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Merge remote-tracking branch 'upstream/develop' into diff_tally3

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Sterling Harper 2015-11-19 20:17:31 -05:00
commit 90e7bd8771
20 changed files with 2372 additions and 1950 deletions
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1
.gitignore vendored
View file

@ -37,6 +37,7 @@ src/xml-fortran/xmlreader
# Test results error file
results_error.dat
inputs_error.dat
results_test.dat
# Test build files
tests/build/

21
docs/source/usersguide/input.rst
View file

@ -1136,6 +1136,15 @@ Each ``material`` element can have the following attributes or sub-elements:
.. note:: If one nuclide is specified in atom percent, all others must also
be given in atom percent. The same applies for weight percentages.
An optional attribute/sub-element for each nuclide is ``scattering``. This
attribute may be set to "data" to use the scattering laws specified by the
cross section library (default). Alternatively, when set to "iso-in-lab",
the scattering laws are used to sample the outgoing energy but an
isotropic-in-lab distribution is used to sample the outgoing angle at each
scattering interaction. The ``scattering`` attribute may be most useful
when using OpenMC to compute multi-group cross-sections for deterministic
transport codes and to quantify the effects of anisotropic scattering.
*Default*: None
:element:
@ -1162,6 +1171,16 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: None
An optional attribute/sub-element for each element is ``scattering``. This
attribute may be set to "data" to use the scattering laws specified by the
cross section library (default). Alternatively, when set to "iso-in-lab",
the scattering laws are used to sample the outgoing energy but an
isotropic-in-lab distribution is used to sample the outgoing angle at each
scattering interaction. The ``scattering`` attribute may be most useful
when using OpenMC to compute multi-group cross-sections for deterministic
transport codes and to quantify the effects of anisotropic scattering.
*Default*: None
:sab:
Associates an S(a,b) table with the material. This element has
@ -1488,7 +1507,7 @@ The ``<tally>`` element accepts the following sub-elements:
:inverse-velocity:
The flux-weighted inverse velocity where the velocity is in units of
meters per second.
centimeters per second.
.. note::
The ``analog`` estimator is actually identical to the ``collision``

23
openmc/element.py
View file

@ -24,6 +24,8 @@ class Element(object):
Chemical symbol of the element, e.g. Pu
xs : str
Cross section identifier, e.g. 71c
scattering : 'data' or 'iso-in-lab' or None
The type of angular scattering distribution to use
"""
@ -31,6 +33,7 @@ class Element(object):
# Initialize class attributes
self._name = ''
self._xs = None
self._scattering = None
# Set class attributes
self.name = name
@ -60,6 +63,10 @@ class Element(object):
def __repr__(self):
string = 'Element - {0}\n'.format(self._name)
string += '{0: <16}{1}{2}\n'.format('\tXS', '=\t', self._xs)
if self.scattering is not None:
string += '{0: <16}{1}{2}\n'.format('\tscattering', '=\t',
self.scattering)
return string
@property
@ -70,6 +77,10 @@ class Element(object):
def name(self):
return self._name
@property
def scattering(self):
return self._scattering
@xs.setter
def xs(self, xs):
check_type('cross section identifier', xs, basestring)
@ -78,4 +89,14 @@ class Element(object):
@name.setter
def name(self, name):
check_type('name', name, basestring)
self._name = name
self._name = name
@scattering.setter
def scattering(self, scattering):
if not scattering in ['data', 'iso-in-lab']:
msg = 'Unable to set scattering for Element to {0} ' \
'which is not "data" or "iso-in-lab"'.format(scattering)
raise ValueError(msg)
self._scattering = scattering

24
openmc/material.py
View file

@ -33,8 +33,8 @@ NO_DENSITY = 99999.
class Material(object):
"""A material composed of a collection of nuclides/elements that can be assigned
to a region of space.
"""A material composed of a collection of nuclides/elements that can be
assigned to a region of space.
Parameters
----------
@ -371,6 +371,12 @@ class Material(object):
self._sab.append((name, xs))
def make_isotropic_in_lab(self):
for nuclide_name in self._nuclides:
self._nuclides[nuclide_name][0].scattering = 'iso-in-lab'
for element_name in self._elements:
self._element[element_name][0].scattering = 'iso-in-lab'
def get_all_nuclides(self):
"""Returns all nuclides in the material
@ -401,8 +407,11 @@ class Material(object):
else:
xml_element.set("wo", str(nuclide[1]))
if nuclide[0]._xs is not None:
xml_element.set("xs", nuclide[0]._xs)
if nuclide[0].xs is not None:
xml_element.set("xs", nuclide[0].xs)
if not nuclide[0].scattering is None:
xml_element.set("scattering", nuclide[0].scattering)
return xml_element
@ -416,6 +425,9 @@ class Material(object):
else:
xml_element.set("wo", str(element[1]))
if not element[0].scattering is None:
xml_element.set("scattering", element[0].scattering)
return xml_element
def _get_nuclides_xml(self, nuclides, distrib=False):
@ -590,6 +602,10 @@ class MaterialsFile(object):
self._materials.remove(material)
def make_isotropic_in_lab(self):
for material in self._materials:
material.make_isotropic_in_lab()
def _create_material_subelements(self):
subelement = ET.SubElement(self._materials_file, "default_xs")

62
openmc/mgxs/library.py
View file

@ -53,6 +53,8 @@ class Library(object):
The types of cross sections in the library (e.g., ['total', 'scatter'])
domain_type : {'material', 'cell', 'distribcell', 'universe'}
Domain type for spatial homogenization
domains : Iterable of Material, Cell or Universe
The spatial domain(s) for which MGXS in the Library are computed
correction : 'P0' or None
Apply the P0 correction to scattering matrices if set to 'P0'
energy_groups : EnergyGroups
@ -80,6 +82,7 @@ class Library(object):
self._by_nuclide = None
self._mgxs_types = []
self._domain_type = None
self._domains = 'all'
self._correction = 'P0'
self._energy_groups = None
self._tally_trigger = None
@ -105,6 +108,7 @@ class Library(object):
clone._by_nuclide = self.by_nuclide
clone._mgxs_types = self.mgxs_types
clone._domain_type = self.domain_type
clone._domains = self.domains
clone._correction = self.correction
clone._energy_groups = copy.deepcopy(self.energy_groups, memo)
clone._tally_trigger = copy.deepcopy(self.tally_trigger, memo)
@ -153,22 +157,24 @@ class Library(object):
def by_nuclide(self):
return self._by_nuclide
@property
def domains(self):
if self.domain_type is None:
raise ValueError('Unable to get all domains without a domain type')
if self.domain_type == 'material':
return self.openmc_geometry.get_all_materials()
elif self.domain_type == 'cell' or self.domain_type == 'distribcell':
return self.openmc_geometry.get_all_material_cells()
elif self.domain_type == 'universe':
return self.openmc_geometry.get_all_universes()
@property
def domain_type(self):
return self._domain_type
@property
def domains(self):
if self._domains == 'all':
if self.domain_type == 'material':
return self.openmc_geometry.get_all_materials()
elif self.domain_type in ['cell', 'distribcell']:
return self.openmc_geometry.get_all_material_cells()
elif self.domain_type == 'universe':
return self.openmc_geometry.get_all_universes()
else:
raise ValueError('Unable to get domains without a domain type')
else:
return self._domains
@property
def correction(self):
return self._correction
@ -223,6 +229,38 @@ class Library(object):
cv.check_value('domain type', domain_type, tuple(openmc.mgxs.DOMAIN_TYPES))
self._domain_type = domain_type
@domains.setter
def domains(self, domains):
# Use all materials, cells or universes in the geometry as domains
if domains == 'all':
self._domains = domains
# User specified a list of material, cell or universe domains
else:
if self.domain_type == 'material':
cv.check_iterable_type('domain', domains, openmc.Material)
all_domains = self.openmc_geometry.get_all_materials()
elif self.domain_type in ['cell', 'distribcell']:
cv.check_iterable_type('domain', domains, openmc.Cell)
all_domains = self.openmc_geometry.get_all_material_cells()
elif self.domain_type == 'universe':
cv.check_iterable_type('domain', domains, openmc.Universe)
all_domains = self.openmc_geometry.get_all_universes()
else:
msg = 'Unable to set domains with ' \
'domain type "{}"'.format(self.domain_type)
raise ValueError(msg)
# Check that each domain can be found in the geometry
for domain in domains:
if domain not in all_domains:
msg = 'Domain "{}" could not be found in the ' \
'geometry.'.format(domain)
raise ValueError(msg)
self._domains = domains
@correction.setter
def correction(self, correction):
cv.check_value('correction', correction, ('P0', None))

