mirror of
https://github.com/openmc-dev/openmc.git
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Greatly expanded documentation.
This commit is contained in:
Paul Romano
parent
54feb3fe20
commit
84dde9b1e3
19 changed files with 877 additions and 356 deletions
|
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@ -1,8 +1,8 @@
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.. _usersguide_input:
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========================
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Creating XML Input Files
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========================
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=======================
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Writing XML Input Files
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=======================
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Unlike many other Monte Carlo codes which use an arbitrary-format ASCII file
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with "cards" to specify a particular geometry, materials, and associated run
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@ -12,7 +12,9 @@ to be exchanged efficiently between different programs and interfaces.
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Anyone who has ever seen webpages written in HTML will be familiar with the
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structure of XML whereby "tags" enclosed in angle brackets denote that a
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particular piece of data will follow. Let us examine the follow example::
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particular piece of data will follow. Let us examine the follow example:
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.. code-block:: xml
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<person>
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<firstname>John</firstname>
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@ -35,14 +37,190 @@ Overview of Files
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-----------------
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To assemble a complete model for OpenMC, one needs to create separate XML files
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for the geometry, materials, and settings. Additionally, an optional tallies XML
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file specifies physical quantities to be tallied. OpenMC expects that these
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files are called:
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for the geometry, materials, and settings. Additionally, there are two optional
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input files. The first is a tallies XML file that specifies physical quantities
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to be tallied. The second is a plots XML file that specifies regions of geometry
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which should be plotted. OpenMC expects that these files are called:
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* ``geometry.xml``
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* ``materials.xml``
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* ``setings.xml``
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* ``tallies.xml``
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* ``plots.xml``
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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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``<criticality>`` Element
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-------------------------
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The ``<criticality>`` element indicates that a criticality calculation should be
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performed. It has the following attributes/sub-elements:
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:batches:
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The total number of batches, where each batch corresponds to multiple
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fission source iterations. Batching is done to eliminate correlation between
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realizations of random variables.
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*Default*: None
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:generations_per_batch:
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The number of total fission source iterations per batch.
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*Default*: 1
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:inactive:
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The number of inactive batches. In general, the starting cycles in a
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criticality 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.
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*Default*: None
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:particles:
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The number of neutrons to simulate per fission source iteration.
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*Default*: None
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.. _cross_sections:
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``<cross_sections>`` Element
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----------------------------
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The ``<cross_sections>`` element has no attributes and simply indicates the path
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to an XML cross section listing file (usually named cross_sections.xml). If this
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element is absent from the settings.xml file, the :envvar:`CROSS_SECTIONS`
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environment variable will be used to find the path to the XML cross section
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listing.
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``<cutoff>`` Element
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--------------------
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The ``<cutoff>`` element indicates the weight cutoff used below which particles
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undergo Russian roulette. Surviving particles are assigned a user-determined
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weight. Note that weight cutoffs and Russian rouletting are not turned on by
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default. 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_grid>`` Element
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-------------------------
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The ``<energy_grid>`` element determines the treatment of the energy grid during
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a simulation. Setting this element to "nuclide" will cause OpenMC to use a
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nuclide's energy grid when determining what points to interpolate between for
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determining cross sections (i.e. non-unionized energy grid). To use a unionized
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energy grid, set this element to "union". Note that the unionized energy grid
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treatment is slightly different than that employed in Serpent.
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*Default*: union
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``<entropy>`` Element
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---------------------
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The ``<entropy>`` element describes a mesh that is used for calculting Shannon
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entropy. This mesh should cover all possible fissionable materials in the
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problem. It has the following attributes/sub-elements:
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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 Cartersian coordinates of the lower-left corner of the mesh.
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*Default*: None
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|
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:upper_right:
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The Cartersian coordinates of the upper-right corner of the mesh.
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*Default*: None
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|
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``<ptables>`` Element
|
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---------------------
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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 "off" or "on".
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*Default*: on
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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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``<source>`` Element
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--------------------
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The ``source`` element gives information on an initial source guess for
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criticality calculations. It takes the following attributes:
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:type:
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The type of source distribution. Setting this to "box" indicates that the
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starting source should be sampled uniformly in a parallelepiped. Setting
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this to "point" indicates that the starting source should be sampled from an
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isotropic point source. Setting this to "file" indicates that the starting
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source should be sampled from a ``source.binary`` file.
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:coeffs:
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For a "box" source distribution, ``coeffs`` should be given as six real
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numbers, the first three of which specify the lower-left corner of a
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parallelepiped and the last three of which specify the upper-right
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corner. Source sites are sampled uniformly through that parallelepiped.
