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Merge pull request #771 from samuelshaner/decay-rate-tracklength

Browse source 
Track length estimator for decay rate tally and other fixes
This commit is contained in:
Will Boyd 2016-12-20 15:42:09 -05:00 • committed by GitHub
commit 580f081795
27 changed files with 644 additions and 338 deletions
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11
examples/python/basic/build-xml.py
View file

@ -15,21 +15,16 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
moderator = openmc.Material(material_id=41, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])

14
examples/python/boxes/build-xml.py
View file

@ -15,25 +15,19 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
u238 = openmc.Nuclide('U238')
# Instantiate some Materials and register the appropriate Nuclides
fuel1 = openmc.Material(material_id=1, name='fuel')
fuel1.set_density('g/cc', 4.5)
fuel1.add_nuclide(u235, 1.)
fuel1.add_nuclide('U235', 1.)
fuel2 = openmc.Material(material_id=2, name='depleted fuel')
fuel2.set_density('g/cc', 4.5)
fuel2.add_nuclide(u238, 1.)
fuel2.add_nuclide('U238', 1.)
moderator = openmc.Material(material_id=3, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

14
examples/python/lattice/hexagonal/build-xml.py
View file

@ -14,26 +14,20 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
fe56 = openmc.Nuclide('Fe56')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
iron.add_nuclide(fe56, 1.)
iron.add_element('Fe', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel, iron])

11
examples/python/lattice/nested/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

11
examples/python/lattice/simple/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

79
examples/python/pincell/build-xml.py
View file

@ -14,86 +14,29 @@ particles = 1000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
h2 = openmc.Nuclide('H2')
he4 = openmc.Nuclide('He4')
b10 = openmc.Nuclide('B10')
b11 = openmc.Nuclide('B11')
o16 = openmc.Nuclide('O16')
o17 = openmc.Nuclide('O17')
cr50 = openmc.Nuclide('Cr50')
cr52 = openmc.Nuclide('Cr52')
cr53 = openmc.Nuclide('Cr53')
cr54 = openmc.Nuclide('Cr54')
fe54 = openmc.Nuclide('Fe54')
fe56 = openmc.Nuclide('Fe56')
fe57 = openmc.Nuclide('Fe57')
fe58 = openmc.Nuclide('Fe58')
zr90 = openmc.Nuclide('Zr90')
zr91 = openmc.Nuclide('Zr91')
zr92 = openmc.Nuclide('Zr92')
zr94 = openmc.Nuclide('Zr94')
zr96 = openmc.Nuclide('Zr96')
sn112 = openmc.Nuclide('Sn112')
sn114 = openmc.Nuclide('Sn114')
sn115 = openmc.Nuclide('Sn115')
sn116 = openmc.Nuclide('Sn116')
sn117 = openmc.Nuclide('Sn117')
sn118 = openmc.Nuclide('Sn118')
sn119 = openmc.Nuclide('Sn119')
sn120 = openmc.Nuclide('Sn120')
sn122 = openmc.Nuclide('Sn122')
sn124 = openmc.Nuclide('Sn124')
u = openmc.Element('U')
o = openmc.Element('O')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_element(u, 1., enrichment=2.4)
uo2.add_element(o, 2.)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_nuclide(he4, 2.4044e-4)
helium.add_element('He', 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_nuclide(o16, 3.0743e-4)
zircaloy.add_nuclide(o17, 7.4887e-7)
zircaloy.add_nuclide(cr50, 3.2962e-6)
zircaloy.add_nuclide(cr52, 6.3564e-5)
zircaloy.add_nuclide(cr53, 7.2076e-6)
zircaloy.add_nuclide(cr54, 1.7941e-6)
zircaloy.add_nuclide(fe54, 8.6699e-6)
zircaloy.add_nuclide(fe56, 1.3610e-4)
zircaloy.add_nuclide(fe57, 3.1431e-6)
zircaloy.add_nuclide(fe58, 4.1829e-7)
zircaloy.add_nuclide(zr90, 2.1827e-2)
zircaloy.add_nuclide(zr91, 4.7600e-3)
zircaloy.add_nuclide(zr92, 7.2758e-3)
zircaloy.add_nuclide(zr94, 7.3734e-3)
zircaloy.add_nuclide(zr96, 1.1879e-3)
zircaloy.add_nuclide(sn112, 4.6735e-6)
zircaloy.add_nuclide(sn114, 3.1799e-6)
zircaloy.add_nuclide(sn115, 1.6381e-6)
zircaloy.add_nuclide(sn116, 7.0055e-5)
zircaloy.add_nuclide(sn117, 3.7003e-5)
zircaloy.add_nuclide(sn118, 1.1669e-4)
zircaloy.add_nuclide(sn119, 4.1387e-5)
zircaloy.add_nuclide(sn120, 1.5697e-4)
zircaloy.add_nuclide(sn122, 2.2308e-5)
zircaloy.add_nuclide(sn124, 2.7897e-5)
zircaloy.add_element('Sn', 0.014 , 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001 , 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
borated_water = openmc.Material(material_id=4, name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_nuclide(b10, 8.0042e-6)
borated_water.add_nuclide(b11, 3.2218e-5)
borated_water.add_nuclide(h1, 4.9457e-2)
borated_water.add_nuclide(h2, 7.4196e-6)
borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
borated_water.add_element('O', 2.4e-2)
borated_water.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