44
openmc/mgxs/mgxs.py
View file

@ -824,9 +824,9 @@ class MGXS(object):
mean = tally.get_reshaped_data(value='mean')
std_dev = tally.get_reshaped_data(value='std_dev')
# Get the mean of the mean, std. dev. across requested subdomains
mean = np.mean(mean[subdomains, ...], axis=0)
std_dev = np.mean(std_dev[subdomains, ...]**2, axis=0)
# Get the mean, std. dev. across requested subdomains
mean = np.sum(mean[subdomains, ...], axis=0)
std_dev = np.sum(std_dev[subdomains, ...]**2, axis=0)
std_dev = np.sqrt(std_dev)
# If domain is distribcell, make subdomain-averaged a 'cell' domain
@ -1224,23 +1224,31 @@ class MGXS(object):
else:
df = df.drop('score', axis=1)
# Rename energy(out) columns
columns = []
if 'energy [MeV]' in df:
# Override energy groups bounds with indices
groups = np.arange(self.num_groups, 0, -1, dtype=np.int)
groups = np.repeat(groups, self.num_nuclides)
if 'energy [MeV]' in df and 'energyout [MeV]' in df:
df.rename(columns={'energy [MeV]': 'group in'}, inplace=True)
columns.append('group in')
if 'energyout [MeV]' in df:
df.rename(columns={'energyout [MeV]': 'group out'}, inplace=True)
columns.append('group out')
in_groups = np.tile(groups, self.num_subdomains)
in_groups = np.repeat(in_groups, self.num_groups)
df['group in'] = in_groups
# Loop over all energy groups and override the bounds with indices
template = '({0:.1e} - {1:.1e})'
bins = self.energy_groups.group_edges
for column in columns:
for i in range(self.num_groups):
group = template.format(bins[i], bins[i+1])
row_indices = df[column] == group
df.loc[row_indices, column] = self.num_groups - i
df.rename(columns={'energyout [MeV]': 'group out'}, inplace=True)
out_groups = np.tile(groups, self.num_subdomains * self.num_groups)
df['group out'] = out_groups
columns = ['group in', 'group out']
elif 'energyout [MeV]' in df:
df.rename(columns={'energyout [MeV]': 'group out'}, inplace=True)
in_groups = np.tile(groups, self.num_subdomains)
df['group out'] = in_groups
columns = ['group out']
elif 'energy [MeV]' in df:
df.rename(columns={'energy [MeV]': 'group in'}, inplace=True)
in_groups = np.tile(groups, self.num_subdomains)
df['group in'] = in_groups
columns = ['group in']
# Select out those groups the user requested
if groups != 'all':

29
openmc/nuclide.py
View file

@ -26,6 +26,8 @@ class Nuclide(object):
zaid : int
1000*(atomic number) + mass number. As an example, the zaid of U-235
would be 92235.
scattering : 'data' or 'iso-in-lab' or None
The type of angular scattering distribution to use
"""
@ -34,6 +36,7 @@ class Nuclide(object):
self._name = ''
self._xs = None
self._zaid = None
self._scattering = None
# Set the Material class attributes
self.name = name
@ -79,6 +82,10 @@ class Nuclide(object):
def zaid(self):
return self._zaid
@property
def scattering(self):
return self._scattering
@name.setter
def name(self, name):
check_type('name', name, basestring)
@ -92,4 +99,24 @@ class Nuclide(object):
@zaid.setter
def zaid(self, zaid):
check_type('zaid', zaid, Integral)
self._zaid = zaid
self._zaid = zaid
@scattering.setter
def scattering(self, scattering):
if not scattering in ['data', 'iso-in-lab']:
msg = 'Unable to set scattering for Nuclide to {0} ' \
'which is not "data" or "iso-in-lab"'.format(scattering)
raise ValueError(msg)
self._scattering = scattering
def __repr__(self):
string = 'Nuclide - {0}\n'.format(self._name)
string += '{0: <16}{1}{2}\n'.format('\tXS', '=\t', self.xs)
if self.zaid is not None:
string += '{0: <16}{1}{2}\n'.format('\tZAID', '=\t', self.zaid)
if self.scattering is not None:
string += '{0: <16}{1}{2}\n'.format('\tscattering', '=\t',
self.scattering)
return string

34
openmc/opencg_compatible.py
View file

@ -486,20 +486,22 @@ def get_opencg_cell(openmc_cell):
# works if the region is a single half-space or an intersection of
# half-spaces, i.e., no complex cells.
region = openmc_cell.region
if isinstance(region, Halfspace):
surface = region.surface
halfspace = -1 if region.side == '-' else 1
opencg_cell.add_surface(get_opencg_surface(surface), halfspace)
elif isinstance(region, Intersection):
for node in region.nodes:
if not isinstance(node, Halfspace):
raise NotImplementedError("Complex cells not yet supported "
"in OpenCG.")
surface = node.surface
halfspace = -1 if node.side == '-' else 1
if region is not None:
if isinstance(region, Halfspace):
surface = region.surface
halfspace = -1 if region.side == '-' else 1
opencg_cell.add_surface(get_opencg_surface(surface), halfspace)
else:
raise NotImplementedError("Complex cells not yet supported in OpenCG.")
elif isinstance(region, Intersection):
for node in region.nodes:
if not isinstance(node, Halfspace):
raise NotImplementedError("Complex cells not yet "
"supported in OpenCG.")
surface = node.surface
halfspace = -1 if node.side == '-' else 1
opencg_cell.add_surface(get_opencg_surface(surface), halfspace)
else:
raise NotImplementedError("Complex cells not yet supported "
"in OpenCG.")
# Add the OpenMC Cell to the global collection of all OpenMC Cells
OPENMC_CELLS[cell_id] = openmc_cell
@ -877,6 +879,7 @@ def get_opencg_lattice(openmc_lattice):
pitch = openmc_lattice.pitch
lower_left = openmc_lattice.lower_left
universes = openmc_lattice.universes
outer = openmc_lattice.outer
if len(pitch) == 2:
new_pitch = np.ones(3, dtype=np.float64)
@ -909,6 +912,8 @@ def get_opencg_lattice(openmc_lattice):
opencg_lattice.dimension = dimension
opencg_lattice.width = pitch
opencg_lattice.universes = universe_array
if outer is not None:
opencg_lattice.outside = get_opencg_universe(outer)
offset = np.array(lower_left, dtype=np.float64) - \
((np.array(pitch, dtype=np.float64) *
@ -955,6 +960,7 @@ def get_openmc_lattice(opencg_lattice):
width = opencg_lattice.width
offset = opencg_lattice.offset
universes = opencg_lattice.universes
outer = opencg_lattice.outside
# Initialize an empty array for the OpenMC nested Universes in this Lattice
universe_array = np.ndarray(tuple(np.array(dimension)),
@ -985,6 +991,8 @@ def get_openmc_lattice(opencg_lattice):
openmc_lattice.pitch = width
openmc_lattice.universes = universe_array
openmc_lattice.lower_left = lower_left
if outer is not None:
openmc_lattice.outer = get_openmc_universe(outer)
# Add the OpenMC Lattice to the global collection of all OpenMC Lattices
OPENMC_LATTICES[lattice_id] = openmc_lattice

