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Code Issues Projects Releases Packages Wiki Activity Actions 5

Changes per comments of @smharper and @paulromano

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This commit is contained in:
Adam Nelson 2016-01-16 13:48:06 -05:00
parent 08fdb8a91d
commit 473d3220dc
19 changed files with 221 additions and 207 deletions
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5
docs/source/usersguide/input.rst
View file

@ -546,7 +546,8 @@ attributes/sub-elements:
*Default*: 0.988 2.249
.. note:: The above format should be used even when using the multi-group :ref:`energy_mode`.
.. note:: The above format should be used even when using the multi-group
:ref:`energy_mode`.
:write_initial:
An element specifying whether to write out the initial source bank used at
@ -1517,6 +1518,8 @@ The ``<tally>`` element accepts the following sub-elements:
.. note:: This score type is not used in the multi-group :ref:`energy_mode`.
.. _kappa_fission:
:kappa-fission:
The recoverable energy production rate due to fission. The recoverable
energy is defined as the fission product kinetic energy, prompt and

2
docs/source/usersguide/install.rst
View file

@ -385,7 +385,7 @@ ACE data as described below. The TALYS-based evaluated nuclear data library,
TENDL_, is also openly available in ACE format.
In multi-group mode, OpenMC utilizes an XML-based library format which can be
used to describe nuclidic- or material-specific quantities.
used to describe nuclide- or material-specific quantities.
Using ENDF/B-VII.1 Cross Sections from NNDC
-------------------------------------------

73
docs/source/usersguide/mgxs_library.rst
View file

@ -87,18 +87,19 @@ attributes/sub-elements required to describe the meta-data:
*Default*: None, this must be provided.
:alias:
The number of total fission source iterations per batch.
An alternative name to use for the microscopic or macroscopic data set.
*Default*: If no alias is provided, it will adopt the value of ``name``.
:kT:
The temperature the data was generated at.
The temperature times Boltzmann's constant (in units of MeV) at which the
data was generated.
*Default*: Room temperature, 2.53E-8 MeV
:fissionable:
This element states whether or not the data in question is fissionable.
Accepted values are ``true`` or ``false``.
Accepted values are "true" or "false".
*Default*: None, this element must be provided.
@ -108,42 +109,42 @@ attributes/sub-elements required to describe the meta-data:
scalar flux weighting (or reduced to an equivalent representation)
and thus are angle-independent, or if the data was generated with angular
dependent fluxes and thus the data is angle-dependent. The options are
either ``isotropic`` or ``angle``.
either "isotropic" or "angle".
*Default*: ``isotropic``
*Default*: "isotropic"
:num_azimuthal:
This element provides the number of equi-width bins that the azimuthal
angular domain is subdivided in the case of angle-dependent cross sections
(i.e., ``angle`` is passed to the ``representation`` element).
(i.e., "angle" is passed to the ``representation`` element).
*Default*: If ``representation`` is ``angle``, this must be provided. If
*Default*: If ``representation`` is "angle", this must be provided. If
not, this parameter is not used.
:num_polar:
This element provides the number of equi-width bins that the polar angular
domain is subdivided in the case of angle-dependent cross sections
(i.e., ``angle`` is passed to the ``representation`` element).
(i.e., "angle" is passed to the ``representation`` element).
*Default*: If ``representation`` is ``angle``, this must be provided. If
*Default*: If ``representation`` is "angle", this must be provided. If
not, this parameter is not used.
:scatt_type:
This element provides the representation of the angular distribution
associated with each group-to-group transfer probability. The options are
either ``legendre``, ``histogram``, or ``tabular``.
The ``legendre`` option means the angular distribution has been
expanded via Legendre polynomials of the order provided in the ``order``
either "legendre", "histogram", or "tabular".
The "legendre" option means the angular distribution has been
expanded via Legendre polynomials of the order provided in the "order"
element.
The ``histogram`` option means the angular distribution is provided in
The "histogram" option means the angular distribution is provided in
an equi-width histogram format with a number of bins as provided in the
``order`` element. This is useful when the angular distribution was
"order" element. This is useful when the angular distribution was
obtained from a Monte Carlo tally and thus is natively in the histogram
format.
The ``tabular`` option means the angular distribution is provided in an
The "tabular" option means the angular distribution is provided in an
equi-spaced point-wise representation.
*Default*: ``legendre``
*Default*: "legendre"
:order:
This element provides either the Legendre order, number of bins, or number
@ -165,17 +166,17 @@ attributes/sub-elements required to describe the meta-data:
:enable:
This attribute/sub-element denotes whether or not the conversion to the
tabular format should be performed or not. A value of ``true`` means
the conversion should be performed, ``false`` means it should not.
tabular format should be performed or not. A value of "true" means
the conversion should be performed, "false" means it should not.
*Default*: ``true``
*Default*: "true"
:num_points:
If the conversion is to take place the number of tabular points is
required. This attribute/sub-element allows the user to set the desired
number of points.
*Default*: ``33``
*Default*: 33
The following attributes/sub-elements are the cross section values to
be used during the transport process.
@ -183,9 +184,9 @@ attributes/sub-elements required to describe the meta-data:
:total:
This element requires the group-wise total cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
``isotropic``, then the length of this list should equal the number of
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
``angle``, then the length of this list should equal the number of groups
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
@ -196,9 +197,9 @@ attributes/sub-elements required to describe the meta-data:
:absorption:
This element requires the group-wise absorption cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
``isotropic``, then the length of this list should equal the number of
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
``angle``, then the length of this list should equal the number of groups
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
@ -210,9 +211,9 @@ attributes/sub-elements required to describe the meta-data:
columns representing incoming group and rows representing the outgoing
group. That is, down-scatter will be above the diagonal of the resultant
matrix. This matrix is repeated for every Legendre order (in order of
increasing orders) if ``scatt_type`` is ``legendre``; otherwise, this
increasing orders) if ``scatt_type`` is "legendre"; otherwise, this
matrix is repeated for every bin of the histogram or tabular
representation. Finally, if ``representation`` is ``angle``, the above
representation. Finally, if ``representation`` is "angle", the above
is repeated for every azimuthal angle and every polar angle, in that
order.
@ -232,32 +233,32 @@ attributes/sub-elements required to describe the meta-data:
neglected).
The following fission-specific data are only needed should ``fissionable``
be ``true``.
be "true".
:fission:
This element requires the group-wise fission cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
``isotropic``, then the length of this list should equal the number of
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
``angle``, then the length of this list should equal the number of groups
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, this is required only if ``fission`` tallies are
*Default*: None, this is required only if fission tallies are
requested and the material is fissionable.
:k_fission:
:kappa_fission:
This element requires the group-wise kappa-fission cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
``isotropic``, then the length of this list should equal the number of
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
``angle``, then the length of this list should equal the number of groups
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, this is required only if ``kappa-fission`` tallies are
*Default*: None, this is required only if :ref:`kappa_fission` tallies are
requested and the material is fissionable.
:chi:
@ -267,9 +268,9 @@ attributes/sub-elements required to describe the meta-data:
not depend on incoming energy. If the user does not wish to make this
approximation, then this should not be provided and this information
included in the ``nu_fission`` element instead. If ``representation`` is
``isotropic``, then the length of this list should equal the number of
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
``angle``, then the length of this list should equal the number of groups
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.