5
examples/python/reflective/build-xml.py
View file

@ -15,13 +15,10 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate a Nuclides
u235 = openmc.Nuclide('U235')
# Instantiate a Material and register the Nuclide
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel])

36
openmc/material.py
View file

@ -129,7 +129,7 @@ class Material(object):
string = 'Material\n'
string += '{0: <16}{1}{2}\n'.format('\tID', '=\t', self._id)
string += '{0: <16}{1}{2}\n'.format('\tName', '=\t', self._name)
string += '{0: <16}{1}{2}\n'.format('\Temperature', '=\t',
string += '{0: <16}{1}{2}\n'.format('\tTemperature', '=\t',
self._temperature)
string += '{0: <16}{1}{2}'.format('\tDensity', '=\t', self._density)
@ -251,7 +251,7 @@ class Material(object):
Parameters
----------
units : {'g/cm3', 'g/cc', 'km/cm3', 'atom/b-cm', 'atom/cm3', 'sum', 'macro'}
units : {'g/cm3', 'g/cc', 'kg/cm3', 'atom/b-cm', 'atom/cm3', 'sum', 'macro'}
Physical units of density.
density : float, optional
Value of the density. Must be specified unless units is given as
@ -641,37 +641,39 @@ class Material(object):
nucs = []
nuc_densities = []
nuc_density_types = []
for nuclide in nuclides.items():
nuc, nuc_density, nuc_density_type = nuclide[1]
nucs.append(nuc)
nuc_densities.append(nuc_density)
nuc_density_types.append(nuc_density_type)
nucs = np.array(nucs)
nuc_densities = np.array(nuc_densities)
nuc_density_types = np.array(nuc_density_types)
if sum_density:
density = np.sum(nuc_densities)
percent_in_atom = np.all(nuc_density_types == 'ao')
density_in_atom = density > 0.
sum_percent = 0.
awrs = []
for n, nuclide in enumerate(nuclides.items()):
awr = openmc.data.atomic_mass(nuclide[0])
if awr is not None:
awrs.append(awr / openmc.data.NEUTRON_MASS)
else:
raise ValueError(nuclide[0] + " is invalid")
# Convert the weight amounts to atomic amounts
if not percent_in_atom:
for n, nuc in enumerate(nucs):
nuc_densities[n] *= self.average_molar_mass / \
openmc.data.atomic_mass(nuc)
# Now that we have the awr, lets finish calculating densities
# Now that we have the atomic amounts, lets finish calculating densities
sum_percent = np.sum(nuc_densities)
nuc_densities = nuc_densities / sum_percent
# Convert the mass density to an atom density
if not density_in_atom:
sum_percent = 0.
for n, nuc in enumerate(nucs):
x = nuc_densities[n]
sum_percent += x * awrs[n]
sum_percent = 1. / sum_percent
density = -density * sum_percent * \
openmc.data.AVOGADRO / openmc.data.NEUTRON_MASS * 1.E-24
density = -density / self.average_molar_mass * 1.E-24 \
* openmc.data.AVOGADRO
nuc_densities = density * nuc_densities
nuclides = OrderedDict()

8
openmc/mgxs/library.py
View file

@ -524,6 +524,14 @@ class Library(object):
for domain in self.domains:
for mgxs_type in self.mgxs_types:
mgxs = self.get_mgxs(domain, mgxs_type)
if mgxs_type in openmc.mgxs.MDGXS_TYPES:
if self.num_delayed_groups == 0:
mgxs.delayed_groups = None
else:
mgxs.delayed_groups \
= list(range(1, self.num_delayed_groups+1))
for tally in mgxs.tallies.values():
tallies_file.append(tally, merge=merge)