196
openmc/plots.py
View file

@ -5,9 +5,9 @@ import sys
import numpy as np
import openmc
import openmc.checkvalue as cv
from openmc.clean_xml import *
from openmc.checkvalue import (check_type, check_value, check_length,
check_greater_than, check_less_than)
if sys.version_info[0] >= 3:
basestring = str
@ -143,70 +143,70 @@ class Plot(object):
self._id = AUTO_PLOT_ID
AUTO_PLOT_ID += 1
else:
check_type('plot ID', plot_id, Integral)
check_greater_than('plot ID', plot_id, 0, equality=True)
cv.check_type('plot ID', plot_id, Integral)
cv.check_greater_than('plot ID', plot_id, 0, equality=True)
self._id = plot_id
@name.setter
def name(self, name):
check_type('plot name', name, basestring)
cv.check_type('plot name', name, basestring)
self._name = name
@width.setter
def width(self, width):
check_type('plot width', width, Iterable, Real)
check_length('plot width', width, 2, 3)
cv.check_type('plot width', width, Iterable, Real)
cv.check_length('plot width', width, 2, 3)
self._width = width
@origin.setter
def origin(self, origin):
check_type('plot origin', origin, Iterable, Real)
check_length('plot origin', origin, 3)
cv.check_type('plot origin', origin, Iterable, Real)
cv.check_length('plot origin', origin, 3)
self._origin = origin
@pixels.setter
def pixels(self, pixels):
check_type('plot pixels', pixels, Iterable, Integral)
check_length('plot pixels', pixels, 2, 3)
cv.check_type('plot pixels', pixels, Iterable, Integral)
cv.check_length('plot pixels', pixels, 2, 3)
for dim in pixels:
check_greater_than('plot pixels', dim, 0)
cv.check_greater_than('plot pixels', dim, 0)
self._pixels = pixels
@filename.setter
def filename(self, filename):
check_type('filename', filename, basestring)
cv.check_type('filename', filename, basestring)
self._filename = filename
@color.setter
def color(self, color):
check_type('plot color', color, basestring)
check_value('plot color', color, ['cell', 'mat'])
cv.check_type('plot color', color, basestring)
cv.check_value('plot color', color, ['cell', 'mat'])
self._color = color
@type.setter
def type(self, plottype):
check_type('plot type', plottype, basestring)
check_value('plot type', plottype, ['slice', 'voxel'])
cv.check_type('plot type', plottype, basestring)
cv.check_value('plot type', plottype, ['slice', 'voxel'])
self._type = plottype
@basis.setter
def basis(self, basis):
check_type('plot basis', basis, basestring)
check_value('plot basis', basis, ['xy', 'xz', 'yz'])
cv.check_type('plot basis', basis, basestring)
cv.check_value('plot basis', basis, ['xy', 'xz', 'yz'])
self._basis = basis
@background.setter
def background(self, background):
check_type('plot background', background, Iterable, Integral)
check_length('plot background', background, 3)
cv.check_type('plot background', background, Iterable, Integral)
cv.check_length('plot background', background, 3)
for rgb in background:
check_greater_than('plot background',rgb, 0, True)
check_less_than('plot background', rgb, 256)
cv.check_greater_than('plot background',rgb, 0, True)
cv.check_less_than('plot background', rgb, 256)
self._background = background
@col_spec.setter
def col_spec(self, col_spec):
check_type('plot col_spec parameter', col_spec, dict, Integral)
cv.check_type('plot col_spec parameter', col_spec, dict, Integral)
for key in col_spec:
if key < 0:
@ -229,18 +229,18 @@ class Plot(object):
@mask_componenets.setter
def mask_components(self, mask_components):
check_type('plot mask_components', mask_components, Iterable, Integral)
cv.check_type('plot mask_components', mask_components, Iterable, Integral)
for component in mask_components:
check_greater_than('plot mask_components', component, 0, True)
cv.check_greater_than('plot mask_components', component, 0, True)
self._mask_components = mask_components
@mask_background.setter
def mask_background(self, mask_background):
check_type('plot mask background', mask_background, Iterable, Integral)
check_length('plot mask background', mask_background, 3)
cv.check_type('plot mask background', mask_background, Iterable, Integral)
cv.check_length('plot mask background', mask_background, 3)
for rgb in mask_background:
check_greater_than('plot mask background', rgb, 0, True)
check_less_than('plot mask background', rgb, 256)
cv.check_greater_than('plot mask background', rgb, 0, True)
cv.check_less_than('plot mask background', rgb, 256)
self._mask_background = mask_background
def __repr__(self):
@ -261,6 +261,97 @@ class Plot(object):
string += '{0: <16}{1}{2}\n'.format('\tCol Spec', '=\t', self._col_spec)
return string
def colorize(self, geometry, seed=1):
"""Generate a color scheme for each domain in the plot.
This routine may be used to generate random, reproducible color schemes.
The colors generated are based upon cell/material IDs in the geometry.
Parameters
----------
geometry : openmc.Geometry
The geometry for which the plot is defined
seed : Integral
The random number seed used to generate the color scheme
"""
cv.check_type('geometry', geometry, openmc.Geometry)
cv.check_type('seed', seed, Integral)
cv.check_greater_than('seed', seed, 1, equality=True)
# Get collections of the domains which will be plotted
if self.color is 'mat':
domains = geometry.get_all_materials()
else:
domains = geometry.get_all_cells()
# Set the seed for the random number generator
np.random.seed(seed)
# Generate random colors for each feature
self.col_spec = {}
for domain in domains:
r = np.random.randint(0, 256)
g = np.random.randint(0, 256)
b = np.random.randint(0, 256)
self.col_spec[domain] = (r, g, b)
def highlight_domains(self, geometry, domains, seed=1,
alpha=0.5, background='gray'):
"""Use alpha compositing to highlight one or more domains in the plot.
This routine generates a color scheme and applies alpha compositing
to make all domains except the highlighted ones appear partially
transparent.
Parameters
----------
geometry : openmc.Geometry
The geometry for which the plot is defined
domains : Iterable of Integral
A collection of the domain IDs to highlight in the plot
seed : Integral
The random number seed used to generate the color scheme
alpha : Real in [0,1]
The value to apply in alpha compisiting
background : 3-tuple of Integral or 'white' or 'black' or 'gray'
The background color to apply in alpha compisiting
"""
cv.check_iterable_type('domains', domains, Integral)
cv.check_type('alpha', alpha, Real)
cv.check_greater_than('alpha', alpha, 0., equality=True)
cv.check_less_than('alpha', alpha, 1., equality=True)
# Get a background (R,G,B) tuple to apply in alpha compositing
if isinstance(background, basestring):
if background == 'white':
background = (255, 255, 255)
elif background == 'black':
background = (0, 0, 0)
elif background == 'gray':
background = (160, 160, 160)
else:
msg = 'The background "{}" is not defined'.format(background)
raise ValueError(msg)
cv.check_iterable_type('background', background, Integral)
# Generate a color scheme
self.colorize(geometry, seed)
# Apply alpha compositing to the colors for all domains
# other than those the user wishes to highlight
for domain_id in self.col_spec:
if domain_id not in domains:
r, g, b = self.col_spec[domain_id]
r = int(((1-alpha) * background[0]) + (alpha * r))
g = int(((1-alpha) * background[1]) + (alpha * g))
b = int(((1-alpha) * background[2]) + (alpha * b))
self._col_spec[domain_id] = (r, g, b)
def get_plot_xml(self):
"""Return XML representation of the plot
@ -349,6 +440,51 @@ class PlotsFile(object):
self._plots.remove(plot)
def colorize(self, geometry, seed=1):
"""Generate a consistent color scheme for each domain in each plot.
This routine may be used to generate random, reproducible color schemes.
The colors generated are based upon cell/material IDs in the geometry.
The color schemes will be consistent for all plots in "plots.xml".
Parameters
----------
geometry : openmc.Geometry
The geometry for which the plots are defined
seed : Integral
The random number seed used to generate the color scheme
"""
for plot in self._plots:
plot.colorize(geometry, seed)
def highlight_domains(self, geometry, domains, seed=1,
alpha=0.5, background='gray'):
"""Use alpha compositing to highlight one or more domains in the plot.
This routine generates a color scheme and applies alpha compositing
to make all domains except the highlighted ones partially transparent.
Parameters
----------
geometry : openmc.Geometry
The geometry for which the plot is defined
domains : Iterable of Integral
A collection of the domain IDs to highlight in the plot
seed : Integral
The random number seed used to generate the color scheme
alpha : Real in [0,1]
The value to apply in alpha compisiting
background : 3-tuple of Integral or 'white' or 'black' or 'gray'
The background color to apply in alpha compisiting
"""
for plot in self._plots:
plot.highlight_domains(geometry, domains, seed, alpha, background)
def _create_plot_subelements(self):
for plot in self._plots:
xml_element = plot.get_plot_xml()