5
docs/source/usersguide/output/source.rst
View file

@ -15,5 +15,6 @@ is that documented here.
**/source_bank** (Compound type)
Source bank information for each particle. The compound type has fields
``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight, position,
direction, and energy of the source particle, respectively.
``wgt``, ``xyz``, ``uvw``, ``E``, ``g``, and ``delayed_group``, which
represent the weight, position, direction, energy, energy group, and
delayed_group of the source particle, respectively.

11
docs/source/usersguide/output/statepoint.rst
View file

@ -39,6 +39,12 @@ The current revision of the statepoint file format is 14.
Pseudo-random number generator seed.
**/run_CE** (*int*)
Flag to denote continuous-energy or multi-group mode. A value of 1
indicates a continuous-energy run while a value of 0 indicates a
multi-group run.
**/run_mode** (*char[]*)
Run mode used. A value of 1 indicates a fixed-source run and a value of 2
@ -251,5 +257,6 @@ if (run_mode == 'k-eigenvalue' and source_present > 0)
**/source_bank** (Compound type)
Source bank information for each particle. The compound type has fields
``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight,
position, direction, and energy of the source particle, respectively.
``wgt``, ``xyz``, ``uvw``, ``E``, ``g``, and ``delayed_group``, which
represent the weight, position, direction, energy, energy group, and
delayed_group of the source particle, respectively.

17
examples/python/pincell_multigroup/build-xml.py
View file

@ -20,7 +20,7 @@ groups = openmc.mgxs.EnergyGroups(group_edges=[1E-11, 0.0635E-6, 10.0E-6,
1.0E-4, 1.0E-3, 0.5, 1.0, 20.0])
# Instantiate the 7-group (C5G7) cross section data
uo2_xsdata = openmc.XSdata('UO2.300k', groups)
uo2_xsdata = openmc.XSdata('UO2.300K', groups)
uo2_xsdata.order = 0
uo2_xsdata.total = np.array([0.1779492, 0.3298048, 0.4803882, 0.5543674,
0.3118013, 0.3951678, 0.5644058])
@ -43,7 +43,7 @@ uo2_xsdata.nu_fission = np.array([2.005998E-02, 2.027303E-03, 1.570599E-02,
uo2_xsdata.chi = np.array([5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07,
0.0000E+00, 0.0000E+00, 0.0000E+00])
h2o_xsdata = openmc.XSdata('LWTR.300k', groups)
h2o_xsdata = openmc.XSdata('LWTR.300K', groups)
h2o_xsdata.order = 0
h2o_xsdata.total = np.array([0.15920605, 0.412969593, 0.59030986, 0.58435,
0.718, 1.2544497, 2.650379])
@ -57,7 +57,7 @@ scatter = [[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0
[0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290],
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200],
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]]
h2o_xsdata.scatter = np.array(scatter[:][:])
h2o_xsdata.scatter = np.array(scatter)
mg_cross_sections_file = openmc.MGXSLibraryFile(groups)
mg_cross_sections_file.add_xsdatas([uo2_xsdata,h2o_xsdata])
@ -69,8 +69,8 @@ mg_cross_sections_file.export_to_xml()
###############################################################################
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('UO2', '300k')
h2o_data = openmc.Macroscopic('LWTR', '300k')
uo2_data = openmc.Macroscopic('UO2', '300K')
h2o_data = openmc.Macroscopic('LWTR', '300K')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel')
@ -83,7 +83,7 @@ water.add_macroscopic(h2o_data)
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '300k'
materials_file.default_xs = '300K'
materials_file.add_materials([uo2, water])
materials_file.export_to_xml()
@ -145,11 +145,6 @@ settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-0.63, -0.63, -1, \
0.63, 0.63, 1])
settings_file.entropy_lower_left = [-0.54, -0.54, -1.e50]
settings_file.entropy_upper_right = [0.54, 0.54, 1.e50]
settings_file.entropy_dimension = [10, 10, 1]
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File

4
examples/xml/pincell_multigroup/materials.xml
View file

@ -1,7 +1,7 @@
<?xml version="1.0"?>
<materials>
<!-- Set default xs set to 71c, which is ENDF-B/VII.0 at 600K -->
<default_xs>71c</default_xs>
<!-- Set default xs set to 300 Kelvin -->
<default_xs>300K</default_xs>
<!-- UO2 -->
<material id="1">

6
examples/xml/pincell_multigroup/settings.xml
View file

@ -27,12 +27,6 @@
</space>
</source>
<!-- Establish statepoints to aid in examining convergence -->
<state_point>
<interval>2000</interval>
<source_write>true</source_write>
</state_point>
<output>
<cross_sections>true</cross_sections>
<summary>true</summary>

