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Allowed nu=True in DiffusionCoefficient

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Added nu-diffusion-coefficient to mgxs_types.

Relying on TransportXS to call the nu-dependent tallies
and assuming those are being used in DiffusionCoefficient
when called with nu=True.
This commit is contained in:
Miriam 2020-10-07 18:40:28 +00:00
parent 7dd35754a3
commit 3182ebf66a
1 changed files with 18 additions and 10 deletions
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28
openmc/mgxs/mgxs.py
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@ -38,7 +38,8 @@ MGXS_TYPES = (
'prompt-nu-fission',
'prompt-nu-fission matrix',
'current',
'diffusion-coefficient'
'diffusion-coefficient',
'nu-diffusion-coefficient'
)
# Supported domain types
@ -697,7 +698,7 @@ class MGXS:
Parameters
----------
mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'current', 'diffusion-coefficient'}
mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'current', 'diffusion-coefficient', 'nu-diffusion-coefficient'}
The type of multi-group cross section object to return
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
The domain for spatial homogenization
@ -779,6 +780,8 @@ class MGXS:
mgxs = Current(domain, domain_type, energy_groups)
elif mgxs_type == 'diffusion-coefficient':
mgxs = DiffusionCoefficient(domain, domain_type, energy_groups)
elif mgxs_type == 'nu-diffusion-coefficient':
mgxs = DiffusionCoefficient(domain, domain_type, energy_groups, nu=True)
mgxs.by_nuclide = by_nuclide
mgxs.name = name
@ -2647,7 +2650,7 @@ class TransportXS(MGXS):
The energy group structure for energy condensation
nu : bool
If True, the cross section data will include neutron multiplication;
defaults to True.
defaults to False.
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
@ -2870,6 +2873,9 @@ class DiffusionCoefficient(TransportXS):
D = \frac{1}{3 \sigma_{tr}}
\end{aligned}
To incorporate the effect of scattering multiplication in the above
relation, the `nu` parameter can be set to `True`.
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
@ -2880,7 +2886,7 @@ class DiffusionCoefficient(TransportXS):
The energy group structure for energy condensation
nu : bool
If True, the cross section data will include neutron multiplication;
must be False in DiffusionCoefficient.
defaults to False.
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
@ -2966,8 +2972,13 @@ class DiffusionCoefficient(TransportXS):
super(DiffusionCoefficient, self).__init__(domain, domain_type, groups,
nu, by_nuclide, name,
num_polar, num_azimuthal)
self._rxn_type = 'diffusion-coefficient'
cv.check_value('nu for diffusion-coefficient', nu, [False])
#self._rxn_type = 'diffusion-coefficient'
#cv.check_value('nu for diffusion-coefficient', nu, [False])
if not nu:
self._rxn_type = 'diffusion-coefficient'
else:
self._rxn_type = 'nu-diffision-coefficient'
@property
def rxn_rate_tally(self):
@ -3012,15 +3023,12 @@ class DiffusionCoefficient(TransportXS):
return self._xs_tally
def get_condensed_xs(self, coarse_groups, condense_dif_coef=True):
def get_condensed_xs(self, coarse_groups):
"""Construct an energy-condensed version of this cross section.
Parameters
----------
coarse_groups : openmc.mgxs.EnergyGroups
The coarse energy group structure of interest
condense_diff_coef: bool
A boolean representing whether a condensed DiffusionCoefficient
mgxs will be calculated.
Returns
-------