59
src/input_xml.F90
View file

@ -7,7 +7,7 @@ module input_xml
use error, only: fatal_error, warning
use geometry_header, only: Cell, Lattice, RectLattice, HexLattice
use global
use list_header, only: ListChar, ListReal
use list_header, only: ListChar, ListInt, ListReal
use mesh_header, only: RegularMesh
use output, only: write_message
use plot_header
@ -994,7 +994,7 @@ contains
logical :: boundary_exists
character(MAX_LINE_LEN) :: filename
character(MAX_WORD_LEN) :: word
character(MAX_LINE_LEN) :: region_spec
character(1000) :: region_spec
type(Cell), pointer :: c
class(Surface), pointer :: s
class(Lattice), pointer :: lat
@ -1759,8 +1759,10 @@ contains
integer :: i ! loop index for materials
integer :: j ! loop index for nuclides
integer :: k ! loop index for elements
integer :: n ! number of nuclides
integer :: n_sab ! number of sab tables for a material
integer :: n_nuc_ele ! number of nuclides in an element
integer :: index_list ! index in xs_listings array
integer :: index_nuclide ! index in nuclides
integer :: index_sab ! index in sab_tables
@ -1775,6 +1777,7 @@ contains
character(MAX_LINE_LEN) :: temp_str ! temporary string when reading
type(ListChar) :: list_names ! temporary list of nuclide names
type(ListReal) :: list_density ! temporary list of nuclide densities
type(ListInt) :: list_iso_lab ! temporary list of isotropic lab scatterers
type(Material), pointer :: mat => null()
type(Node), pointer :: doc => null()
type(Node), pointer :: node_mat => null()
@ -1934,6 +1937,21 @@ contains
end if
end if
! Check enforced isotropic lab scattering
if (check_for_node(node_nuc, "scattering")) then
call get_node_value(node_nuc, "scattering", temp_str)
if (adjustl(to_lower(temp_str)) == "iso-in-lab") then
call list_iso_lab % append(1)
else if (adjustl(to_lower(temp_str)) == "data") then
call list_iso_lab % append(0)
else
call fatal_error("Scattering must be isotropic in lab or follow&
& the ACE file data")
end if
else
call list_iso_lab % append(0)
end if
! store full name
call get_node_value(node_nuc, "name", temp_str)
if (check_for_node(node_nuc, "xs")) &
@ -2005,6 +2023,9 @@ contains
&element: " // trim(name))
end if
! Get current number of nuclides
n_nuc_ele = list_names % size()
! Expand element into naturally-occurring isotopes
if (check_for_node(node_ele, "ao")) then
call get_node_value(node_ele, "ao", temp_dble)
@ -2014,6 +2035,29 @@ contains
call fatal_error("The ability to expand a natural element based on &
&weight percentage is not yet supported.")
end if
! Compute number of new nuclides from the natural element expansion
n_nuc_ele = list_names % size() - n_nuc_ele
! Check enforced isotropic lab scattering
if (check_for_node(node_ele, "scattering")) then
call get_node_value(node_ele, "scattering", temp_str)
else
temp_str = "data"
end if
! Set ace or iso-in-lab scattering for each nuclide in element
do k = 1, n_nuc_ele
if (adjustl(to_lower(temp_str)) == "iso-in-lab") then
call list_iso_lab % append(1)
else if (adjustl(to_lower(temp_str)) == "data") then
call list_iso_lab % append(0)
else
call fatal_error("Scattering must be isotropic in lab or follow&
& the ACE file data")
end if
end do
end do NATURAL_ELEMENTS
! ========================================================================
@ -2025,6 +2069,7 @@ contains
allocate(mat % names(n))
allocate(mat % nuclide(n))
allocate(mat % atom_density(n))
allocate(mat % p0(n))
ALL_NUCLIDES: do j = 1, mat % n_nuclides
! Check that this nuclide is listed in the cross_sections.xml file
@ -2061,6 +2106,14 @@ contains
! Copy name and atom/weight percent
mat % names(j) = name
mat % atom_density(j) = list_density % get_item(j)
! Cast integer isotropic lab scattering flag to boolean
if (list_iso_lab % get_item(j) == 1) then
mat % p0(j) = .true.
else
mat % p0(j) = .false.
end if
end do ALL_NUCLIDES
! Check to make sure either all atom percents or all weight percents are
@ -2077,6 +2130,7 @@ contains
! Clear lists
call list_names % clear()
call list_density % clear()
call list_iso_lab % clear()
! =======================================================================
! READ AND PARSE <sab> TAG FOR S(a,b) DATA
@ -4245,7 +4299,6 @@ contains
call list_density % append(density * 0.999885_8)
call list_names % append('1002.' // xs)
call list_density % append(density * 0.000115_8)
case ('he')
call list_names % append('2003.' // xs)
call list_density % append(density * 0.00000134_8)

3
src/material_header.F90
View file

@ -34,6 +34,9 @@ module material_header
! Does this material contain fissionable nuclides?
logical :: fissionable = .false.
! enforce isotropic scattering in lab
logical, allocatable :: p0(:)
end type Material
end module material_header