11
openmc/material.py
View file

@ -50,8 +50,8 @@ class Material(object):
Density of the material (units defined separately)
density_units : str
Units used for `density`. Can be one of 'g/cm3', 'g/cc', 'kg/cm3',
'atom/b-cm', 'atom/cm3', 'sum', or 'macro' (the latter only applies
if in multi-group mode).
'atom/b-cm', 'atom/cm3', 'sum', or 'macro'. The 'macro' unit only
applies in the case of a multi-group calculation.
"""
@ -346,7 +346,7 @@ class Material(object):
'has already been added'.format(self._id, macroscopic)
raise ValueError(msg)
if not isinstance(macroscopic, (openmc.Macroscopic, str)):
if not isinstance(macroscopic, (openmc.Macroscopic, basestring)):
msg = 'Unable to add a Macroscopic to Material ID="{0}" with a ' \
'non-Macroscopic value "{1}"'.format(self._id, macroscopic)
raise ValueError(msg)
@ -511,7 +511,7 @@ class Material(object):
return xml_element
def _get_macroscopic_xml(self, macroscopic, distrib=False):
def _get_macroscopic_xml(self, macroscopic):
xml_element = ET.Element("macroscopic")
xml_element.set("name", macroscopic._name)
@ -626,8 +626,7 @@ class Material(object):
subelement.append(subsubelement)
else:
# Create macroscopic XML subelements
subsubelement = self._get_macroscopic_xml(self,
self._macroscopic,
subsubelement = self._get_macroscopic_xml(self._macroscopic,
distrib=True)
subelement.append(subsubelement)

2
openmc/mgxs/groups.py
View file

@ -54,7 +54,7 @@ class EnergyGroups(object):
def __eq__(self, other):
if not isinstance(other, EnergyGroups):
return False
elif self.group_edges != other.group_edges:
elif (self.group_edges != other.group_edges).all():
return False
else:
return True

6
openmc/mgxs_library.py
View file

@ -15,7 +15,7 @@ from openmc.checkvalue import check_type, check_value, check_greater_than, \
from openmc.clean_xml import *
# Supported incoming particle MGXS angular treatment representations
REPRESENTATIONS = ['isotropic', 'angle']
_REPRESENTATIONS = ['isotropic', 'angle']
def ndarray_to_string(arr):
"""Converts a numpy ndarray in to a join with spaces between entries
@ -87,7 +87,7 @@ class XSdata(object):
name : str, optional
Name of the mgxs data set.
representation : {'isotropic' or 'angle'}
representation : {'isotropic', 'angle'}
Method used in generating the MGXS (isotropic or angle-dependent flux
weighting). Defaults to 'isotropic'
@ -241,7 +241,7 @@ class XSdata(object):
@representation.setter
def representation(self, representation):
# Check it is of valid type.
check_value('representation', representation, REPRESENTATIONS)
check_value('representation', representation, _REPRESENTATIONS)
self._representation = representation
@alias.setter

10
openmc/settings.py
View file

@ -72,12 +72,10 @@ class SettingsFile(object):
environment variable will be used for continuous-energy calculations
and :envvar:`MG_CROSS_SECTIONS` will be used for multi-group
calculations to find the path to the XML cross section file.
energy_grid : str
Set the method used to search energy grids. Acceptable values are
'nuclide', 'logarithm', and 'material-union'.
energy_mode : str
energy_grid : {'nuclide', 'logarithm', 'material-union'}
Set the method used to search energy grids.
energy_mode : {'continuous-energy', 'multi-group'}
Set whether the calculation should be continuous-energy or multi-group.
Acceptable values are 'continuous-energy' or 'multi-group'
max_order : int
Maximum scattering order to apply globally when in multi-group mode.
ptables : bool
@ -495,7 +493,7 @@ class SettingsFile(object):
@max_order.setter
def max_order(self, max_order):
check_type('maximum scattering order', max_order, Integral)
check_greater_than('maximum scattering order', max_order, 0)
check_greater_than('maximum scattering order', max_order, 0, True)
self._max_order = max_order
@source_file.setter

4
openmc/summary.py
View file

@ -60,7 +60,7 @@ class Summary(object):
self.date_and_time = self._f['date_and_time'][...]
# Read if continuous-energy or multi-group
self.run_CE = bool(self._f['run_CE'].value)
self.run_CE = (self._f['run_CE'].value == 1)
self.n_batches = self._f['n_batches'].value
self.n_particles = self._f['n_particles'].value
@ -278,7 +278,7 @@ class Summary(object):
# Get the distribcell index
ind = self._f['geometry/cells'][key]['distribcell_index'].value
if ind != 0:
cell.distribcell_index = ind
cell.distribcell_index = ind
# Add the Cell to the global dictionary of all Cells
self.cells[index] = cell

2
src/cmfd_input.F90
View file

@ -108,7 +108,7 @@ contains
do i = 1, ng
found = .false.
do g = 1, energy_groups + 1
if (cmfd%egrid(i) == energy_bins(g)) then
if (cmfd % egrid(i) == energy_bins(g)) then
found = .true.
exit
end if

2
src/initialize.F90
View file

@ -16,7 +16,7 @@ module initialize
hdf5_tallyresult_t, hdf5_integer8_t
use input_xml, only: read_input_xml, cells_in_univ_dict, read_plots_xml
use material_header, only: Material
use mgxs_data
use mgxs_data, only: read_mgxs, same_nuclide_mg_list, create_macro_xs
use output, only: title, header, print_version, write_message, &
print_usage, write_xs_summary, print_plot
use random_lcg, only: initialize_prng

