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Merge pull request #2174 from paulromano/prepare-0.13.1
Prepare for version 0.13.1
This commit is contained in:
Paul Romano
GitHub
commit
48fb58a05d
28 changed files with 296 additions and 50 deletions
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@ -51,3 +51,10 @@ Spatial Distributions
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openmc.stats.SphericalIndependent
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openmc.stats.Box
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openmc.stats.Point
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.. autosummary::
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:toctree: generated
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:nosignatures:
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:template: myfunction.rst
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openmc.stats.spherical_uniform
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182
docs/source/releasenotes/0.13.1.rst
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182
docs/source/releasenotes/0.13.1.rst
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@ -0,0 +1,182 @@
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====================
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What's New in 0.13.1
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====================
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.. currentmodule:: openmc
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-------
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Summary
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-------
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This release of OpenMC includes many bug fixes as well as improvements in
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geometry modeling, mesh functionality, source specification, depletion
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capabilities, and other general enhancements. The depletion module features a
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new transport operator, :class:`openmc.deplete.IndependentOperator`, that allows
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a depletion calculation to be performed using arbitrary one-group cross sections
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(e.g., generated by an external solver) along with a
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:class:`openmc.deplete.MicroXS` class for managing one-group cross sections. The
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track file generation capability has been significantly overhauled and a new
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:class:`openmc.Tracks` class was introduced to allow access to information in
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track files from the Python API. Support has been added for new ENDF thermal
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scattering evaluations that use mixed coherent/incoherent elastic scattering.
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------------------------------------
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Compatibility Notes and Deprecations
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------------------------------------
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- The ``openmc.deplete.Operator`` class has been renamed
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:class:`openmc.deplete.CoupledOperator`.
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- The ``openmc.deplete.ResultsList`` class has been renamed to
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:class:`openmc.deplete.Results` and no longer requires you to call the
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``from_hdf5()`` method in order to create it; instead, you can directly
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instantiate it.
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- A few methods that represent k-effective have been renamed for the sake of
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consistency:
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- ``openmc.StatePoint.k_combined`` is now :attr:`openmc.StatePoint.keff`
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- ``openmc.deplete.ResultsList.get_eigenvalue`` is now
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:meth:`openmc.deplete.Results.get_keff`
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- The :class:`openmc.stats.SphericalIndependent` class, which used to
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accept a distribution for ``theta`` now accepts a distribution for ``cos_theta``
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instead in order to more easily handle the common case of specifying a uniform
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spatial distribution over a sphere (also see the new
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:func:`openmc.stats.spherical_uniform` function).
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- If you are building OpenMC from source, note that several of our CMake options
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have been changed:
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========== ======================
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Old option New option
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========== ======================
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debug ---
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optimize ---
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profile OPENMC_ENABLE_PROFILE
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coverage OPENMC_ENABLE_COVERAGE
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openmp OPENMC_USE_OPENMP
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--- OPENMC_USE_MPI
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dagmc OPENMC_USE_DAGMC
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libmesh OPENMC_USE_LIBMESH
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========== ======================
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The ``debug`` and ``optimize`` options have been removed; instead, use the
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standard `CMAKE_BUILD_TYPE
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<https://cmake.org/cmake/help/latest/variable/CMAKE_BUILD_TYPE.html>`_
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variable.
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------------
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New Features
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------------
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- Two new composite surfaces: :class:`openmc.model.IsogonalOctagon` and
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:class:`openmc.model.CylinderSector`.
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- The :class:`~openmc.DAGMCUniverse` class now has a
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:attr:`~openmc.DAGMCUniverse.bounding_box` attribute and a
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:meth:`~openmc.DAGMCUniverse.bounding_region` method.
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- When translating a :class:`~openmc.Region` using the
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:meth:`~openmc.Region.translate` method, there is now an ``inplace`` argument.
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- The :class:`~openmc.Material` class has several new methods and attributes:
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- The :meth:`~openmc.Material.add_components` methods allows you to add
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multiple nuclides/elements to a material with a single call by passing a
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dictionary.