46
src/physics.F90
View file

@ -73,10 +73,11 @@ contains
type(Particle), intent(inout) :: p
integer :: i_nuclide ! index in nuclides array
integer :: i_nuc_mat ! index in material's nuclides array
integer :: i_reaction ! index in nuc % reactions array
type(Nuclide), pointer :: nuc
i_nuclide = sample_nuclide(p, 'total ')
call sample_nuclide(p, 'total ', i_nuclide, i_nuc_mat)
! Get pointer to table
nuc => nuclides(i_nuclide)
@ -106,7 +107,7 @@ contains
! Sample a scattering reaction and determine the secondary energy of the
! exiting neutron
call scatter(p, i_nuclide)
call scatter(p, i_nuclide, i_nuc_mat)
! Play russian roulette if survival biasing is turned on
@ -121,13 +122,13 @@ contains
! SAMPLE_NUCLIDE
!===============================================================================
function sample_nuclide(p, base) result(i_nuclide)
subroutine sample_nuclide(p, base, i_nuclide, i_nuc_mat)
type(Particle), intent(in) :: p
character(7), intent(in) :: base ! which reaction to sample based on
integer :: i_nuclide
integer, intent(out) :: i_nuclide
integer, intent(out) :: i_nuc_mat
integer :: i
real(8) :: prob
real(8) :: cutoff
real(8) :: atom_density ! atom density of nuclide in atom/b-cm
@ -147,20 +148,20 @@ contains
cutoff = prn() * material_xs % fission
end select
i = 0
i_nuc_mat = 0
prob = ZERO
do while (prob < cutoff)
i = i + 1
i_nuc_mat = i_nuc_mat + 1
! Check to make sure that a nuclide was sampled
if (i > mat % n_nuclides) then
if (i_nuc_mat > mat % n_nuclides) then
call write_particle_restart(p)
call fatal_error("Did not sample any nuclide during collision.")
end if
! Find atom density
i_nuclide = mat % nuclide(i)
atom_density = mat % atom_density(i)
i_nuclide = mat % nuclide(i_nuc_mat)
atom_density = mat % atom_density(i_nuc_mat)
! Determine microscopic cross section
select case (base)
@ -177,7 +178,7 @@ contains
prob = prob + sigma
end do
end function sample_nuclide
end subroutine sample_nuclide
!===============================================================================
! SAMPLE_FISSION
@ -300,10 +301,11 @@ contains
! SCATTER
!===============================================================================
subroutine scatter(p, i_nuclide)
subroutine scatter(p, i_nuclide, i_nuc_mat)
type(Particle), intent(inout) :: p
integer, intent(in) :: i_nuclide
integer, intent(in) :: i_nuc_mat
integer :: i
integer :: i_grid
@ -312,6 +314,12 @@ contains
real(8) :: cutoff
type(Nuclide), pointer :: nuc
type(Reaction), pointer :: rxn
real(8) :: uvw_new(3) ! outgoing uvw for iso-in-lab scattering
real(8) :: uvw_old(3) ! incoming uvw for iso-in-lab scattering
real(8) :: phi ! azimuthal angle for iso-in-lab scattering
! copy incoming direction
uvw_old(:) = p % coord(1) % uvw
! Get pointer to nuclide and grid index/interpolation factor
nuc => nuclides(i_nuclide)
@ -390,6 +398,20 @@ contains
! Set event component
p % event = EVENT_SCATTER
! sample new outgoing angle for isotropic in lab scattering
if (materials(p % material) % p0(i_nuc_mat)) then
! sample isotropic-in-lab outgoing direction
uvw_new(1) = TWO * prn() - ONE
phi = TWO * PI * prn()
uvw_new(2) = cos(phi) * sqrt(ONE - uvw_new(1)*uvw_new(1))
uvw_new(3) = sin(phi) * sqrt(ONE - uvw_new(1)*uvw_new(1))
p % mu = dot_product(uvw_old, uvw_new)
! change direction of particle
p % coord(1) % uvw = uvw_new
end if
end subroutine scatter
!===============================================================================

4
src/relaxng/materials.rnc
View file

@ -15,6 +15,8 @@ element materials {
attribute name { xsd:string { maxLength = "7" } }) &
(element xs { xsd:string { maxLength = "3" } } |
attribute xs { xsd:string { maxLength = "3" } })? &
(element scattering { ( "data" | "iso-in-lab" ) } |
attribute scattering { ( "data" | "iso-in-lab" ) })? &
(
(element ao { xsd:double } | attribute ao { xsd:double }) |
(element wo { xsd:double } | attribute wo { xsd:double })
@ -26,6 +28,8 @@ element materials {
attribute name { xsd:string { maxLength = "2" } }) &
(element xs { xsd:string { maxLength = "3" } } |
attribute xs { xsd:string { maxLength = "3" } })? &
(element scattering { ( "data" | "iso-in-lab" ) } |
attribute scattering { ( "data" | "iso-in-lab" ) })? &
(
(element ao { xsd:double } | attribute ao { xsd:double }) |
(element wo { xsd:double } | attribute wo { xsd:double })

46
src/relaxng/materials.rng
View file

@ -12,6 +12,20 @@
<data type="int"/>
</attribute>
</choice>
<optional>
<choice>
<element name="name">
<data type="string">
<param name="maxLength">52</param>
</data>
</element>
<attribute name="name">
<data type="string">
<param name="maxLength">52</param>
</data>
</attribute>
</choice>
</optional>
<element name="density">
<interleave>
<optional>
@ -67,6 +81,22 @@
</attribute>
</choice>
</optional>
<optional>
<choice>
<element name="scattering">
<choice>
<value>data</value>
<value>iso-in-lab</value>
</choice>
</element>
<attribute name="scattering">
<choice>
<value>data</value>
<value>iso-in-lab</value>
</choice>
</attribute>
</choice>
</optional>
<choice>
<choice>
<element name="ao">
@ -117,6 +147,22 @@
</attribute>
</choice>
</optional>
<optional>
<choice>
<element name="scattering">
<choice>
<value>data</value>
<value>iso-in-lab</value>
</choice>
</element>
<attribute name="scattering">
<choice>
<value>data</value>
<value>iso-in-lab</value>
</choice>
</attribute>
</choice>
</optional>
<choice>
<choice>
<element name="ao">

44
src/tally.F90
View file

@ -65,6 +65,7 @@ contains
real(8) :: macro_total ! material macro total xs
real(8) :: macro_scatt ! material macro scatt xs
real(8) :: uvw(3) ! particle direction
real(8) :: E ! particle energy
type(Material), pointer :: mat
type(Reaction), pointer :: rxn
type(Nuclide), pointer :: nuc
@ -128,6 +129,14 @@ contains
case (SCORE_INVERSE_VELOCITY)
! make sure the correct energy is used
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
E = p % E
else
E = p % last_E
end if
if (t % estimator == ESTIMATOR_ANALOG) then
! All events score to an inverse velocity bin. We actually use a
! collision estimator in place of an analog one since there is no way
@ -139,12 +148,16 @@ contains
else
score = p % last_wgt
end if
! Score the flux weighted inverse velocity with velocity in units of
! cm/s
score = score / material_xs % total &
/ (sqrt(TWO * p % E / (MASS_NEUTRON_MEV)) * C_LIGHT)
/ (sqrt(TWO * E / (MASS_NEUTRON_MEV)) * C_LIGHT * 100.0_8)
else
! For inverse velocity, we need no cross section
score = flux / (sqrt(TWO * p % E / (MASS_NEUTRON_MEV)) * C_LIGHT)
! For inverse velocity, we don't need a cross section. The velocity is
! in units of cm/s.
score = flux / (sqrt(TWO * E / (MASS_NEUTRON_MEV)) * C_LIGHT * 100.0_8)
end if
@ -425,6 +438,13 @@ contains
case (SCORE_DELAYED_NU_FISSION)
! make sure the correct energy is used
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
E = p % E
else
E = p % last_E
end if
! Set the delayedgroup filter index and the number of delayed group bins
dg_filter = t % find_filter(FILTER_DELAYEDGROUP)
@ -460,11 +480,11 @@ contains
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the yield for this delayed group
yield = yield_delayed(nuc, p % E, d)
yield = yield_delayed(nuc, E, d)
! Compute the score and tally to bin
score = p % absorb_wgt * yield * micro_xs(p % event_nuclide) &
% fission * nu_delayed(nuc, p % E) / &
% fission * nu_delayed(nuc, E) / &
micro_xs(p % event_nuclide) % absorption
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
@ -474,7 +494,7 @@ contains
! by multiplying the absorbed weight by the fraction of the
! delayed-nu-fission xs to the absorption xs
score = p % absorb_wgt * micro_xs(p % event_nuclide) &
% fission * nu_delayed(nuc, p % E) / &
% fission * nu_delayed(nuc, E) / &
micro_xs(p % event_nuclide) % absorption
end if
end if
@ -528,11 +548,11 @@ contains
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the yield for this delayed group
yield = yield_delayed(nuc, p % E, d)
yield = yield_delayed(nuc, E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuclide) % fission * yield &
* nu_delayed(nuc, p % E) * atom_density * flux
* nu_delayed(nuc, E) * atom_density * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
@ -540,7 +560,7 @@ contains
! If the delayed group filter is not present, compute the score
! by multiplying the delayed-nu-fission macro xs by the flux
score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, p % E)&
score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, E)&
* atom_density * flux
end if
@ -572,11 +592,11 @@ contains
nuc => nuclides(i_nuc)
! Get the yield for the desired nuclide and delayed group
yield = yield_delayed(nuc, p % E, d)
yield = yield_delayed(nuc, E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuc) % fission * yield &
* nu_delayed(nuc, p % E) * atom_density_ * flux
* nu_delayed(nuc, E) * atom_density_ * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
end do
@ -596,7 +616,7 @@ contains
! Accumulate the contribution from each nuclide
score = score + micro_xs(i_nuc) % fission &
* nu_delayed(nuclides(i_nuc), p % E) * atom_density_ * flux
* nu_delayed(nuclides(i_nuc), E) * atom_density_ * flux
end do
end if
end if

