mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-26 05:05:30 -04:00
1596 lines
54 KiB
ReStructuredText
1596 lines
54 KiB
ReStructuredText
.. _io_settings:
|
|
|
|
======================================
|
|
Settings Specification -- settings.xml
|
|
======================================
|
|
|
|
All simulation parameters and miscellaneous options are specified in the
|
|
settings.xml file.
|
|
|
|
---------------------
|
|
``<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.
|
|
|
|
*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
|
|
|
|
----------------------
|
|
``<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_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
|
|
|
|
---------------------
|
|
``<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
|
|
|
|
: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
|
|
|
|
: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
|
|
|
|
--------------------
|
|
``<stride>`` Element
|
|
--------------------
|
|
|
|
The ``stride`` element is used to specify how many random numbers are allocated
|
|
for each source particle history.
|
|
|
|
*Default*: 152,917
|
|
|
|
.. _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, either ``neutron`` or ``photon``.
|
|
|
|
*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.
|
|
|
|
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`).
|
|
|
|
: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.
|
|
|
|
: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 distrubtions, 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
|
|
|
|
: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
|
|
|
|
: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.
|
|
|
|
: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", 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.
|
|
|
|
---------------------------------------
|
|
``<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
|
|
|
|
-------------------------
|
|
``<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``. 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.
|
|
|
|
------------------------------
|
|
``<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:
|
|
|
|
: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 (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 (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
|
|
|
|
---------------------------------------
|
|
``<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.
|