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Added legendre moment parameter to Pandas DF construction for ScatterMatrixXS

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This commit is contained in:
Will Boyd 2016-05-09 12:17:55 -04:00
parent f5f26bada0
commit 2dae4f9a69
1 changed files with 62 additions and 5 deletions
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67
openmc/mgxs/mgxs.py
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@ -1833,14 +1833,15 @@ class NuScatterXS(MGXS):
class ScatterMatrixXS(MGXS):
"""A scattering matrix multi-group cross section.
"""A scattering matrix multi-group cross section for one or more Legendre
moments.
Attributes
----------
correction : 'P0' or None
Apply the P0 correction to scattering matrices if set to 'P0'
order : int
The highest order in the scattering matrix (default is 0)
The highest legendre moment in the scattering matrix (default is 0)
"""
@ -1869,8 +1870,8 @@ class ScatterMatrixXS(MGXS):
def tallies(self):
"""Construct the OpenMC tallies needed to compute this cross section.
This method constructs three analog tallies to compute the 'flux',
'scatter' and 'scatter-P1' reaction rates in the spatial domain and
This method constructs three analog tallies to compute the 'flux'
and Legendre scattering moment reaction rates in the spatial domain and
energy groups of interest.
"""
@ -1955,7 +1956,6 @@ class ScatterMatrixXS(MGXS):
self._order = order
# FIXME: Add order param
def get_slice(self, nuclides=[], in_groups=[], out_groups=[]):
"""Build a sliced ScatterMatrix for the specified nuclides and
energy groups.
@ -2151,6 +2151,63 @@ class ScatterMatrixXS(MGXS):
return xs
def get_pandas_dataframe(self, groups='all', nuclides='all', moment='all',
xs_type='macro', summary=None):
"""Build a Pandas DataFrame for the MGXS data.
This method leverages :meth:`openmc.Tally.get_pandas_dataframe`, but
renames the columns with terminology appropriate for cross section data.
Parameters
----------
groups : Iterable of Integral or 'all'
Energy groups of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
The nuclides of the cross-sections to include in the dataframe. This
may be a list of nuclide name strings (e.g., ['U-235', 'U-238']).
The special string 'all' will include the cross sections for all
nuclides in the spatial domain. The special string 'sum' will
include the cross sections summed over all nuclides. Defaults
to 'all'.
moment : int or 'all'
The scattering matrix moment to return. All moments will be
returned if the moment is 'all' (default); otherwise, a specific
moment will be returned.
xs_type: {'macro', 'micro'}
Return macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
summary : None or openmc.Summary
An optional Summary object to be used to construct columns for
distribcell tally filters (default is None). The geometric
information in the Summary object is embedded into a multi-index
column with a geometric "path" to each distribcell intance.
NOTE: This option requires the OpenCG Python package.
Returns
-------
pandas.DataFrame
A Pandas DataFrame for the cross section data.
Raises
------
ValueError
When this method is called before the multi-group cross section is
computed from tally data.
"""
df = super(ScatterMatrixXS, self).get_pandas_dataframe(
groups, nuclides, xs_type, summary)
# Select rows corresponding to requested scattering moment
if moment != 'all':
cv.check_type('moment', moment, Integral)
cv.check_greater_than('moment', moment, 0, equality=True)
cv.check_less_than('moment', moment, self.order, equality=True)
df = df[df['score'] == str(self.xs_tally.scores[moment])]
return df
def print_xs(self, subdomains='all', nuclides='all',
xs_type='macro', moment=0):
"""Prints a string representation for the multi-group cross section.