crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-27 05:35:49 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions

Merge pull request #825 from paulromano/make-docs-look-better

Browse source 
Improve layout for notebooks in documentation
This commit is contained in:
Will Boyd 2017-03-08 18:56:04 -05:00 • committed by GitHub
commit e9b9db477c
4 changed files with 151 additions and 146 deletions
Download patch file Download diff file Expand all files Collapse all files

7
docs/source/_static/theme_overrides.css
View file

@ -20,3 +20,10 @@
.wy-side-nav-search {
background-color: #343131;
}
/* Make embedded Jupyter notebooks look better */
div#notebook-container.container {
padding: 0px;
width: auto;
box-shadow: none;
}

2
docs/source/usersguide/index.rst
View file

@ -9,7 +9,7 @@ essential aspects of using OpenMC to perform simulations.
.. toctree::
:numbered:
:maxdepth: 3
:maxdepth: 2
beginners
install

280
docs/source/usersguide/input.rst
View file

@ -1494,145 +1494,13 @@ The ``<tally>`` element accepts the following sub-elements:
:type:
The type of the filter. Accepted options are "cell", "cellborn",
"material", "universe", "energy", "energyout", "mesh", "distribcell",
"delayedgroup", and "energyfunction".
"material", "universe", "energy", "energyout", "mu", "polar",
"azimuthal", "mesh", "distribcell", "delayedgroup", and
"energyfunction".
:bins:
For each filter type, the corresponding ``bins`` entry is given as
follows:
:cell:
A list of cells in which the tally should be accumulated.
:cellborn:
This filter allows the tally to be scored to only when particles were
originally born in a specified cell.
:surface:
A list of surfaces for which the tally should be accumulated.
:material:
A list of materials for which the tally should be accumulated.
:universe:
A list of universes for which the tally should be accumulated.
:energy:
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **pre-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
<filter type="energy" bins="0.0 1.0e6 20.0e6" />
then two energy bins will be created, one with energies between 0 and
1 MeV and the other with energies between 1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:energyout:
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **post-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
<filter type="energyout" bins="0.0 1.0e6 20.0e6" />
then two post-collision energy bins will be created, one with
energies between 0 and 1 MeV and the other with energies between
1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:mu:
A monotonically increasing list of bounding **post-collision** cosines
of the change in a particle's angle (i.e., :math:`\mu = \hat{\Omega}
\cdot \hat{\Omega}'`), which represents a portion of the possible
values of :math:`[-1,1]`. For example, spanning all of :math:`[-1,1]`
with five equi-width bins can be specified as:
.. code-block:: xml
<filter type="mu" bins="-1.0 -0.6 -0.2 0.2 0.6 1.0" />
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[-1,1]` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[-1,1]` can be instead written as:
.. code-block:: xml
<filter type="mu" bins="5" />
:polar:
A monotonically increasing list of bounding particle polar angles
which represents a portion of the possible values of :math:`[0,\pi]`.
For example, spanning all of :math:`[0,\pi]` with five equi-width
bins can be specified as:
.. code-block:: xml
<filter type="polar" bins="0.0 0.6283 1.2566 1.8850 2.5132 3.1416"/>
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[0,\pi]` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[0,\pi]` can be instead written as:
.. code-block:: xml
<filter type="polar" bins="5" />
:azimuthal:
A monotonically increasing list of bounding particle azimuthal angles
which represents a portion of the possible values of :math:`[-\pi,\pi)`.
For example, spanning all of :math:`[-\pi,\pi)` with two equi-width
bins can be specified as:
.. code-block:: xml
<filter type="azimuthal" bins="0.0 3.1416 6.2832" />
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[-\pi,\pi)` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[-\pi,\pi)` can be instead written as:
.. code-block:: xml
<filter type="azimuthal" bins="2" />
:mesh:
The ``id`` of a structured mesh to be tallied over.
:distribcell:
The single cell which should be tallied uniquely for all instances.
.. note::
The distribcell filter will take a single cell ID and will tally
each unique occurrence of that cell separately. This filter will
not accept more than one cell ID. It is not recommended to combine
this filter with a cell or mesh filter.
:delayedgroup:
A list of delayed neutron precursor groups for which the tally should
be accumulated. For instance, to tally to all 6 delayed groups in the
ENDF/B-VII.1 library the filter is specified as:
.. code-block:: xml
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
:energyfunction:
``energyfunction`` filters do not use the ``bins`` entry. Instead
they use ``energy`` and ``y``.
A description of the bins for each type of filter can be found in
:ref:`filter_types`.
:energy:
``energyfunction`` filters multiply tally scores by an arbitrary
@ -1941,6 +1809,144 @@ The ``<tally>`` element accepts the following sub-elements:
*Default*: None
.. _filter_types:
Filter Types
++++++++++++
For each filter type, the following table describes what the ``bins`` attribute
should be set to:
:cell:
A list of unique IDs for cells in which the tally should be accumulated.
:cellborn:
This filter allows the tally to be scored to only when particles were
originally born in a specified cell. A list of cell IDs should be given.
:material:
A list of unique IDs for matreials in which the tally should be accumulated.
:universe:
A list of unique IDs for universes in which the tally should be accumulated.
:energy:
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **pre-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
<filter type="energy" bins="0.0 1.0e6 20.0e6" />
then two energy bins will be created, one with energies between 0 and
1 MeV and the other with energies between 1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:energyout:
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **post-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
<filter type="energyout" bins="0.0 1.0e6 20.0e6" />
then two post-collision energy bins will be created, one with
energies between 0 and 1 MeV and the other with energies between
1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:mu:
A monotonically increasing list of bounding **post-collision** cosines
of the change in a particle's angle (i.e., :math:`\mu = \hat{\Omega}
\cdot \hat{\Omega}'`), which represents a portion of the possible
values of :math:`[-1,1]`. For example, spanning all of :math:`[-1,1]`
with five equi-width bins can be specified as:
.. code-block:: xml
<filter type="mu" bins="-1.0 -0.6 -0.2 0.2 0.6 1.0" />
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[-1,1]` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[-1,1]` can be instead written as:
.. code-block:: xml
<filter type="mu" bins="5" />
:polar:
A monotonically increasing list of bounding particle polar angles
which represents a portion of the possible values of :math:`[0,\pi]`.
For example, spanning all of :math:`[0,\pi]` with five equi-width
bins can be specified as:
.. code-block:: xml
<filter type="polar" bins="0.0 0.6283 1.2566 1.8850 2.5132 3.1416"/>
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[0,\pi]` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[0,\pi]` can be instead written as:
.. code-block:: xml
<filter type="polar" bins="5" />
:azimuthal:
A monotonically increasing list of bounding particle azimuthal angles
which represents a portion of the possible values of :math:`[-\pi,\pi)`.
For example, spanning all of :math:`[-\pi,\pi)` with two equi-width
bins can be specified as:
.. code-block:: xml
<filter type="azimuthal" bins="0.0 3.1416 6.2832" />
Alternatively, if only one value is provided as a bin, OpenMC will
interpret this to mean the complete range of :math:`[-\pi,\pi)` should
be automatically subdivided in to the provided value for the bin.
That is, the above example of five equi-width bins spanning
:math:`[-\pi,\pi)` can be instead written as:
.. code-block:: xml
<filter type="azimuthal" bins="2" />
:mesh:
The unique ID of a structured mesh to be tallied over.
:distribcell:
The single cell which should be tallied uniquely for all instances.
.. note:: The distribcell filter will take a single cell ID and will tally
each unique occurrence of that cell separately. This filter will not
accept more than one cell ID. It is not recommended to combine this
filter with a cell or mesh filter.
:delayedgroup:
A list of delayed neutron precursor groups for which the tally should
be accumulated. For instance, to tally to all 6 delayed groups in the
ENDF/B-VII.1 library the filter is specified as:
.. code-block:: xml
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
:energyfunction:
``energyfunction`` filters do not use the ``bins`` entry. Instead
they use ``energy`` and ``y``.
``<mesh>`` Element
------------------

8
openmc/__init__.py
View file

@ -28,11 +28,3 @@ from openmc.summary import *
from openmc.particle_restart import *
from openmc.mixin import *
from openmc.plotter import *
try:
# Ignore matplotlib warning caused by OpenMOC calling matplotlib.use()
with warnings.catch_warnings():
warnings.simplefilter("ignore")
from openmc.openmoc_compatible import *
except ImportError:
pass