Merge pull request #2174 from paulromano/prepare-0.13.1

Prepare for version 0.13.1

Author: paulromano

docs/source/pythonapi/stats.rst

@@ -51,3 +51,10 @@ Spatial Distributions
    openmc.stats.SphericalIndependent
    openmc.stats.Box
    openmc.stats.Point
+
+.. autosummary::
+   :toctree: generated
+   :nosignatures:
+   :template: myfunction.rst
+
+   openmc.stats.spherical_uniform

docs/source/releasenotes/0.13.1.rst

@@ -0,0 +1,182 @@
+====================
+What's New in 0.13.1
+====================
+
+.. currentmodule:: openmc
+
+-------
+Summary
+-------
+
+This release of OpenMC includes many bug fixes as well as improvements in
+geometry modeling, mesh functionality, source specification, depletion
+capabilities, and other general enhancements. The depletion module features a
+new transport operator, :class:`openmc.deplete.IndependentOperator`, that allows
+a depletion calculation to be performed using arbitrary one-group cross sections
+(e.g., generated by an external solver) along with a
+:class:`openmc.deplete.MicroXS` class for managing one-group cross sections. The
+track file generation capability has been significantly overhauled and a new
+:class:`openmc.Tracks` class was introduced to allow access to information in
+track files from the Python API. Support has been added for new ENDF thermal
+scattering evaluations that use mixed coherent/incoherent elastic scattering.
+
+------------------------------------
+Compatibility Notes and Deprecations
+------------------------------------
+
+- The ``openmc.deplete.Operator`` class has been renamed
+  :class:`openmc.deplete.CoupledOperator`.
+- The ``openmc.deplete.ResultsList`` class has been renamed to
+  :class:`openmc.deplete.Results` and no longer requires you to call the
+  ``from_hdf5()`` method in order to create it; instead, you can directly
+  instantiate it.
+- A few methods that represent k-effective have been renamed for the sake of
+  consistency:
+
+  - ``openmc.StatePoint.k_combined`` is now :attr:`openmc.StatePoint.keff`
+  - ``openmc.deplete.ResultsList.get_eigenvalue`` is now
+    :meth:`openmc.deplete.Results.get_keff`
+
+- The :class:`openmc.stats.SphericalIndependent` class, which used to
+  accept a distribution for ``theta`` now accepts a distribution for ``cos_theta``
+  instead in order to more easily handle the common case of specifying a uniform
+  spatial distribution over a sphere (also see the new
+  :func:`openmc.stats.spherical_uniform` function).
+
+- If you are building OpenMC from source, note that several of our CMake options
+  have been changed:
+
+  ==========  ======================
+  Old option  New option
+  ==========  ======================
+  debug       ---
+  optimize    ---
+  profile     OPENMC_ENABLE_PROFILE
+  coverage    OPENMC_ENABLE_COVERAGE
+  openmp      OPENMC_USE_OPENMP
+  ---         OPENMC_USE_MPI
+  dagmc       OPENMC_USE_DAGMC
+  libmesh     OPENMC_USE_LIBMESH
+  ==========  ======================
+
+  The ``debug`` and ``optimize`` options have been removed; instead, use the
+  standard `CMAKE_BUILD_TYPE
+  <https://cmake.org/cmake/help/latest/variable/CMAKE_BUILD_TYPE.html>`_
+  variable.
+
+------------
+New Features
+------------
+
+- Two new composite surfaces: :class:`openmc.model.IsogonalOctagon` and
+  :class:`openmc.model.CylinderSector`.
+- The :class:`~openmc.DAGMCUniverse` class now has a
+  :attr:`~openmc.DAGMCUniverse.bounding_box` attribute and a
+  :meth:`~openmc.DAGMCUniverse.bounding_region` method.
+- When translating a :class:`~openmc.Region` using the
+  :meth:`~openmc.Region.translate` method, there is now an ``inplace`` argument.
+- The :class:`~openmc.Material` class has several new methods and attributes:
+
+  - The :meth:`~openmc.Material.add_components` methods allows you to add
+    multiple nuclides/elements to a material with a single call by passing a
+    dictionary.
+  - The :meth:`~openmc.Material.get_activity` method returns the activity of a
+    material in Bq, Bq/g, or Bq/cm³.
+  - The :meth:`~openmc.Material.remove_element` method removes an element from a
+    material
+  - The :meth:`~openmc.Material.get_nuclide_atoms` method gives the number of
+    atoms of each nuclide in a material
+
+- All mesh classes now have a ``volumes`` property that provides the volume of
+  each mesh element as well as ``write_data_to_vtk`` methods.
+- Support for externally managed MOAB meshes or libMesh meshes.
+- Multiple discrete distributions can be merged with the new
+  :meth:`~openmc.stats.Discrete.merge` method.
+- The :func:`openmc.stats.spherical_uniform` function creates a uniform
+  distribution over a sphere using the
+  :class:`~openmc.stats.SphericalIndependent` class.
+- Univariate distributions in the :mod:`openmc.stats` module now have
+  ``sample()`` methods.
+- An ``openmc_sample_external_source`` function has been added to the C API with
+  a corresponding Python binding :func:`openmc.lib.sample_external_source`.
+- The track file generation capability has been completely overhauled. Track
+  files now include much more information, and a new :class:`~openmc.Tracks`
+  class allows access to track file information from the Python API and has a