98
tests/test_mgxs_library_condense/results_true.dat
View file

@ -1,49 +1,49 @@
material group in nuclide mean std. dev.
0 1 1 total 0.419289 0.01638 material group in nuclide mean std. dev.
0 1 1 total 0.07774 0.003273 material group in group out nuclide mean std. dev.
0 1 1 1 total 0.352665 0.015654 material group out nuclide mean std. dev.
0 1 1 total 1 0.119622 material group in nuclide mean std. dev.
0 2 1 total 0.247316 0.009562 material group in nuclide mean std. dev.
0 2 1 total 0 0 material group in group out nuclide mean std. dev.
0 2 1 1 total 0.244838 0.009996 material group out nuclide mean std. dev.
0 2 1 total 0 0 material group in nuclide mean std. dev.
0 3 1 total 0.409938 0.042262 material group in nuclide mean std. dev.
0 3 1 total 0 0 material group in group out nuclide mean std. dev.
0 3 1 1 total 0.403354 0.041386 material group out nuclide mean std. dev.
0 3 1 total 0 0 material group in nuclide mean std. dev.
0 4 1 total 0.344007 0.05352 material group in nuclide mean std. dev.
0 4 1 total 0 0 material group in group out nuclide mean std. dev.
0 4 1 1 total 0.340438 0.052067 material group out nuclide mean std. dev.
0 4 1 total 0 0 material group in nuclide mean std. dev.
0 5 1 total 0 0 material group in nuclide mean std. dev.
0 5 1 total 0 0 material group in group out nuclide mean std. dev.
0 5 1 1 total 0 0 material group out nuclide mean std. dev.
0 5 1 total 0 0 material group in nuclide mean std. dev.
0 6 1 total 0 0 material group in nuclide mean std. dev.
0 6 1 total 0 0 material group in group out nuclide mean std. dev.
0 6 1 1 total 0 0 material group out nuclide mean std. dev.
0 6 1 total 0 0 material group in nuclide mean std. dev.
0 7 1 total 0 0 material group in nuclide mean std. dev.
0 7 1 total 0 0 material group in group out nuclide mean std. dev.
0 7 1 1 total 0 0 material group out nuclide mean std. dev.
0 7 1 total 0 0 material group in nuclide mean std. dev.
0 8 1 total 0 0 material group in nuclide mean std. dev.
0 8 1 total 0 0 material group in group out nuclide mean std. dev.
0 8 1 1 total 0 0 material group out nuclide mean std. dev.
0 8 1 total 0 0 material group in nuclide mean std. dev.
0 9 1 total 0.751873 0.559701 material group in nuclide mean std. dev.
0 9 1 total 0 0 material group in group out nuclide mean std. dev.
0 9 1 1 total 0.695491 0.50757 material group out nuclide mean std. dev.
0 9 1 total 0 0 material group in nuclide mean std. dev.
0 10 1 total 0 0 material group in nuclide mean std. dev.
0 10 1 total 0 0 material group in group out nuclide mean std. dev.
0 10 1 1 total 0 0 material group out nuclide mean std. dev.
0 10 1 total 0 0 material group in nuclide mean std. dev.
0 11 1 total 0.457329 0.403578 material group in nuclide mean std. dev.
0 11 1 total 0 0 material group in group out nuclide mean std. dev.
0 11 1 1 total 0.446737 0.392775 material group out nuclide mean std. dev.
0 11 1 total 0 0 material group in nuclide mean std. dev.
0 12 1 total 0.574978 0.38864 material group in nuclide mean std. dev.
0 12 1 total 0 0 material group in group out nuclide mean std. dev.
0 12 1 1 total 0.559478 0.377512 material group out nuclide mean std. dev.
0 12 1 total 0 0
material group in nuclide mean std. dev.
0 1 1 total 0.419289 0.01638 material group in nuclide mean std. dev.
0 1 1 total 0.07774 0.003273 material group in group out nuclide mean std. dev.
0 1 1 1 total 0.352665 0.015654 material group out nuclide mean std. dev.
0 1 1 total 1 0.119622 material group in nuclide mean std. dev.
0 2 1 total 0.247316 0.009562 material group in nuclide mean std. dev.
0 2 1 total 0 0 material group in group out nuclide mean std. dev.
0 2 1 1 total 0.244838 0.009996 material group out nuclide mean std. dev.
0 2 1 total 0 0 material group in nuclide mean std. dev.
0 3 1 total 0.409938 0.042262 material group in nuclide mean std. dev.
0 3 1 total 0 0 material group in group out nuclide mean std. dev.
0 3 1 1 total 0.403354 0.041386 material group out nuclide mean std. dev.
0 3 1 total 0 0 material group in nuclide mean std. dev.
0 4 1 total 0.344007 0.05352 material group in nuclide mean std. dev.
0 4 1 total 0 0 material group in group out nuclide mean std. dev.
0 4 1 1 total 0.340438 0.052067 material group out nuclide mean std. dev.
0 4 1 total 0 0 material group in nuclide mean std. dev.
0 5 1 total 0 0 material group in nuclide mean std. dev.
0 5 1 total 0 0 material group in group out nuclide mean std. dev.
0 5 1 1 total 0 0 material group out nuclide mean std. dev.
0 5 1 total 0 0 material group in nuclide mean std. dev.
0 6 1 total 0 0 material group in nuclide mean std. dev.
0 6 1 total 0 0 material group in group out nuclide mean std. dev.
0 6 1 1 total 0 0 material group out nuclide mean std. dev.
0 6 1 total 0 0 material group in nuclide mean std. dev.
0 7 1 total 0 0 material group in nuclide mean std. dev.
0 7 1 total 0 0 material group in group out nuclide mean std. dev.
0 7 1 1 total 0 0 material group out nuclide mean std. dev.
0 7 1 total 0 0 material group in nuclide mean std. dev.
0 8 1 total 0 0 material group in nuclide mean std. dev.
0 8 1 total 0 0 material group in group out nuclide mean std. dev.
0 8 1 1 total 0 0 material group out nuclide mean std. dev.
0 8 1 total 0 0 material group in nuclide mean std. dev.
0 9 1 total 0.751873 0.559701 material group in nuclide mean std. dev.
0 9 1 total 0 0 material group in group out nuclide mean std. dev.
0 9 1 1 total 0.695491 0.50757 material group out nuclide mean std. dev.
0 9 1 total 0 0 material group in nuclide mean std. dev.
0 10 1 total 0 0 material group in nuclide mean std. dev.
0 10 1 total 0 0 material group in group out nuclide mean std. dev.
0 10 1 1 total 0 0 material group out nuclide mean std. dev.
0 10 1 total 0 0 material group in nuclide mean std. dev.
0 11 1 total 0.457329 0.403578 material group in nuclide mean std. dev.
0 11 1 total 0 0 material group in group out nuclide mean std. dev.
0 11 1 1 total 0.446737 0.392775 material group out nuclide mean std. dev.
0 11 1 total 0 0 material group in nuclide mean std. dev.
0 12 1 total 0.574978 0.38864 material group in nuclide mean std. dev.
0 12 1 total 0 0 material group in group out nuclide mean std. dev.
0 12 1 1 total 0.559478 0.377512 material group out nuclide mean std. dev.
0 12 1 total 0 0

220
tests/test_mgxs_library_no_nuclides/results_true.dat
View file

@ -1,121 +1,121 @@
material group in nuclide mean std. dev.
1 1 1 total 0.384379 0.01649
0 1 2 total 0.812087 0.07419 material group in nuclide mean std. dev.
1 1 1 total 0.02127 0.000894
0 1 2 total 0.69604 0.053458 material group in group out nuclide mean std. dev.
3 1 1 1 total 0.349924 0.016649
2 1 1 2 total 0.000173 0.000173
1 1 2 1 total 0.001948 0.001952
0 1 2 2 total 0.379607 0.040078 material group out nuclide mean std. dev.
1 1 1 total 1 0.119622
0 1 2 total 0 0.000000 material group in nuclide mean std. dev.
1 2 1 total 0.245043 0.008827
0 2 2 total 0.266458 0.052209 material group in nuclide mean std. dev.
1 2 1 total 0 0
0 2 2 total 0 0 material group in group out nuclide mean std. dev.
3 2 1 1 total 0.243657 0.009083
2 2 1 2 total 0.000000 0.000000
1 2 2 1 total 0.000000 0.000000
0 2 2 2 total 0.254787 0.055563 material group out nuclide mean std. dev.
material group in nuclide mean std. dev.
1 1 1 total 0.384379 0.01649
0 1 2 total 0.812087 0.07419 material group in nuclide mean std. dev.
1 1 1 total 0.02127 0.000894
0 1 2 total 0.69604 0.053458 material group in group out nuclide mean std. dev.
3 1 1 1 total 0.349924 0.016649
2 1 1 2 total 0.000173 0.000173
1 1 2 1 total 0.001948 0.001952
0 1 2 2 total 0.379607 0.040078 material group out nuclide mean std. dev.
1 1 1 total 1 0.119622
0 1 2 total 0 0.000000 material group in nuclide mean std. dev.
1 2 1 total 0.245043 0.008827
0 2 2 total 0.266458 0.052209 material group in nuclide mean std. dev.
1 2 1 total 0 0
0 2 2 total 0 0 material group in nuclide mean std. dev.
1 3 1 total 0.282277 0.037242
0 3 2 total 1.427320 0.247127 material group in nuclide mean std. dev.
1 3 1 total 0 0
0 3 2 total 0 0 material group in group out nuclide mean std. dev.
3 3 1 1 total 0.253967 0.036173
2 3 1 2 total 0.027273 0.001807
1 3 2 1 total 0.000000 0.000000
0 3 2 2 total 1.376527 0.240257 material group out nuclide mean std. dev.
0 2 2 total 0 0 material group in group out nuclide mean std. dev.
3 2 1 1 total 0.243657 0.009083
2 2 1 2 total 0.000000 0.000000
1 2 2 1 total 0.000000 0.000000
0 2 2 2 total 0.254787 0.055563 material group out nuclide mean std. dev.
1 2 1 total 0 0
0 2 2 total 0 0 material group in nuclide mean std. dev.
1 3 1 total 0.282277 0.037242
0 3 2 total 1.427320 0.247127 material group in nuclide mean std. dev.
1 3 1 total 0 0
0 3 2 total 0 0 material group in nuclide mean std. dev.
1 4 1 total 0.255723 0.051917
0 4 2 total 1.179767 0.229380 material group in nuclide mean std. dev.
1 4 1 total 0 0
0 4 2 total 0 0 material group in group out nuclide mean std. dev.
3 4 1 1 total 0.232978 0.049771
2 4 1 2 total 0.022281 0.002625
1 4 2 1 total 0.000000 0.000000
0 4 2 2 total 1.146809 0.222198 material group out nuclide mean std. dev.
0 3 2 total 0 0 material group in group out nuclide mean std. dev.
3 3 1 1 total 0.253967 0.036173
2 3 1 2 total 0.027273 0.001807
1 3 2 1 total 0.000000 0.000000
0 3 2 2 total 1.376527 0.240257 material group out nuclide mean std. dev.
1 3 1 total 0 0
0 3 2 total 0 0 material group in nuclide mean std. dev.
1 4 1 total 0.255723 0.051917
0 4 2 total 1.179767 0.229380 material group in nuclide mean std. dev.
1 4 1 total 0 0
0 4 2 total 0 0 material group in nuclide mean std. dev.
1 5 1 total 0 0
0 5 2 total 0 0 material group in nuclide mean std. dev.
1 5 1 total 0 0
0 5 2 total 0 0 material group in group out nuclide mean std. dev.
3 5 1 1 total 0 0
2 5 1 2 total 0 0
1 5 2 1 total 0 0
0 5 2 2 total 0 0 material group out nuclide mean std. dev.
0 4 2 total 0 0 material group in group out nuclide mean std. dev.
3 4 1 1 total 0.232978 0.049771
2 4 1 2 total 0.022281 0.002625
1 4 2 1 total 0.000000 0.000000
0 4 2 2 total 1.146809 0.222198 material group out nuclide mean std. dev.
1 4 1 total 0 0
0 4 2 total 0 0 material group in nuclide mean std. dev.
1 5 1 total 0 0
0 5 2 total 0 0 material group in nuclide mean std. dev.
1 6 1 total 0 0
0 6 2 total 0 0 material group in nuclide mean std. dev.
1 6 1 total 0 0
0 6 2 total 0 0 material group in group out nuclide mean std. dev.
3 6 1 1 total 0 0
2 6 1 2 total 0 0
1 6 2 1 total 0 0
0 6 2 2 total 0 0 material group out nuclide mean std. dev.
0 5 2 total 0 0 material group in nuclide mean std. dev.
1 5 1 total 0 0
0 5 2 total 0 0 material group in group out nuclide mean std. dev.
3 5 1 1 total 0 0
2 5 1 2 total 0 0
1 5 2 1 total 0 0
0 5 2 2 total 0 0 material group out nuclide mean std. dev.
1 5 1 total 0 0
0 5 2 total 0 0 material group in nuclide mean std. dev.
1 6 1 total 0 0
0 6 2 total 0 0 material group in nuclide mean std. dev.
1 7 1 total 0 0
0 7 2 total 0 0 material group in nuclide mean std. dev.
1 7 1 total 0 0
0 7 2 total 0 0 material group in group out nuclide mean std. dev.
3 7 1 1 total 0 0
2 7 1 2 total 0 0
1 7 2 1 total 0 0
0 7 2 2 total 0 0 material group out nuclide mean std. dev.
0 6 2 total 0 0 material group in nuclide mean std. dev.
1 6 1 total 0 0
0 6 2 total 0 0 material group in group out nuclide mean std. dev.
3 6 1 1 total 0 0
2 6 1 2 total 0 0
1 6 2 1 total 0 0
0 6 2 2 total 0 0 material group out nuclide mean std. dev.
1 6 1 total 0 0
0 6 2 total 0 0 material group in nuclide mean std. dev.
1 7 1 total 0 0
0 7 2 total 0 0 material group in nuclide mean std. dev.
1 8 1 total 0 0
0 8 2 total 0 0 material group in nuclide mean std. dev.
1 8 1 total 0 0
0 8 2 total 0 0 material group in group out nuclide mean std. dev.
3 8 1 1 total 0 0
2 8 1 2 total 0 0
1 8 2 1 total 0 0
0 8 2 2 total 0 0 material group out nuclide mean std. dev.
0 7 2 total 0 0 material group in nuclide mean std. dev.
1 7 1 total 0 0
0 7 2 total 0 0 material group in group out nuclide mean std. dev.
3 7 1 1 total 0 0
2 7 1 2 total 0 0
1 7 2 1 total 0 0
0 7 2 2 total 0 0 material group out nuclide mean std. dev.
1 7 1 total 0 0
0 7 2 total 0 0 material group in nuclide mean std. dev.
1 8 1 total 0 0
0 8 2 total 0 0 material group in nuclide mean std. dev.
1 9 1 total 0.504036 0.379624
0 9 2 total 1.687095 2.536622 material group in nuclide mean std. dev.
1 9 1 total 0 0
0 9 2 total 0 0 material group in group out nuclide mean std. dev.
3 9 1 1 total 0.504036 0.379624
2 9 1 2 total 0.000000 0.000000
1 9 2 1 total 0.000000 0.000000
0 9 2 2 total 1.417955 2.158027 material group out nuclide mean std. dev.
0 8 2 total 0 0 material group in nuclide mean std. dev.
1 8 1 total 0 0
0 8 2 total 0 0 material group in group out nuclide mean std. dev.
3 8 1 1 total 0 0
2 8 1 2 total 0 0
1 8 2 1 total 0 0
0 8 2 2 total 0 0 material group out nuclide mean std. dev.
1 8 1 total 0 0
0 8 2 total 0 0 material group in nuclide mean std. dev.
1 9 1 total 0.504036 0.379624
0 9 2 total 1.687095 2.536622 material group in nuclide mean std. dev.
1 9 1 total 0 0
0 9 2 total 0 0 material group in nuclide mean std. dev.
1 10 1 total 0 0
0 10 2 total 0 0 material group in nuclide mean std. dev.
1 10 1 total 0 0
0 10 2 total 0 0 material group in group out nuclide mean std. dev.
3 10 1 1 total 0 0
2 10 1 2 total 0 0
1 10 2 1 total 0 0
0 10 2 2 total 0 0 material group out nuclide mean std. dev.
0 9 2 total 0 0 material group in group out nuclide mean std. dev.
3 9 1 1 total 0.504036 0.379624
2 9 1 2 total 0.000000 0.000000
1 9 2 1 total 0.000000 0.000000
0 9 2 2 total 1.417955 2.158027 material group out nuclide mean std. dev.
1 9 1 total 0 0
0 9 2 total 0 0 material group in nuclide mean std. dev.
1 10 1 total 0 0
0 10 2 total 0 0 material group in nuclide mean std. dev.
1 11 1 total 0.302826 0.401311
0 11 2 total 1.006145 1.091638 material group in nuclide mean std. dev.
1 11 1 total 0 0
0 11 2 total 0 0 material group in group out nuclide mean std. dev.
3 11 1 1 total 0.275679 0.385676
2 11 1 2 total 0.027147 0.020009
1 11 2 1 total 0.000000 0.000000
0 11 2 2 total 0.957929 1.051959 material group out nuclide mean std. dev.
0 10 2 total 0 0 material group in nuclide mean std. dev.
1 10 1 total 0 0
0 10 2 total 0 0 material group in group out nuclide mean std. dev.
3 10 1 1 total 0 0
2 10 1 2 total 0 0
1 10 2 1 total 0 0
0 10 2 2 total 0 0 material group out nuclide mean std. dev.
1 10 1 total 0 0
0 10 2 total 0 0 material group in nuclide mean std. dev.
1 11 1 total 0.302826 0.401311
0 11 2 total 1.006145 1.091638 material group in nuclide mean std. dev.
1 11 1 total 0 0
0 11 2 total 0 0 material group in nuclide mean std. dev.
1 12 1 total 0.255933 0.268426
0 12 2 total 1.113345 0.988676 material group in nuclide mean std. dev.
1 12 1 total 0 0
0 12 2 total 0 0 material group in group out nuclide mean std. dev.
3 12 1 1 total 0.226310 0.254872
2 12 1 2 total 0.029622 0.017760
1 12 2 1 total 0.000000 0.000000
0 12 2 2 total 1.071690 0.958290 material group out nuclide mean std. dev.
0 11 2 total 0 0 material group in group out nuclide mean std. dev.
3 11 1 1 total 0.275679 0.385676
2 11 1 2 total 0.027147 0.020009
1 11 2 1 total 0.000000 0.000000
0 11 2 2 total 0.957929 1.051959 material group out nuclide mean std. dev.
1 11 1 total 0 0
0 11 2 total 0 0 material group in nuclide mean std. dev.
1 12 1 total 0.255933 0.268426
0 12 2 total 1.113345 0.988676 material group in nuclide mean std. dev.
1 12 1 total 0 0
0 12 2 total 0 0
0 12 2 total 0 0 material group in group out nuclide mean std. dev.
3 12 1 1 total 0.226310 0.254872
2 12 1 2 total 0.029622 0.017760
1 12 2 1 total 0.000000 0.000000
0 12 2 2 total 1.071690 0.958290 material group out nuclide mean std. dev.
1 12 1 total 0 0
0 12 2 total 0 0

