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408
docs/source/io_formats/mgxs_library.rst
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@ -6,21 +6,19 @@ Multi-Group Cross Section Library Format
OpenMC can be run in continuous-energy mode or multi-group mode, provided the
nuclear data is available. In continuous-energy mode, the
``cross_sections.xml`` file contains necessary meta-data for each data set,
``cross_sections.xml`` file contains necessary meta-data for each dataset,
including the name and a file system location where the complete library
can be found. In multi-group mode, this ``mgxs.xml`` file contains
this same meta-data describing the nuclide or material, but also contains the
group-wise nuclear data. This portion of the manual describes the format of
the multi-group data library required to be used in the ``mgxs.xml``
file.
can be found. In multi-group mode, the multi-group meta-data and the
nuclear data itself is contained within an ``mgxs.h5`` file. This portion of
the manual describes the format of the multi-group data library required
to be used in the ``mgxs.h5`` file.
Similar to the other input file types, the multi-group library is provided in
the XML_ format. This library must provide some meta-data about the library
itself (such as the number of groups and the group structure, etc.) as well as
the actual cross section data itself for each of the necessary nuclides or
materials.
The multi-group library is provided in the HDF5_ format. This library must
provide some meta-data about the library itself (such as the number of
groups and the group structure, etc.) as well as the actual cross section
data itself for each of the necessary nuclides or materials.
.. _XML: http://www.w3.org/XML/
.. _HDF5: http://www.hdfgroup.org/HDF5/
.. _mgxs_lib_spec:
@ -28,277 +26,139 @@ materials.
MGXS Library Specification
--------------------------
The multi-group library meta-data is contained within the groups_,
group_structure_, and inverse_velocities_ elements.
The actual multi-group data itself is contained within the xsdata_ element.
**/**
.. _groups:
:Attributes: - **groups** (*int*) -- Number of energy groups
- **group structure** (*double[]*) -- Monotonically increasing
list of group boundaries, in units of MeV. The length of this
array should be the number of groups plus 1.
``<groups>`` Element
--------------------
**/<library name>/**
The ``<groups>`` element has no attributes and simply provides the number of
energy groups contained within the library.
The data within <library name> contains the temperature-dependent multi-group
data for the nuclide or material that it represents.
*Default*: None, this must be provided.
:Attributes: - **atomic_weight_ratio** (*double*) -- The atomic weight ratio
(optional, i.e. it is not meaningful for material-wise data).
- **fissionable** (*bool*) -- Whether the dataset is fissionable
(True) or not (False).
- **representation** (*char[]*) -- The method used to generate and
represent the multi-group cross sections. That is, whether they
were generated with scalar flux weighting (or reduced to a
similar representation) and thus are angle-independent, or if the
data was generated with angular dependent fluxes and thus the
data is angle-dependent. Valid values are either "isotropic" or
"angle".
- **num_azimuthal** (*int*) -- Number of equal width angular bins
that the azimuthal angular domain is subdivided if the
`representation` attribute is "angle". This parameter is
ignored otherwise.
- **num_polar** (*int*) -- Number of equal width angular bins
that the polar angular domain is subdivided if the
`representation` attribute is "angle". This parameter is
ignored otherwise.
- **scatter_format** (*char[]*) -- The representation of the
scattering angular distribution. The options are either
"legendre", "histogram", or "tabular". If not provided, the
default of "legendre" will be assumed.
- **order** (*int*) -- Either the Legendre order, number of bins,
or number of points (depending on the value of `scatter_format`)
used to describe the angular distribution associated with each
group-to-group transfer probability.
- **scatter_shape** (*char[]*) -- The shape of the provided
scatter and multiplicity matrix. The values provided are strings
describing the ordering the scattering array is provided in
row-major (i.e., C/C++ and Python) indexing. Valid values are
"[Order][G][G']" or "[Order][G'][G]" where "G'" denotes the
secondary/outgoing energy groups, "G" denotes the incoming
energy groups, and "Order" is the angular distribution index.
This value is not required; if not the default value of
"[Order][G][G']" will be assumed.
.. _group_structure:
**/<library name>/kTs/**
``<group_structure>`` Element
-----------------------------
:Datasets:
- **<TTT>K** (*double*) -- kT values (in MeV) for each Temperature
TTT (in Kelvin), rounded to the nearest integer
The ``<group_structure>`` element has no attributes and should be provided as a
monotonically increasing list of bounding energies, in MeV, for a number of
groups. To provide proper energy boundaries, the length of the data within the
``<group_structure>`` element should be one more than the number of groups in
the problem. For example, a two-group problem could be specified as:
**/<library name>/<TTT>K/**
.. code-block:: xml
Temperature-dependent data, provided for temperature <TTT>K.
<group_structure> 0.0 0.625E-6 20.0 </group_structure>
:Datasets: - **total** (*double[]* or *double[][][]*) -- Total cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups][azimuthal][polar].
- **absorption** (*double[]* or *double[][][]*) -- Absorption
cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups][azimuthal][polar].
- **fission** (*double[]* or *double[][][]*) -- Fission
cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups][azimuthal][polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **kappa-fission** (*double[]* or *double[][][]*) -- Kappa-Fission
(energy-release from fission) cross section.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups][azimuthal][polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **chi** (*double[]* or *double[][][]*) -- Fission neutron energy
spectra.
This is a 1-D vector if `representation` is "isotropic", or a 3-D
vector if `representation` is "angle" with dimensions of
[groups][azimuthal][polar]. This is only required if the dataset
is fissionable and fission-tallies are expected to be used.
- **nu-fission** (*double[]* to *double[][][][]*) -- Nu-Fission
cross section.
If **chi** is provided, then `nu-fission` has the same
dimensionality as `fission`. If **chi** is not provided, then
the `nu-fission` data must represent the fission neutron energy
spectra as well and thus will have one additional dimension
for the outgoing energy group. In this case, `nu-fission` has the
same dimensionality as `multiplicity_matrix`.
- **inverse_velocities** (*double[]*) -- Average inverse velocity
for each of the groups in the library. This dataset is optional.
*Default*: None, this must be provided.
**/<library name>/<TTT>K/scatter_data/**
.. _inverse_velocities:
``<inverse_velocities>`` Element
--------------------------------
The ``<inverse_velocities>`` element optionally indicates the average
inverse velocity corresponding to each of the groups in the problem.
This element should therefore be an array with a length which matches the
number of groups set in the groups_ element.
*Default*: Should this be needed by the presence of an ``inverse-velocity``
score in the ``tallies.xml`` file and not provided in this element, OpenMC
will simply convert the group mid-point energy to an inverse of the velocity
and use this information for tallying.
.. _xsdata:
``<xsdata>`` Element
--------------------
The ``<xsdata>`` element contains the nuclide or material-specific meta-data as
well as the actual cross section data. The following are the
attributes/sub-elements required to describe the meta-data:
:name:
The name of the microscopic or macroscopic data set. An extension to the
name must be provided (e.g., the ``.300K`` in ``UO2.300K``). The name and
extension together must be twelve or less characters in length. This
extension must follow a period and be five characters or less in length.
similar to the equivalent in the continuous-energy ``cross_sections.xml``
file, is used to denote variants of the particular nuclide or material of
interest (i.e. the ``UO2`` data in this example could have been generated
at a temperature of 300K).
*Default*: None, this must be provided.
:alias:
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 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".
*Default*: None, this element must be provided.
:representation:
This element provides the method used to generate and represent the
multi-group cross sections. That is, whether they were generated with
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".
*Default*: "isotropic"
:num_azimuthal:
This element provides the number of equal width angular 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).
Note that these bins are equal in azimuthal angle widths, not equal in the
cosine of the azimuthal angle widths.
*Default*: If ``representation`` is "angle", this must be provided. This
parameter is not used for other ``representation`` types.
:num_polar:
This element provides the number of equal width angular 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).
Note that these bins are equal in polar angle widths, not equal in the
cosine of the polar angle widths.
*Default*: If ``representation`` is "angle", this must be provided. This
parameter is not used for other ``representation`` types.
: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"
element.
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
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
equi-spaced point-wise representation.
*Default*: "legendre"
:order:
This element provides either the Legendre order, number of bins, or number
of points used to describe the angular distribution associated with each
group-to-group transfer probability. The specific meaning of this bin
depends upon the value of ``scatt_type`` as discussed above.
*Default*: None, this element must be provided.
:tabular_legendre:
This optional element is used to set how the Legendre scattering kernel, if
provided via the ``scatt_type`` element above, is represented and thus used
during the scattering process. Specifically, the options are to either
convert the Legendre expansion to a tabular representation or leave it as
a set of Legendre coefficients. Converting to a tabular representation
will cost memory but can allow for a decrease in runtime compared to
leaving as a set of Legendre coefficients. This element has the following
attributes/sub-elements:
: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.
*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
The following attributes/sub-elements are the cross section values to
be used during the transport process.
: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
groups described in the ``groups`` element. If ``representation`` is
"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*: If not provided, it will be determined by summing the
absorption and scattering cross sections.
: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
groups described in the ``groups`` element. If ``representation`` is
"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 must be provided.
:scatter:
This element requires the scattering moment matrices presented with the
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
matrix is repeated for every bin of the histogram or tabular
representation. Finally, if ``representation`` is "angle", the above
is repeated for every azimuthal angle and every polar angle, in that
order.
*Default*: None, this must be provided.
:multiplicity:
This element provides the ratio of neutrons produced in scattering
collisions to the neutrons which undergo scattering collisions; that is,
the multiplicity provides the code with a scaling factor to account for
neutrons being produced in (n,xn) reactions. This information is assumed
isotropic and therefore does not need to be repeated for every Legendre
moment or histogram/tabular bin. This matrix follows the same arrangement
as described for the ``scatter`` element, with the exception of the
data needed to provide the scattering type information.
*Default*: Multiplicities of 1.0 are assumed (i.e., (n,xn) reactions are
neglected).
The following fission-specific data are only needed should ``fissionable``
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
groups described in the ``groups`` element. If ``representation`` is
"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
requested and the material is fissionable.
: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
groups described in the ``groups`` element. If ``representation`` is
"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
requested and the material is fissionable.
:chi:
This element requires the group-wise fission spectra ordered by
increasing group index (i.e., fast to thermal). This element should be
used if making the common approximation that the fission spectra does
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
groups described in the ``groups`` element. If ``representation`` is
"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, either this element is provided or ``nu_fission`` is
provided in fission matrix form, or the material is not fissionable.
:nu_fission:
This element provides either the group-wise fission production cross
section vector (i.e., if ``chi`` is provided), or is the group-wise fission
production matrix. If providing the vector, it should be ordered the same
as the ``fission`` data. If providing the matrix, it should be ordered
the same as the ``multiplicity`` matrix.
*Default*: None, either this element must be provided if the material
is fissionable.
Data specific to neutron scattering for the temperature <TTT>K
:Datasets: - **g_min** (*int[]* or *int[][][]*) --
Minimum (most energetic) groups with non-zero values of
the scattering matrix provided. If `scatter_shape` is
"[Order][G][G']" then `g_min` will describe the minimum values
of "G'" for each "G"; if `scatter_shape` is "[Order][G'][G]"
then `g_min` will describe the minimum values of "G" for each "G'".
These group numbers use the standard
ordering where the fastest neutron energy group is group 1 while
the slowest neutron energy group is group G.
The dimensionality of `g_min` is:
`g_min[g]`, or `g_min[num_polar][num_azimuthal][g]`.
The former is used when `representation` is "isotropic", and the
latter when `representation` is "angle".
- **g_max** (*int[]* or *int[][][]*) --
Similar to `g_min`, except this dataset describes the maximum
(least energetic) groups with non-zero values of
the scattering matrix.
- **scatter_matrix** (*double[]*) -- Flattened representation of the
scattering moment matrices. The pre-flattened array corresponds to
the shape provied in `scatter_shape`, but if `representation` is
"angle" the dimensionality in `scatter_shape` is prepended by
"[num_polar][num_azimuthal]" dimensions. The right-most energy
group dimension will only include the entries between `g_min` and
`g_max`.
dimension has a dimensionality of `g_min` to `g_max`.
- **multiplicity_matrix** (*double[]*) -- Flattened representation of
the scattering moment matrices. This dataset provides the code with
a scaling factor to account for neutrons being produced in (n,xn)
reactions. This is assumed isotropic and therefore is not repeated
for every Legendre moment or histogram/tabular bin. This dataset is
optional, if it is not provided no multiplication (i.e., values of
1.0) will be assumed.
The pre-flattened array is shapes consistent with `scatter_matrix`
except the "[Order]" dimension in `scatter_shape` is ignored since
this data is assumed isotropic.

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docs/source/pythonapi/examples/mgxs-part-iv.ipynb
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72
docs/source/pythonapi/index.rst
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@ -345,6 +345,7 @@ Functions
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.model.create_triso_lattice
openmc.model.pack_trisos
@ -353,16 +354,6 @@ Functions
:mod:`openmc.data` -- Nuclear Data Interface
--------------------------------------------
Physical Data
-------------
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.data.atomic_mass
Core Classes
------------
@ -375,11 +366,23 @@ Core Classes
openmc.data.Reaction
openmc.data.Product
openmc.data.Tabulated1D
openmc.data.FissionEnergyRelease
openmc.data.ThermalScattering
openmc.data.CoherentElastic
openmc.data.FissionEnergyRelease
openmc.data.DataLibrary
Core Functions
--------------
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.data.atomic_mass
openmc.data.write_compact_458_library
Angle-Energy Distributions
--------------------------
@ -393,6 +396,7 @@ Angle-Energy Distributions
openmc.data.CorrelatedAngleEnergy
openmc.data.UncorrelatedAngleEnergy
openmc.data.NBodyPhaseSpace
openmc.data.LaboratoryAngleEnergy
openmc.data.AngleDistribution
openmc.data.EnergyDistribution
openmc.data.ArbitraryTabulated
@ -405,6 +409,24 @@ Angle-Energy Distributions
openmc.data.LevelInelastic
openmc.data.ContinuousTabular
Resonance Data
--------------
.. autosummary::
:toctree: generated
:nosignatures:
:template: myclass.rst
openmc.data.Resonances
openmc.data.ResonanceRange
openmc.data.SingleLevelBreitWigner
openmc.data.MultiLevelBreitWigner
openmc.data.ReichMoore
openmc.data.RMatrixLimited
openmc.data.ParticlePair
openmc.data.SpinGroup
openmc.data.Unresolved
ACE Format
----------
@ -425,9 +447,37 @@ Functions
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.data.ace.ascii_to_binary
openmc.data.write_compact_458_library
ENDF Format
-----------
Classes
+++++++
.. autosummary::
:toctree: generated
:nosignatures:
:template: myclass.rst
openmc.data.endf.Evaluation
Functions
+++++++++
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.data.endf.float_endf
openmc.data.endf.get_cont_record
openmc.data.endf.get_head_record
openmc.data.endf.get_tab1_record
openmc.data.endf.get_tab2_record
openmc.data.endf.get_text_record
.. _Jupyter: https://jupyter.org/
.. _NumPy: http://www.numpy.org/

20
docs/source/quickinstall.rst
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@ -8,6 +8,26 @@ This quick install guide outlines the basic steps needed to install OpenMC on
your computer. For more detailed instructions on configuring and installing
OpenMC, see :ref:`usersguide_install` in the User's Manual.
----------------------------------------
Installing on Linux/Mac with conda-forge
----------------------------------------
`Conda <http://conda.pydata.org/docs/>`_ is an open source package management
system and environment management system for installing multiple versions of
software packages and their dependencies and switching easily between them. If
you have `conda` installed on your system, OpenMC can be installed via the
`conda-forge` channel. First, add the `conda-forge` channel with:
.. code-block:: sh
conda config --add channels conda-forge
OpenMC can then be installed with:
.. code-block:: sh
conda install openmc
--------------------------------
Installing on Ubuntu through PPA
--------------------------------

59
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@ -121,10 +121,13 @@ multi-group mode.
``<cutoff>`` Element
--------------------
The ``<cutoff>`` element indicates the weight cutoff used below which particles
undergo Russian roulette. Surviving particles are assigned a user-determined
weight. Note that weight cutoffs and Russian rouletting are not turned on by
default. This element has the following attributes/sub-elements:
The ``<cutoff>`` element indicates two kinds of cutoffs. The first is the weight
cutoff used below which particles undergo Russian roulette. Surviving particles
are assigned a user-determined weight. Note that weight cutoffs and Russian
rouletting are not turned on by default. The second is the energy cutoff which
is used to kill particles under certain energy. The energy cutoff should not be
used unless you know particles under the energy are of no importance to results
you care. This element has the following attributes/sub-elements:
:weight:
The weight below which particles undergo Russian roulette.
@ -137,6 +140,11 @@ default. This element has the following attributes/sub-elements:
*Default*: 1.0
:energy:
The energy under which particles will be killed.
*Default*: 0.0
.. _eigenvalue:
``<eigenvalue>`` Element
@ -709,6 +717,36 @@ survival biasing, otherwise known as implicit capture or absorption.
*Default*: false
.. _tabular_legendre:
``<tabular_legendre>`` Element
---------------------------------
The optional ``<tabular_legendre>`` element specifies how the multi-group
Legendre scattering kernel is represented if encountered in a multi-group
problem. Specifically, the options are to either convert the Legendre
expansion to a tabular representation or leave it as a set of Legendre
coefficients. Converting to a tabular representation will cost memory but can
allow for a decrease in runtime compared to leaving as a set of Legendre
coefficients. This element has the following attributes/sub-elements:
:enable:
This attribute/sub-element denotes whether or not the conversion of a
Legendre scattering expansion to the tabular format should be performed or
not. A value of “true” means the conversion should be performed, “false”
means it will not.
*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
.. note:: This element is only used in the multi-group :ref:`energy_mode`.
.. _temperature_default:
``<temperature_default>`` Element
@ -872,6 +910,19 @@ displayed. This element takes the following attributes:
*Default*: 5
``<create_fission_neutrons>`` Element
-------------------------------------
The ``<create_fission_neutrons>`` element indicates whether fission neutrons
should be created or not. If this element is set to "true", fission neutrons
will be created; otherwise the fission is treated as capture and no fission
neutron will be created. Note that this option is only applied to fixed source
calculation. For eigenvalue calculation, fission will always be treated as real
fission.
*Default*: true
``<volume_calc>`` Element
-------------------------

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@ -4,6 +4,38 @@
Installation and Configuration
==============================
----------------------------------------
Installing on Linux/Mac with conda-forge
----------------------------------------
`Conda <http://conda.pydata.org/docs/>`_ is an open source package management
system and environment management system for installing multiple versions of
software packages and their dependencies and switching easily between
them. `conda-forge <https://conda-forge.github.io/>`_ is a community-led conda
channel of installable packages. For instructions on installing conda, please
consult their `documentation
<http://conda.pydata.org/docs/install/quick.html>`_.
Once you have `conda` installed on your system, add the `conda-forge` channel to
your configuration with:
.. code-block:: sh
conda config --add channels conda-forge
Once the `conda-forge` channel has been enabled, OpenMC can then be installed
with:
.. code-block:: sh
conda install openmc
It is possible to list all of the versions of OpenMC available on your platform with:
.. code-block:: sh
conda search openmc --channel conda-forge
-----------------------------
Installing on Ubuntu with PPA
-----------------------------
@ -51,7 +83,7 @@ Prerequisites
installed on your machine. Since a number of Fortran 2003/2008 features
are used in the code, it is recommended that you use the latest version of
whatever compiler you choose. For gfortran_, it is necessary to use
version 4.6.0 or above.
version 4.8.0 or above.
If you are using Debian or a Debian derivative such as Ubuntu, you can
install the gfortran compiler using the following command::
@ -407,13 +439,11 @@ extract the ACE data, fix any deficiencies, and create an HDF5 library:
.. code-block:: sh
cd openmc/data
python get_nndc_data.py
openmc-get-nndc-data
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
variable to the absolute path of the file
``openmc/data/nndc_hdf5/cross_sections.xml``. This cross section set is used by
the test suite.
variable to the absolute path of the file ``nndc_hdf5/cross_sections.xml``. This
cross section set is used by the test suite.
Using JEFF Cross Sections from OECD/NEA
---------------------------------------
@ -424,12 +454,10 @@ and extract the ACE data, fix any deficiencies, and create an HDF5 library.
.. code-block:: sh
cd openmc/data
python get_jeff_data.py
openmc-get-jeff-data
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
variable to the absolute path of the file
``openmc/data/jeff-3.2-hdf5/cross_sections.xml``.
variable to the absolute path of the file ``jeff-3.2-hdf5/cross_sections.xml``.
Using Cross Sections from MCNP
------------------------------
@ -441,8 +469,7 @@ format, run the following:
.. code-block:: sh
cd openmc/data
python convert_mcnp_endf70.py /path/to/mcnpdata/
openmc-convert-mcnp70-data /path/to/mcnpdata/
where ``/path/to/mcnpdata`` is the directory containing the ``endf70[a-k]``
files.
@ -452,8 +479,7 @@ the following script:
.. code-block:: sh
cd openmc/data
python convert_mcnp_endf71.py /path/to/mcnpdata
openmc-convert-mcnp71-data /path/to/mcnpdata
where ``/path/to/mcnpdata`` is the directory containing the ``endf71x`` and
``ENDF71SaB`` directories.
@ -470,16 +496,16 @@ that are to be converted:
1. List each ACE library as a positional argument. This is very useful in
conjunction with the usual shell utilities (ls, find, etc.).
2. Use the --xml option to specify a pre-v0.9 cross_sections.xml file.
3. Use the --xsdir option to specify a MCNP xsdir file.
4. Use the --xsdata option to specify a Serpent xsdata file.
2. Use the ``--xml`` option to specify a pre-v0.9 cross_sections.xml file.
3. Use the ``--xsdir` option to specify a MCNP xsdir file.
4. Use the ``--xsdata`` option to specify a Serpent xsdata file.
The script does not use any extra information from cross_sections.xml/ xsdir/
xsdata files to determine whether the nuclide is metastable. Instead, the
--metastable argument can be used to specify whether the ZAID naming convention
follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data
convention (essentially the same as NNDC, except that the first metastable state
of Am242 is 95242 and the ground state is 95642).
``--metastable`` argument can be used to specify whether the ZAID naming
convention follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the
MCNP data convention (essentially the same as NNDC, except that the first
metastable state of Am242 is 95242 and the ground state is 95642).
The ``openmc-ace-to-hdf5`` script has the following command-line flags: