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Merge remote-tracking branch 'upstream/develop' into minimalist-imports

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wbinventor@gmail.com 2016-04-09 13:44:00 -04:00
commit a596ba3ec3
8 changed files with 136 additions and 42 deletions
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2
.travis.yml
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@ -27,7 +27,7 @@ before_install:
- conda config --set always_yes yes --set changeps1 no
- conda update -q conda
- conda info -a
- conda create -q -n test-environment python=$TRAVIS_PYTHON_VERSION numpy scipy h5py pandas
- conda create -q -n test-environment python=$TRAVIS_PYTHON_VERSION numpy scipy h5py=2.5 pandas
- source activate test-environment
# Install GCC, MPICH, HDF5, PHDF5

8
docs/source/pythonapi/energy_groups.rst
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@ -1,8 +0,0 @@
.. _pythonapi_energy_groups:
=============
Energy Groups
=============
.. automodule:: openmc.mgxs.groups
:members:

3
docs/source/pythonapi/index.rst
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@ -19,6 +19,7 @@ on a given module or class.
:maxdepth: 1
ace
mgxs_library
**Creating input files:**
@ -65,8 +66,6 @@ on a given module or class.
:maxdepth: 1
mgxs
energy_groups
mgxs_library
**Example Jupyter Notebooks:**

57
docs/source/pythonapi/mgxs.rst
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@ -4,31 +4,57 @@
Multi-Group Cross Sections
==========================
.. currentmodule:: openmc.mgxs.mgxs
----------------------------
Summary of Available Classes
----------------------------
Energy Groups
-------------
.. currentmodule:: openmc.mgxs.groups
.. autosummary::
MGXS
AbsorptionXS
CaptureXS
Chi
FissionXS
NuFissionXS
NuScatterXS
NuScatterMatrixXS
ScatterXS
ScatterMatrixXS
TotalXS
TransportXS
EnergyGroups
Multi-group Cross Sections
--------------------------
.. currentmodule:: openmc.mgxs.mgxs
.. autosummary::
MGXS
AbsorptionXS
CaptureXS
Chi
FissionXS
NuFissionXS
NuScatterXS
NuScatterMatrixXS
ScatterXS
ScatterMatrixXS
TotalXS
TransportXS
Multi-group Cross Section Libraries
-----------------------------------
.. currentmodule:: openmc.mgxs.library
.. autosummary::
Library
-------------------
Class Documentation
-------------------
.. automodule:: openmc.mgxs.groups
:members:
.. currentmodule:: openmc.mgxs.mgxs
.. autoclass:: MGXS
:members:
@ -64,3 +90,6 @@ Class Documentation
.. autoclass:: TransportXS
:members:
.. automodule:: openmc.mgxs.library
:members:

8
docs/source/pythonapi/mgxs_library.rst
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@ -1,8 +1,8 @@
.. _pythonapi_mgxs_library:
============
MGXS Library
============
==============================
Multi-group Cross Section Data
==============================
.. automodule:: openmc.mgxs.library
.. automodule:: openmc.mgxs_library
:members:

3
openmc/material.py
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@ -723,9 +723,8 @@ class MaterialsFile(object):
material.make_isotropic_in_lab()
def _create_material_subelements(self):
subelement = ET.SubElement(self._materials_file, "default_xs")
if self._default_xs is not None:
subelement = ET.SubElement(self._materials_file, "default_xs")
subelement.text = self._default_xs
for material in self._materials:

95
openmc/mgxs_library.py
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@ -87,8 +87,9 @@ class XSdata(object):
----------
name : str, optional
Name of the mgxs data set.
representation : {'isotropic', 'angle'}
energy_groups : openmc.mgxs.EnergyGroups
Energygroup structure
representation : {'isotropic', 'angle'}, optional
Method used in generating the MGXS (isotropic or angle-dependent flux
weighting). Defaults to 'isotropic'
@ -99,10 +100,10 @@ class XSdata(object):
alias : str
Separate unique identifier for the xsdata object
kT : float
Temperature (in units of MeV) of this data set.
energy_groups : EnergyGroups
Temperature (in units of MeV).
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure
fissionable : boolean
fissionable : bool
Whether or not this is a fissionable data set.
scatt_type : {'legendre', 'histogram', or 'tabular'}
Angular distribution representation (legendre, histogram, or tabular)
@ -115,6 +116,85 @@ class XSdata(object):
Legendre polynomial form). Dict contains two keys: 'enable' and
'num_points'. 'enable' is a boolean and 'num_points' is the
number of points to use, if 'enable' is True.
num_azimuthal : int
Number of equal width angular bins that the azimuthal angular domain is
subdivided into. This only applies when ``representation`` is "angle".
num_polar : int
Number of equal width angular bins that the polar angular domain is
subdivided into. This only applies when ``representation`` is "angle".
total : numpy.ndarray
Group-wise total cross section ordered by increasing group index (i.e.,
fast to thermal). If ``representation`` is "isotropic", then the length
of this list should equal the number of groups described in the
``groups`` element. If ``representation`` is "angle", then the length
of this list should equal the number of groups times the number of
azimuthal angles times the number of polar angles, with the
inner-dimension being groups, intermediate-dimension being azimuthal
angles and outer-dimension being the polar angles.
absorption : numpy.ndarray
Group-wise absorption cross section ordered by increasing group index
(i.e., fast to thermal). If ``representation`` is "isotropic", then the
length of this list should equal the number of groups described in the
``groups`` attribute. If ``representation`` is "angle", then the length
of this list should equal the number of groups times the number of
azimuthal angles times the number of polar angles, with the
inner-dimension being groups, intermediate-dimension being azimuthal
angles and outer-dimension being the polar angles.
scatter : numpy.ndarray
Scattering moment matrices presented with the columns representing
incoming group and rows representing the outgoing group. That is,
down-scatter will be above the diagonal of the resultant matrix. This
matrix is repeated for every Legendre order (in order of increasing
orders) if ``scatt_type`` is "legendre"; otherwise, this matrix is
repeated for every bin of the histogram or tabular representation.
Finally, if ``representation`` is "angle", the above is repeated for
every azimuthal angle and every polar angle, in that order.
multiplicity : numpy.ndarray
Ratio of neutrons produced in scattering collisions to the neutrons
which undergo scattering collisions; that is, the multiplicity provides
the code with a scaling factor to account for neutrons being produced in
(n,xn) reactions. This information is assumed isotropic and therefore
does not need to be repeated for every Legendre moment or
histogram/tabular bin. This matrix follows the same arrangement as
described for the ``scatter`` attribute, with the exception of the data
needed to provide the scattering type information.
fission : numpy.ndarray
Group-wise fission cross section ordered by increasing group index
(i.e., fast to thermal). If ``representation`` is "isotropic", then the
length of this list should equal the number of groups described in the
``groups`` attribute. If ``representation`` is "angle", then the length
of this list should equal the number of groups times the number of
azimuthal angles times the number of polar angles, with the
inner-dimension being groups, intermediate-dimension being azimuthal
angles and outer-dimension being the polar angles.
k_fission : numpy.ndarray
Group-wise kappa-fission cross section ordered by increasing group index
(i.e., fast to thermal). If ``representation`` is "isotropic", then the
length of this list should equal the number of groups described in the
``groups`` attribute. If ``representation`` is "angle", then the length
of this list should equal the number of groups times the number of
azimuthal angles times the number of polar angles, with the
inner-dimension being groups, intermediate-dimension being azimuthal
angles and outer-dimension being the polar angles.
chi : numpy.ndarray
Group-wise fission spectra ordered by increasing group index (i.e., fast
to thermal). This attribute should be used if making the common
approximation that the fission spectra does not depend on incoming
energy. If the user does not wish to make this approximation, then this
should not be provided and this information included in the
``nu_fission`` element instead. If ``representation`` is "isotropic",
then the length of this list should equal the number of groups described
in the ``groups`` element. If ``representation`` is "angle", then the
length of this list should equal the number of groups times the number
of azimuthal angles times the number of polar angles, with the
inner-dimension being groups, intermediate-dimension being azimuthal
angles and outer-dimension being the polar angles.
nu_fission : numpy.ndarray
Group-wise fission production cross section vector (i.e., if ``chi`` is
provided), or is the group-wise fission production matrix. If providing
the vector, it should be ordered the same as the ``fission`` data. If
providing the matrix, it should be ordered the same as the
``multiplicity`` matrix.
"""
def __init__(self, name, energy_groups, representation="isotropic"):
@ -577,8 +657,6 @@ class MGXSLibraryFile(object):
Energy group structure.
inverse_velocities : Iterable of Real
Inverse of velocities, units of sec/cm
filename : str
XML file to write to.
xsdatas : Iterable of XSdata
Iterable of multi-Group cross section data objects
"""
@ -717,6 +795,3 @@ class MGXSLibraryFile(object):
tree = ET.ElementTree(self._cross_sections_file)
tree.write(filename, xml_declaration=True,
encoding='utf-8', method="xml")

2
tests/test_source/inputs_true.dat
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@ -1 +1 @@
5c2fdde85affcd44c1b02c07c300acb8e5c189c1adbf7aa079e37a68e8b8313678fc292bd7f6e0d0957f723e05b8146bd165cf3315dde5f6b2f88ebc954cd65e
526c91551d9a80dc01216e5cb04162253f12ec684cc2b4912ca18cfc510f1ea2e5303029f1c1607882082b0c2c8a47f25dd5be14678f449a1579e3601d1bdec5