191
src/input_xml.F90
View file

@ -109,7 +109,7 @@ contains
temp_str = trim(to_lower(temp_str))
if (temp_str == "mg" .or. temp_str == "multi-group") then
run_CE = .false.
else if (temp_str == "ce" .or. temp_str == "continuous") then
else if (temp_str == "ce" .or. temp_str == "continuous-energy") then
run_CE = .true.
end if
end if
@ -406,7 +406,7 @@ contains
inquire(FILE=path_source, EXIST=file_exists)
if (.not. file_exists) then
call fatal_error("Binary source file '" // trim(path_source) &
&// "' does not exist!")
// "' does not exist!")
end if
else
@ -433,7 +433,7 @@ contains
coeffs_reqd = 3
case default
call fatal_error("Invalid spatial distribution for external source: "&
&// trim(type))
// trim(type))
end select
! Determine number of parameters specified
@ -481,7 +481,7 @@ contains
external_source % type_angle = SRC_ANGLE_TABULAR
case default
call fatal_error("Invalid angular distribution for external source: "&
&// trim(type))
// trim(type))
end select
! Determine number of parameters specified
@ -532,7 +532,7 @@ contains
external_source % type_energy = SRC_ENERGY_TABULAR
case default
call fatal_error("Invalid energy distribution for external source: " &
&// trim(type))
// trim(type))
end select
! Determine number of parameters specified
@ -926,7 +926,7 @@ contains
! check to make sure a nuclide is specified
if (.not. check_for_node(node_scatterer, "nuclide")) then
call fatal_error("No nuclide specified for scatterer " &
&// trim(to_str(i)) // " in settings.xml file!")
// trim(to_str(i)) // " in settings.xml file!")
end if
call get_node_value(node_scatterer, "nuclide", &
nuclides_0K(i) % nuclide)
@ -940,7 +940,7 @@ contains
if (.not. check_for_node(node_scatterer, "xs_label")) then
call fatal_error("Must specify the temperature dependent name of &
&scatterer " // trim(to_str(i)) &
&// " given in cross_sections.xml")
// " given in cross_sections.xml")
end if
call get_node_value(node_scatterer, "xs_label", &
nuclides_0K(i) % name)
@ -948,7 +948,7 @@ contains
! check to make sure 0K xs name for which method is applied is given
if (.not. check_for_node(node_scatterer, "xs_label_0K")) then
call fatal_error("Must specify the 0K name of scatterer " &
&// trim(to_str(i)) // " given in cross_sections.xml")
// trim(to_str(i)) // " given in cross_sections.xml")
end if
call get_node_value(node_scatterer, "xs_label_0K", &
nuclides_0K(i) % name_0K)
@ -1008,7 +1008,7 @@ contains
default_expand = JENDL_40
case default
call fatal_error("Unknown natural element expansion option: " &
&// trim(temp_str))
// trim(temp_str))
end select
end if
@ -1125,7 +1125,7 @@ contains
! Check to make sure 'id' hasn't been used
if (cell_dict % has_key(c % id)) then
call fatal_error("Two or more cells use the same unique ID: " &
&// to_str(c % id))
// to_str(c % id))
end if
! Read material
@ -1146,14 +1146,14 @@ contains
! Check for error
if (c % material == ERROR_INT) then
call fatal_error("Invalid material specified on cell " &
&// to_str(c % id))
// to_str(c % id))
end if
end select
! Check to make sure that either material or fill was specified
if (c % material == NONE .and. c % fill == NONE) then
call fatal_error("Neither material nor fill was specified for cell " &
&// trim(to_str(c % id)))
// trim(to_str(c % id)))
end if
! Check to make sure that both material and fill haven't been
@ -1213,7 +1213,7 @@ contains
n = get_arraysize_double(node_cell, "rotation")
if (n /= 3) then
call fatal_error("Incorrect number of rotation parameters on cell " &
&// to_str(c % id))
// to_str(c % id))
end if
! Copy rotation angles in x,y,z directions
@ -1241,7 +1241,7 @@ contains
! another universe
if (c % fill == NONE) then
call fatal_error("Cannot apply a translation to cell " &
&// trim(to_str(c % id)) // " because it is not filled with &
// trim(to_str(c % id)) // " because it is not filled with &
&another universe")
end if
@ -1357,7 +1357,7 @@ contains
! Check to make sure 'id' hasn't been used
if (surface_dict % has_key(s%id)) then
call fatal_error("Two or more surfaces use the same unique ID: " &
&// to_str(s%id))
// to_str(s%id))
end if
! Copy surface name
@ -1372,10 +1372,10 @@ contains
n = get_arraysize_double(node_surf, "coeffs")
if (n < coeffs_reqd) then
call fatal_error("Not enough coefficients specified for surface: " &
&// trim(to_str(s%id)))
// trim(to_str(s%id)))
elseif (n > coeffs_reqd) then
call fatal_error("Too many coefficients specified for surface: " &
&// trim(to_str(s%id)))
// trim(to_str(s%id)))
end if
allocate(coeffs(n))
@ -1501,7 +1501,7 @@ contains
! Check to make sure 'id' hasn't been used
if (lattice_dict % has_key(lat % id)) then
call fatal_error("Two or more lattices use the same unique ID: " &
&// to_str(lat % id))
// to_str(lat % id))
end if
! Copy lattice name
@ -1629,7 +1629,7 @@ contains
! Check to make sure 'id' hasn't been used
if (lattice_dict % has_key(lat % id)) then
call fatal_error("Two or more lattices use the same unique ID: " &
&// to_str(lat % id))
// to_str(lat % id))
end if
! Copy lattice name
@ -1883,7 +1883,7 @@ contains
! Check to make sure 'id' hasn't been used
if (material_dict % has_key(mat % id)) then
call fatal_error("Two or more materials use the same unique ID: " &
&// to_str(mat % id))
// to_str(mat % id))
end if
! Copy material name
@ -1906,7 +1906,7 @@ contains
call get_node_ptr(node_mat, "density", node_dens)
else
call fatal_error("Must specify density element in material " &
&// trim(to_str(mat % id)))
// trim(to_str(mat % id)))
end if
! Initialize value to zero
@ -1943,7 +1943,7 @@ contains
sum_density = .false.
