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1106 lines
39 KiB
ReStructuredText
1106 lines
39 KiB
ReStructuredText
.. _io_settings:
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======================================
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Settings Specification -- settings.xml
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======================================
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All simulation parameters and miscellaneous options are specified in the
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settings.xml file.
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---------------------
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``<batches>`` Element
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---------------------
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The ``<batches>`` element indicates the total number of batches to execute,
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where each batch corresponds to a tally realization. In a fixed source
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calculation, each batch consists of a number of source particles. In an
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eigenvalue calculation, each batch consists of one or many fission source
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iterations (generations), where each generation itself consists of a number of
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source neutrons.
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*Default*: None
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----------------------------------
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``<confidence_intervals>`` Element
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----------------------------------
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The ``<confidence_intervals>`` element has no attributes and has an accepted
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value of "true" or "false". If set to "true", uncertainties on tally results
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will be reported as the half-width of the 95% two-sided confidence interval. If
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set to "false", uncertainties on tally results will be reported as the sample
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standard deviation.
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*Default*: false
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-------------------------------------
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``<create_delayed_neutrons>`` Element
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-------------------------------------
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The ``<create_delayed_neutrons>`` element indicates whether delayed neutrons
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are created in fission. If this element is set to "true", delayed neutrons
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will be created in fission events; otherwise only prompt neutrons will be
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created.
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*Default*: true
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-------------------------------------
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``<create_fission_neutrons>`` Element
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-------------------------------------
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The ``<create_fission_neutrons>`` element indicates whether fission neutrons
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should be created or not. If this element is set to "true", fission neutrons
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will be created; otherwise the fission is treated as capture and no fission
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neutron will be created. Note that this option is only applied to fixed source
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calculation. For eigenvalue calculation, fission will always be treated as real
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fission.
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*Default*: true
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--------------------
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``<cutoff>`` Element
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--------------------
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The ``<cutoff>`` element indicates two kinds of cutoffs. The first is the weight
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cutoff used below which particles undergo Russian roulette. Surviving particles
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are assigned a user-determined weight. Note that weight cutoffs and Russian
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rouletting are not turned on by default. The second is the energy cutoff which
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is used to kill particles under certain energy. The energy cutoff should not be
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used unless you know particles under the energy are of no importance to results
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you care. This element has the following attributes/sub-elements:
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:weight:
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The weight below which particles undergo Russian roulette.
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*Default*: 0.25
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:weight_avg:
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The weight that is assigned to particles that are not killed after Russian
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roulette.
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*Default*: 1.0
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:energy_neutron:
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The energy under which neutrons will be killed.
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*Default*: 0.0
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:energy_photon:
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The energy under which photons will be killed.
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*Default*: 1000.0
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:energy_electron:
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The energy under which electrons will be killed.
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*Default*: 0.0
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:energy_positron:
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The energy under which positrons will be killed.
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*Default*: 0.0
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----------------------------
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``<delayed_photon_scaling>``
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----------------------------
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Determines whether to scale the fission photon yield to account for delayed
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photon energy. The photon yields are scaled as (EGP + EGD)/EGP where EGP and EGD
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are the prompt and delayed photon components of energy release, respectively,
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from MF=1, MT=458 on an ENDF evaluation.
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*Default*: true
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--------------------------------
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``<electron_treatment>`` Element
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--------------------------------
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When photon transport is enabled, the ``<electron_treatment>`` element tells
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OpenMC whether to deposit all energy from electrons locally (``led``) or create
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secondary bremsstrahlung photons (``ttb``).
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*Default*: ttb
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.. _energy_mode:
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-------------------------
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``<energy_mode>`` Element
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-------------------------
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The ``<energy_mode>`` element tells OpenMC if the run-mode should be
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continuous-energy or multi-group. Options for entry are: ``continuous-energy``
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or ``multi-group``.
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*Default*: continuous-energy
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--------------------------
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``<entropy_mesh>`` Element
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--------------------------
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The ``<entropy_mesh>`` element indicates the ID of a mesh that is to be used for
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calculating Shannon entropy. The mesh should cover all possible fissionable
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materials in the problem and is specified using a :ref:`mesh_element`.
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----------------------------
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``<event_based>``
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----------------------------
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Determines whether to use event-based parallelism instead of the default
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history-based parallelism.
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*Default*: false
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-----------------------------------
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``<generations_per_batch>`` Element
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-----------------------------------
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The ``<generations_per_batch>`` element indicates the number of total fission
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source iterations per batch for an eigenvalue calculation. This element is
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ignored for all run modes other than "eigenvalue".
