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updated documentation and comments in Python API

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Sam Shaner 2016-10-24 09:12:57 -04:00
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117
docs/source/usersguide/input.rst
View file

@ -105,19 +105,6 @@ standard deviation.
*Default*: false
.. _cross_sections:
``<cross_sections>`` Element
----------------------------
The ``<cross_sections>`` element has no attributes and simply indicates the path
to an XML cross section listing file (usually named cross_sections.xml). If this
element is absent from the settings.xml file, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used to find the
path to the XML cross section listing when in continuous-energy mode, and the
:envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable will be used in
multi-group mode.
``<cutoff>`` Element
--------------------
@ -289,20 +276,6 @@ based on the recommended value in LA-UR-14-24530_.
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
.. _multipole_library:
``<multipole_library>`` Element
-------------------------------
The ``<multipole_library>`` element indicates the directory containing a
windowed multipole library. If a windowed multipole library is available,
OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
cross sections. If this element is absent from the settings.xml file, the
:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
.. note:: The <temperature_multipole> element must also be set to "true" for
windowed multipole functionality.
``<max_order>`` Element
---------------------------
@ -316,29 +289,6 @@ then, OpenMC will only use up to the :math:`P_1` data.
.. note:: This element is not used in the continuous-energy
:ref:`energy_mode`.
.. _natural_elements:
``<natural_elements>`` Element
------------------------------
The ``<natural_elements>`` element indicates to OpenMC what nuclides are
available in the cross section library when expanding an ``<element>`` into
separate isotopes (see :ref:`material`). The accepted values are:
- ENDF/B-VII.0
- ENDF/B-VII.1
- JEFF-3.1.1
- JEFF-3.1.2
- JEFF-3.2
- JENDL-3.2
- JENDL-3.3
- JENDL-4.0
Note that the value is case-insensitive, so "ENDF/B-VII.1" is equivalent to
"endf/b-vii.1".
*Default*: ENDF/B-VII.1
``<no_reduce>`` Element
-----------------------
@ -1368,6 +1318,33 @@ Here is an example of a properly defined 2d hexagonal lattice:
Materials Specification -- materials.xml
----------------------------------------
.. _cross_sections:
``<cross_sections>`` Element
----------------------------
The ``<cross_sections>`` element has no attributes and simply indicates the path
to an XML cross section listing file (usually named cross_sections.xml). If this
element is absent from the settings.xml file, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used to find the
path to the XML cross section listing when in continuous-energy mode, and the
:envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable will be used in
multi-group mode.
.. _multipole_library:
``<multipole_library>`` Element
-------------------------------
The ``<multipole_library>`` element indicates the directory containing a
windowed multipole library. If a windowed multipole library is available,
OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
cross sections. If this element is absent from the settings.xml file, the
:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
.. note:: The <temperature_multipole> element must also be set to "true" for
windowed multipole functionality.
.. _material:
``<material>`` Element
@ -1440,43 +1417,6 @@ Each ``material`` element can have the following attributes or sub-elements:
.. note:: The ``scattering`` attribute/sub-element is not used in the
multi-group :ref:`energy_mode`.
:element:
Specifies that a natural element is present in the material. The natural
element is split up into individual isotopes based on `IUPAC Isotopic
Compositions of the Elements 2009`_. This element has
attributes/sub-elements called ``name``, and ``ao``. The ``name``
attribute is the atomic symbol of the element. Finally, the ``ao``
attribute specifies the atom percent of the element within the material,
respectively. One example would be as follows:
.. code-block:: xml
<element name="Al" ao="8.7115e-03" />
<element name="Mg" ao="1.5498e-04" />
<element name="Mn" ao="2.7426e-05" />
<element name="Cu" ao="1.6993e-04" />
In some cross section libraries, certain naturally occurring isotopes do not
have cross sections. The :ref:`natural_elements` option determines how a
natural element is split into isotopes in these cases.
*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
.. note:: The ``scattering`` attribute/sub-element is not used in the
multi-group :ref:`energy_mode`.
:sab:
Associates an S(a,b) table with the material. This element has one
attribute/sub-element called ``name``. The ``name`` attribute
@ -1503,9 +1443,6 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: None
.. _IUPAC Isotopic Compositions of the Elements 2009:
http://pac.iupac.org/publications/pac/pdf/2011/pdf/8302x0397.pdf
------------------------------------
Tallies Specification -- tallies.xml
------------------------------------

37
openmc/element.py
View file

@ -84,9 +84,9 @@ class Element(object):
@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)
if not scattering in ['data', 'iso-in-lab', None]:
msg = 'Unable to set scattering for Element to {0} which ' \
'is not "data", "iso-in-lab", or None'.format(scattering)
raise ValueError(msg)
self._scattering = scattering
@ -123,11 +123,14 @@ class Element(object):
"""
# Get the length of this elements atomic symbol
name_len = len(self.name)
# Get the nuclides present in nature
natural_nuclides = set()
for nuclide in sorted(NATURAL_ABUNDANCE.keys()):
if re.match(r'{}\d+'.format(self.name), nuclide):
natural_nuclides.add(int(nuclide[len(self.name):]))
natural_nuclides.add(int(nuclide[name_len:]))
# Create lists to store the expanded nuclides and abundances
nuclides = []
@ -149,7 +152,7 @@ class Element(object):
nuclide = child.attrib['materials']
if re.match(r'{}\d+'.format(self.name), nuclide) and \
'_m' not in nuclide:
library_nuclides.add(int(nuclide[len(self.name):]))
library_nuclides.add(int(nuclide[name_len:]))
# Get a set of the mutual and absent nuclides
mutual_nuclides = natural_nuclides.intersection(library_nuclides)
@ -162,7 +165,9 @@ class Element(object):
if len(absent_nuclides) == 0:
for nuclide, abundance in sorted(NATURAL_ABUNDANCE.items()):
if re.match(r'{}\d+'.format(self.name), nuclide):
nuclides.append(openmc.Nuclide(nuclide))
nuc = openmc.Nuclide(nuclide)
nuc.scattering = self.scattering
nuclides.append(nuc)
abundances.append(abundance)
# If no natural elements are present in the library, check if the
@ -170,7 +175,9 @@ class Element(object):
# nuclide. Else, raise an error.
elif len(mutual_nuclides) == 0:
if 0 in library_nuclides:
nuclides.append(openmc.Nuclide(self.name + '0'))
nuc = openmc.Nuclide(self.name + '0')
nuc.scattering = self.scattering
nuclides.append(nuc)
abundances.append(1.0)
else:
msg = 'Unable to expand element {0} because the cross '\
@ -188,18 +195,20 @@ class Element(object):
nuclides_a = []
for nuclide, abundance in sorted(NATURAL_ABUNDANCE.items()):
if re.match(r'{}\d+'.format(self.name), nuclide) and \
int(nuclide[len(self.name):]) in mutual_nuclides:
nuclides.append(openmc.Nuclide(nuclide))
nuclides_a.append(int(nuclide[len(self.name):]))
int(nuclide[name_len:]) in mutual_nuclides:
nuc = openmc.Nuclide(nuclide)
nuc.scattering = self.scattering
nuclides.append(nuc)
nuclides_a.append(int(nuclide[name_len:]))
abundances.append(abundance)
# Adjust the abundances for the absent nuclides
for nuclide, abundance in sorted(NATURAL_ABUNDANCE.items()):
if re.match(r'{}\d+'.format(self.name), nuclide) and \
int(nuclide[len(self.name):]) in absent_nuclides:
int(nuclide[name_len:]) in absent_nuclides:
# Get index to the nearest nuclide
a = int(nuclide[len(self.name):])
a = int(nuclide[name_len:])
i = min(list(range(len(nuclides_a))), key=lambda j: \
abs(nuclides_a[j] - a))
@ -264,9 +273,9 @@ class Element(object):
for i in range(n_nuclides):
abundances[i] /= sum_abundances
# Create a list of the isotopes in this element
isotopes = []
for nuclide, abundance in zip(nuclides, abundances):
pct = float('{:2.10f}'.format(percent*abundance))
isotopes.append((nuclide, pct, percent_type))
isotopes.append((nuclide, percent*abundance, percent_type))
return isotopes

4
openmc/material.py
View file

@ -461,6 +461,10 @@ class Material(object):
self._id)
raise ValueError(msg)
# Check that the enrichment is in the valid range
cv.check_less_than('enrichment', enrichment, 1/1.008, equality=True)
cv.check_greater_than('enrichment', enrichment, 0., equality=True)
if enrichment > 0.05:
msg = 'A uranium enrichment of {0} was given for Material ID='\
'"{1}". OpenMC assumes the U234/U235 mass ratio is '\

6
openmc/nuclide.py
View file

@ -89,9 +89,9 @@ class Nuclide(object):
@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)
if not scattering in ['data', 'iso-in-lab', None]:
msg = 'Unable to set scattering for Nuclide to {0} which ' \
'is not "data", "iso-in-lab", or None'.format(scattering)
raise ValueError(msg)
self._scattering = scattering

86
src/input_xml.F90
View file

@ -139,10 +139,10 @@ contains
! Look for deprecated cross_sections.xml file in settings.xml
if (check_for_node(doc, "cross_sections")) then
call warning("Setting cross_sections in settings.xml has been deprecated. &
&The cross_sections are now set in materials.xml and the &
&cross_sections input to materials.xml and OPENMC_CROSS_SECTIONS &
&environment variable will take precendent over setting &
call warning("Setting cross_sections in settings.xml has been deprecated.&
& The cross_sections are now set in materials.xml and the &
&cross_sections input to materials.xml and the OPENMC_CROSS_SECTIONS&
& environment variable will take precendent over setting &
&cross_sections in settings.xml.")
call get_node_value(doc, "cross_sections", path_cross_sections)
end if
@ -152,7 +152,7 @@ contains
if (check_for_node(doc, "multipole_library")) then
call warning("Setting multipole_library in settings.xml has been &
&deprecated. The multipole_library is now set in materials.xml and&
& the multipole_library input to materials.xml and &
& the multipole_library input to materials.xml and the &
&OPENMC_MULTIPOLE_LIBRARY environment variable will take &
&precendent over setting multipole_library in settings.xml.")
call get_node_value(doc, "multipole_library", path_multipole)
@ -1033,17 +1033,9 @@ contains
! Get temperature settings
if (check_for_node(doc, "temperature_default")) then
call warning("Setting temperature_default in settings.xml has been &
&deprecated. temperature_default is now set in materials.xml and &
&its input to materials.xml will take precendent &
&over setting it in settings.xml.")
call get_node_value(doc, "temperature_default", temperature_default)
end if
if (check_for_node(doc, "temperature_method")) then
call warning("Setting temperature_method in settings.xml has been &
&deprecated. temperature_method is now set in materials.xml and &
&its input to materials.xml will take precendent &
&over setting it in settings.xml.")
call get_node_value(doc, "temperature_method", temp_str)
select case (to_lower(temp_str))
case ('nearest')
@ -1055,17 +1047,9 @@ contains
end select
end if
if (check_for_node(doc, "temperature_tolerance")) then
call warning("Setting temperature_tolerance in settings.xml has been &
&deprecated. temperature_tolerance is now set in materials.xml and &
&its input to materials.xml will take precendent &
&over setting it in settings.xml.")
call get_node_value(doc, "temperature_tolerance", temperature_tolerance)
end if
if (check_for_node(doc, "temperature_multipole")) then
call warning("Setting temperature_multipole in settings.xml has been &
&deprecated. temperature_multipole is now set in materials.xml and &
&its input to materials.xml will take precendent &
&over setting it in settings.xml.")
call get_node_value(doc, "temperature_multipole", temp_str)
select case (to_lower(temp_str))
case ('true', '1')
@ -2076,7 +2060,7 @@ contains
! Find cross_sections.xml file -- the first place to look is the
! materials.xml file. If no file is found there, then we check the
! CROSS_SECTIONS environment variable
! OPENMC_CROSS_SECTIONS environment variable
if (.not. check_for_node(doc, "cross_sections")) then
! No cross_sections.xml file specified in settings.xml, check
! environment variable
@ -2084,6 +2068,8 @@ contains
call get_environment_variable("OPENMC_CROSS_SECTIONS", env_variable)
if (len_trim(env_variable) == 0) then
call get_environment_variable("CROSS_SECTIONS", env_variable)
! FIXME: When deprecated option of setting the cross sections in
! settings.xml is removed, remove ".and. path_cross_sections == ''"
if (len_trim(env_variable) == 0 .and. path_cross_sections == '') then
call fatal_error("No cross_sections.xml file was specified in &
&materials.xml, settings.xml, or in the OPENMC_CROSS_SECTIONS&
@ -2100,13 +2086,15 @@ contains
path_cross_sections = trim(env_variable)
else
call get_environment_variable("OPENMC_MG_CROSS_SECTIONS", env_variable)
! FIXME: When deprecated option of setting the mg cross sections in
! settings.xml is removed, remove ".and. path_cross_sections == ''"
if (len_trim(env_variable) == 0 .and. path_cross_sections == '') then
call fatal_error("No mgxs.xml file was specified in &
call fatal_error("No mgxs.h5 file was specified in &
&materials.xml or in the OPENMC_MG_CROSS_SECTIONS environment &
&variable. OpenMC needs such a file to identify where to &
&find ACE cross section libraries. Please consult the user's &
&find MG cross section libraries. Please consult the user's &
&guide at http://mit-crpg.github.io/openmc for information on &
&how to set up ACE cross section libraries.")
&how to set up MG cross section libraries.")
else
path_cross_sections = trim(env_variable)
end if
@ -2129,40 +2117,10 @@ contains
path_multipole = trim(path_multipole) // "/"
end if
! Get temperature settings
if (check_for_node(doc, "temperature_default")) then
call get_node_value(doc, "temperature_default", temperature_default)
end if
if (check_for_node(doc, "temperature_method")) then
call get_node_value(doc, "temperature_method", temp_str)
select case (to_lower(temp_str))
case ('nearest')
temperature_method = TEMPERATURE_NEAREST
case ('interpolation')
temperature_method = TEMPERATURE_INTERPOLATION
case default
call fatal_error("Unknown temperature method: " // trim(temp_str))
end select
end if
if (check_for_node(doc, "temperature_tolerance")) then
call get_node_value(doc, "temperature_tolerance", temperature_tolerance)
end if
if (check_for_node(doc, "temperature_multipole")) then
call get_node_value(doc, "temperature_multipole", temp_str)
select case (to_lower(temp_str))
case ('true', '1')
temperature_multipole = .true.
case ('false', '0')
temperature_multipole = .false.
case default
call fatal_error("Unrecognized value for <use_windowed_multipole> in &
&materials.xml")
end select
end if
! Close materials XML file
call close_xmldoc(doc)
! Now that the cross_sections.xml or mgxs.h5 has been located, read it in
if (run_CE) then
call read_ce_cross_sections_xml(libraries)
else
@ -2225,12 +2183,12 @@ contains
integer :: index_nuclide ! index in nuclides
integer :: index_sab ! index in sab_tables
logical :: file_exists ! does materials.xml exist?
character(20) :: name ! name of nuclide, e.g. 92235.03c
character(20) :: name ! name of nuclide, e.g. 92235.03c
character(MAX_LINE_LEN) :: filename ! absolute path to materials.xml
character(MAX_LINE_LEN) :: temp_str ! temporary string when reading
type(VectorChar) :: names ! temporary list of nuclide names
type(VectorInt) :: list_iso_lab ! temporary list of isotropic lab scatterers
type(Material), pointer :: mat => null()
type(Material), pointer :: mat => null()
type(Node), pointer :: doc => null()
type(Node), pointer :: node_mat => null()
type(Node), pointer :: node_nuc => null()
@ -4674,7 +4632,8 @@ contains
end subroutine generate_rpn
!===============================================================================
! ASSIGN_NUCLIDE_DENSITIES Assign and normalize nuclide densities
! ASSIGN_NUCLIDE_DENSITIES Read in the nuclide densities from materials.xml
! and assign to the corresponding nuclides.
!===============================================================================
subroutine assign_nuclide_densities()
@ -4731,16 +4690,13 @@ contains
! Copy units
call get_node_value(node_dens, "units", units)
! If the units is 'sum', then the total density of the material is taken
! to be the sum of the atom fractions listed on the nuclides
if (units == 'sum') then
! If the user gave the units as 'sum', then the total density of the
! material is taken to be the sum of the atom fractions listed on the
! nuclides
sum_density = .true.
else if (units == 'macro') then
if (check_for_node(node_dens, "value")) then
! Copy value
call get_node_value(node_dens, "value", val)
else
val = ONE
@ -4752,7 +4708,6 @@ contains
sum_density = .false.
else
! Copy value
call get_node_value(node_dens, "value", val)
! Check for erroneous density
@ -4761,6 +4716,7 @@ contains
call fatal_error("Need to specify a positive density on material " &
// trim(to_str(mat % id)) // ".")
end if
! Adjust material density based on specified units
select case(to_lower(units))
case ('g/cc', 'g/cm3')

2
tests/test_element_wo/inputs_true.dat
View file

@ -1 +1 @@
c472877dd96f51dcfd7b45ec1c230c3847b26eb1f0a97751bb175b70eafff2ad8ee97e3ed8691a3bc8bf8f27c074a9d34021322988185bb04c3e55958f40589b
f270ae43a29c53065678b419629ad713f5a4cc23c75825fee4da049d4159756f463fec86263a8dea6ce10fff6d60b24ebf100e8f5cc80e6d4fc31297ae1709ed

2
tests/test_enrichment/inputs_true.dat
View file

@ -1 +1 @@
6f8d495cd7537f8ca29355664b31d429fe13f656a868852769b27ef9566e7bb683973322e41a92bab65beb1ba6cf477698b30a7ae6bc17d2e34cb44751e8c81d
835357f7b3c10f1153bb5842978075d9bb788d639759f143dfe58064865712fdf4e17ce50747949411e7c5215f33371e883e48b670509e222fa338bbfe613b80