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docs/source/usersguide/settings.rst
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docs/source/usersguide/settings.rst
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.. _usersguide_settings:
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==================
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Execution Settings
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==================
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.. currentmodule:: openmc
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Once you have created the materials and geometry for your simulation, the last
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step to have a complete model is to specify execution settings through the
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:class:`openmc.Settings` class. At a minimum, you need to specify a :ref:`source
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distribution <usersguide_source>` and :ref:`how many particles to run
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<usersguide_particles>`. Many other execution settings can be set using the
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:class:`openmc.Settings` object, but they are generally optional.
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.. _usersguide_run_modes:
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---------
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Run Modes
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---------
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The :attr:`Settings.run_mode` attribute controls what run mode is used when
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:ref:`scripts_openmc` is executed. There are five different run modes that can
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be specified:
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'eigenvalue'
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Runs a :math:`k` eigenvalue simulation. See :ref:`methods_eigenvalue` for a
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full description of eigenvalue calculations. In this mode, the
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:attr:`Settings.source` specifies a starting source that is only used for the
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first fission generation.
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'fixed source'
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Runs a fixed-source calculation with a specified external source, specified in
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the :attr:`Settings.source` attribute.
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'volume'
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Runs a stochastic volume calculation.
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'plot'
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Generates slice or voxel plots (see :ref:`usersguide_plots`).
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'particle restart'
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Simulate a single source particle using a particle restart file.
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So, for example, to specify that OpenMC should be run in fixed source mode, you
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would need to instantiate a :class:`openmc.Settings` object and assign the
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:attr:`Settings.run_mode` attribute::
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settings = openmc.Settings()
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settings.run_mode = 'fixed source'
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If you don't specify a run mode, the default run mode is 'eigenvalue'.
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.. _usersguide_particles:
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-------------------
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Number of Particles
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-------------------
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For a fixed source simulation, the total number of source particle histories
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simulated is broken up into a number of *batches*, each corresponding to a
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:ref:`realization <methods_tallies>` of the tally random variables. Thus, you
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need to specify both the number of batches (:attr:`Settings.batches`) as well as
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the number of particles per batch (:attr:`Settings.particles`).
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For a :math:`k` eigenvalue simulation, particles are grouped into *fission
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generations*, as described in :ref:`methods_eigenvalue`. Successive fission
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generations can be combined into a batch for statistical purposes. By default, a
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batch will consist of only a single fission generation, but this can be changed
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with the :attr:`Settings.generations_per_batch` attribute. For problems with a
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high dominance ratio, using multiple generations per batch can help reduce
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underprediction of variance, thereby leading to more accurate confidence
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intervals. Tallies should not be scored to until the source distribution
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converges, as described in :ref:`method-successive-generations`, which may take
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many generations. To specify the number of batches that should be discarded
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before tallies begin to accumulate, use the :attr:`Settings.inactive` attribute.
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The following example shows how one would simulate 10000 particles per
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generation, using 10 generations per batch, 150 total batches, and discarding 5
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batches. Thus, a total of 145 active batches (or 1450 generations) will be used
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for accumulating tallies.
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::
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settings.particles = 10000
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settings.generations_per_batch = 10
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settings.batches = 150
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settings.inactive = 5
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.. _usersguide_source:
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-----------------------------
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External Source Distributions
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-----------------------------
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External source distributions can be specified through the
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:attr:`Settings.source` attribute. If you have a single external source, you can
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create an instance of :class:`openmc.Source` and use it to set the
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:attr:`Settings.source` attribute. If you have multiple external sources with
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varying source strengths, :attr:`Settings.source` should be set to a list of
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:class:`openmc.Source` objects.
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The :class:`openmc.Source` class has three main attributes that one can set:
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:attr:`Source.space`, which defines the spatial distribution,
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:attr:`Source.angle`, which defines the angular distribution, and
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:attr:`Source.energy`, which defines the energy distribution.
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The spatial distribution can be set equal to a sub-class of
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:class:`openmc.stats.Spatial`; common choices are :class:`openmc.stats.Point` or
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:class:`openmc.stats.Box`. To independently specify distributions in the
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:math:`x`, :math:`y`, and :math:`z` coordinates, you can use
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:class:`openmc.stats.CartesianIndependent`.
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The angular distribution can be set equal to a sub-class of
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:class:`openmc.stats.UnitSphere` such as :class:`openmc.stats.Isotropic`,
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:class:`openmc.stats.Monodirectional`, or
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:class:`openmc.stats.PolarAzimuthal`. By default, if no angular distribution is
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specified, an isotropic angular distribution is used.
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The energy distribution can be set equal to any univariate probability
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distribution. This could be a probability mass function
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(:class:`openmc.stats.Discrete`), a Watt fission spectrum
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(:class:`openmc.stats.Watt`), or a tabular distribution
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(:class:`openmc.stats.Tabular`). By default, if no energy distribution is
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specified, a Watt fission spectrum with :math:`a` = 0.988 MeV and :math:`b` =
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2.249 MeV :sup:`-1` is used.