+  :meth:`~openmc.Tracks.write_to_vtk` method for writing a VTK file. Multiple
+  tracks are now written to a single file (one per MPI rank).
+- A new :func:`openmc.wwinp_to_wws` function that converts weight windows from a
+  ``wwinp`` file to a list of :class:`~openmc.WeightWindows` objects.
+- The new :meth:`openmc.EnergyFilter.from_group_structure` method provides a
+  way of creating an energy filter with a group structure identified by name.
+- The :class:`openmc.data.Decay` class now has a
+  :attr:`~openmc.data.Decay.sources` property that provides radioactive decay
+  source distributions.
+- A :class:`openmc.mgxs.ReducedAbsorptionXS` class produces a multigroup cross
+  section representing "reduced" absorption (absorption less neutron production
+  from (n,xn) reactions).
+- Added support in the Python API and HDF5 nuclear data format for new ENDF
+  thermal neutron scattering evaluations with mixed coherent elastic and
+  incoherent elastic.
+- CMake now relies on ``find_package(MPI)`` for a more standard means of
+  identifying an MPI compiler configuration.
+
+---------
+Bug Fixes
+---------
+
+- `Fix bug when a rotation matrix is passed to Halfspace.rotate <https://github.com/openmc-dev/openmc/pull/1974>`_
+- `Fix bug for spherical mesh string repr <https://github.com/openmc-dev/openmc/pull/1983>`_
+- `Fix package_data specification to include pyx files <https://github.com/openmc-dev/openmc/pull/1989>`_
+- `Allow meshes with same ID to appear in multiple files <https://github.com/openmc-dev/openmc/pull/1990>`_
+- `Fix overwritten variable in get_libraries_from_xsdata <https://github.com/openmc-dev/openmc/pull/1994>`_
+- `Write output files to correct directory <https://github.com/openmc-dev/openmc/pull/1998>`_
+- `Allow CMake to properly find third-party packages <https://github.com/openmc-dev/openmc/pull/2006>`_
+- `Fix Region.from_expression when ")(" appears in specification <https://github.com/openmc-dev/openmc/pull/2012>`_
+- `Move lost particle reset from finalize() to reset() <https://github.com/openmc-dev/openmc/pull/2015>`_
+- `Minor typo fixes in test_lattice.py <https://github.com/openmc-dev/openmc/pull/2024>`_
+- `Fix color assignment in Universe.plot <https://github.com/openmc-dev/openmc/pull/2034>`_
+- `Several depletion-related fixes <https://github.com/openmc-dev/openmc/pull/2037>`_
+- `Allow control of C++ standard used by compiler <https://github.com/openmc-dev/openmc/pull/2040/files>`_
+- `Fix IO format documentation for surface source read/write <https://github.com/openmc-dev/openmc/pull/2048>`_
+- `Make sure basis gets set in Plot.from_geometry <https://github.com/openmc-dev/openmc/pull/2050>`_
+- `Improve robustness of torus distance calculation <https://github.com/openmc-dev/openmc/pull/2053>`_
+- `Allow use of redundant fission when adjusting KERMA in from_njoy <https://github.com/openmc-dev/openmc/pull/2055>`_
+- `Disable GNU extensions for CMake target <https://github.com/openmc-dev/openmc/pull/2072>`_
+- `Two from_xml fixes <https://github.com/openmc-dev/openmc/pull/2074>`_
+- `Fix for rare infinite loop when finding cell <https://github.com/openmc-dev/openmc/pull/2075>`_
+- `Allow photon heating to be tallied by nuclide <https://github.com/openmc-dev/openmc/pull/2076>`_
+- `Use UTF-8 encoding when reading dose coefficients <https://github.com/openmc-dev/openmc/pull/2078>`_
+- `Fix a corner case in Region.from_expression <https://github.com/openmc-dev/openmc/pull/2081>`_
+- `Fix bug in spherical and cylindrical meshes <https://github.com/openmc-dev/openmc/pull/2108>`_
+- `Ensure weight window bounds are flattened when writing to XML <https://github.com/openmc-dev/openmc/pull/2113>`_
+- `Fix for std::cout sync bug in output.cpp <https://github.com/openmc-dev/openmc/pull/2122>`_
+- `Allow compiling against fmt v9 <https://github.com/openmc-dev/openmc/pull/2136>`_
+- `Fix TimeFilter for small time intervals <https://github.com/openmc-dev/openmc/pull/2168>`_
+
+------------
+Contributors
+------------
+
+- `David Andrs <https://github.com/andrsd>`_
+- `Hunter Belanger <https://github.com/HunterBelanger>`_
+- `Helen Brooks <https://github.com/helen-brooks>`_
+- `Rémi Delaporte-Mathurin <https://github.com/RemDelaporteMathurin>`_
+- `Joffrey Dorville <https://github.com/JoffreyDorville>`_
+- `Christopher Fichtlscherer <https://github.com/cfichtlscherer>`_
+- `Lewis Gross <https://github.com/lewisgross1296>`_
+- `Andrew Johnson <https://github.com/drewejohnson>`_
+- `Kalin Kiesling <https://github.com/kkiesling>`_
+- `Amanda Lund <https://github.com/amandalund>`_
+- `Richard Morrison <https://github.com/richmorrison>`_
+- `Patrick Myers <https://github.com/myerspat>`_
+- `Adam Nelson <https://github.com/nelsonag>`_
+- `April Novak <https://github.com/aprilnovak>`_
+- `Ethan Peterson <https://github.com/eepeterson>`_
+- `Gavin Ridley <https://github.com/gridley>`_
+- `Paul Romano <https://github.com/paulromano>`_
+- `Jonathan Shimwell <https://github.com/Shimwell>`_
+- `Patrick Shriwise <https://github.com/pshriwise>`_
+- `Amelia Trainer <https://github.com/ameliojo>`_
+- `John Tramm <https://github.com/jtramm>`_
+- `Bob Urberger <https://github.com/burberger>`_
+- `Olek Yardas <https://github.com/yardasol>`_

