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Merge remote-tracking branch 'upstream/develop' into pyapi_filters
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Sterling Harper
commit
e464e22690
23 changed files with 687 additions and 696 deletions
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@ -4,7 +4,7 @@
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Summary File Format
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===================
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The current revision of the summary file format is 1.
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The current revision of the summary file format is 4.
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**/filetype** (*char[]*)
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@ -129,7 +129,16 @@ The current revision of the summary file format is 1.
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**/geometry/cells/cell <uid>/distribcell_index** (*int*)
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Index of this cell in distribcell filter arrays.
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Index of this cell in distribcell arrays. Only present if this cell is
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listed in a distribcell filter or if it uses distributed materials.
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**/geometry/cells/cell <uid>/paths** (*char[][]*)
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The paths traversed through the CSG tree to reach each distribcell
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instance. This consists of the integer IDs for each universe, cell and
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lattice delimited by '->'. Each lattice cell is specified by its (x,y) or
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(x,y,z) indices. Only present if this cell is listed in a distribcell filter
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or if it uses distributed materials.
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**/geometry/surfaces/surface <uid>/index** (*int*)
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@ -244,90 +253,6 @@ The current revision of the summary file format is 1.
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Names of S(:math:`\alpha`,:math:`\beta`) tables assigned to the material.
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**/tallies/n_tallies** (*int*)
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Number of tallies in the problem.
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**/tallies/n_meshes** (*int*)
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Number of meshes in the problem.
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**/tallies/mesh <uid>/index** (*int*)
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Index in the meshes array used internally in OpenMC.
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**/tallies/mesh <uid>/type** (*char[]*)
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Type of the mesh. The only valid option is currently 'regular'.
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**/tallies/mesh <uid>/dimension** (*int[]*)
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Number of mesh cells in each direction.
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**/tallies/mesh <uid>/lower_left** (*double[]*)
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Coordinates of the lower-left corner of the mesh.
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**/tallies/mesh <uid>/upper_right** (*double[]*)
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Coordinates of the upper-right corner of the mesh.
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**/tallies/mesh <uid>/width** (*double[]*)
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Width of a single mesh cell in each direction.
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**/tallies/tally <uid>/index** (*int*)
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Index in tallies array used internally in OpenMC.
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**/tallies/tally <uid>/name** (*char[]*)
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Name of the tally.
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**/tallies/tally <uid>/n_filters** (*int*)
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Number of filters applied to the tally.
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**/tallies/tally <uid>/filter <j>/type** (*char[]*)
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Type of the j-th filter. Can be 'universe', 'material', 'cell', 'cellborn',
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'surface', 'mesh', 'energy', 'energyout', or 'distribcell'.
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**/tallies/tally <uid>/filter <j>/offset** (*int*)
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Filter offset (used for distribcell filter).
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**/tallies/tally <uid>/filter <j>/paths** (*char[][]*)
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The paths traversed through the CSG tree to reach each distribcell
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instance (for 'distribcell' filters only). This consists of the integer
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IDs for each universe, cell and lattice delimited by '->'. Each lattice
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cell is specified by its (x,y) or (x,y,z) indices.
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**/tallies/tally <uid>/filter <j>/n_bins** (*int*)
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Number of bins for the j-th filter.
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**/tallies/tally <uid>/filter <j>/bins** (*int[]* or *double[]*)
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Value for each filter bin of this type.
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**/tallies/tally <uid>/nuclides** (*char[][]*)
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Array of nuclides to tally. Note that if no nuclide is specified in the user
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input, a single 'total' nuclide appears here.
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**/tallies/tally <uid>/n_score_bins** (*int*)
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Number of scoring bins for a single nuclide. In general, this can be greater
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than the number of user-specified scores since each score might have
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multiple scoring bins, e.g., scatter-PN.
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**/tallies/tally <uid>/moment_orders** (*char[][]*)
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Tallying moment orders for Legendre and spherical harmonic tally expansions
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(*e.g.*, 'P2', 'Y1,2', etc.).
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**/tallies/tally <uid>/score_bins** (*char[][]*)
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Scoring bins for the tally.
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File diff suppressed because one or more lines are too long
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@ -366,6 +366,7 @@ Core Classes
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openmc.data.ThermalScattering
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openmc.data.CoherentElastic
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openmc.data.FissionEnergyRelease
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openmc.data.DataLibrary
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Angle-Energy Distributions
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--------------------------
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@ -383,14 +383,16 @@ Cross Section Configuration
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---------------------------
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In order to run a simulation with OpenMC, you will need cross section data for
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each nuclide or material in your problem. OpenMC can be run in
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continuous-energy or multi-group mode.
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each nuclide or material in your problem. OpenMC can be run in continuous-energy
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or multi-group mode.
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In continuous-energy mode OpenMC uses ACE format cross sections; in this case
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you can use nuclear data that was processed with NJOY_, such as that
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distributed with MCNP_ or Serpent_. Several sources provide free processed
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ACE data as described below. The TALYS-based evaluated nuclear data library,
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TENDL_, is also openly available in ACE format.
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In continuous-energy mode, OpenMC uses a native HDF5 format to store all nuclear
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data. If you have ACE format data that was produced with NJOY_, such as that
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distributed with MCNP_ or Serpent_, it can be converted to the HDF5 format using
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the :ref:`openmc-ace-to-hdf5 <other_cross_sections>` script distributed with
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OpenMC. Several sources provide openly available ACE data as described
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below. The TALYS-based evaluated nuclear data library, TENDL_, is also available
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in ACE format.
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In multi-group mode, OpenMC utilizes an XML-based library format which can be
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used to describe nuclide- or material-specific quantities.
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@ -400,8 +402,8 @@ Using ENDF/B-VII.1 Cross Sections from NNDC
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The NNDC_ provides ACE data from the ENDF/B-VII.1 neutron and thermal scattering
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sublibraries at four temperatures processed using NJOY_. To use this data with
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OpenMC, a script is provided with OpenMC that will automatically download,
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extract, and set up a confiuration file:
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OpenMC, a script is provided with OpenMC that will automatically download and
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extract the ACE data, fix any deficiencies, and create an HDF5 library:
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.. code-block:: sh
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@ -410,56 +412,99 @@ extract, and set up a confiuration file:
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At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
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variable to the absolute path of the file
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``openmc/data/nndc/cross_sections.xml``. This cross section set is used by the
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test suite.
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``openmc/data/nndc_hdf5/cross_sections.xml``. This cross section set is used by
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the test suite.
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Using JEFF Cross Sections from OECD/NEA
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---------------------------------------
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The NEA_ provides processed ACE data from the JEFF_ nuclear library upon
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request. A DVD of the data can be requested here_. To use this data with OpenMC,
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the following steps must be taken:
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The NEA_ provides processed ACE data from the JEFF_ library. To use this data
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with OpenMC, a script is provided with OpenMC that will automatically download
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and extract the ACE data, fix any deficiencies, and create an HDF5 library.
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1. Copy and unzip the data on the DVD to a directory on your computer.
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2. In the root directory, a file named ``xsdir``, or some variant thereof,
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should be present. This file contains a listing of all the cross sections and
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is used by MCNP. This file should be converted to a ``cross_sections.xml``
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file for use with OpenMC. A utility is provided in the OpenMC distribution
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for this purpose:
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.. code-block:: sh
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.. code-block:: sh
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cd openmc/data
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python get_jeff_data.py
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openmc/scripts/openmc-xsdir-to-xml xsdir31 cross_sections.xml
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3. In the converted ``cross_sections.xml`` file, change the contents of the
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<directory> element to the absolute path of the directory containing the
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actual ACE files.
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4. Additionally, you may need to change any occurrences of upper-case "ACE"
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within the ``cross_sections.xml`` file to lower-case.
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5. Either set the :ref:`cross_sections` in a settings.xml file or the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
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the ``cross_sections.xml`` file.
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At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
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variable to the absolute path of the file
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``openmc/data/jeff-3.2-hdf5/cross_sections.xml``.
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Using Cross Sections from MCNP
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------------------------------
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To use cross sections distributed with MCNP, change the <directory> element in
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the ``cross_sections.xml`` file in the root directory of the OpenMC distribution
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to the location of the MCNP cross sections. Then, either set the
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:ref:`cross_sections` in a settings.xml file or the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
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the ``cross_sections.xml`` file.
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OpenMC is provided with a script that will automatically convert ENDF/B-VII.0
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and ENDF/B-VII.1 ACE data that is provided with MCNP5 or MCNP6. To convert the
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ENDF/B-VII.0 ACE files (``endf70[a-k]`` and ``endf70sab``) into the native HDF5
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format, run the following:
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Using Cross Sections from Serpent
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---------------------------------
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.. code-block:: sh
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cd openmc/data
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python convert_mcnp_endf70.py /path/to/mcnpdata/
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where ``/path/to/mcnpdata`` is the directory containing the ``endf70[a-k]``
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files.
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To convert the ENDF/B-VII.1 ACE files (the endf71x and ENDF71SaB libraries), use
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the following script:
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.. code-block:: sh
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cd openmc/data
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python convert_mcnp_endf71.py /path/to/mcnpdata
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where ``/path/to/mcnpdata`` is the directory containing the ``endf71x`` and
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``ENDF71SaB`` directories.
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.. _other_cross_sections:
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Using Other Cross Sections
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--------------------------
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If you have a library of ACE format cross sections other than those listed above
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that you need to convert to OpenMC's HDF5 format, the ``openmc-ace-to-hdf5``
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script can be used. There are four different ways you can specify ACE libraries
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that are to be converted:
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1. List each ACE library as a positional argument. This is very useful in
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conjunction with the usual shell utilities (ls, find, etc.).
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2. Use the --xml option to specify a pre-v0.9 cross_sections.xml file.
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3. Use the --xsdir option to specify a MCNP xsdir file.
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4. Use the --xsdata option to specify a Serpent xsdata file.
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The script does not use any extra information from cross_sections.xml/ xsdir/
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xsdata files to determine whether the nuclide is metastable. Instead, the
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--metastable argument can be used to specify whether the ZAID naming convention
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follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data
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convention (essentially the same as NNDC, except that the first metastable state
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of Am242 is 95242 and the ground state is 95642).
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The ``openmc-ace-to-hdf5`` script has the following command-line flags:
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-h, --help show this help message and exit
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-d DESTINATION, --destination DESTINATION
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Directory to create new library in (default: .)
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-m META, --metastable META
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How to interpret ZAIDs for metastable nuclides. META
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can be either 'nndc' or 'mcnp'. (default: nndc)
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--xml XML Old-style cross_sections.xml that lists ACE libraries
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(default: None)
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--xsdir XSDIR MCNP xsdir file that lists ACE libraries (default:
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None)
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--xsdata XSDATA Serpent xsdata file that lists ACE libraries (default:
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None)
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--fission_energy_release FISSION_ENERGY_RELEASE
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HDF5 file containing fission energy release data
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(default: None)
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To use cross sections distributed with Serpent, change the <directory> element
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in the ``cross_sections_serpent.xml`` file in the root directory of the OpenMC
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distribution to the location of the Serpent cross sections. Then, either set the
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:ref:`cross_sections` in a settings.xml file or the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
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the ``cross_sections_serpent.xml``
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file.
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Using Multi-Group Cross Sections
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--------------------------------
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@ -471,14 +516,13 @@ However, if the user has obtained or generated their own library, the user
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should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
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to the absolute path of the file library expected to used most frequently.
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.. _NJOY: http://t2.lanl.gov/nis/codes.shtml
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.. _NJOY: http://t2.lanl.gov/nis/codes/NJOY12/
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.. _NNDC: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
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.. _NEA: http://www.oecd-nea.org
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.. _JEFF: http://www.oecd-nea.org/dbdata/jeff/
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.. _here: http://www.oecd-nea.org/dbdata/pubs/jeff312-cd.html
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.. _JEFF: https://www.oecd-nea.org/dbforms/data/eva/evatapes/jeff_32/
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.. _MCNP: http://mcnp.lanl.gov
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.. _Serpent: http://montecarlo.vtt.fi
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.. _TENDL: ftp://ftp.nrg.eu/pub/www/talys/tendl2012/tendl2012.html
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.. _TENDL: https://tendl.web.psi.ch/tendl_2015/tendl2015.html
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--------------
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Running OpenMC
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|
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