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As an example, to create an isotropic, 10 MeV monoenergetic source uniformly
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distributed over a cube centered at the origin with an edge length of 10 cm, one
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would run::
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source = openmc.Source()
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source.space = openmc.stats.Box((-5, -5, -5), (5, 5, 5))
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source.angle = openmc.stats.Isotropic()
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source.energy = openmc.stats.Discrete([10.0e6], [1.0])
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settings.source = source
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The :class:`openmc.Source` class also has a :attr:`Source.strength` attribute
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that indicates the relative strength of a source distribution if multiple are
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used. For example, to create two sources, one that should be sampled 70% of the
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time and another that should be sampled 30% of the time::
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src1 = openmc.Source()
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src1.strength = 0.7
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...
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src2 = openmc.Source()
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src2.strength = 0.3
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...
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settings.source = [src1, src2]
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Finally, the :attr:`Source.particle` attribute can be used to indicate the
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source should be composed of particles other than neutrons. For example, the
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following would generate a photon source::
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source = openmc.Source()
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source.particle = 'photon'
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...
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settings.source = source
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For a full list of all classes related to statistical distributions, see
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:ref:`pythonapi_stats`.
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---------------
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Shannon Entropy
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---------------
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To assess convergence of the source distribution, the scalar Shannon entropy
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metric is often used in Monte Carlo codes. OpenMC also allows you to calculate
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Shannon entropy at each generation over a specified mesh, created using the
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:class:`openmc.Mesh` class. After instantiating a :class:`Mesh`, you need to
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specify the lower-left coordinates of the mesh (:attr:`Mesh.lower_left`), the
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number of mesh cells in each direction (:attr:`Mesh.dimension`) and either the
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upper-right coordinates of the mesh (:attr:`Mesh.upper_right`) or the width of
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each mesh cell (:attr:`Mesh.width`). Once you have a mesh, simply assign it to
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the :attr:`Settings.entropy_mesh` attribute.
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::
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entropy_mesh = openmc.Mesh()
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entropy_mesh.lower_left = (-50, -50, -25)
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entropy_mesh.upper_right = (50, 50, 25)
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entropy_mesh.dimension = (8, 8, 8)
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settings.entropy_mesh = entropy_mesh
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If you're unsure of what bounds to use for the entropy mesh, you can try getting
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a bounding box for the entire geometry using the :attr:`Geometry.bounding_box`
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property::
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geom = openmc.Geometry()
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...
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m = openmc.Mesh()
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m.lower_left, m.upper_right = geom.bounding_box
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m.dimension = (8, 8, 8)
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settings.entropy_mesh = m
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----------------
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Photon Transport
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----------------
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In addition to neutrons, OpenMC is also capable of simulating the passage of
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photons through matter. This allows the modeling of photon production from
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neutrons as well as pure photon calculations. The
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:attr:`Settings.photon_transport` attribute can be used to enable photon
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transport::
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settings.photon_transport = True
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The way in which OpenMC handles secondary charged particles can be specified
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with the :attr:`Settings.electron_treatment` attribute. By default, the
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:ref:`thick-target bremsstrahlung <ttb>` (TTB) approximation is used to generate
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bremsstrahlung radiation emitted by electrons and positrons created in photon
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interactions. To neglect secondary bremsstrahlung photons and instead deposit
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all energy from electrons locally, the local energy deposition option can be
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selected::
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settings.electron_treatment = 'led'
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.. note::
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Some features related to photon transport are not currently implemented,
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including:
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* Tallying photon energy deposition.
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* Generating a photon source from a neutron calculation that can be used
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for a later fixed source photon calculation.
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* Photoneutron reactions.
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--------------------------
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Generation of Output Files
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--------------------------
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A number of attributes of the :class:`openmc.Settings` class can be used to
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control what files are output and how often. First, there is the
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:attr:`Settings.output` attribute which takes a dictionary having keys
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'summary', 'tallies', and 'path'. The first two keys controls whether a
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``summary.h5`` and ``tallies.out`` file are written, respectively (see
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:ref:`result_files` for a description of those files). By default, output files
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are written to the current working directory; this can be changed by setting the
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'path' key. For example, if you want to disable the ``tallies.out`` file and
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write the ``summary.h5`` to a directory called 'results', you'd specify the
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:attr:`Settings.output` dictionary as::
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settings.output = {
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'tallies': False,
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'path': 'results'
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}
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Generation of statepoint and source files is handled separately through the
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:attr:`Settings.statepoint` and :attr:`Settings.sourcepoint` attributes. Both of
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those attributes expect dictionaries and have a 'batches' key which indicates at
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which batches statepoints and source files should be written. Note that by
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default, the source is written as part of the statepoint file; this behavior can
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be changed by the 'separate' and 'write' keys of the
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:attr:`Settings.sourcepoint` dictionary, the first of which indicates whether
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the source should be written to a separate file and the second of which
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indicates whether the source should be written at all.
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As an example, to write a statepoint file every five batches::
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settings.batches = n
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settings.statepoint = {'batches': range(5, n + 5, 5)}
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.. _NIST ESTAR database: https://physics.nist.gov/PhysRefData/Star/Text/ESTAR.html
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