docs/source/releasenotes/index.rst

@@ -7,6 +7,7 @@ Release Notes
 .. toctree::
   :maxdepth: 1
 
+  0.13.1
   0.13.0
   0.12.2
   0.12.1

docs/source/usersguide/depletion.rst

@@ -14,7 +14,7 @@ are requested.
 
 The depletion module is designed such that the reaction rate solution (the
 transport "operator") is completely isolated from the solution of the
-transmutation equations and the method used for advancing time. 
+transmutation equations and the method used for advancing time.
 
 :mod:`openmc.deplete` supports multiple time-integration methods for determining
 material compositions over time. Each method appears as a different class.
@@ -40,7 +40,7 @@ time::
 Note that the coupling between the reaction rate solver and the transmutation
 solver happens in-memory rather than by reading/writing files on disk. OpenMC
 has two categories of transport operators for obtaining transmutation reaction
-rates. 
+rates.
 
 .. _coupled-depletion:
 
@@ -65,7 +65,7 @@ Any material that contains a fissionable nuclide is depleted by default, but
 this can behavior can be changed with the :attr:`Material.depletable` attribute.
 
 .. important::
-   
+
    The volume must be specified for each material that is depleted by setting
    the :attr:`Material.volume` attribute. This is necessary in order to
    calculate the proper normalization of tally results based on the source rate.
@@ -198,8 +198,8 @@ Transport-independent depletion
 ===============================
 
 .. warning::
-   
-   This feature is still under heavy development and has yet to be rigorously 
+
+   This feature is still under heavy development and has yet to be rigorously
    verified. API changes and feature additions are possible and likely in
    the near future.
 
@@ -237,7 +237,7 @@ object::
                 'U236': 4.57e18,
                 'O16': 4.64e22,
                 'O17': 1.76e19}
-    volume = 0.5 
+    volume = 0.5
     op = openmc.deplete.IndependentOperator.from_nuclides(volume,
                                                           nuclides,
                                                           micro_xs,
@@ -250,8 +250,8 @@ transport-depletion calculation and follow the same steps from there.
 .. note::
 
    Ideally, one-group cross section data should be available for every
-   reaction in the depletion chain. If a nuclide that has a reaction 
-   associated with it in the depletion chain is present in the `nuclides` 
+   reaction in the depletion chain. If a nuclide that has a reaction
+   associated with it in the depletion chain is present in the `nuclides`
    parameter but not the cross section data, that reaction will not be
    simulated.
 
