------------------------------------

The ``<weight_window_checkpoints>`` element indicates the checkpoints for weight
window split/roulette (surface, collision or both). This element has the
following sub-elements/attributes:

  :surface:
    If set to "true", weight window checks will be performed at surface
    crossings.

    *Default*: False

  :collision:
    If set to "true", weight window checks will be performed at collisions.

    *Default*: True

--------------------------------------
``<weight_windows_file>`` Element
--------------------------------------

  The ``weight_windows_file`` element has no attributes and contains the path to
  a weight windows HDF5 file to load during simulation initialization.

-------------------------------
``<weight_windows_on>`` Element
-------------------------------

  The ``weight_windows_on`` element indicates whether weight windows are
  enabled.

  *Default*: False

----------------------------------
``<write_initial_source>`` Element
----------------------------------

  The ``write_initial_source`` element indicates whether to write the initial
  source distribution to file.

  *Default*: False