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*Default*: 1
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----------------------
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``<inactive>`` Element
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----------------------
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The ``<inactive>`` element indicates the number of inactive batches used in a
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k-eigenvalue calculation. In general, the starting fission source iterations in
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an eigenvalue calculation can not be used to contribute to tallies since the
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fission source distribution and eigenvalue are generally not converged
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immediately. This element is ignored for all run modes other than "eigenvalue".
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*Default*: 0
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--------------------------
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``<keff_trigger>`` Element
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--------------------------
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The ``<keff_trigger>`` element (ignored for all run modes other than
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"eigenvalue".) specifies a precision trigger on the combined
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:math:`k_{eff}`. The trigger is a convergence criterion on the uncertainty of
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the estimated eigenvalue. It has the following attributes/sub-elements:
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:type:
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The type of precision trigger. Accepted options are "variance", "std_dev",
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and "rel_err".
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:variance:
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Variance of the batch mean :math:`\sigma^2`
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:std_dev:
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Standard deviation of the batch mean :math:`\sigma`
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:rel_err:
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Relative error of the batch mean :math:`\frac{\sigma}{\mu}`
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*Default*: None
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:threshold:
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The precision trigger's convergence criterion for the
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combined :math:`k_{eff}`.
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*Default*: None
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.. note:: See section on the :ref:`trigger` for more information.
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---------------------------
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``<log_grid_bins>`` Element
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---------------------------
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The ``<log_grid_bins>`` element indicates the number of bins to use for the
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logarithmic-mapped energy grid. Using more bins will result in energy grid
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searches over a smaller range at the expense of more memory. The default is
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based on the recommended value in LA-UR-14-24530_.
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*Default*: 8000
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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.. _LA-UR-14-24530: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-14-24530.pdf
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---------------------------
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``<material_cell_offsets>``
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---------------------------
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By default, OpenMC will count the number of instances of each cell filled with a
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material and generate "offset tables" that are used for cell instance tallies.
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The ``<material_cell_offsets>`` element allows a user to override this default
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setting and turn off the generation of offset tables, if desired, by setting it
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to false.
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*Default*: true
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----------------------------------------
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``<max_particles_in_flight>`` Element
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----------------------------------------
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This element indicates the number of neutrons to run in flight concurrently
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when using event-based parallelism. A higher value uses more memory, but
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may be more efficient computationally.
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*Default*: 100000
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---------------------------
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``<max_order>`` Element
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---------------------------
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The ``<max_order>`` element allows the user to set a maximum scattering order
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to apply to every nuclide/material in the problem. That is, if the data
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library has :math:`P_3` data available, but ``<max_order>`` was set to ``1``,
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then, OpenMC will only use up to the :math:`P_1` data.
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*Default*: Use the maximum order in the data library
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.. note:: This element is not used in the continuous-energy
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:ref:`energy_mode`.
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.. _mesh_element:
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------------------
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``<mesh>`` Element
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------------------
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The ``<mesh>`` element describes a mesh that is used either for calculating
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Shannon entropy, applying the uniform fission site method, or in tallies. For
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Shannon entropy meshes, the mesh should cover all possible fissionable materials
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in the problem. It has the following attributes/sub-elements:
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:id:
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A unique integer that is used to identify the mesh.
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:dimension:
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The number of mesh cells in the x, y, and z directions, respectively.
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*Default*: If this tag is not present, the number of mesh cells is
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automatically determined by the code.
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:lower_left:
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The Cartesian coordinates of the lower-left corner of the mesh.
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*Default*: None
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:upper_right:
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The Cartesian coordinates of the upper-right corner of the mesh.
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*Default*: None
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-----------------------
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``<no_reduce>`` Element
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-----------------------
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The ``<no_reduce>`` element has no attributes and has an accepted value of
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"true" or "false". If set to "true", all user-defined tallies and global tallies
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will not be reduced across processors in a parallel calculation. This means that
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the accumulate score in one batch on a single processor is considered as an
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independent realization for the tally random variable. For a problem with large
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tally data, this option can significantly improve the parallel efficiency.
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*Default*: false
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--------------------
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``<output>`` Element
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--------------------
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The ``<output>`` element determines what output files should be written to disk
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during the run. The sub-elements are described below, where "true" will write
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out the file and "false" will not.
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:summary:
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Writes out an HDF5 summary file describing all of the user input files that
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were read in.
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*Default*: true
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:tallies:
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Write out an ASCII file of tally results.