92
openmc/mgxs/mdgxs.py
View file

@ -923,7 +923,7 @@ class ChiDelayed(MDGXS):
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'}
estimator : '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
@ -966,6 +966,7 @@ class ChiDelayed(MDGXS):
super(ChiDelayed, self).__init__(domain, domain_type, energy_groups,
delayed_groups, by_nuclide, name)
self._rxn_type = 'chi-delayed'
self._estimator = 'analog'
@property
def scores(self):
@ -987,10 +988,6 @@ class ChiDelayed(MDGXS):
def tally_keys(self):
return ['delayed-nu-fission-in', 'delayed-nu-fission-out']
@property
def estimator(self):
return 'analog'
@property
def rxn_rate_tally(self):
if self._rxn_rate_tally is None:
@ -1018,6 +1015,34 @@ class ChiDelayed(MDGXS):
return self._xs_tally
def get_homogenized_mgxs(self, other_mgxs):
"""Construct a homogenized MGXS with other MGXS objects.
This method constructs a new MGXS object that is the flux-weighted
combination of two MGXS objects. It is equivalent to what one would
obtain if the tally spatial domain were designed to encompass the
individual domains for both MGXS objects.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
return self._get_homogenized_mgxs(other_mgxs, 'delayed-nu-fission-in')
def get_slice(self, nuclides=[], groups=[], delayed_groups=[]):
"""Build a sliced ChiDelayed for the specified nuclides and energy
groups.
@ -1586,6 +1611,33 @@ class Beta(MDGXS):
return self._xs_tally
def get_homogenized_mgxs(self, other_mgxs):
"""Construct a homogenized MGXS with other MGXS objects.
This method constructs a new MGXS object that is the flux-weighted
combination of two MGXS objects. It is equivalent to what one would
obtain if the tally spatial domain were designed to encompass the
individual domains for both MGXS objects.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
return self._get_homogenized_mgxs(other_mgxs, 'nu-fission')
class DecayRate(MDGXS):
r"""The decay rate for delayed neutron precursors.
@ -1703,7 +1755,6 @@ class DecayRate(MDGXS):
super(DecayRate, self).__init__(domain, domain_type, energy_groups,
delayed_groups, by_nuclide, name)
self._rxn_type = 'decay-rate'
self._estimator = 'analog'
@property
def scores(self):
@ -1738,6 +1789,33 @@ class DecayRate(MDGXS):
return self._xs_tally
def get_homogenized_mgxs(self, other_mgxs):
"""Construct a homogenized MGXS with other MGXS objects.
This method constructs a new MGXS object that is the flux-weighted
combination of two MGXS objects. It is equivalent to what one would
obtain if the tally spatial domain were designed to encompass the
individual domains for both MGXS objects.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
return self._get_homogenized_mgxs(other_mgxs, 'delayed-nu-fission')
@add_metaclass(ABCMeta)
class MatrixMDGXS(MDGXS):
@ -2295,7 +2373,7 @@ class DelayedNuFissionMatrixXS(MatrixMDGXS):
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'}
estimator : '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

131
openmc/mgxs/mgxs.py
View file

@ -996,6 +996,111 @@ class MGXS(object):
avg_xs.sparse = self.sparse
return avg_xs
def _get_homogenized_mgxs(self, other_mgxs, denom_score='flux'):
"""Construct a homogenized MGXS with other MGXS objects.
This method constructs a new MGXS object that is the flux-weighted
combination of two MGXS objects. It is equivalent to what one would
obtain if the tally spatial domain were designed to encompass the
individual domains for both MGXS objects. This is accomplished by
summing the rxn rate (numerator) tally and the denominator tally
(often a tally of the flux over the spatial domain) that are used to
compute a multi-group cross-section.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
denom_score : str
The denominator score in the denominator of computing the MGXS.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
# Check type of denom score
cv.check_type('denom_score', denom_score, str)
# Construct a collection of the subdomain filter bins to homogenize
# across
if isinstance(other_mgxs, openmc.mgxs.MGXS):
other_mgxs = [other_mgxs]
cv.check_iterable_type('other_mgxs', other_mgxs, openmc.mgxs.MGXS)
for mgxs in other_mgxs:
if mgxs.rxn_type != self.rxn_type:
msg = 'Not able to homogenize two MGXS with different rxn types'
raise ValueError(msg)
# Clone this MGXS to initialize the homogenized version
homogenized_mgxs = copy.deepcopy(self)
homogenized_mgxs._derived = True
name = 'hom({}, '.format(self.domain.name)
# Get the domain filter
filter_type = _DOMAIN_TO_FILTER[self.domain_type]
self_filter = self.rxn_rate_tally.find_filter(filter_type)
# Get the rxn rate and denom tallies
rxn_rate_tally = self.rxn_rate_tally
denom_tally = self.tallies[denom_score]
for mgxs in other_mgxs:
# Swap the domain filter bins for the other mgxs rxn rate tally
other_rxn_rate_tally = copy.deepcopy(mgxs.rxn_rate_tally)
other_filter = other_rxn_rate_tally.find_filter(filter_type)
other_filter._bins = self_filter._bins
# Swap the domain filter bins for the denom tally
other_denom_tally = copy.deepcopy(mgxs.tallies[denom_score])
other_filter = other_denom_tally.find_filter(filter_type)
other_filter._bins = self_filter._bins
# Add the rxn rate and denom tallies
rxn_rate_tally += other_rxn_rate_tally
denom_tally += other_denom_tally
# Update the name for the homogenzied MGXS
name += '{}, '.format(mgxs.domain.name)
# Set the properties of the homogenized MGXS
homogenized_mgxs._rxn_rate_tally = rxn_rate_tally
homogenized_mgxs.tallies[denom_score] = denom_tally
homogenized_mgxs._domain.name = name[:-2] + ')'
return homogenized_mgxs
def get_homogenized_mgxs(self, other_mgxs):
"""Construct a homogenized mgxs with other MGXS objects.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
return self._get_homogenized_mgxs(other_mgxs, 'flux')
def get_slice(self, nuclides=[], groups=[]):
"""Build a sliced MGXS for the specified nuclides and energy groups.
@ -3182,7 +3287,7 @@ class NuScatterXS(MGXS):
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'collision', 'analog'}
estimator : '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
@ -4561,6 +4666,28 @@ class Chi(MGXS):
return self._xs_tally
def get_homogenized_mgxs(self, other_mgxs):
"""Construct a homogenized mgxs with other MGXS objects.
Parameters
----------
other_mgxs : openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS
The MGXS to homogenize with this one.
Returns
-------
openmc.mgxs.MGXS
A new homogenized MGXS
Raises
------
ValueError
If the other_mgxs is of a different type.
"""
return self._get_homogenized_mgxs(other_mgxs, 'nu-fission-in')
def get_slice(self, nuclides=[], groups=[]):
"""Build a sliced Chi for the specified nuclides and energy groups.
@ -5338,7 +5465,7 @@ class PromptNuFissionMatrixXS(MatrixMGXS):
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'collision', 'analog'}
estimator : '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

