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Merge pull request #592 from friedmud/kappa_fission

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Add KappaFission capability to MGXS
This commit is contained in:
Will Boyd 2016-02-23 14:16:54 -05:00
commit 94387397f7
1 changed files with 63 additions and 76 deletions
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139
openmc/mgxs/mgxs.py
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@ -25,6 +25,7 @@ MGXS_TYPES = ['total',
'capture',
'fission',
'nu-fission',
'kappa-fission',
'scatter',
'nu-scatter',
'scatter matrix',
@ -315,7 +316,7 @@ class MGXS(object):
Parameters
----------
mgxs_type : {'total', 'transport', 'absorption', 'capture', 'fission', 'nu-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'chi'}
mgxs_type : {'total', 'transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'chi'}
The type of multi-group cross section object to return
domain : Material or Cell or Universe
The domain for spatial homogenization
@ -352,6 +353,8 @@ class MGXS(object):
mgxs = FissionXS(domain, domain_type, energy_groups)
elif mgxs_type == 'nu-fission':
mgxs = NuFissionXS(domain, domain_type, energy_groups)
elif mgxs_type == 'kappa-fission':
mgxs = KappaFissionXS(domain, domain_type, energy_groups)
elif mgxs_type == 'scatter':
mgxs = ScatterXS(domain, domain_type, energy_groups)
elif mgxs_type == 'nu-scatter':
@ -1484,96 +1487,80 @@ class CaptureXS(MGXS):
self._rxn_rate_tally.sparse = self.sparse
return self._rxn_rate_tally
class FissionXSBase(MGXS):
"""A fission production multi-group cross section base class
for NuFission and KappaFission
"""
class FissionXS(MGXS):
# This is an abstract class which cannot be instantiated
__metaclass__ = abc.ABCMeta
def __init__(self, rxn_type, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(FissionXSBase, self).__init__(domain, domain_type,
groups, by_nuclide, name)
self._rxn_type = rxn_type
@property
def tallies(self):
"""Construct the OpenMC tallies needed to compute this cross section.
This method constructs two tracklength tallies to compute the 'flux'
and 'rxn_type' reaction rates in the spatial domain and energy
groups of interest.
"""
# Instantiate tallies if they do not exist
if self._tallies is None:
# Create a list of scores for each Tally to be created
scores = ['flux', self._rxn_type]
estimator = 'tracklength'
keys = scores
# Create the non-domain specific Filters for the Tallies
group_edges = self.energy_groups.group_edges
energy_filter = openmc.Filter('energy', group_edges)
filters = [[energy_filter], [energy_filter]]
# Initialize the Tallies
self._create_tallies(scores, filters, keys, estimator)
return self._tallies
@property
def rxn_rate_tally(self):
if self._rxn_rate_tally is None:
self._rxn_rate_tally = self.tallies[self._rxn_type]
self._rxn_rate_tally.sparse = self.sparse
return self._rxn_rate_tally
class FissionXS(FissionXSBase):
"""A fission multi-group cross section."""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(FissionXS, self).__init__(domain, domain_type,
super(FissionXS, self).__init__('fission', domain, domain_type,
groups, by_nuclide, name)
self._rxn_type = 'fission'
@property
def tallies(self):
"""Construct the OpenMC tallies needed to compute this cross section.
This method constructs two tracklength tallies to compute the 'flux'
and 'fission' reaction rates in the spatial domain and energy
groups of interest.
"""
# Instantiate tallies if they do not exist
if self._tallies is None:
# Create a list of scores for each Tally to be created
scores = ['flux', 'fission']
estimator = 'tracklength'
keys = scores
# Create the non-domain specific Filters for the Tallies
group_edges = self.energy_groups.group_edges
energy_filter = openmc.Filter('energy', group_edges)
filters = [[energy_filter], [energy_filter]]
# Initialize the Tallies
self._create_tallies(scores, filters, keys, estimator)
return self._tallies
@property
def rxn_rate_tally(self):
if self._rxn_rate_tally is None:
self._rxn_rate_tally = self.tallies['fission']
self._rxn_rate_tally.sparse = self.sparse
return self._rxn_rate_tally
class NuFissionXS(MGXS):
class NuFissionXS(FissionXSBase):
"""A fission production multi-group cross section."""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(NuFissionXS, self).__init__(domain, domain_type,
super(NuFissionXS, self).__init__('nu-fission', domain, domain_type,
groups, by_nuclide, name)
self._rxn_type = 'nu-fission'
@property
def tallies(self):
"""Construct the OpenMC tallies needed to compute this cross section.
This method constructs two tracklength tallies to compute the 'flux'
and 'nu-fission' reaction rates in the spatial domain and energy
groups of interest.
"""
# Instantiate tallies if they do not exist
if self._tallies is None:
# Create a list of scores for each Tally to be created
scores = ['flux', 'nu-fission']
estimator = 'tracklength'
keys = scores
# Create the non-domain specific Filters for the Tallies
group_edges = self.energy_groups.group_edges
energy_filter = openmc.Filter('energy', group_edges)
filters = [[energy_filter], [energy_filter]]
# Initialize the Tallies
self._create_tallies(scores, filters, keys, estimator)
return self._tallies
@property
def rxn_rate_tally(self):
if self._rxn_rate_tally is None:
self._rxn_rate_tally = self.tallies['nu-fission']
self._rxn_rate_tally.sparse = self.sparse
return self._rxn_rate_tally
class KappaFissionXS(FissionXSBase):
"""A recoverable fission energy production rate multi-group cross section."""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(KappaFissionXS, self).__init__('kappa-fission', domain, domain_type,
groups, by_nuclide, name)
class ScatterXS(MGXS):
"""A scatter multi-group cross section."""