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For a "point" source distribution, ``coeffs`` should be given as three real
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numbers which specify the (x,y,z) location of an isotropic point source
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For a "file" source distribution, ``coeffs`` should not be specified.
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``<survival_biasing>`` Element
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------------------------------
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The ``<survival_biasing>`` element has no attributes and assumes wither the
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value ``on`` or ``off``. If turned on, this option will enable the use of
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survival biasing, otherwise known as implicit capture or absorption.
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*Default*: off
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``<trace>`` Element
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-------------------
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The ``<trace>`` element can be used to print out detailed information about a
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single particle during a simulation. This element should be followed by three
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integers: the batch number, generation number, and particle number.
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*Default*: None
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``<verbosity>`` Element
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-----------------------
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The ``<verbosity>`` element tells the code how much information to display to
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the standard output. A higher verbosity corresponds to more information being
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displayed. This element takes the following attributes:
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:value:
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The specified verbosity between 1 and 10.
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*Default*: 5
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--------------------------------------
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Geometry Specification -- geometry.xml
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@ -57,9 +235,11 @@ bounding surfaces.
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Every geometry.xml must have an XML declaration at the beginning of the file and
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a root element named geometry. Within the root element the user can define any
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number of cells, surfaces, and lattices. Let us look at the following example::
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||||
number of cells, surfaces, and lattices. Let us look at the following example:
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||||
|
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<?xml version="1.0">
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.. code-block:: xml
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||||
|
||||
<?xml version="1.0"?>
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||||
<geometry>
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<!-- This is a comment -->
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||||
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|
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@ -83,7 +263,9 @@ At the beginning of this file is a comment, denoted by a tag starting with
|
|||
may span multiple lines. One convenient feature of the XML input format is that
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||||
sub-elements of the ``cell`` and ``surface`` elements can also be equivalently
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expressed of attributes of the original element, e.g. the geometry file above
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could be written as::
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||||
could be written as:
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||||
|
||||
.. code-block:: xml
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||||
|
||||
<?xml version="1.0">
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<geometry>
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|
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@ -94,10 +276,10 @@ could be written as::
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||||
</geometry>
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||||
|
||||
``surface`` Element
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-------------------
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``<surface>`` Element
|
||||
---------------------
|
||||
|
||||
Each ``surface`` element can have the following attributes or sub-elements:
|
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Each ``<surface>`` element can have the following attributes or sub-elements:
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|
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:id:
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A unique integer that can be used to identify the surface.
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||||
|
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@ -105,8 +287,8 @@ Each ``surface`` element can have the following attributes or sub-elements:
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*Default*: None
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:type:
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The type of the surfaces. This can be ``x-plane``, ``y-plane``, ``z-plane``,
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``plane``, ``x-cylinder``, ``y-cylinder``, ``z-cylinder``, or ``sphere``.
|
||||
The type of the surfaces. This can be "x-plane", "y-plane", "z-plane",
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"plane", "x-cylinder", "y-cylinder", "z-cylinder", or "sphere".
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||||
*Default*: None
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|
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@ -117,10 +299,10 @@ Each ``surface`` element can have the following attributes or sub-elements:
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*Default*: None
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:boundary:
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The boundary condition for the surface. This can be ``transmission``,
|
||||
``vacuum``, or ``reflective``.
|
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The boundary condition for the surface. This can be "transmission",
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||||
"vacuum", or "reflective".
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*Default*: ``transmission``
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*Default*: "transmission"
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The following quadratic surfaces can be modeled:
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|
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@ -159,10 +341,10 @@ The following quadratic surfaces can be modeled:
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A sphere of the form :math:`(x - x_0)^2 + (y - y_0)^2 + (z - z_0)^2 =
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R^2`. The coefficients specified are ":math:`x_0 \: y_0 \: z_0 \: R`".
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||||
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||||
``cell`` Element
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----------------
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``<cell>`` Element
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||||
------------------
|
||||
|
||||
Each ``cell`` element can have the following attributes or sub-elements:
|
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Each ``<cell>`` element can have the following attributes or sub-elements:
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||||
:id:
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A unique integer that can be used to identify the surface.