if (val <= ZERO) then
call fatal_error("Need to specify a positive density on material " &
&// trim(to_str(mat % id)) // ".")
// trim(to_str(mat % id)) // ".")
end if
! Adjust material density based on specified units
@ -1958,7 +1958,7 @@ contains
mat % density = 1.0e-24_8 * val
case default
call fatal_error("Unkwown units '" // trim(units) &
&// "' specified on material " // trim(to_str(mat % id)))
// "' specified on material " // trim(to_str(mat % id)))
end select
end if
@ -1981,7 +1981,7 @@ contains
call get_node_list(node_mat, "macroscopic", node_macro_list)
if (get_list_size(node_macro_list) > 1) then
call fatal_error("Only one macroscopic object permitted per material, " &
&// trim(to_str(mat % id)))
// trim(to_str(mat % id)))
else if (get_list_size(node_macro_list) == 1) then
call get_list_item(node_macro_list, 1, node_nuc)
@ -1989,7 +1989,7 @@ contains
! Check for empty name on nuclide
if (.not.check_for_node(node_nuc, "name")) then
call fatal_error("No name specified on macroscopic data in material " &
&// trim(to_str(mat % id)))
// trim(to_str(mat % id)))
end if
! Check for cross section
@ -2033,7 +2033,7 @@ contains
call list_density % append(ONE)
else
call fatal_error("Units can only be macro for macroscopic data " &
&// trim(name))
// trim(name))
end if
else
@ -2048,7 +2048,7 @@ contains
! Check for empty name on nuclide
if (.not.check_for_node(node_nuc, "name")) then
call fatal_error("No name specified on nuclide in material " &
&// trim(to_str(mat % id)))
// trim(to_str(mat % id)))
end if
! Check for cross section
@ -2079,7 +2079,7 @@ contains
if (.not.check_for_node(node_nuc, "ao") .and. &
.not.check_for_node(node_nuc, "wo")) then
call fatal_error("No atom or weight percent specified for nuclide " &
&// trim(name))
// trim(name))
elseif (check_for_node(node_nuc, "ao") .and. &
check_for_node(node_nuc, "wo")) then
call fatal_error("Cannot specify both atom and weight percents for a &
@ -2110,7 +2110,7 @@ contains
! Check for empty name on natural element
if (.not.check_for_node(node_ele, "name")) then
call fatal_error("No name specified on nuclide in material " &
&// trim(to_str(mat % id)))
// trim(to_str(mat % id)))
end if
call get_node_value(node_ele, "name", name)
@ -2131,7 +2131,7 @@ contains
if (.not.check_for_node(node_ele, "ao") .and. &
.not.check_for_node(node_ele, "wo")) then
call fatal_error("No atom or weight percent specified for element " &
&// trim(name))
// trim(name))
elseif (check_for_node(node_ele, "ao") .and. &
check_for_node(node_ele, "wo")) then
call fatal_error("Cannot specify both atom and weight percents for &
@ -2191,7 +2191,7 @@ contains
name = trim(list_names % get_item(j))
if (.not. xs_listing_dict % has_key(to_lower(name))) then
call fatal_error("Could not find nuclide " // trim(name) &
&// " in cross_sections data file!")
// " in cross_sections data file!")
end if
if (run_CE) then
@ -2199,7 +2199,7 @@ contains
n = len_trim(name)
if (name(n:n) /= 'c') then
call fatal_error("Cross-section table " // trim(name) &
&// " is not a continuous-energy neutron table.")
// " is not a continuous-energy neutron table.")
end if
end if
@ -2238,7 +2238,7 @@ contains
if (.not. (all(mat % atom_density >= ZERO) .or. &
all(mat % atom_density <= ZERO))) then
call fatal_error("Cannot mix atom and weight percents in material " &
&// to_str(mat % id))
// to_str(mat % id))
end if
! Determine density if it is a sum value
@ -2286,7 +2286,7 @@ contains
! Check that this nuclide is listed in the cross_sections.xml file
if (.not. xs_listing_dict % has_key(to_lower(name))) then
call fatal_error("Could not find S(a,b) table " // trim(name) &
&// " in cross_sections.xml file!")
// " in cross_sections.xml file!")
end if
! Find index in xs_listing and set the name and alias according to the
@ -2448,7 +2448,7 @@ contains
! Check to make sure 'id' hasn't been used
if (mesh_dict % has_key(m % id)) then
call fatal_error("Two or more meshes use the same unique ID: " &
&// to_str(m % id))
// to_str(m % id))
end if
! Read mesh type
@ -2589,7 +2589,7 @@ contains
! Check to make sure 'id' hasn't been used
if (tally_dict % has_key(t % id)) then
call fatal_error("Two or more tallies use the same unique ID: " &
&// to_str(t % id))
// to_str(t % id))
end if
! Copy tally name
@ -2630,7 +2630,7 @@ contains
end if
else
call fatal_error("Bins not set in filter on tally " &
&// trim(to_str(t % id)))
// trim(to_str(t % id)))
end if
! Determine type of filter
@ -2722,7 +2722,7 @@ contains
m => meshes(i_mesh)
else
call fatal_error("Could not find mesh " // trim(to_str(id)) &
&// " specified on tally " // trim(to_str(t % id)))
// " specified on tally " // trim(to_str(t % id)))
end if
! Determine number of bins -- this is assuming that the tally is
@ -2906,8 +2906,8 @@ contains
case default
! Specified tally filter is invalid, raise error
call fatal_error("Unknown filter type '" &
&// trim(temp_str) // "' on tally " &
&// trim(to_str(t % id)) // ".")
// trim(temp_str) // "' on tally " &
// trim(to_str(t % id)) // ".")
end select
@ -3003,8 +3003,8 @@ contains
! Check if no nuclide was found
if (.not. associated(pair_list)) then
call fatal_error("Could not find the nuclide " &
&// trim(word) // " specified in tally " &
&// trim(to_str(t % id)) // " in any material.")
// trim(word) // " specified in tally " &
// trim(to_str(t % id)) // " in any material.")
end if
deallocate(pair_list)
@ -3072,12 +3072,12 @@ contains
! maximum order.
! The above scheme will essentially take the absolute value
if (master) call warning("Invalid scattering order of " &
