crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-27 21:55:41 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Merge pull request #680 from samuelshaner/new-mgxs-tallies

Browse source 
New MGXS for chi-prompt, prompt-nu-fission, and velocity
This commit is contained in:
Will Boyd 2016-07-29 16:27:45 -04:00 • committed by GitHub
commit 97fd71b3e8
26 changed files with 681 additions and 152 deletions
Download patch file Download diff file Expand all files Collapse all files

13
docs/source/pythonapi/examples/mgxs-part-i.ipynb
View file

@ -383,6 +383,9 @@
"* `ScatterMatrixXS`\n",
"* `NuScatterMatrixXS`\n",
"* `Chi`\n",
"* `ChiPrompt`\n",
"* `InverseVelocity`\n",
"* `PromptNuFissionXS`\n",
"\n",
"These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure."
]
@ -1164,21 +1167,21 @@
],
"metadata": {
"kernelspec": {
"display_name": "Python 3",
"display_name": "Python 2",
"language": "python",
"name": "python3"
"name": "python2"
},
"language_info": {
"codemirror_mode": {
"name": "ipython",
"version": 3
"version": 2
},
"file_extension": ".py",
"mimetype": "text/x-python",
"name": "python",
"nbconvert_exporter": "python",
"pygments_lexer": "ipython3",
"version": "3.5.2"
"pygments_lexer": "ipython2",
"version": "2.7.11"
}
},
"nbformat": 4,

13
docs/source/pythonapi/examples/mgxs-part-iii.ipynb
View file

@ -549,6 +549,9 @@
"* `ScatterMatrixXS` (`\"scatter matrix\"`)\n",
"* `NuScatterMatrixXS` (`\"nu-scatter matrix\"`)\n",
"* `Chi` (`\"chi\"`)\n",
"* `ChiPrompt` (`\"chi prompt\"`)\n",
"* `InverseVelocity` (`\"inverse-velocity\"`)\n",
"* `PromptNuFissionXS` (`\"prompt-nu-fission\"`)\n",
"\n",
"In this case, let's create the multi-group cross sections needed to run an OpenMOC simulation to verify the accuracy of our cross sections. In particular, we will define `\"transport\"`, `\"nu-fission\"`, `'\"fission\"`, `\"nu-scatter matrix\"` and `\"chi\"` cross sections for our `Library`.\n",
"\n",
@ -1596,21 +1599,21 @@
],
"metadata": {
"kernelspec": {
"display_name": "Python 3",
"display_name": "Python 2",
"language": "python",
"name": "python3"
"name": "python2"
},
"language_info": {
"codemirror_mode": {
"name": "ipython",
"version": 3
"version": 2
},
"file_extension": ".py",
"mimetype": "text/x-python",
"name": "python",
"nbconvert_exporter": "python",
"pygments_lexer": "ipython3",
"version": "3.5.2"
"pygments_lexer": "ipython2",
"version": "2.7.11"
}
},
"nbformat": 4,

3
docs/source/pythonapi/index.rst
View file

@ -269,13 +269,16 @@ Multi-group Cross Sections
openmc.mgxs.AbsorptionXS
openmc.mgxs.CaptureXS
openmc.mgxs.Chi
openmc.mgxs.ChiPrompt
openmc.mgxs.FissionXS
openmc.mgxs.InverseVelocity
openmc.mgxs.KappaFissionXS
openmc.mgxs.MultiplicityMatrixXS
openmc.mgxs.NuFissionXS
openmc.mgxs.NuFissionMatrixXS
openmc.mgxs.NuScatterXS
openmc.mgxs.NuScatterMatrixXS
openmc.mgxs.PromptNuFissionXS
openmc.mgxs.ScatterXS
openmc.mgxs.ScatterMatrixXS
openmc.mgxs.TotalXS

4
docs/source/usersguide/input.rst
View file

@ -1764,6 +1764,10 @@ The ``<tally>`` element accepts the following sub-elements:
| |fission. This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|prompt-nu-fission |Total production of prompt neutrons due to |
| |fission. This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|nu-fission |Total production of neutrons due to fission. |
+----------------------+---------------------------------------------------+
|nu-scatter, |These scores are similar in functionality to their |

2
openmc/checkvalue.py
View file

@ -212,7 +212,7 @@ def check_less_than(name, value, maximum, equality=False):
raise ValueError(msg)
def check_greater_than(name, value, minimum, equality=False):
"""Ensure that an object's value is less than a given value.
"""Ensure that an object's value is greater than a given value.
Parameters
----------

2
openmc/filter.py
View file

@ -563,7 +563,7 @@ class Filter(object):
# Initialize dictionary to build Pandas Multi-index column
filter_dict = {}
# Append Mesh ID as outermost index of mult-index
# Append Mesh ID as outermost index of multi-index
mesh_key = 'mesh {0}'.format(self.mesh.id)
# Find mesh dimensions - use 3D indices for simplicity