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|
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@ -195,13 +377,13 @@ Each ``cell`` element can have the following attributes or sub-elements:
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|||
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*Default*: None
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``lattice`` Element
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-------------------
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``<lattice>`` Element
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||||
---------------------
|
||||
|
||||
The ``lattice`` can be used to represent repeating structures (e.g. fuel pins in
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an assembly) or other geometry which naturally fits into a two-dimensional
|
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The ``<lattice>`` can be used to represent repeating structures (e.g. fuel pins
|
||||
in an assembly) or other geometry which naturally fits into a two-dimensional
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structured mesh. Each cell within the lattice is filled with a specified
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universe. A ``lattice`` accepts the following attributes or sub-elements:
|
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universe. A ``<lattice>`` accepts the following attributes or sub-elements:
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:id:
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A unique integer that can be used to identify the surface.
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|
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@ -240,8 +422,8 @@ universe. A ``lattice`` accepts the following attributes or sub-elements:
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Materials Specification -- materials.xml
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----------------------------------------
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``material`` Element
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||||
--------------------
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||||
``<material>`` Element
|
||||
----------------------
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||||
|
||||
Each ``material`` element can have the following attributes or sub-elements:
|
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|
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@ -249,12 +431,13 @@ Each ``material`` element can have the following attributes or sub-elements:
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A unique integer that can be used to identify the material.
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:density:
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|
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An element with attributes/sub-elements called ``value`` and ``units``. The
|
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``value`` attribute is the numeric value of the density while the ``units``
|
||||
can be "g/cm3", "kg/m3", "atom/b-cm", or "atom/cm3". For example, this could
|
||||
be specified as::
|
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|
||||
<density value="4.5" units="g/cm3" />
|
||||
can be "g/cm3", "kg/m3", "atom/b-cm", "atom/cm3", or "sum". The "sum" unit
|
||||
indicates that the density should be calculated as the sum of the atom
|
||||
fractions for each nuclide in the material. This should not be used in
|
||||
conjunction with weight percents.
|
||||
|
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*Default*: None
|
||||
|
||||
|
|
@ -282,182 +465,18 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
|
||||
*Default*: None
|
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|
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``default_xs`` Element
|
||||
----------------------
|
||||
``<default_xs>`` Element
|
||||
------------------------
|
||||
|
||||
In some circumstances, the cross-section identifier may be the same for many or
|
||||
all nuclides in a given problem. In this case, rather than specifying the
|
||||
``xs=...`` attribute on every nuclide, a ``default_xs`` element can be used to
|
||||
``xs=...`` attribute on every nuclide, a ``<default_xs>`` element can be used to
|
||||
set the default cross-section identifier for any nuclide without an identifier
|
||||
explicitly listed. This element has no attributes and accepts a 3-letter string
|
||||
that indicates the default cross-section identifier, e.g. "70c".
|
||||
|
||||
*Default*: None
|
||||
|
||||
--------------------------------------
|
||||
Settings Specification -- settings.xml
|
||||
--------------------------------------
|
||||
|
||||
All simulation parameters and miscellaneous options are specified in the
|
||||
settings.xml file.
|
||||
|
||||
``criticality`` Element
|
||||
-----------------------
|
||||
|
||||
The ``criticality`` element indicates that a criticality calculation should be
|
||||
performed. It has the following attributes/sub-elements:
|
||||
|
||||
:batches:
|
||||
The total number of batches, where each batch corresponds to multiple
|
||||
fission source iterations. Batching is done to eliminate correlation between
|
||||
realizations of random variables.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:generations_per_batch:
|
||||
The number of total fission source iterations per batch.
|
||||
|
||||
*Default*: 1
|
||||
|
||||
:inactive:
|
||||
The number of inactive batches. In general, the starting cycles in a
|
||||
criticality calculation can not be used to contribute to tallies since the
|
||||
fission source distribution and eigenvalue are generally not converged
|
||||
immediately.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:particles:
|
||||
The number of neutrons to simulate per fission source iteration.
|
||||
|
||||
*Default*: None
|
||||
|
||||
``cross_sections`` Element
|
||||
--------------------------
|
||||
|
||||
The ``cross_sections`` element has no attributes and simply indicates the path
|
||||
to an XML cross section listing file (usually named cross_sections.xml). If this
|
||||
element is absent from the settings.xml file, the environment variable
|
||||
``CROSS_SECTIONS`` will be used to find the path to the XML cross section
|
||||
listing.