&// trim(to_str(n_order)) // " requested. Setting to the &
// trim(to_str(n_order)) // " requested. Setting to the &
&maximum permissible value, " &
&// trim(to_str(MAX_ANG_ORDER)))
// trim(to_str(MAX_ANG_ORDER)))
n_order = MAX_ANG_ORDER
sarray(j) = trim(MOMENT_STRS(imomstr)) &
&// trim(to_str(MAX_ANG_ORDER))
// trim(to_str(MAX_ANG_ORDER))
end if
! Find total number of bins for this case
if (imomstr >= YN_LOC) then
@ -3139,9 +3139,9 @@ contains
! maximum order.
! The above scheme will essentially take the absolute value
if (master) call warning("Invalid scattering order of " &
&// trim(to_str(n_order)) // " requested. Setting to &
// trim(to_str(n_order)) // " requested. Setting to &
&the maximum permissible value, " &
&// trim(to_str(MAX_ANG_ORDER)))
// trim(to_str(MAX_ANG_ORDER)))
n_order = MAX_ANG_ORDER
end if
score_name = trim(MOMENT_N_STRS(imomstr)) // "n"
@ -3293,9 +3293,21 @@ contains
case ('n2n', '(n,2n)')
t % score_bins(j) = N_2N
! Disallow for MG mode since data not present
if (.not. run_CE) then
call fatal_error("Cannot tally (n,2n) reaction rate in &
&multi-group mode")
end if
case ('n3n', '(n,3n)')
t % score_bins(j) = N_3N
! Disallow for MG mode since data not present
if (.not. run_CE) then
call fatal_error("Cannot tally (n,3n) reaction rate in &
&multi-group mode")
end if
case ('n4n', '(n,4n)')
t % score_bins(j) = N_4N
@ -3480,13 +3492,13 @@ contains
t % score_bins(j) = MT
else
call fatal_error("Invalid MT on <scores>: " &
&// trim(sarray(l)))
// trim(sarray(l)))
end if
else
! Specified score was not an integer
call fatal_error("Unknown scoring function: " &
&// trim(sarray(l)))
// trim(sarray(l)))
end if
end select
@ -3507,7 +3519,7 @@ contains
deallocate(sarray)
else
call fatal_error("No <scores> specified on tally " &
&// trim(to_str(t % id)) // ".")
// trim(to_str(t % id)) // ".")
end if
! If settings.xml trigger is turned on, create tally triggers
@ -3768,7 +3780,7 @@ contains
inquire(FILE=filename, EXIST=file_exists)
if (.not. file_exists) then
call fatal_error("Plots XML file '" // trim(filename) &
&// "' does not exist!")
// "' does not exist!")
end if
! Display output message
@ -3800,7 +3812,7 @@ contains
! Check to make sure 'id' hasn't been used
if (plot_dict % has_key(pl % id)) then
call fatal_error("Two or more plots use the same unique ID: " &
&// to_str(pl % id))
// to_str(pl % id))
end if
! Copy plot type
@ -3815,7 +3827,7 @@ contains
pl % type = PLOT_TYPE_VOXEL
case default
call fatal_error("Unsupported plot type '" // trim(temp_str) &
&// "' in plot " // trim(to_str(pl % id)))
// "' in plot " // trim(to_str(pl % id)))
end select
! Set output file path
@ -3835,14 +3847,14 @@ contains
call get_node_array(node_plot, "pixels", pl % pixels(1:2))
else
call fatal_error("<pixels> must be length 2 in slice plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
else if (pl % type == PLOT_TYPE_VOXEL) then
if (get_arraysize_integer(node_plot, "pixels") == 3) then
call get_node_array(node_plot, "pixels", pl % pixels(1:3))
else
call fatal_error("<pixels> must be length 3 in voxel plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
end if
@ -3850,13 +3862,13 @@ contains
if (check_for_node(node_plot, "background")) then
if (pl % type == PLOT_TYPE_VOXEL) then
if (master) call warning("Background color ignored in voxel plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
if (get_arraysize_integer(node_plot, "background") == 3) then
call get_node_array(node_plot, "background", pl % not_found % rgb)
else
call fatal_error("Bad background RGB in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
else
pl % not_found % rgb = (/ 255, 255, 255 /)
@ -3877,7 +3889,7 @@ contains
pl % basis = PLOT_BASIS_YZ
case default
call fatal_error("Unsupported plot basis '" // trim(temp_str) &
&// "' in plot " // trim(to_str(pl % id)))
// "' in plot " // trim(to_str(pl % id)))
end select
end if
@ -3886,7 +3898,7 @@ contains
call get_node_array(node_plot, "origin", pl % origin)
else
call fatal_error("Origin must be length 3 in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
! Copy plotting width
@ -3895,14 +3907,14 @@ contains
call get_node_array(node_plot, "width", pl % width(1:2))
else
call fatal_error("<width> must be length 2 in slice plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
else if (pl % type == PLOT_TYPE_VOXEL) then
if (get_arraysize_double(node_plot, "width") == 3) then
call get_node_array(node_plot, "width", pl % width(1:3))
else
call fatal_error("<width> must be length 3 in voxel plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
end if
@ -3912,7 +3924,7 @@ contains
if (pl % level < 0) then
call fatal_error("Bad universe level in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
else
pl % level = PLOT_LEVEL_LOWEST
@ -3946,7 +3958,7 @@ contains
case default
call fatal_error("Unsupported plot color type '" // trim(temp_str) &
&// "' in plot " // trim(to_str(pl % id)))
// "' in plot " // trim(to_str(pl % id)))
end select
! Get the number of <col_spec> nodes and get a list of them
@ -3969,7 +3981,7 @@ contains
! Check and make sure 3 values are specified for RGB
if (get_arraysize_double(node_col, "rgb") /= 3) then
call fatal_error("Bad RGB in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
! Ensure that there is an id for this color specification
@ -3988,7 +4000,7 @@ contains
call get_node_array(node_col, "rgb", pl % colors(col_id) % rgb)
else
call fatal_error("Could not find cell " // trim(to_str(col_id)) &
&// " specified in plot " // trim(to_str(pl % id)))
// " specified in plot " // trim(to_str(pl % id)))
end if