106
openmc/mgxs_library.py
View file

@ -15,8 +15,8 @@ from openmc.checkvalue import check_type, check_value, check_greater_than, \
# Supported incoming particle MGXS angular treatment representations
_REPRESENTATIONS = ['isotropic', 'angle']
_SCATTER_TYPES = ['tabular', 'legendre', 'histogram']
_XS_SHAPES = ["[G][G'][Order]", "[G]", "[G']", "[G][G']", "[DG]", "[G][DG]",
"[G'][DG]", "[G][G'][DG]"]
_XS_SHAPES = ["[G][G'][Order]", "[G]", "[G']", "[G][G']", "[DG]", "[DG][G]",
"[DG][G']", "[DG][G][G']"]
class XSdata(object):
@ -145,11 +145,11 @@ class XSdata(object):
[DG]: beta, decay_rate
[G][DG]: delayed_nu_fission, beta, decay_rate
[DG][G]: delayed_nu_fission, beta, decay_rate
[G'][DG]: chi_delayed
[DG][G']: chi_delayed
[G][G'][DG]: delayed_nu_fission
[DG][G][G']: delayed_nu_fission
"""
@ -296,13 +296,13 @@ class XSdata(object):
self._xs_shapes["[G][G']"] = (self.energy_groups.num_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG]"] = (self.num_delayed_groups,)
self._xs_shapes["[G][DG]"] = (self.energy_groups.num_groups,
self.num_delayed_groups)
self._xs_shapes["[G'][DG]"] = (self.energy_groups.num_groups,
self.num_delayed_groups)
self._xs_shapes["[G][G'][DG]"] = (self.energy_groups.num_groups,
self._xs_shapes["[DG][G]"] = (self.num_delayed_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG'][G']"] = (self.num_delayed_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG][G][G']"] = (self.num_delayed_groups,
self.energy_groups.num_groups,
self.num_delayed_groups)
self.energy_groups.num_groups)
self._xs_shapes["[G][G'][Order]"] \
= (self.energy_groups.num_groups,
@ -634,7 +634,7 @@ class XSdata(object):
"""
# Get the accepted shapes for this xs
shapes = [self.xs_shapes["[G']"], self.xs_shapes["[G'][DG]"]]
shapes = [self.xs_shapes["[G']"], self.xs_shapes["[DG][G']"]]
# Convert to a numpy array so we can easily get the shape for checking
chi_delayed = np.asarray(chi_delayed)
@ -664,7 +664,7 @@ class XSdata(object):
"""
# Get the accepted shapes for this xs
shapes = [self.xs_shapes["[DG]"], self.xs_shapes["[G][DG]"]]
shapes = [self.xs_shapes["[DG]"], self.xs_shapes["[DG][G]"]]
# Convert to a numpy array so we can easily get the shape for checking
beta = np.asarray(beta)
@ -694,7 +694,7 @@ class XSdata(object):
"""
# Get the accepted shapes for this xs
shapes = [self.xs_shapes["[DG]"], self.xs_shapes["[G][DG]"]]
shapes = [self.xs_shapes["[DG]"], self.xs_shapes["[DG][G]"]]
# Convert to a numpy array so we can easily get the shape for checking
decay_rate = np.asarray(decay_rate)
@ -856,7 +856,7 @@ class XSdata(object):
"""
# Get the accepted shapes for this xs
shapes = [self.xs_shapes["[G][DG]"], self.xs_shapes["[G][G'][DG]"]]
shapes = [self.xs_shapes["[DG][G]"], self.xs_shapes["[DG][G][G']"]]
# Convert to a numpy array so we can easily get the shape for checking
delayed_nu_fission = np.asarray(delayed_nu_fission)
@ -925,7 +925,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -970,7 +970,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1017,7 +1017,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1064,7 +1064,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1119,7 +1119,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1174,7 +1174,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1231,7 +1231,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1277,7 +1277,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1320,7 +1320,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1366,7 +1366,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1414,7 +1414,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1459,7 +1459,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1509,7 +1509,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1598,7 +1598,7 @@ class XSdata(object):
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata
openmc.mgxs.Library.get_xsdata()
"""
@ -1644,6 +1644,54 @@ class XSdata(object):
self._multiplicity_matrix[i] = \
np.nan_to_num(self._multiplicity_matrix[i])
def set_inverse_velocity_mgxs(self, inverse_velocity, temperature=294.,
nuclide='total', xs_type='macro',
subdomain=None):
"""This method allows for an openmc.mgxs.InverseVelocity
to be used to set the inverse velocity for this XSdata object.
Parameters
----------
inverse_velocity : openmc.mgxs.InverseVelocity
MGXS object containing the inverse velocity for the domain of
interest.
temperature : float
Temperature (in Kelvin) of the data. Defaults to room temperature
(294K).
nuclide : str
Individual nuclide (or 'total' if obtaining material-wise data)
to gather data for. Defaults to 'total'.
xs_type: {'macro', 'micro'}
Provide the macro or micro cross section in units of cm^-1 or
barns. Defaults to 'macro'.
subdomain : iterable of int
If the MGXS contains a mesh domain type, the subdomain parameter
specifies which mesh cell (i.e., [i, j, k] index) to use.
See also
--------
openmc.mgxs.Library.create_mg_library()
openmc.mgxs.Library.get_xsdata()
"""
check_type('inverse_velocity', inverse_velocity, openmc.mgxs.InverseVelocity)
check_value('energy_groups', inverse_velocity.energy_groups,
[self.energy_groups])
check_value('domain_type', inverse_velocity.domain_type,
openmc.mgxs.DOMAIN_TYPES)
check_type('temperature', temperature, Real)
check_value('temperature', temperature, self.temperatures)
i = np.where(self.temperatures == temperature)[0][0]
if self.representation == 'isotropic':
self._inverse_velocity[i] = inverse_velocity.get_xs\
(nuclides=nuclide, xs_type=xs_type,
subdomains=subdomain)
elif self.representation == 'angle':
msg = 'Angular-Dependent MGXS have not yet been implemented'
raise ValueError(msg)
def to_hdf5(self, file):
"""Write XSdata to an HDF5 file