2
openmc/mgxs/library.py
View file

@ -452,7 +452,7 @@ class Library(object):
----------
domain : Material or Cell or Universe or Integral
The material, cell, or universe object of interest (or its ID)
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'}
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'}
The type of multi-group cross section object to return
Returns

461
openmc/mgxs/mgxs.py
View file

@ -34,7 +34,10 @@ MGXS_TYPES = ['total',
'nu-scatter matrix',
'multiplicity matrix',
'nu-fission matrix',
'chi']
'chi',
'chi-prompt',
'inverse-velocity',
'prompt-nu-fission']
# Supported domain types
@ -427,7 +430,7 @@ class MGXS(object):
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'}
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'}
The type of multi-group cross section object to return
domain : openmc.Material or openmc.Cell or openmc.Universe
The domain for spatial homogenization
@ -482,6 +485,12 @@ class MGXS(object):
mgxs = NuFissionMatrixXS(domain, domain_type, energy_groups)
elif mgxs_type == 'chi':
mgxs = Chi(domain, domain_type, energy_groups)
elif mgxs_type == 'chi-prompt':
mgxs = ChiPrompt(domain, domain_type, energy_groups)
elif mgxs_type == 'inverse-velocity':
mgxs = InverseVelocity(domain, domain_type, energy_groups)
elif mgxs_type == 'prompt-nu-fission':
mgxs = PromptNuFissionXS(domain, domain_type, energy_groups)
mgxs.by_nuclide = by_nuclide
mgxs.name = name
@ -704,7 +713,7 @@ class MGXS(object):
def get_xs(self, groups='all', subdomains='all', nuclides='all',
xs_type='macro', order_groups='increasing',
value='mean', **kwargs):
"""Returns an array of multi-group cross sections.
r"""Returns an array of multi-group cross sections.
This method constructs a 2D NumPy array for the requested multi-group
cross section data data for one or more energy groups and subdomains.
@ -1148,6 +1157,9 @@ class MGXS(object):
string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
# Generate the header for an individual XS
xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
# If cross section data has not been computed, only print string header
if self.tallies is None:
print(string)
@ -1167,11 +1179,7 @@ class MGXS(object):
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
# Build header for cross section type
if xs_type == 'macro':
string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
else:
string += '{0: <16}\n'.format('\tCross Sections [barns]:')
string += '{0: <16}\n'.format(xs_header)
template = '{0: <12}Group {1} [{2: <10} - {3: <10}MeV]:\t'
# Loop over energy groups ranges
@ -1530,6 +1538,26 @@ class MGXS(object):
df.sort_values(by=[self.domain_type] + columns, inplace=True)
return df
def get_units(self, xs_type='macro'):
"""This method returns the units of a MGXS based on a desired xs_type.
Parameters
----------
xs_type: {'macro', 'micro'}
Return the macro or micro cross section units.
Defaults to 'macro'.
Returns
-------
str
A string representing the units of the MGXS.
"""
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
return 'cm^-1' if xs_type == 'macro' else 'barns'
class MatrixMGXS(MGXS):
"""An abstract multi-group cross section for some energy group structure
@ -1878,6 +1906,9 @@ class MatrixMGXS(MGXS):
string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
# Generate the header for an individual XS
xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
# If cross section data has not been computed, only print string header
if self.tallies is None:
print(string)
@ -1906,11 +1937,7 @@ class MatrixMGXS(MGXS):
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
# Build header for cross section type
if xs_type == 'macro':
string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
else:
string += '{0: <16}\n'.format('\tCross Sections [barns]:')
string += '{0: <16}\n'.format(xs_header)
template = '{0: <12}Group {1} -> Group {2}:\t\t'
# Loop over incoming/outgoing energy groups ranges
@ -3723,6 +3750,9 @@ class ScatterMatrixXS(MatrixMGXS):
string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
# Generate the header for an individual XS
xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
# If cross section data has not been computed, only print string header
if self.tallies is None:
print(string)
@ -3751,11 +3781,7 @@ class ScatterMatrixXS(MatrixMGXS):
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
# Build header for cross section type
if xs_type == 'macro':
string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
else:
string += '{0: <16}\n'.format('\tCross Sections [barns]:')
string += '{0: <16}\n'.format(xs_header)
template = '{0: <12}Group {1} -> Group {2}:\t\t'
# Loop over incoming/outgoing energy groups ranges
@ -4183,14 +4209,14 @@ class Chi(MGXS):
.. math::
\langle \nu\sigma_{f,\rightarrow g} \phi \rangle &= \int_{r \in V} dr
\langle \nu\sigma_{f,g' \rightarrow g} \phi \rangle &= \int_{r \in V} dr
\int_{4\pi} d\Omega' \int_0^\infty dE' \int_{E_g}^{E_{g-1}} dE \; \chi(E)
\nu\sigma_f (r, E') \psi(r, E', \Omega')\\
\langle \nu\sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu\sigma_f (r,
E') \psi(r, E', \Omega') \\
\chi_g &= \frac{\langle \nu\sigma_{f,\rightarrow g} \phi \rangle}{\langle
\nu\sigma_f \phi \rangle}
\chi_g &= \frac{\langle \nu\sigma_{f,g' \rightarrow g} \phi \rangle}
{\langle \nu\sigma_f \phi \rangle}
Parameters
----------
@ -4395,7 +4421,7 @@ class Chi(MGXS):
"""
if not self.can_merge(other):
raise ValueError('Unable to merge Chi')
raise ValueError('Unable to merge a Chi MGXS')
# Create deep copy of tally to return as merged tally
merged_mgxs = copy.deepcopy(self)
@ -4414,7 +4440,7 @@ class Chi(MGXS):
# The nuclides must be mutually exclusive
for nuclide in self.nuclides:
if nuclide in other.nuclides:
msg = 'Unable to merge Chi with shared nuclides'
msg = 'Unable to merge a Chi MGXS with shared nuclides'
raise ValueError(msg)
# Concatenate lists of nuclides for the merged MGXS
@ -4618,3 +4644,392 @@ class Chi(MGXS):
df['std. dev.'] *= np.tile(densities, tile_factor)
return df
def get_units(self, xs_type='macro'):
"""Returns the units of Chi.
This method returns the units of Chi, which is "%" for both macro
and micro xs types.
Parameters
----------
xs_type: {'macro', 'micro'}
Return the macro or micro cross section units.
Defaults to 'macro'.
Returns
-------
str
A string representing the units of Chi.
"""
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
# Chi has the same units (%) for both macro and micro
return '%'
class ChiPrompt(Chi):
r"""The prompt fission spectrum.
This class can be used for both OpenMC input generation and tally data
post-processing to compute spatially-homogenized and energy-integrated
multi-group cross sections for multi-group neutronics calculations. At a
minimum, one needs to set the :attr:`ChiPrompt.energy_groups` and
:attr:`ChiPrompt.domain` properties. Tallies for the flux and appropriate
reaction rates over the specified domain are generated automatically via the
:attr:`ChiPrompt.tallies` property, which can then be appended to a
:class:`openmc.Tallies` instance.
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
necessary data to compute multi-group cross sections from a
:class:`openmc.StatePoint` instance. The derived multi-group cross section
can then be obtained from the :attr:`ChiPrompt.xs_tally` property.
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
fission spectrum is calculated as:
.. math::
\langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle &= \int_{r \in V}
dr \int_{4\pi} d\Omega' \int_0^\infty dE' \int_{E_g}^{E_{g-1}} dE \;
\chi(E) \nu^p \sigma_f (r, E') \psi(r, E', \Omega')\\
\langle \nu^p \sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu^p \sigma_f (r,
E') \psi(r, E', \Omega') \\
\chi_g^p &= \frac{\langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle}
{\langle \nu^p \sigma_f \phi \rangle}
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
Attributes
----------
name : str, optional
Name of the multi-group cross section
rxn_type : str
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
tally_trigger : openmc.Trigger
An (optional) tally precision trigger given to each tally used to
compute the cross section
scores : list of str
The scores in each tally used to compute the multi-group cross section
filters : list of openmc.Filter
The filters in each tally used to compute the multi-group cross section
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'analog'}
The tally estimator used to compute the multi-group cross section
tallies : collections.OrderedDict
OpenMC tallies needed to compute the multi-group cross section. The keys
are strings listed in the :attr:`ChiPrompt.tally_keys` property and
values are instances of :class:`openmc.Tally`.
rxn_rate_tally : openmc.Tally
Derived tally for the reaction rate tally used in the numerator to
compute the multi-group cross section. This attribute is None
unless the multi-group cross section has been computed.
xs_tally : openmc.Tally
Derived tally for the multi-group cross section. This attribute
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage
loaded_sp : bool
Whether or not a statepoint file has been loaded with tally data
derived : bool
Whether or not the MGXS is merged from one or more other MGXS
hdf5_key : str
The key used to index multi-group cross sections in an HDF5 data store
"""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(ChiPrompt, self).__init__(domain, domain_type, groups, by_nuclide, name)
self._rxn_type = 'chi-prompt'
@property
def scores(self):
return ['prompt-nu-fission', 'prompt-nu-fission']
class InverseVelocity(MGXS):
r"""An inverse velocity multi-group cross section.
This class can be used for both OpenMC input generation and tally data
post-processing to compute spatially-homogenized and energy-integrated
multi-group neutron inverse velocities for multi-group neutronics
calculations. The units of inverse velocity are seconds per centimeter. At a
minimum, one needs to set the :attr:`InverseVelocity.energy_groups` and
:attr:`InverseVelocity.domain` properties. Tallies for the flux and
appropriate reaction rates over the specified domain are generated
automatically via the :attr:`InverseVelocity.tallies` property, which can
then be appended to a :class:`openmc.Tallies` instance.
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
necessary data to compute multi-group cross sections from a
:class:`openmc.StatePoint` instance. The derived multi-group cross section
can then be obtained from the :attr:`InverseVelocity.xs_tally` property.
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
neutron inverse velocities are calculated by tallying the flux-weighted
inverse velocity and the flux. The inverse velocity is then the
flux-weighted inverse velocity divided by the flux:
.. math::
\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
\frac{\psi (r, E, \Omega)}{v (r, E)}}{\int_{r \in V} dr \int_{4\pi}
d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
Attributes
----------
name : str, optional
Name of the multi-group cross section
rxn_type : str
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
tally_trigger : openmc.Trigger
An (optional) tally precision trigger given to each tally used to
compute the cross section
scores : list of str
The scores in each tally used to compute the multi-group cross section
filters : list of openmc.Filter
The filters in each tally used to compute the multi-group cross section
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'analog'}
The tally estimator used to compute the multi-group cross section
tallies : collections.OrderedDict
OpenMC tallies needed to compute the multi-group cross section. The keys
are strings listed in the :attr:`InverseVelocity.tally_keys` property
and values are instances of :class:`openmc.Tally`.
rxn_rate_tally : openmc.Tally
Derived tally for the reaction rate tally used in the numerator to
compute the multi-group cross section. This attribute is None
unless the multi-group cross section has been computed.
xs_tally : openmc.Tally
Derived tally for the multi-group cross section. This attribute
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage
loaded_sp : bool
Whether or not a statepoint file has been loaded with tally data
derived : bool
Whether or not the MGXS is merged from one or more other MGXS
hdf5_key : str
The key used to index multi-group cross sections in an HDF5 data store
"""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(InverseVelocity, self).__init__(domain, domain_type,
groups, by_nuclide, name)
self._rxn_type = 'inverse-velocity'
def get_units(self, xs_type='macro'):
"""Returns the units of InverseVelocity.
This method returns the units of an InverseVelocity based on a desired
xs_type.
Parameters
----------
xs_type: {'macro', 'micro'}
Return the macro or micro cross section units.
Defaults to 'macro'.
Returns
-------
str
A string representing the units of the InverseVelocity.
"""
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
if xs_type == 'macro':
return 'second/cm'
else:
raise ValueError('Unable to return the units of InverseVelocity for'
' xs_type other than "macro"')
class PromptNuFissionXS(MGXS):
r"""A prompt fission neutron production multi-group cross section.
This class can be used for both OpenMC input generation and tally data
post-processing to compute spatially-homogenized and energy-integrated
multi-group cross sections for multi-group neutronics calculations. At a
minimum, one needs to set the :attr:`PromptNuFissionXS.energy_groups` and
:attr:`PromptNuFissionXS.domain` properties. Tallies for the flux and
appropriate reaction rates over the specified domain are generated
automatically via the :attr:`PromptNuFissionXS.tallies` property, which can
then be appended to a :class:`openmc.Tallies` instance.
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
necessary data to compute multi-group cross sections from a
:class:`openmc.StatePoint` instance. The derived multi-group cross section
can then be obtained from the :attr:`PromptNuFissionXS.xs_tally` property.
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
fission spectrum is calculated as:
.. math::
\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
\nu\sigma_f^p (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
Attributes
----------
name : str, optional
Name of the multi-group cross section
rxn_type : str
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
tally_trigger : openmc.Trigger
An (optional) tally precision trigger given to each tally used to
compute the cross section
scores : list of str
The scores in each tally used to compute the multi-group cross section
filters : list of openmc.Filter
The filters in each tally used to compute the multi-group cross section
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'analog'}
The tally estimator used to compute the multi-group cross section
tallies : collections.OrderedDict
OpenMC tallies needed to compute the multi-group cross section. The keys
are strings listed in the :attr:`PromptNuFissionXS.tally_keys` property
and values are instances of :class:`openmc.Tally`.
rxn_rate_tally : openmc.Tally
Derived tally for the reaction rate tally used in the numerator to
compute the multi-group cross section. This attribute is None
unless the multi-group cross section has been computed.
xs_tally : openmc.Tally
Derived tally for the multi-group cross section. This attribute
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage
loaded_sp : bool
Whether or not a statepoint file has been loaded with tally data
derived : bool
Whether or not the MGXS is merged from one or more other MGXS
hdf5_key : str
The key used to index multi-group cross sections in an HDF5 data store
"""
def __init__(self, domain=None, domain_type=None,
groups=None, by_nuclide=False, name=''):
super(PromptNuFissionXS, self).__init__(domain, domain_type, groups,
by_nuclide, name)
self._rxn_type = 'prompt-nu-fission'

5
src/constants.F90
View file

@ -289,7 +289,7 @@ module constants
EVENT_ABSORB = 2
! Tally score type
integer, parameter :: N_SCORE_TYPES = 20
integer, parameter :: N_SCORE_TYPES = 21
integer, parameter :: &
SCORE_FLUX = -1, & ! flux
SCORE_TOTAL = -2, & ! total reaction rate
@ -310,7 +310,8 @@ module constants
SCORE_NU_SCATTER_YN = -17, & ! angular flux-weighted nu-scattering moment (0:N)
SCORE_EVENTS = -18, & ! number of events
SCORE_DELAYED_NU_FISSION = -19, & ! delayed neutron production rate
SCORE_INVERSE_VELOCITY = -20 ! flux-weighted inverse velocity
SCORE_PROMPT_NU_FISSION = -20, & ! prompt neutron production rate
SCORE_INVERSE_VELOCITY = -21 ! flux-weighted inverse velocity
! Maximum scattering order supported
integer, parameter :: MAX_ANG_ORDER = 10

2
src/endf.F90
View file

@ -42,6 +42,8 @@ contains
string = "nu-fission"
case (SCORE_DELAYED_NU_FISSION)
string = "delayed-nu-fission"
case (SCORE_PROMPT_NU_FISSION)
string = "prompt-nu-fission"
case (SCORE_KAPPA_FISSION)
string = "kappa-fission"
case (SCORE_CURRENT)

6
src/input_xml.F90
View file

@ -3619,6 +3619,12 @@ contains
! Set tally estimator to analog
t % estimator = ESTIMATOR_ANALOG
end if
case ('prompt-nu-fission')
t % score_bins(j) = SCORE_PROMPT_NU_FISSION
if (t % find_filter(FILTER_ENERGYOUT) > 0) then
! Set tally estimator to analog
t % estimator = ESTIMATOR_ANALOG
end if
! Disallow for MG mode since data not present
if (.not. run_CE) then

1
src/output.F90
View file

@ -790,6 +790,7 @@ contains
score_names(abs(SCORE_NU_SCATTER_PN)) = "Scattering Prod. Rate Moment"
score_names(abs(SCORE_NU_SCATTER_YN)) = "Scattering Prod. Rate Moment"
score_names(abs(SCORE_DELAYED_NU_FISSION)) = "Delayed-Nu-Fission Rate"
score_names(abs(SCORE_PROMPT_NU_FISSION)) = "Prompt-Nu-Fission Rate"
score_names(abs(SCORE_INVERSE_VELOCITY)) = "Flux-Weighted Inverse Velocity"
! Create filename for tally output

222
src/tally.F90
View file

@ -407,7 +407,7 @@ contains
! neutrons were emitted with different energies, multiple
! outgoing energy bins may have been scored to. The following
! logic treats this special case and results to multiple bins
call score_fission_eout_ce(p, t, score_index)
call score_fission_eout_ce(p, t, score_index, score_bin)
cycle SCORE_LOOP
end if
end if
@ -441,6 +441,67 @@ contains
end if
case (SCORE_PROMPT_NU_FISSION)
if (t % estimator == ESTIMATOR_ANALOG) then
if (survival_biasing .or. p % fission) then
if (t % find_filter(FILTER_ENERGYOUT) > 0) then
! Normally, we only need to make contributions to one scoring
! bin. However, in the case of fission, since multiple fission
! neutrons were emitted with different energies, multiple
! outgoing energy bins may have been scored to. The following
! logic treats this special case and results to multiple bins
call score_fission_eout_ce(p, t, score_index, score_bin)
cycle SCORE_LOOP
end if
end if
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! prompt-nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO) then
score = p % absorb_wgt * micro_xs(p % event_nuclide) % fission &
* nuclides(p % event_nuclide) % nu(E, EMISSION_PROMPT) &
/ micro_xs(p % event_nuclide) % absorption
else
score = ZERO
end if
else
! Skip any non-fission events
if (.not. p % fission) cycle SCORE_LOOP
! If there is no outgoing energy filter, than we only need to
! score to one bin. For the score to be 'analog', we need to
! score the number of particles that were banked in the fission
! bank as prompt neutrons. Since this was weighted by 1/keff, we
! multiply by keff to get the proper score.
score = keff * p % wgt_bank * (ONE - sum(p % n_delayed_bank) &
/ real(p % n_bank, 8))
end if
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % fission * nuclides(i_nuclide) % &
nu(E, EMISSION_PROMPT) * atom_density * flux
else
score = ZERO
! Loop over all nuclides in the current material
do l = 1, materials(p % material) % n_nuclides
! Get atom density
atom_density_ = materials(p % material) % atom_density(l)
! Get index in nuclides array
i_nuc = materials(p % material) % nuclide(l)
! Accumulate the contribution from each nuclide
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) % &
nu(E, EMISSION_PROMPT) * atom_density_ * flux
end do
end if
end if
case (SCORE_DELAYED_NU_FISSION)
! make sure the correct energy is used
@ -461,7 +522,7 @@ contains
! neutrons were emitted with different energies, multiple
! outgoing energy bins may have been scored to. The following
! logic treats this special case and results to multiple bins
call score_fission_delayed_eout(p, t, score_index)
call score_fission_eout_ce(p, t, score_index, score_bin)
cycle SCORE_LOOP
end if
end if
@ -1515,12 +1576,18 @@ contains
! neutrons produced with different energies.
!===============================================================================
subroutine score_fission_eout_ce(p, t, i_score)
type(Particle), intent(in) :: p
subroutine score_fission_eout_ce(p, t, i_score, score_bin)
type(Particle), intent(in) :: p
type(TallyObject), intent(inout) :: t
integer, intent(in) :: i_score ! index for score
integer, intent(in) :: i_score ! index for score
integer, intent(in) :: score_bin
integer :: i ! index of outgoing energy filter
integer :: j ! index of delayedgroup filter
integer :: d ! delayed group
integer :: g ! another delayed group
integer :: d_bin ! delayed group bin index
integer :: n ! number of energies on filter
integer :: k ! loop index for bank sites
integer :: bin_energyout ! original outgoing energy bin
@ -1542,6 +1609,10 @@ contains
! loop over number of particles banked
do k = 1, p % n_bank
! get the delayed group
g = fission_bank(n_bank - p % n_bank + k) % delayed_group
! determine score based on bank site weight and keff
score = keff * fission_bank(n_bank - p % n_bank + k) % wgt
@ -1555,13 +1626,49 @@ contains
! change outgoing energy bin
matching_bins(i) = binary_search(t % filters(i) % real_bins, n, E_out)
! determine scoring index
i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
! Case for tallying prompt neutrons
if (score_bin == SCORE_NU_FISSION .or. &
(score_bin == SCORE_PROMPT_NU_FISSION .and. g == 0)) then
! Add score to tally
! determine scoring index
i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
! Add score to tally
!$omp atomic
t % results(i_score, i_filter) % value = &
t % results(i_score, i_filter) % value + score
t % results(i_score, i_filter) % value = &
t % results(i_score, i_filter) % value + score
! Case for tallying delayed emissions
else if (score_bin == SCORE_DELAYED_NU_FISSION .and. g /= 0) then
! Get the index of delayed group filter
j = t % find_filter(FILTER_DELAYEDGROUP)
! if the delayed group filter is present, tally to corresponding
! delayed group bin if it exists
if (j > 0) then
! loop over delayed group bins until the corresponding bin is found
do d_bin = 1, t % filters(j) % n_bins
d = t % filters(j) % int_bins(d_bin)
! check whether the delayed group of the particle is equal to the
! delayed group of this bin
if (d == g) call score_fission_delayed_dg(t, d_bin, score, i_score)
end do
! if the delayed group filter is not present, add score to tally
else
! determine scoring index
i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
! Add score to tally
!$omp atomic
t % results(i_score, i_filter) % value = &
t % results(i_score, i_filter) % value + score
end if
end if
end do
! reset outgoing energy bin and score index
@ -1647,101 +1754,6 @@ contains
end subroutine score_fission_eout_mg
!===============================================================================
! SCORE_FISSION_DELAYED_EOUT handles a special case where we need to store
! delayed neutron production rate with an outgoing energy filter (think of a
! fission matrix). In this case, we may need to score to multiple bins if there
! were multiple neutrons produced with different energies.
!===============================================================================
subroutine score_fission_delayed_eout(p, t, i_score)
type(Particle), intent(in) :: p
type(TallyObject), intent(inout) :: t
integer, intent(in) :: i_score ! index for score
integer :: i ! index of outgoing energy filter
integer :: j ! index of delayedgroup filter
integer :: d ! delayed group
integer :: g ! another delayed group
integer :: d_bin ! delayed group bin index
integer :: n ! number of energies on filter
integer :: k ! loop index for bank sites
integer :: bin_energyout ! original outgoing energy bin
integer :: i_filter ! index for matching filter bin combination
real(8) :: score ! actual score
real(8) :: E_out ! energy of fission bank site
! Save original outgoing energy bin
i = t % find_filter(FILTER_ENERGYOUT)
bin_energyout = matching_bins(i)
! Get the index of delayed group filter
j = t % find_filter(FILTER_DELAYEDGROUP)
! Get number of energies on filter
n = size(t % filters(i) % real_bins)
! Since the creation of fission sites is weighted such that it is
! expected to create n_particles sites, we need to multiply the
! score by keff to get the true delayed-nu-fission rate.
! loop over number of particles banked
do k = 1, p % n_bank
! get the delayed group
g = fission_bank(n_bank - p % n_bank + k) % delayed_group
! check if the particle was born delayed
if (g /= 0) then
! determine score based on bank site weight and keff
score = keff * fission_bank(n_bank - p % n_bank + k) % wgt
! determine outgoing energy from fission bank
E_out = fission_bank(n_bank - p % n_bank + k) % E
! check if outgoing energy is within specified range on filter
if (E_out < t % filters(i) % real_bins(1) .or. &
E_out > t % filters(i) % real_bins(n)) cycle
! change outgoing energy bin
matching_bins(i) = binary_search(t % filters(i) % real_bins, n, E_out)
! if the delayed group filter is present, tally to corresponding
! delayed group bin if it exists
if (j > 0) then
! loop over delayed group bins until the corresponding bin is found
do d_bin = 1, t % filters(j) % n_bins
d = t % filters(j) % int_bins(d_bin)
! check whether the delayed group of the particle is equal to the
! delayed group of this bin
if (d == g) then
call score_fission_delayed_dg(t, d_bin, score, i_score)
end if
end do
! if the delayed group filter is not present, add score to tally
else
! determine scoring index
i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
! Add score to tally
!$omp atomic
t % results(i_score, i_filter) % value = &
t % results(i_score, i_filter) % value + score
end if
end if
end do
! reset outgoing energy bin
matching_bins(i) = bin_energyout
end subroutine score_fission_delayed_eout
!===============================================================================
! SCORE_FISSION_DELAYED_DG helper function used to increment the tally when a
! delayed group filter is present.

2
tests/test_mgxs_library_condense/inputs_true.dat
View file

@ -1 +1 @@
855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e

18
tests/test_mgxs_library_condense/results_true.dat
View file

@ -34,6 +34,12 @@
0 10000 1 1 total 0.085835 0.005592
material group out nuclide mean std. dev.
0 10000 1 total 1.0 0.046071
material group out nuclide mean std. dev.
0 10000 1 total 1.0 0.051471
material group in nuclide mean std. dev.
0 10000 1 total 4.996730e-07 3.650635e-08
material group in nuclide mean std. dev.
0 10000 1 total 0.090004 0.006367
material group in nuclide mean std. dev.
0 10001 1 total 0.311594 0.013793
material group in nuclide mean std. dev.
@ -70,6 +76,12 @@
0 10001 1 1 total 0.0 0.0
material group out nuclide mean std. dev.
0 10001 1 total 0.0 0.0
material group out nuclide mean std. dev.
0 10001 1 total 0.0 0.0
material group in nuclide mean std. dev.
0 10001 1 total 5.454760e-07 4.949800e-08
material group in nuclide mean std. dev.
0 10001 1 total 0.0 0.0
material group in nuclide mean std. dev.
0 10002 1 total 0.904999 0.043964
material group in nuclide mean std. dev.
@ -106,3 +118,9 @@
0 10002 1 1 total 0.0 0.0
material group out nuclide mean std. dev.
0 10002 1 total 0.0 0.0
material group out nuclide mean std. dev.
0 10002 1 total 0.0 0.0
material group in nuclide mean std. dev.
0 10002 1 total 5.773006e-07 5.322132e-08
material group in nuclide mean std. dev.
0 10002 1 total 0.0 0.0

2
tests/test_mgxs_library_distribcell/inputs_true.dat
View file

@ -1 +1 @@
3a3b7f75b326c94a8e5c7efe3046b2fdb887e9f75ecf6eb27587f9450c77cf8fd6acc4198c15bffb4e7ceead6d7b4327c19536bbf9cc35dfaae3f4ce4c26cc1a
2d948f3b12293294eaeca231a3df9d51195379e8bb38dd3e68d3bc512a7d08ed52a1109054ca381684ec127268710f6d6e9210ac8154c9b379608e996627624a

6
tests/test_mgxs_library_distribcell/results_true.dat
View file

@ -34,3 +34,9 @@
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 1 total 0.0 0.0
avg(distribcell) group out nuclide mean std. dev.
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 total 0.0 0.0
avg(distribcell) group out nuclide mean std. dev.
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 total 0.0 0.0
avg(distribcell) group in nuclide mean std. dev.
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 total 0.000001 6.946255e-07
avg(distribcell) group in nuclide mean std. dev.
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 total 0.0 0.0

2
tests/test_mgxs_library_hdf5/inputs_true.dat
View file

@ -1 +1 @@
855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e

27
tests/test_mgxs_library_hdf5/results_true.dat
View file

@ -63,6 +63,15 @@ domain=10000 type=nu-fission matrix
domain=10000 type=chi
[ 1. 0.]
[ 0.04607052 0. ]
domain=10000 type=chi-prompt
[ 1. 0.]
[ 0.05147146 0. ]
domain=10000 type=inverse-velocity
[ 5.70932329e-08 2.85573950e-06]
[ 4.68792969e-09 2.44216503e-07]
domain=10000 type=prompt-nu-fission
[ 0.01923922 0.46671903]
[ 0.00130951 0.04141087]
domain=10001 type=total
[ 0.31373767 0.3008214 ]
[ 0.0155819 0.02805245]
@ -128,6 +137,15 @@ domain=10001 type=nu-fission matrix
domain=10001 type=chi
[ 0. 0.]
[ 0. 0.]
domain=10001 type=chi-prompt
[ 0. 0.]
[ 0. 0.]
domain=10001 type=inverse-velocity
[ 5.99598048e-08 2.98549021e-06]
[ 4.55309296e-09 3.41701554e-07]
domain=10001 type=prompt-nu-fission
[ 0. 0.]
[ 0. 0.]
domain=10002 type=total
[ 0.66457226 2.05238401]
[ 0.03121475 0.22434291]
@ -193,3 +211,12 @@ domain=10002 type=nu-fission matrix
domain=10002 type=chi
[ 0. 0.]
[ 0. 0.]
domain=10002 type=chi-prompt
[ 0. 0.]
[ 0. 0.]
domain=10002 type=inverse-velocity
[ 6.02207831e-08 3.04495537e-06]
[ 3.78043696e-09 3.60007673e-07]
domain=10002 type=prompt-nu-fission
[ 0. 0.]
[ 0. 0.]

2
tests/test_mgxs_library_no_nuclides/inputs_true.dat
View file

@ -1 +1 @@
855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e

27
tests/test_mgxs_library_no_nuclides/results_true.dat
View file

@ -75,6 +75,15 @@
material group out nuclide mean std. dev.
1 10000 1 total 1.0 0.046071
0 10000 2 total 0.0 0.000000
material group out nuclide mean std. dev.
1 10000 1 total 1.0 0.051471
0 10000 2 total 0.0 0.000000
material group in nuclide mean std. dev.
1 10000 1 total 5.709323e-08 4.687930e-09
0 10000 2 total 2.855740e-06 2.442165e-07
material group in nuclide mean std. dev.
1 10000 1 total 0.019239 0.001310
0 10000 2 total 0.466719 0.041411
material group in nuclide mean std. dev.
1 10001 1 total 0.313738 0.015582
0 10001 2 total 0.300821 0.028052
@ -152,6 +161,15 @@
material group out nuclide mean std. dev.
1 10001 1 total 0.0 0.0
0 10001 2 total 0.0 0.0
material group out nuclide mean std. dev.
1 10001 1 total 0.0 0.0
0 10001 2 total 0.0 0.0
material group in nuclide mean std. dev.
1 10001 1 total 5.995980e-08 4.553093e-09
0 10001 2 total 2.985490e-06 3.417016e-07
material group in nuclide mean std. dev.
1 10001 1 total 0.0 0.0
0 10001 2 total 0.0 0.0
material group in nuclide mean std. dev.
1 10002 1 total 0.664572 0.031215
0 10002 2 total 2.052384 0.224343
@ -229,3 +247,12 @@
material group out nuclide mean std. dev.
1 10002 1 total 0.0 0.0
0 10002 2 total 0.0 0.0
material group out nuclide mean std. dev.
1 10002 1 total 0.0 0.0
0 10002 2 total 0.0 0.0
material group in nuclide mean std. dev.
1 10002 1 total 6.022078e-08 3.780437e-09
0 10002 2 total 3.044955e-06 3.600077e-07
material group in nuclide mean std. dev.
1 10002 1 total 0.0 0.0
0 10002 2 total 0.0 0.0

2
tests/test_mgxs_library_nuclides/inputs_true.dat
View file

@ -1 +1 @@
739796983940a1bad601998cf9ea2f90453a994477c7f675c2fd404d1864fe04a7fbfb5a15c5fe7cf9bad016b78432ba0910baba6f9cc026143761e9f526b62a
e4a5f03ab6167e96462c4ef537533fe33b98d7878ae00824c5619356bda8d548b3c71af01ba8c88d5a9b46dd1471d331e6f678a164af922200f2ee3642be6340

2
tests/test_mgxs_library_nuclides/results_true.dat
View file

@ -1 +1 @@
d56c6bae6bf3cd8950d3f50f089458c1c6c807be780fc97570532c6af6eb7e3057968d3345bd3c363f01315271129ef7b2ca028b05767353e601dc37b035f8a7
e494320a213b5704a2ac915a2ba504857be91961ceb6735b6ad05d81eb31c44c9584d5bd9d40baececf1dcb5b030e6ecec63cfbd20639baf69bcb596c5c46591

2
tests/test_tallies/inputs_true.dat
View file

@ -1 +1 @@
a392a7a8f27fd2f959b06a6809df29b3482e215da175b19846c248af44dc2ee7e2b05269b802dd9d6ed434506b05092f349436a0048411adab6b296dba0bd683
ca47172a42f6c13b244a763c990cbe4811662708ee03307d810a4542ee34bb5db7cc29d66aea313dad95b9f38a4ff7943ded527cfd0c7c8825372fec40cfc0d0

2
tests/test_tallies/results_true.dat
View file

@ -1 +1 @@
a0c7d6ca246ecd7dd5fed06373af142390971401c4e97744f29e55810ab9c231c97c4d8947cdf0b3d2df0ae829a9ddf768e5b2d889bbea34f2b6db0e567db884
7d085e38f331083a8c3f7814eb6025f9457a1a8e3899e7d9a65af1ba92ea75c0182fcc605d4eadd3eb7edfc0949df688d39a3f45683e10cbdcfb6d7954f34129

3
tests/test_tallies/test_tallies.py
View file

@ -122,7 +122,8 @@ class TalliesTestHarness(PyAPITestHarness):
t.filters = [cell_filter]
t.scores = ['absorption', 'delayed-nu-fission', 'events', 'fission',
'inverse-velocity', 'kappa-fission', '(n,2n)', '(n,n1)',
'(n,gamma)', 'nu-fission', 'scatter', 'elastic', 'total']
'(n,gamma)', 'nu-fission', 'scatter', 'elastic', 'total',
'prompt-nu-fission']
score_tallies[0].estimator = 'tracklength'
score_tallies[1].estimator = 'analog'
score_tallies[2].estimator = 'collision'