else if (pl % color_by == PLOT_COLOR_MATS) then
@ -3998,8 +4010,8 @@ contains
call get_node_array(node_col, "rgb", pl % colors(col_id) % rgb)
else
call fatal_error("Could not find material " &
&// trim(to_str(col_id)) // " specified in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(col_id)) // " specified in plot " &
// trim(to_str(pl % id)))
end if
end if
@ -4013,7 +4025,7 @@ contains
if (pl % type == PLOT_TYPE_VOXEL) then
call warning("Meshlines ignored in voxel plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
select case(n_meshlines)
@ -4047,7 +4059,7 @@ contains
! Check and make sure 3 values are specified for RGB
if (get_arraysize_double(node_meshlines, "color") /= 3) then
call fatal_error("Bad RGB for meshlines color in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
call get_node_array(node_meshlines, "color", &
@ -4064,7 +4076,7 @@ contains
if (.not. associated(ufs_mesh)) then
call fatal_error("No UFS mesh for meshlines on plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
pl % meshlines_mesh => ufs_mesh
@ -4085,7 +4097,7 @@ contains
if (.not. associated(entropy_mesh)) then
call fatal_error("No entropy mesh for meshlines on plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
if (.not. allocated(entropy_mesh % dimension)) then
@ -4114,18 +4126,18 @@ contains
end if
else
call fatal_error("Could not find mesh " &
&// trim(to_str(meshid)) // " specified in meshlines for &
// trim(to_str(meshid)) // " specified in meshlines for &
&plot " // trim(to_str(pl % id)))
end if
case default
call fatal_error("Invalid type for meshlines on plot " &
&// trim(to_str(pl % id)) // ": " // trim(meshtype))
// trim(to_str(pl % id)) // ": " // trim(meshtype))
end select
case default
call fatal_error("Mutliple meshlines specified in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end select
end if
@ -4137,13 +4149,13 @@ contains
if (pl % type == PLOT_TYPE_VOXEL) then
if (master) call warning("Mask ignored in voxel plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
select case(n_masks)
case default
call fatal_error("Mutliple masks specified in plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
case (1)
! Get pointer to mask
@ -4154,7 +4166,7 @@ contains
n_comp = get_arraysize_integer(node_mask, "components")
if (n_comp == 0) then
call fatal_error("Missing <components> in mask of plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
allocate(iarray(n_comp))
call get_node_array(node_mask, "components", iarray)
@ -4170,7 +4182,7 @@ contains
iarray(j) = cell_dict % get_key(col_id)
else
call fatal_error("Could not find cell " &
&// trim(to_str(col_id)) // " specified in the mask in &
// trim(to_str(col_id)) // " specified in the mask in &
&plot " // trim(to_str(pl % id)))
end if
@ -4180,7 +4192,7 @@ contains
iarray(j) = material_dict % get_key(col_id)
else
call fatal_error("Could not find material " &
&// trim(to_str(col_id)) // " specified in the mask in &
// trim(to_str(col_id)) // " specified in the mask in &
&plot " // trim(to_str(pl % id)))
end if
@ -4194,7 +4206,7 @@ contains
call get_node_array(node_mask, "background", pl % colors(j) % rgb)
else
call fatal_error("Missing <background> in mask of plot " &
&// trim(to_str(pl % id)))
// trim(to_str(pl % id)))
end if
end if
end do
@ -4239,7 +4251,7 @@ contains
if (.not. file_exists) then
! Could not find cross_sections.xml file
call fatal_error("Cross sections XML file '" &
&// trim(path_cross_sections) // "' does not exist!")
// trim(path_cross_sections) // "' does not exist!")
end if
call write_message("Reading cross sections XML file...", 5)
@ -4269,7 +4281,7 @@ contains
filetype = ASCII
else
call fatal_error("Unknown filetype in cross_sections.xml: " &
&// trim(temp_str))
// trim(temp_str))
end if
! copy default record length and entries for binary files
@ -4367,8 +4379,8 @@ contains
do i = 1, n_res_scatterers_total
if (.not. xs_listing_dict % has_key(trim(nuclides_0K(i) % name_0K))) then
call fatal_error("Could not find nuclide " &
&// trim(nuclides_0K(i) % name_0K) &
&// " in cross_sections.xml file!")
// trim(nuclides_0K(i) % name_0K) &
// " in cross_sections.xml file!")
end if
end do
@ -4385,14 +4397,13 @@ contains
type(Node), pointer :: doc => null()
type(Node), pointer :: node_xsdata => null()
type(NodeList), pointer :: node_xsdata_list => null()
! character(MAX_LINE_LEN) :: temp_str
! Check if cross_sections.xml exists
inquire(FILE=path_cross_sections, EXIST=file_exists)
if (.not. file_exists) then
! Could not find cross_sections.xml file
call fatal_error("Cross sections XML file '" &
&// trim(path_cross_sections) // "' does not exist!")
// trim(path_cross_sections) // "' does not exist!")
end if
call write_message("Reading cross sections XML file...", 5)
@ -4417,7 +4428,7 @@ contains
allocate(energy_bin_avg(energy_groups))
do i = 1, energy_groups
energy_bin_avg(i) = 0.5_8 * (energy_bins(i) + energy_bins(i + 1))
energy_bin_avg(i) = HALF * (energy_bins(i) + energy_bins(i + 1))
end do
allocate(inverse_velocities(energy_groups))
@ -4440,8 +4451,8 @@ contains
! Allocate xs_listings array
if (n_listings == 0) then
call fatal_error("No XSDATA listings present in cross_sections.xml &
&file!")
call fatal_error("At least one <xsdata> element must be present in &
&cross_sections.xml file!")
else
allocate(xs_listings(n_listings))
end if
@ -4474,7 +4485,7 @@ contains
if (check_for_node(node_xsdata, "kT")) then
call get_node_value(node_xsdata, "kT", listing % kT)
else
listing % kT = 2.53E-8_8
listing % kT = 293.6_8 * K_BOLTZMANN
end if
! determine type of cross section

18
src/mgxs_data.F90
View file

@ -370,12 +370,12 @@ contains
end if
if (get_kfiss) then
allocate(this % k_fission(groups))
if (check_for_node(node_xsdata, "k_fission")) then
call get_node_array(node_xsdata, "k_fission", this % k_fission)
if (check_for_node(node_xsdata, "kappa_fission")) then
call get_node_array(node_xsdata, "kappa_fission", this % k_fission)
else
error_code = 1
error_text = "k_fission data missing, required due to kappa-fission&
& tallies in tallies.xml file!"
error_text = "kappa_fission data missing, required due to &
&kappa-fission tallies in tallies.xml file!"
return
end if
end if
@ -554,22 +554,22 @@ contains
deallocate(temp_arr)
else
error_code = 1
error_text = "Fission data missing, required due to kappa-fission&
error_text = "Fission data missing, required due to fission&
& tallies in tallies.xml file!"
return
end if
end if
if (get_kfiss) then
if (check_for_node(node_xsdata, "k_fission")) then
if (check_for_node(node_xsdata, "kappa_fission")) then
allocate(temp_arr(groups * this % Nazi * this % Npol))
call get_node_array(node_xsdata, "k_fission", temp_arr)
call get_node_array(node_xsdata, "kappa_fission", temp_arr)
allocate(this % k_fission(groups, this % Nazi, this % Npol))
this % k_fission = reshape(temp_arr, (/groups, this % Nazi, this % Npol/))
deallocate(temp_arr)
else
error_code = 1
error_text = "k_fission data missing, required due to kappa-fission&
& tallies in tallies.xml file!"
error_text = "kappa_fission data missing, required due to &
&kappa-fission tallies in tallies.xml file!"
return
end if
end if

14
src/state_point.F90
View file

@ -727,14 +727,12 @@ contains
! It is not impossible for a state point to be generated from a CE run but
! to be loaded in to an MG run (or vice versa), check to prevent that.
call read_dataset(file_id, "run_CE", sp_run_CE)
if (sp_run_CE == 0) then
if (run_CE) &
call fatal_error("State point file is from multi-group run but &
& current run is continous-energy!")
else if (sp_run_CE == 1) then
if (.not. run_CE) &
call fatal_error("State point file is from continuous-energy run but &
& current run is multi-group!")
if (sp_run_CE == 0 .and. run_CE) then
call fatal_error("State point file is from multi-group run but &
& current run is continous-energy!")
else if (sp_run_CE == 1 .and. .not. run_CE) then
call fatal_error("State point file is from continuous-energy run but &
& current run is multi-group!")
end if
! Read and overwrite run information except number of batches

45
src/tally.F90
View file

@ -778,7 +778,6 @@ contains
end if
end if
end select
!#########################################################################
@ -807,6 +806,14 @@ contains
real(8) :: macro_total ! material macro total xs
real(8) :: macro_scatt ! material macro scatt xs
real(8) :: micro_abs ! nuclidic microscopic abs
real(8) :: p_uvw(3) ! Particle's current uvw
! Set the direction, if needed for nuclidic data, so that nuc % get_xs
! knows wihch direction it should be using for direction-dependent
! mgxs
if (i_nuclide > 0) then
p_uvw = p % coord(p % n_coord) % uvw
end if
i = 0
SCORE_LOOP: do q = 1, t % n_user_score_bins
@ -860,7 +867,7 @@ contains
else
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'total', UVW=p % coord(i) % uvw) * &
score = nuc % get_xs(p % g, 'total', UVW=p_uvw) * &
atom_density * flux
end associate
else
@ -902,7 +909,7 @@ contains
! Note SCORE_SCATTER_N not available for tracklength/collision.
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'scatter', UVW=p % coord(i) % uvw) * &
score = nuc % get_xs(p % g, 'scatter', UVW=p_uvw) * &
atom_density * flux
end associate
else
@ -1069,7 +1076,7 @@ contains
else
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw) &
score = nuc % get_xs(p % g, 'absorption', UVW=p_uvw) &
* atom_density * flux
end associate
else
@ -1085,10 +1092,10 @@ contains
! calculate fraction of absorptions that would have resulted in
! fission
associate (nuc => nuclides_MG(i_nuclide) % obj)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p_uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) &
nuc % get_xs(p % g, 'fission', UVW=p_uvw) &
/ micro_abs
else
score = ZERO
@ -1102,20 +1109,20 @@ contains
! fission reaction rate
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = p % last_wgt &
* nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) &
/ nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
* nuc % get_xs(p % g, 'fission', UVW=p_uvw) &
/ nuc % get_xs(p % g, 'absorption', UVW=p_uvw)
end associate
end if
else
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) * &
score = nuc % get_xs(p % g, 'fission', UVW=p_uvw) * &
atom_density * flux
end associate
else
score = flux * macro_xs(p % material) % obj % get_xs(p % g, &
'fission', UVW=p % coord(i) % uvw)
'fission', UVW=p_uvw)
end if
end if
@ -1139,10 +1146,10 @@ contains
! calculate fraction of absorptions that would have resulted in
! nu-fission
associate (nuc => nuclides_MG(i_nuclide) % obj)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p_uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) / &
nuc % get_xs(p % g, 'fission', UVW=p_uvw) / &
micro_abs
else
score = ZERO
@ -1162,7 +1169,7 @@ contains
else
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'nu_fission', UVW=p % coord(i) % uvw) &
score = nuc % get_xs(p % g, 'nu_fission', UVW=p_uvw) &
* atom_density * flux
end associate
else
@ -1180,10 +1187,10 @@ contains
! calculate fraction of absorptions that would have resulted in
! fission scale by kappa-fission
associate (nuc => nuclides_MG(i_nuclide) % obj)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p_uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) / &
nuc % get_xs(p % g, 'k_fission', UVW=p_uvw) / &
micro_abs
end if
end associate
@ -1195,20 +1202,20 @@ contains
! the fission energy production rate
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = p % last_wgt * &
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) / &
nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
nuc % get_xs(p % g, 'k_fission', UVW=p_uvw) / &
nuc % get_xs(p % g, 'absorption', UVW=p_uvw)
end associate
end if
else
if (i_nuclide > 0) then
associate (nuc => nuclides_MG(i_nuclide) % obj)
score = nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) &
score = nuc % get_xs(p % g, 'k_fission', UVW=p_uvw) &
* atom_density * flux
end associate
else
score = flux * macro_xs(p % material) % obj % get_xs(p % g, &
'k_fission', UVW=p % coord(i) % uvw)
'k_fission', UVW=p_uvw)
end if
end if