2
openmc/settings.py
View file

@ -830,7 +830,7 @@ class Settings(object):
def _create_keff_trigger_subelement(self, run_mode_element):
if self._keff_trigger is not None:
element = ET.SubElement(run_mode_subelement, "keff_trigger")
element = ET.SubElement(run_mode_element, "keff_trigger")
for key in self._keff_trigger:
subelement = ET.SubElement(element, key)

11
openmc/tallies.py
View file

@ -705,7 +705,7 @@ class Tally(object):
"""
# Two tallys must have the same number of filters
# Two tallies must have the same number of filters
if len(self.filters) != len(other.filters):
return False
@ -3486,9 +3486,14 @@ class Tallies(cv.CheckedList):
for d in derivs:
root_element.append(d.to_xml_element())
def export_to_xml(self):
def export_to_xml(self, path='tallies.xml'):
"""Create a tallies.xml file that can be used for a simulation.
Parameters
----------
path : str
Path to file to write. Defaults to 'tallies.xml'.
"""
root_element = ET.Element("tallies")
@ -3501,5 +3506,5 @@ class Tallies(cv.CheckedList):
# Write the XML Tree to the tallies.xml file
tree = ET.ElementTree(root_element)
tree.write("tallies.xml", xml_declaration=True,
tree.write(path, xml_declaration=True,
encoding='utf-8', method="xml")

7
src/input_xml.F90
View file

@ -3668,13 +3668,6 @@ contains
end if
case ('decay-rate')
t % score_bins(j) = SCORE_DECAY_RATE
! Set tally estimator to analog for CE mode
! (MG mode has all data available without a collision being
! necessary)
if (run_CE) then
t % estimator = ESTIMATOR_ANALOG
end if
case ('delayed-nu-fission')
t % score_bins(j) = SCORE_DELAYED_NU_FISSION
if (t % find_filter(FILTER_ENERGYOUT) > 0) then

3
src/nuclide_header.F90
View file

@ -621,7 +621,8 @@ module nuclide_header
case (EMISSION_DELAYED)
if (this % n_precursor > 0) then
if (present(group)) then
if (present(group) .and. group < &
size(this % reactions(this % index_fission(1)) % products)) then
! If delayed group specified, determine yield immediately
associate(p => this % reactions(this % index_fission(1)) % products(1 + group))
nu = p % yield % evaluate(E)

273
src/tally.F90
View file

@ -679,8 +679,8 @@ contains
end if
end if
case (SCORE_DECAY_RATE)
! make sure the correct energy is used
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
E = p % E
@ -691,12 +691,134 @@ contains
! Set the delayedgroup filter index
dg_filter = t % find_filter(FILTER_DELAYEDGROUP)
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! delayed-nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO .and. &
nuclides(p % event_nuclide) % fissionable) then
if (t % estimator == ESTIMATOR_ANALOG) then
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! delayed-nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO .and. &
nuclides(p % event_nuclide) % fissionable) then
! Check if the delayed group filter is present
if (dg_filter > 0) then
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! Loop over all delayed group bins and tally to them
! individually
do d_bin = 1, filt % n_bins
! Get the delayed group for this bin
d = filt % groups(d_bin)
! Compute the yield for this delayed group
yield = nuclides(p % event_nuclide) &
% nu(E, EMISSION_DELAYED, d)
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! Compute the score
score = p % absorb_wgt * yield * &
micro_xs(p % event_nuclide) % fission &
/ micro_xs(p % event_nuclide) % absorption &
* rxn % products(1 + d) % decay_rate
end associate
! Tally to bin
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
end select
else
! If the delayed group filter is not present, compute the score
! by accumulating the absorbed weight times the decay rate times
! the fraction of the delayed-nu-fission xs to the absorption xs
! for all delayed groups.
score = ZERO
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! We need to be careful not to overshoot the number of delayed
! groups since this could cause the range of the rxn % products
! array to be exceeded. Hence, we use the size of this array
! and not the MAX_DELAYED_GROUPS constant for this loop.
do d = 1, size(rxn % products) - 2
score = score + rxn % products(1 + d) % decay_rate * &
p % absorb_wgt * micro_xs(p % event_nuclide) % fission *&
nuclides(p % event_nuclide) % nu(E, EMISSION_DELAYED, d)&
/ micro_xs(p % event_nuclide) % absorption
end do
end associate
end if
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. Since this was weighted by 1/keff, we multiply by keff
! to get the proper score. Loop over the neutrons produced from
! fission and check which ones are delayed. If a delayed neutron is
! encountered, add its contribution to the fission bank to the
! score.
score = ZERO
! 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
! Case for tallying delayed emissions
if (g /= 0) then
! Accumulate the decay rate times delayed nu fission score
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! determine score based on bank site weight and keff.
score = score + keff * fission_bank(n_bank - p % n_bank + k) &
% wgt * rxn % products(1 + g) % decay_rate
end associate
! if the delayed group filter is present, tally to corresponding
! delayed group bin if it exists
if (dg_filter > 0) then
! declare the delayed group filter type
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! loop over delayed group bins until the corresponding bin is
! found
do d_bin = 1, filt % n_bins
d = filt % groups(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, score_index)
end if
end do
end select
! Reset the score to zero
score = ZERO
end if
end if
end do
end if
else
! Check if tally is on a single nuclide
if (i_nuclide > 0) then
! Check if the delayed group filter is present
if (dg_filter > 0) then
@ -711,17 +833,14 @@ contains
d = filt % groups(d_bin)
! Compute the yield for this delayed group
yield = nuclides(p % event_nuclide) &
% nu(E, EMISSION_DELAYED, d)
yield = nuclides(i_nuclide) % nu(E, EMISSION_DELAYED, d)
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
associate (rxn => nuclides(i_nuclide) % &
reactions(nuclides(i_nuclide) % index_fission(1)))
! Compute the score
score = p % absorb_wgt * yield * &
micro_xs(p % event_nuclide) % fission &
/ micro_xs(p % event_nuclide) % absorption &
* rxn % products(1 + d) % decay_rate
! Compute the score and tally to bin
score = micro_xs(i_nuclide) % fission * yield * flux * &
atom_density * rxn % products(1 + d) % decay_rate
end associate
! Tally to bin
@ -746,78 +865,90 @@ contains
! and not the MAX_DELAYED_GROUPS constant for this loop.
do d = 1, size(rxn % products) - 2
score = score + rxn % products(1 + d) % decay_rate * &
p % absorb_wgt * micro_xs(p % event_nuclide) % fission *&
nuclides(p % event_nuclide) % nu(E, EMISSION_DELAYED, d)&
/ micro_xs(p % event_nuclide) % absorption
score = score + micro_xs(i_nuclide) % fission * flux * &
nuclides(i_nuclide) % nu(E, EMISSION_DELAYED) * &
atom_density * rxn % products(1 + d) % decay_rate
end do
end associate
end if
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. Since this was weighted by 1/keff, we multiply by keff
! to get the proper score. Loop over the neutrons produced from
! fission and check which ones are delayed. If a delayed neutron is
! encountered, add its contribution to the fission bank to the
! score.
! Tally is on total nuclides
else
score = ZERO
! Check if the delayed group filter is present
if (dg_filter > 0) then
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! loop over number of particles banked
do k = 1, p % n_bank
! Loop over all nuclides in the current material
do l = 1, materials(p % material) % n_nuclides
! get the delayed group
g = fission_bank(n_bank - p % n_bank + k) % delayed_group
! Get atom density
atom_density_ = materials(p % material) % atom_density(l)
! Case for tallying delayed emissions
if (g /= 0) then
! Get index in nuclides array
i_nuc = materials(p % material) % nuclide(l)
! Accumulate the decay rate times delayed nu fission score
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
if (nuclides(i_nuc) % fissionable) then
! determine score based on bank site weight and keff.
score = score + keff * fission_bank(n_bank - p % n_bank + k) &
% wgt * rxn % products(1 + g) % decay_rate
end associate
! Loop over all delayed group bins and tally to them
! individually
do d_bin = 1, filt % n_bins
! if the delayed group filter is present, tally to corresponding
! delayed group bin if it exists
if (dg_filter > 0) then
! Get the delayed group for this bin
d = filt % groups(d_bin)
! declare the delayed group filter type
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! Get the yield for the desired nuclide and delayed group
yield = nuclides(i_nuc) % nu(E, EMISSION_DELAYED, d)
! loop over delayed group bins until the corresponding bin is
! found
do d_bin = 1, filt % n_bins
d = filt % groups(d_bin)
associate (rxn => nuclides(i_nuc) % &
reactions(nuclides(i_nuc) % index_fission(1)))
! check whether the delayed group of the particle is equal to
! the delayed group of this bin
if (d == g) then
! Compute the score
score = micro_xs(i_nuc) % fission * yield * flux * &
atom_density_ * rxn % products(1 + d) % decay_rate
end associate
! Tally to bin
call score_fission_delayed_dg(t, d_bin, score, score_index)
end if
end do
end select
end do
end if
end do
cycle SCORE_LOOP
end select
else
! Reset the score to zero
score = ZERO
end if
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)
if (nuclides(i_nuc) % fissionable) then
associate (rxn => nuclides(i_nuc) % &
reactions(nuclides(i_nuc) % index_fission(1)))
! We need to be careful not to overshoot the number of delayed
! groups since this could cause the range of the rxn % products
! array to be exceeded. Hence, we use the size of this array
! and not the MAX_DELAYED_GROUPS constant for this loop.
do d = 1, size(rxn % products) - 2
! Accumulate the contribution from each nuclide
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) %&
nu(E, EMISSION_DELAYED) * atom_density_ * flux * &
rxn % products(1 + d) % decay_rate
end do
end associate
end if
end do
end if
end do
! If the delayed group filter is present, cycle because the
! score_fission_delayed_dg(...) has already tallied the score
if (dg_filter > 0) then
cycle SCORE_LOOP
end if
end if

12
tests/test_mgxs_library_condense/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

12
tests/test_mgxs_library_condense/results_true.dat
View file

@ -64,12 +64,12 @@
4 10000 5 1 total 0.001231 0.000105
5 10000 6 1 total 0.000512 0.000044
material delayedgroup group in nuclide mean std. dev.
0 10000 1 1 total 0.000000 0.000000
1 10000 2 1 total 0.032739 0.028454
2 10000 3 1 total 0.120780 0.170809
3 10000 4 1 total 0.302780 0.109110
4 10000 5 1 total 0.000000 0.000000
5 10000 6 1 total 0.000000 0.000000
0 10000 1 1 total 0.013355 0.001207
1 10000 2 1 total 0.032600 0.002866
2 10000 3 1 total 0.121083 0.010442
3 10000 4 1 total 0.305910 0.025627
4 10000 5 1 total 0.861934 0.068282
5 10000 6 1 total 2.895065 0.230223
material delayedgroup group in group out nuclide mean std. dev.
0 10000 1 1 1 total 0.000000 0.000000
1 10000 2 1 1 total 0.000384 0.000236

4
tests/test_mgxs_library_distribcell/inputs_true.dat
View file

@ -423,7 +423,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="distribcell" />
@ -431,7 +431,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="distribcell" />

12
tests/test_mgxs_library_distribcell/results_true.dat
View file

@ -64,12 +64,12 @@
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.001210 0.000058
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 0.000504 0.000024
sum(distribcell) delayedgroup group in nuclide mean std. dev.
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 total 0.000000 0.000000
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 total 0.032739 0.046300
2 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 3 1 total 0.120780 0.170809
3 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 4 1 total 0.000000 0.000000
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.000000 0.000000
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 2.853000 4.034751
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 total 0.013353 0.000686
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 total 0.032613 0.001627
2 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 3 1 total 0.121054 0.005911
3 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 4 1 total 0.305627 0.014428
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.860892 0.037879
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 2.891521 0.127879
sum(distribcell) delayedgroup group in group out nuclide mean std. dev.
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 1 total 0.000000 0.000000
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 1 total 0.000175 0.000175

12
tests/test_mgxs_library_hdf5/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

24
tests/test_mgxs_library_hdf5/results_true.dat
View file

@ -117,18 +117,18 @@ domain=10000 type=beta
[1.82980497e-04 4.50738567e-04]
[7.48899920e-05 1.88812772e-04]]
domain=10000 type=decay-rate
[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 3.27390000e-02]
[0.00000000e+00 1.20780000e-01]
[0.00000000e+00 3.02780000e-01]
[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]
[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 2.84543737e-02]
[0.00000000e+00 1.70808714e-01]
[0.00000000e+00 1.09109511e-01]
[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]
[[1.34450193e-02 1.33360001e-02]
[3.20638662e-02 3.27389978e-02]
[1.22136025e-01 1.20780007e-01]
[3.15269337e-01 3.02780066e-01]
[8.89232590e-01 8.49490287e-01]
[2.98940410e+00 2.85300088e+00]]
[[1.08439397e-03 1.44623816e-03]
[2.65794939e-03 3.55041884e-03]
[1.02955008e-02 1.30981285e-02]
[2.71748525e-02 3.28353351e-02]
[7.93682876e-02 9.21239479e-02]
[2.66253276e-01 3.09396954e-01]]
domain=10000 type=delayed-nu-fission matrix
[[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]

4
tests/test_mgxs_library_mesh/inputs_true.dat
View file

@ -660,7 +660,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="1" type="mesh" />
@ -668,7 +668,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="1" type="mesh" />

52
tests/test_mgxs_library_mesh/results_true.dat
View file

@ -214,32 +214,32 @@
21 2 2 1 4 1 total 0.002675 0.000783
22 2 2 1 5 1 total 0.001199 0.000341
23 2 2 1 6 1 total 0.000499 0.000142
mesh 1 delayedgroup group in nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.00000 0.00000
1 1 1 1 2 1 total 0.00000 0.00000
2 1 1 1 3 1 total 0.00000 0.00000
3 1 1 1 4 1 total 0.00000 0.00000
4 1 1 1 5 1 total 0.84949 1.20136
5 1 1 1 6 1 total 0.00000 0.00000
6 1 2 1 1 1 total 0.00000 0.00000
7 1 2 1 2 1 total 0.00000 0.00000
8 1 2 1 3 1 total 0.00000 0.00000
9 1 2 1 4 1 total 0.00000 0.00000
10 1 2 1 5 1 total 0.00000 0.00000
11 1 2 1 6 1 total 0.00000 0.00000
12 2 1 1 1 1 total 0.00000 0.00000
13 2 1 1 2 1 total 0.00000 0.00000
14 2 1 1 3 1 total 0.00000 0.00000
15 2 1 1 4 1 total 0.00000 0.00000
16 2 1 1 5 1 total 0.00000 0.00000
17 2 1 1 6 1 total 0.00000 0.00000
18 2 2 1 1 1 total 0.00000 0.00000
19 2 2 1 2 1 total 0.00000 0.00000
20 2 2 1 3 1 total 0.00000 0.00000
21 2 2 1 4 1 total 0.00000 0.00000
22 2 2 1 5 1 total 0.00000 0.00000
23 2 2 1 6 1 total 0.00000 0.00000
mesh 1 delayedgroup group in nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.013362 0.003586
1 1 1 1 2 1 total 0.032554 0.008644
2 1 1 1 3 1 total 0.121179 0.031941
3 1 1 1 4 1 total 0.306858 0.079970
4 1 1 1 5 1 total 0.865303 0.220416
5 1 1 1 6 1 total 2.906554 0.741587
6 1 2 1 1 1 total 0.013362 0.004521
7 1 2 1 2 1 total 0.032556 0.011045
8 1 2 1 3 1 total 0.121174 0.041226
9 1 2 1 4 1 total 0.306816 0.104995
10 1 2 1 5 1 total 0.865155 0.301124
11 1 2 1 6 1 total 2.906049 1.010025
12 2 1 1 1 1 total 0.013353 0.002006
13 2 1 1 2 1 total 0.032614 0.004677
14 2 1 1 3 1 total 0.121052 0.016785
15 2 1 1 4 1 total 0.305600 0.040199
16 2 1 1 5 1 total 0.860792 0.101713
17 2 1 1 6 1 total 2.891182 0.344217
18 2 2 1 1 1 total 0.013356 0.004047
19 2 2 1 2 1 total 0.032596 0.009453
20 2 2 1 3 1 total 0.121091 0.034091
21 2 2 1 4 1 total 0.305990 0.082530
22 2 2 1 5 1 total 0.862223 0.216936
23 2 2 1 6 1 total 2.896051 0.731528
mesh 1 delayedgroup group in group out nuclide mean std. dev.
x y z
0 1 1 1 1 1 1 total 0.000000 0.000000

12
tests/test_mgxs_library_no_nuclides/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

24
tests/test_mgxs_library_no_nuclides/results_true.dat
View file

@ -129,18 +129,18 @@
8 10000 5 2 total 0.004684 0.000451
10 10000 6 2 total 0.001962 0.000189
material delayedgroup group in nuclide mean std. dev.
1 10000 1 1 total 0.000000 0.000000
3 10000 2 1 total 0.000000 0.000000
5 10000 3 1 total 0.000000 0.000000
7 10000 4 1 total 0.000000 0.000000
9 10000 5 1 total 0.000000 0.000000
11 10000 6 1 total 0.000000 0.000000
0 10000 1 2 total 0.000000 0.000000
2 10000 2 2 total 0.032739 0.028454
4 10000 3 2 total 0.120780 0.170809
6 10000 4 2 total 0.302780 0.109110
8 10000 5 2 total 0.000000 0.000000
10 10000 6 2 total 0.000000 0.000000
1 10000 1 1 total 0.013445 0.001084
3 10000 2 1 total 0.032064 0.002658
5 10000 3 1 total 0.122136 0.010296
7 10000 4 1 total 0.315269 0.027175
9 10000 5 1 total 0.889233 0.079368
11 10000 6 1 total 2.989404 0.266253
0 10000 1 2 total 0.013336 0.001446
2 10000 2 2 total 0.032739 0.003550
4 10000 3 2 total 0.120780 0.013098
6 10000 4 2 total 0.302780 0.032835
8 10000 5 2 total 0.849490 0.092124
10 10000 6 2 total 2.853001 0.309397
material delayedgroup group in group out nuclide mean std. dev.
3 10000 1 1 1 total 0.000000 0.000000
7 10000 2 1 1 total 0.000000 0.000000