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- The :meth:`~openmc.Material.get_activity` method returns the activity of a
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material in Bq, Bq/g, or Bq/cm³.
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- The :meth:`~openmc.Material.remove_element` method removes an element from a
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material
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- The :meth:`~openmc.Material.get_nuclide_atoms` method gives the number of
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atoms of each nuclide in a material
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- All mesh classes now have a ``volumes`` property that provides the volume of
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each mesh element as well as ``write_data_to_vtk`` methods.
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- Support for externally managed MOAB meshes or libMesh meshes.
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- Multiple discrete distributions can be merged with the new
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:meth:`~openmc.stats.Discrete.merge` method.
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- The :func:`openmc.stats.spherical_uniform` function creates a uniform
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distribution over a sphere using the
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:class:`~openmc.stats.SphericalIndependent` class.
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- Univariate distributions in the :mod:`openmc.stats` module now have
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``sample()`` methods.
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- An ``openmc_sample_external_source`` function has been added to the C API with
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a corresponding Python binding :func:`openmc.lib.sample_external_source`.
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- The track file generation capability has been completely overhauled. Track
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files now include much more information, and a new :class:`~openmc.Tracks`
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class allows access to track file information from the Python API and has a
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:meth:`~openmc.Tracks.write_to_vtk` method for writing a VTK file. Multiple
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tracks are now written to a single file (one per MPI rank).
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- A new :func:`openmc.wwinp_to_wws` function that converts weight windows from a
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``wwinp`` file to a list of :class:`~openmc.WeightWindows` objects.
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- The new :meth:`openmc.EnergyFilter.from_group_structure` method provides a
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way of creating an energy filter with a group structure identified by name.
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- The :class:`openmc.data.Decay` class now has a
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:attr:`~openmc.data.Decay.sources` property that provides radioactive decay
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source distributions.
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- A :class:`openmc.mgxs.ReducedAbsorptionXS` class produces a multigroup cross
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section representing "reduced" absorption (absorption less neutron production
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from (n,xn) reactions).
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- Added support in the Python API and HDF5 nuclear data format for new ENDF
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thermal neutron scattering evaluations with mixed coherent elastic and
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incoherent elastic.
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- CMake now relies on ``find_package(MPI)`` for a more standard means of
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identifying an MPI compiler configuration.
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---------
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Bug Fixes
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---------
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- `Fix bug when a rotation matrix is passed to Halfspace.rotate <https://github.com/openmc-dev/openmc/pull/1974>`_
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- `Fix bug for spherical mesh string repr <https://github.com/openmc-dev/openmc/pull/1983>`_
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- `Fix package_data specification to include pyx files <https://github.com/openmc-dev/openmc/pull/1989>`_
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- `Allow meshes with same ID to appear in multiple files <https://github.com/openmc-dev/openmc/pull/1990>`_
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- `Fix overwritten variable in get_libraries_from_xsdata <https://github.com/openmc-dev/openmc/pull/1994>`_
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- `Write output files to correct directory <https://github.com/openmc-dev/openmc/pull/1998>`_
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- `Allow CMake to properly find third-party packages <https://github.com/openmc-dev/openmc/pull/2006>`_
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- `Fix Region.from_expression when ")(" appears in specification <https://github.com/openmc-dev/openmc/pull/2012>`_
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- `Move lost particle reset from finalize() to reset() <https://github.com/openmc-dev/openmc/pull/2015>`_
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- `Minor typo fixes in test_lattice.py <https://github.com/openmc-dev/openmc/pull/2024>`_
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- `Fix color assignment in Universe.plot <https://github.com/openmc-dev/openmc/pull/2034>`_
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- `Several depletion-related fixes <https://github.com/openmc-dev/openmc/pull/2037>`_
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- `Allow control of C++ standard used by compiler <https://github.com/openmc-dev/openmc/pull/2040/files>`_
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- `Fix IO format documentation for surface source read/write <https://github.com/openmc-dev/openmc/pull/2048>`_
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- `Make sure basis gets set in Plot.from_geometry <https://github.com/openmc-dev/openmc/pull/2050>`_
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- `Improve robustness of torus distance calculation <https://github.com/openmc-dev/openmc/pull/2053>`_
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- `Allow use of redundant fission when adjusting KERMA in from_njoy <https://github.com/openmc-dev/openmc/pull/2055>`_
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- `Disable GNU extensions for CMake target <https://github.com/openmc-dev/openmc/pull/2072>`_
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- `Two from_xml fixes <https://github.com/openmc-dev/openmc/pull/2074>`_
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- `Fix for rare infinite loop when finding cell <https://github.com/openmc-dev/openmc/pull/2075>`_
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- `Allow photon heating to be tallied by nuclide <https://github.com/openmc-dev/openmc/pull/2076>`_
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- `Use UTF-8 encoding when reading dose coefficients <https://github.com/openmc-dev/openmc/pull/2078>`_
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- `Fix a corner case in Region.from_expression <https://github.com/openmc-dev/openmc/pull/2081>`_
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- `Fix bug in spherical and cylindrical meshes <https://github.com/openmc-dev/openmc/pull/2108>`_
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- `Ensure weight window bounds are flattened when writing to XML <https://github.com/openmc-dev/openmc/pull/2113>`_
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- `Fix for std::cout sync bug in output.cpp <https://github.com/openmc-dev/openmc/pull/2122>`_
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- `Allow compiling against fmt v9 <https://github.com/openmc-dev/openmc/pull/2136>`_
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- `Fix TimeFilter for small time intervals <https://github.com/openmc-dev/openmc/pull/2168>`_
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------------
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Contributors
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------------
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- `David Andrs <https://github.com/andrsd>`_
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- `Hunter Belanger <https://github.com/HunterBelanger>`_
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- `Helen Brooks <https://github.com/helen-brooks>`_
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- `Rémi Delaporte-Mathurin <https://github.com/RemDelaporteMathurin>`_
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- `Joffrey Dorville <https://github.com/JoffreyDorville>`_
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- `Christopher Fichtlscherer <https://github.com/cfichtlscherer>`_
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- `Lewis Gross <https://github.com/lewisgross1296>`_
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- `Andrew Johnson <https://github.com/drewejohnson>`_
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- `Kalin Kiesling <https://github.com/kkiesling>`_
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- `Amanda Lund <https://github.com/amandalund>`_
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- `Richard Morrison <https://github.com/richmorrison>`_
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- `Patrick Myers <https://github.com/myerspat>`_
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- `Adam Nelson <https://github.com/nelsonag>`_
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- `April Novak <https://github.com/aprilnovak>`_
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- `Ethan Peterson <https://github.com/eepeterson>`_
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- `Gavin Ridley <https://github.com/gridley>`_
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- `Paul Romano <https://github.com/paulromano>`_
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- `Jonathan Shimwell <https://github.com/Shimwell>`_
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- `Patrick Shriwise <https://github.com/pshriwise>`_
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- `Amelia Trainer <https://github.com/ameliojo>`_
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- `John Tramm <https://github.com/jtramm>`_
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- `Bob Urberger <https://github.com/burberger>`_
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- `Olek Yardas <https://github.com/yardasol>`_
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@ -7,6 +7,7 @@ Release Notes
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.. toctree::
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:maxdepth: 1
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0.13.1
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0.13.0
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0.12.2
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0.12.1
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@ -14,7 +14,7 @@ are requested.
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The depletion module is designed such that the reaction rate solution (the
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transport "operator") is completely isolated from the solution of the
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transmutation equations and the method used for advancing time.
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transmutation equations and the method used for advancing time.
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:mod:`openmc.deplete` supports multiple time-integration methods for determining
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material compositions over time. Each method appears as a different class.
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@ -40,7 +40,7 @@ time::
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Note that the coupling between the reaction rate solver and the transmutation
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solver happens in-memory rather than by reading/writing files on disk. OpenMC
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has two categories of transport operators for obtaining transmutation reaction
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rates.
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rates.
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.. _coupled-depletion:
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@ -65,7 +65,7 @@ Any material that contains a fissionable nuclide is depleted by default, but
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this can behavior can be changed with the :attr:`Material.depletable` attribute.
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.. important::
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The volume must be specified for each material that is depleted by setting
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the :attr:`Material.volume` attribute. This is necessary in order to
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calculate the proper normalization of tally results based on the source rate.
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@ -198,8 +198,8 @@ Transport-independent depletion
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===============================
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.. warning::
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This feature is still under heavy development and has yet to be rigorously
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This feature is still under heavy development and has yet to be rigorously
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verified. API changes and feature additions are possible and likely in
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the near future.
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@ -237,7 +237,7 @@ object::
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'U236': 4.57e18,
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'O16': 4.64e22,
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'O17': 1.76e19}
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volume = 0.5
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volume = 0.5
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op = openmc.deplete.IndependentOperator.from_nuclides(volume,
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nuclides,
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micro_xs,
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@ -250,8 +250,8 @@ transport-depletion calculation and follow the same steps from there.
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.. note::
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Ideally, one-group cross section data should be available for every
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reaction in the depletion chain. If a nuclide that has a reaction
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associated with it in the depletion chain is present in the `nuclides`
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reaction in the depletion chain. If a nuclide that has a reaction
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associated with it in the depletion chain is present in the `nuclides`
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parameter but not the cross section data, that reaction will not be
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simulated.
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@ -289,7 +289,7 @@ or from data arrays::
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.. important::
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Both :meth:`~openmc.deplete.MicroXS.from_csv()` and
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:meth:`~openmc.deplete.MicroXS.from_array()` assume the cross section values
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:meth:`~openmc.deplete.MicroXS.from_array()` assume the cross section values
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provided are in barns by defualt, but have no way of verifying this. Make
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sure your cross sections are in the correct units before passing to a
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:class:`~openmc.deplete.IndependentOperator` object.
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@ -315,26 +315,28 @@ normalizing reaction rates:
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the cross-sections by the ``source-rate``.
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2. ``fission-q`` normalization, which uses the ``power`` or ``power_density``
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provided by the time integrator to obtain reaction rates by computing a value
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for the flux based on this power. The general equation for the flux is
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for the flux based on this power. The general equation for the flux is
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.. math::
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\phi = \frac{P}{V \cdot \sum_i (Q_i \cdot \sigma^f_i \cdot \n_i)}
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\phi = \frac{P}{\sum\limits_i (Q_i \sigma^f_i N_i)}
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where :math:`\sum_i` is the sum over all nuclides :math:`i`. This equation
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makes the same assumptions and issues as discussed in
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:ref:`energy-deposition`. Unfortunately, the proposed solution in that
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where :math:`P` is the power, :math:`Q_i` is the fission Q value for nuclide
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:math:`i`, :math:`\sigma_i^f` is the microscopic fission cross section for
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nuclide :math:`i`, and :math:`N_i` is the number of atoms of nuclide
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:math:`i`. This equation makes the same assumptions and issues as discussed
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in :ref:`energy-deposition`. Unfortunately, the proposed solution in that
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section does not apply here since we are decoupled from transport code.
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However, there is a method to converge to a more accurate value for flux by
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using substeps during time integration.
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`This paper <https://doi.org/10.1016/j.anucene.2016.05.031>`_ provides a
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good discussion of this method.
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using substeps during time integration. `This paper
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<https://doi.org/10.1016/j.anucene.2016.05.031>`_ provides a good discussion
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of this method.
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.. warning::
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The accuracy of results when using ``fission-q`` is entirely dependent on
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your depletion chain. Make sure it has sufficient data to resolve the
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dynamics of your particular scenario.
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dynamics of your particular scenario.
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Multiple Materials
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~~~~~~~~~~~~~~~~~~
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