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*Default*: true
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.. note:: The tally results will always be written to a binary/HDF5 state
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point file.
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:path:
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Absolute or relative path where all output files should be written to. The
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specified path must exist or else OpenMC will abort.
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*Default*: Current working directory
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-----------------------
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``<particles>`` Element
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-----------------------
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This element indicates the number of neutrons to simulate per fission source
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iteration when a k-eigenvalue calculation is performed or the number of
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particles per batch for a fixed source simulation.
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*Default*: None
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------------------------------
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``<photon_transport>`` Element
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------------------------------
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The ``<photon_transport>`` element determines whether photon transport is
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enabled. This element has no attributes or sub-elements and can be set to
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either "false" or "true".
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*Default*: false
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---------------------
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``<ptables>`` Element
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---------------------
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The ``<ptables>`` element determines whether probability tables should be used
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in the unresolved resonance range if available. This element has no attributes
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or sub-elements and can be set to either "false" or "true".
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*Default*: true
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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----------------------------------
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``<resonance_scattering>`` Element
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----------------------------------
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The ``resonance_scattering`` element indicates to OpenMC that a method be used
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to properly account for resonance elastic scattering (typically for nuclides
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with Z > 40). This element can contain one or more of the following attributes
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or sub-elements:
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:enable:
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Indicates whether a resonance elastic scattering method should be turned
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on. Accepts values of "true" or "false".
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*Default*: If the ``<resonance_scattering>`` element is present, "true".
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:method:
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Which resonance elastic scattering method is to be applied: "rvs" (relative
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velocity sampling) or "dbrc" (Doppler broadening rejection correction).
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Descriptions of each of these methods are documented here_.
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.. _here: https://doi.org/10.1016/j.anucene.2017.12.044
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*Default*: "rvs"
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:energy_min:
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The energy in eV above which the resonance elastic scattering method should
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be applied.
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*Default*: 0.01 eV
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:energy_max:
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The energy in eV below which the resonance elastic scattering method should
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be applied.
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*Default*: 1000.0 eV
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:nuclides:
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A list of nuclides to which the resonance elastic scattering method should
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be applied.
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*Default*: If ``<resonance_scattering>`` is present but the ``<nuclides>``
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sub-element is not given, the method is applied to all nuclides with 0 K
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elastic scattering data present.
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.. note:: If the ``resonance_scattering`` element is not given, the free gas,
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constant cross section scattering model, which has historically been
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used by Monte Carlo codes to sample target velocities, is used to
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treat the target motion of all nuclides. If
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``resonance_scattering`` is present, the constant cross section
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method is applied below ``energy_min`` and the target-at-rest
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(asymptotic) kernel is used above ``energy_max``.
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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----------------------
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``<run_mode>`` Element
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----------------------
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The ``<run_mode>`` element indicates which run mode should be used when OpenMC
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is executed. This element has no attributes or sub-elements and can be set to
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"eigenvalue", "fixed source", "plot", "volume", or "particle restart".
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*Default*: None
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------------------
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``<seed>`` Element
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------------------
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The ``seed`` element is used to set the seed used for the linear congruential
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pseudo-random number generator.
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*Default*: 1
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--------------------
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``<source>`` Element
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--------------------
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The ``source`` element gives information on an external source distribution to
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be used either as the source for a fixed source calculation or the initial
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source guess for criticality calculations. Multiple ``<source>`` elements may be
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specified to define different source distributions. Each one takes the following
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attributes/sub-elements:
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:strength:
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The strength of the source. If multiple sources are present, the source
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strength indicates the relative probability of choosing one source over the
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other.
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*Default*: 1.0
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:particle:
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The source particle type, either ``neutron`` or ``photon``.
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*Default*: neutron
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:file:
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If this attribute is given, it indicates that the source is to be read from
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a binary source file whose path is given by the value of this element. Note,
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the number of source sites needs to be the same as the number of particles
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simulated in a fission source generation.
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*Default*: None
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:library:
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If this attribute is given, it indicates that the source is to be
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instantiated from an externally compiled source function. This source can be
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as complex as is required to define the source for your problem. The library
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has a few basic requirements:
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* It must contain a class that inherits from ``openmc::Source``;
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* The class must implement a function called ``sample()``;
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* There must be an ``openmc_create_source()`` function that creates the source
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as a unique pointer. This function can be used to pass parameters through to
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the source from the XML, if needed.
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More documentation on how to build sources can be found in :ref:`custom_source`.
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*Default*: None
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:parameters:
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If this attribute is given, it provides the parameters to pass through to the
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class generated using the ``library`` parameter . More documentation on how to
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build parametrized sources can be found in :ref:`parameterized_custom_source`.
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*Default*: None
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:space:
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An element specifying the spatial distribution of source sites. This element
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has the following attributes:
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:type:
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The type of spatial distribution. Valid options are "box", "fission",
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"point", "cartesian", "cylindrical", and "spherical". A "box" spatial
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distribution has coordinates sampled uniformly in a parallelepiped. A
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"fission" spatial distribution samples locations from a "box"
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distribution but only locations in fissionable materials are accepted.
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A "point" spatial distribution has coordinates specified by a triplet.
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A "cartesian" spatial distribution specifies independent distributions of
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x-, y-, and z-coordinates. A "cylindrical" spatial distribution specifies
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independent distributions of r-, phi-, and z-coordinates where phi is the
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azimuthal angle and the origin for the cylindrical coordinate system is
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specified by origin. A "spherical" spatial distribution specifies
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independent distributions of r-, cos_theta-, and phi-coordinates where
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cos_theta is the cosine of the angle with respect to the z-axis, phi is
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the azimuthal angle, and the sphere is centered on the coordinate
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(x0,y0,z0). A "mesh" spatial distribution samples source sites from a mesh element
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based on the relative strengths provided in the node. Source locations
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within an element are sampled isotropically. If no strengths are provided,
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the space within the mesh is uniformly sampled.
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*Default*: None
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:parameters:
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For a "box" or "fission" spatial distribution, ``parameters`` should be
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given as six real numbers, the first three of which specify the lower-left
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corner of a parallelepiped and the last three of which specify the
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upper-right corner. Source sites are sampled uniformly through that
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parallelepiped.
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For a "point" spatial distribution, ``parameters`` should be given as
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three real numbers which specify the (x,y,z) location of an isotropic
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point source.
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For an "cartesian" distribution, no parameters are specified. Instead,
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the ``x``, ``y``, and ``z`` elements must be specified.
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For a "cylindrical" distribution, no parameters are specified. Instead,
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the ``r``, ``phi``, ``z``, and ``origin`` elements must be specified.
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For a "spherical" distribution, no parameters are specified. Instead,
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the ``r``, ``theta``, ``phi``, and ``origin`` elements must be specified.
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*Default*: None
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:x:
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For an "cartesian" distribution, this element specifies the distribution
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of x-coordinates. The necessary sub-elements/attributes are those of a
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univariate probability distribution (see the description in
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:ref:`univariate`).
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:y:
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For an "cartesian" distribution, this element specifies the distribution
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of y-coordinates. The necessary sub-elements/attributes are those of a
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univariate probability distribution (see the description in
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:ref:`univariate`).
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:z:
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For both "cartesian" and "cylindrical" distributions, this element
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specifies the distribution of z-coordinates. The necessary
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sub-elements/attributes are those of a univariate probability
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distribution (see the description in :ref:`univariate`).
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|
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:r:
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For "cylindrical" and "spherical" distributions, this element specifies
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the distribution of r-coordinates (cylindrical radius and spherical
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radius, respectively). The necessary sub-elements/attributes are those
|
|
of a univariate probability distribution (see the description in
|
|
:ref:`univariate`).
|
|
|
|
: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`).
|
|
|
|
:origin:
|
|
For "cylindrical and "spherical" distributions, this element specifies
|
|
the coordinates for the origin of the coordinate system.
|
|
|
|
: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`).
|
|
|
|
: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
|
|
|
|
.. _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", and "mixture". 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.
|
|
|
|
*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 "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.
|
|
|
|
-------------------------
|
|
``<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
|
|
|
|
--------------------------
|
|
``<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
|
|
|
|
------------------------------
|
|
``<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``. 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
|
|
|
|
: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
|
|
|
|
------------------------------
|
|
``<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.
|
|
|
|
|
|
------------------------
|
|
``<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`.
|
|
|
|
.. _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:
|
|
|
|
:cells:
|
|
The unique IDs of cells 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
|
|
|
|
----------------------------
|
|
``<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 (e.g., 'neutron')
|
|
|
|
*Default*: None
|
|
|
|
: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
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bound. A factor will be applied to raise the weight window to be lower than
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the particle's weight by a factor of max_lower_bound_ratio during transport
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if exceeded.
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:max_split:
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Maximum allowable number of particles when splitting
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*Default*: 10
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:weight_cutoff:
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Threshold below which particles will be terminated
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*Default*: :math:`10^{-38}`
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