|
||||
|
||||
``cutoff`` Element
|
||||
------------------
|
||||
|
||||
The ``cutoff`` element indicates 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. This element has the following attributes/sub-elements:
|
||||
|
||||
: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
|
||||
|
||||
``energy_grid`` Element
|
||||
-----------------------
|
||||
|
||||
The ``energy_grid`` element determines the treatment of the energy grid during a
|
||||
simulation. Setting this element to "nuclide" will cause OpenMC to use a
|
||||
nuclide's energy grid when determining what points to interpolate between for
|
||||
determining cross sections (i.e. non-unionized energy grid). To use a unionized
|
||||
energy grid, set this element to "union". Note that the unionized energy grid
|
||||
treatment is slightly different than that employed in Serpent.
|
||||
|
||||
*Default*: union
|
||||
|
||||
``entropy`` Element
|
||||
-------------------
|
||||
|
||||
This element describes a mesh that is used for calculting Shannon entropy. This
|
||||
mesh should cover all possible fissionable materials in the problem. It has the
|
||||
following attributes/sub-elements:
|
||||
|
||||
: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 Cartersian coordinates of the lower-left corner of the mesh.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:upper_right:
|
||||
The Cartersian coordinates of the upper-right corner of the mesh.
|
||||
|
||||
*Default*: None
|
||||
|
||||
``ptables`` Element
|
||||
-------------------
|
||||
|
||||
The ``ptables`` element determines whether probability tables should be used in
|
||||
the unresolved resonance range if available. This element has no attributes or
|
||||
sub-elements and can be set to either "off" or "on".
|
||||
|
||||
*Default*: on
|
||||
|
||||
``seed`` Element
|
||||
----------------
|
||||
|
||||
The ``seed`` element is used to set the seed used for the linear congruential
|
||||
pseudo-random number generator.
|
||||
|
||||
*Default*: 1
|
||||
|
||||
``source`` Element
|
||||
------------------
|
||||
|
||||
The ``source`` element gives information on an initial source guess for
|
||||
criticality calculations. It takes the following attributes:
|
||||
|
||||
:type:
|
||||
The type of source distribution. Currently, the only accepted option is
|
||||
"box"
|
||||
|
||||
:coeffs:
|
||||
For a "box" source distribution, ``coeffs`` should be given as six integers,
|
||||
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.
|
||||
|
||||
``survival_biasing`` Element
|
||||
----------------------------
|
||||
|
||||
The ``survival_biasing`` element has no attributes and assumes wither the
|
||||
value ``on`` or ``off``. If turned on, this option will enable the use of
|
||||
survival biasing, otherwise known as implicit capture or absorption.
|
||||
|
||||
*Default*: off
|
||||
|
||||
``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 two
|
||||
integers, the cycle and one for the particle number.
|
||||
|
||||
*Default*: None
|
||||
|
||||
``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. This element takes the following attributes:
|
||||
|
||||
:value:
|
||||
The specified verbosity between 1 and 10.
|
||||
|
||||
*Default*: 5
|
||||
|
||||
------------------------------------
|
||||
Tallies Specification -- tallies.xml
|
||||
------------------------------------
|
||||
|
|
@ -476,12 +495,12 @@ filters can be used for a tally. The following types of filter are available:
|
|||
cell, universe, material, surface, birth region, pre-collision energy,
|
||||
post-collision energy, and an arbitrary structured mesh.
|
||||
|
||||
The two valid elements in the tallies.xml file are ``tally`` and ``mesh``.
|
||||
The two valid elements in the tallies.xml file are ``<tally>`` and ``<mesh>``.
|
||||
|
||||
``tally`` Element
|
||||
-----------------
|
||||
``<tally>`` Element
|
||||
-------------------
|
||||
|
||||
The ``tally`` element accepts the following sub-elements:
|
||||
The ``<tally>`` element accepts the following sub-elements:
|
||||
|
||||
:filters:
|
||||
A list of filters to specify what region of phase space should contribute to
|
||||
|
|
@ -561,11 +580,11 @@ The following responses can be tallied.
|
|||
:nu-fission:
|
||||
Total production of neutrons due to fission
|
||||
|
||||
``mesh`` Element
|
||||
----------------
|
||||
``<mesh>`` Element
|
||||
------------------
|
||||
|
||||
If a structured mesh is desired as a filter for a tally, it must be specified in
|
||||
a separate element with the tag name ``mesh``. This element has the following
|
||||
a separate element with the tag name ``<mesh>``. This element has the following
|
||||
attributes/sub-elements:
|
||||
|
||||
:type:
|
||||
|
|
@ -582,8 +601,8 @@ attributes/sub-elements:
|
|||
:width:
|
||||
The width of mesh cells in each direction.