3312
tests/test_mgxs_library_nuclides/results_true.dat
View file

File diff suppressed because it is too large Load diff

8
tests/test_score_delayed_nufission/results_true.dat
View file

@ -19,11 +19,11 @@ tally 2:
1.976462E-02
1.953328E-04
tally 3:
1.687894E-02
5.776176E-05
1.686299E-02
5.765477E-05
0.000000E+00
0.000000E+00
0.000000E+00
0.000000E+00
1.061803E-02
2.308636E-05
1.061240E-02
2.305936E-05

48
tests/test_score_inverse_velocity/results_true.dat
View file

@ -1,29 +1,29 @@
k-combined:
9.903196E-01 4.279617E-02
tally 1:
1.049628E-03
2.261930E-07
4.056389E-04
3.411247E-08
2.243766E-03
1.069671E-06
6.354432E-04
8.370608E-08
1.049628E-05
2.261930E-11
4.056389E-06
3.411247E-12
2.243766E-05
1.069671E-10
6.354432E-06
8.370608E-12
tally 2:
1.048031E-03
2.263789E-07
4.276093E-04
4.136358E-08
2.667795E-03
1.528407E-06
6.199140E-04
7.840402E-08
1.029200E-05
2.180706E-11
4.353200E-06
4.363747E-12
2.211790E-05
1.055892E-10
6.086777E-06
7.589579E-12
tally 3:
1.048031E-03
2.263789E-07
4.276093E-04
4.136358E-08
2.667795E-03
1.528407E-06
6.199140E-04
7.840402E-08
1.029200E-05
2.180706E-11
4.353200E-06
4.363747E-12
2.211790E-05
1.055892E-10
6.086777E-06
7.589579E-12