@@ -289,7 +289,7 @@ or from data arrays::
 .. important::
 
    Both :meth:`~openmc.deplete.MicroXS.from_csv()` and
-   :meth:`~openmc.deplete.MicroXS.from_array()` assume the cross section values 
+   :meth:`~openmc.deplete.MicroXS.from_array()` assume the cross section values
    provided are in barns by defualt, but have no way of verifying this. Make
    sure your cross sections are in the correct units before passing to a
    :class:`~openmc.deplete.IndependentOperator` object.
@@ -315,26 +315,28 @@ normalizing reaction rates:
    the cross-sections by the ``source-rate``.
 2. ``fission-q`` normalization, which uses the ``power`` or ``power_density``
    provided by the time integrator to obtain reaction rates by computing a value
-   for the flux based on this power. The general equation for the flux is 
+   for the flux based on this power. The general equation for the flux is
 
    .. math::
 
-      \phi = \frac{P}{V \cdot \sum_i (Q_i \cdot \sigma^f_i \cdot \n_i)}
+      \phi = \frac{P}{\sum\limits_i (Q_i \sigma^f_i N_i)}
 
-   where :math:`\sum_i` is the sum over all nuclides :math:`i`. This equation
-   makes the same assumptions and issues as discussed in
-   :ref:`energy-deposition`. Unfortunately, the proposed solution in that
+   where :math:`P` is the power, :math:`Q_i` is the fission Q value for nuclide
+   :math:`i`, :math:`\sigma_i^f` is the microscopic fission cross section for
+   nuclide :math:`i`, and :math:`N_i` is the number of atoms of nuclide
+   :math:`i`. This equation makes the same assumptions and issues as discussed
+   in :ref:`energy-deposition`. Unfortunately, the proposed solution in that
    section does not apply here since we are decoupled from transport code.
    However, there is a method to converge to a more accurate value for flux by
-   using substeps during time integration.
-   `This paper <https://doi.org/10.1016/j.anucene.2016.05.031>`_ provides a
-   good discussion of this method. 
+   using substeps during time integration. `This paper
+   <https://doi.org/10.1016/j.anucene.2016.05.031>`_ provides a good discussion
+   of this method.
 
 .. warning::
 
    The accuracy of results when using ``fission-q`` is entirely dependent on
    your depletion chain. Make sure it has sufficient data to resolve the
-   dynamics of your particular scenario. 
+   dynamics of your particular scenario.
 
 Multiple Materials
 ~~~~~~~~~~~~~~~~~~

examples/pincell_depletion/restart_depletion.py

@@ -43,7 +43,7 @@
 
 # Create depletion "operator"
 chain_file = 'chain_simple.xml'
-op = openmc.deplete.Operator(model, chain_file, previous_results)
+op = openmc.deplete.CoupledOperator(model, chain_file, previous_results)
 
 # Perform simulation using the predictor algorithm
 time_steps = [1.0, 1.0, 1.0, 1.0, 1.0]  # days

examples/pincell_depletion/run_depletion.py

@@ -88,7 +88,7 @@
 
 # Create depletion "operator"
 chain_file = 'chain_simple.xml'
-op = openmc.deplete.Operator(model, chain_file)
+op = openmc.deplete.CoupledOperator(model, chain_file)
 
 # Perform simulation using the predictor algorithm
 time_steps = [1.0, 1.0, 1.0, 1.0, 1.0]  # days

openmc/data/kalbach_mann.py

@@ -178,6 +178,8 @@ def kalbach_slope(energy_projectile, energy_emitted, za_projectile,
     energies are not calculated with the AWR number, but approximated with
     the number of mass instead.
 
+    .. versionadded:: 0.13.1
+
     Parameters
     ----------
     energy_projectile : float
@@ -593,6 +595,9 @@ def from_endf(cls, file_obj, za_emitted, za_target, projectile_mass):
         If the projectile is a neutron, the slope is calculated when it is
         not given explicitly.
 
+        .. versionchanged:: 0.13.1
+            Arguments changed to accommodate slope calculation
+
         Parameters
         ----------
         file_obj : file-like object

openmc/deplete/coupled_operator.py

@@ -56,6 +56,7 @@ def _find_cross_sections(model):
         )
     return cross_sections
 
+
 def _get_nuclides_with_data(cross_sections):
     """Loads cross_sections.xml file to find nuclides with neutron data
 
@@ -81,6 +82,7 @@ def _get_nuclides_with_data(cross_sections):
 
     return nuclides
 
+
 class CoupledOperator(OpenMCOperator):
     """Transport-coupled transport operator.
 
@@ -93,6 +95,9 @@ class CoupledOperator(OpenMCOperator):
         The geometry and settings parameters have been replaced with a
         model parameter that takes a :class:`~openmc.model.Model` object
 
+    .. versionchanged:: 0.13.1
+        Name changed from ``Operator`` to ``CoupledOperator``
+
     Parameters
     ----------
     model : openmc.model.Model
@@ -534,6 +539,7 @@ def finalize(self):
         if self.cleanup_when_done:
             openmc.lib.finalize()
 
+
 # Retain deprecated name for the time being
 def Operator(*args, **kwargs):
     # warn of name change