|
||||
|
||||
``assume_separate`` Element
|
||||
---------------------------
|
||||
``<assume_separate>`` Element
|
||||
-----------------------------
|
||||
|
||||
In cases where the user needs to specify many different tallies each of which
|
||||
are spatially separate, this tag can be used to cut down on some of the tally
|
||||
|
|
@ -596,16 +615,16 @@ tallies. This element should be followed by "yes" or "no"
|
|||
|
||||
*Default*: no
|
||||
|
||||
-------------------------------------------
|
||||
Geometry Plotting Specification -- plot.xml
|
||||
-------------------------------------------
|
||||
--------------------------------------------
|
||||
Geometry Plotting Specification -- plots.xml
|
||||
--------------------------------------------
|
||||
|
||||
A basic 2D plotting capability is available in OpenMC by creating a
|
||||
plots.xml file and subsequently running with the command-line flag ``-plot``. The
|
||||
root element of the plot.xml is simply ``<plots>`` and any number output
|
||||
figures can be defined with ``<plot>`` sub-elements.
|
||||
A basic 2D plotting capability is available in OpenMC by creating a plots.xml
|
||||
file and subsequently running with the command-line flag ``-plot``. The root
|
||||
element of the plots.xml is simply ``<plots>`` and any number output figures can
|
||||
be defined with ``<plot>`` sub-elements.
|
||||
|
||||
``plot`` Element
|
||||
``<plot>`` Element
|
||||
------------------
|
||||
|
||||
Each plot must contain a combination of the following attributes or sub-elements:
|
||||
|
|
@ -627,14 +646,14 @@ Each plot must contain a combination of the following attributes or sub-elements
|
|||
*Default*: ``cell``
|
||||
|
||||
:origin:
|
||||
Specifies the XYZ coordinate of the center of the plot. Should be 3 floats
|
||||
separated by spaces.
|
||||
Specifies the (x,y,z) coordinate of the center of the plot. Should be three
|
||||
floats separated by spaces.
|
||||
|
||||
*Default*: None - Required entry
|
||||
|
||||
:width:
|
||||
Specifies the width of the plot along each of the basis directions.
|
||||
Should be 2 or 3 floats separated by spaces for 2D plots and 3D plots,
|
||||
Specifies the width of the plot along each of the basis directions. Should
|
||||
be two or three floats separated by spaces for 2D plots and 3D plots,
|
||||
respectively.
|
||||
|
||||
*Default*: None - Required entry
|
||||
|
|
@ -652,18 +671,18 @@ Each plot must contain a combination of the following attributes or sub-elements
|
|||
|
||||
*Default*: "slice"
|
||||
|
||||
``plot`` elements of ``type`` ``slice`` also contain the following attributes or
|
||||
sub-elements:
|
||||
``<plot>`` elements of ``type`` "slice" also contain the following attributes or
|
||||
sub-elements:
|
||||
|
||||
:basis:
|
||||
Keyword specifying the plane of the plot for ``slice`` type plots. Can be
|
||||
one of: ``xy``, ``xz``, ``yz``.
|
||||
one of: "xy", "xz", "yz".
|
||||
|
||||
*Default*: ``xy``
|
||||
*Default*: "xy"
|
||||
|
||||
:pixels:
|
||||
Specifies the number of pixes to be used along each of the basis directions for
|
||||
``slice`` plots. Should be 2 integers separated by spaces.
|
||||
Specifies the number of pixes to be used along each of the basis directions
|
||||
for "slice" plots. Should be two integers separated by spaces.
|
||||
|
||||
.. warning:: The ``pixels`` input determines the output file size. For the PPM
|
||||
format, 10 million pixels will result in a file just under 30 MB in
|
||||
|
|
@ -675,16 +694,16 @@ sub-elements:
|
|||
.. warning:: Geometry features along a basis direction smaller than ``width``/``pixels``
|
||||
along that basis direction may not appear in the plot.
|
||||
|
||||
*Default*: None - Required entry for ``slice`` plots
|
||||
*Default*: None - Required entry for "slice" plots
|
||||
|
||||
:background:
|
||||
Specifies the RGB color of the regions where no OpenMC cell can be found. Should
|
||||
be 3 integers deparated by spaces.
|
||||
be three integers separated by spaces.
|
||||
|
||||
*Default*: 0 0 0 (white)
|
||||
|
||||
:col_spec:
|
||||
Any number of this optional tag may be included in each ``plot`` element, which can
|
||||
Any number of this optional tag may be included in each ``<plot>`` element, which can
|
||||
override the default random colors for cells or materials. Each ``col_spec``
|
||||
element must contain ``id`` and ``rgb`` sub-elements.
|
||||
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue