Merge branch 'develop' into mixed_ncrystal_pr

docs/source/conf.py

@@ -69,9 +69,9 @@ copyright = '2011-2022, Massachusetts Institute of Technology, UChicago Argonne
 # built documents.
 #
 # The short X.Y version.
-version = "0.14"
+version = "0.13"
 # The full version, including alpha/beta/rc tags.
-release = "0.14.0"
+release = "0.13.3"
 
 # The language for content autogenerated by Sphinx. Refer to documentation
 # for a list of supported languages.
@@ -247,7 +247,7 @@ napoleon_use_ivar = True
 intersphinx_mapping = {
     'python': ('https://docs.python.org/3', None),
     'numpy': ('https://numpy.org/doc/stable/', None),
-    'scipy': ('https://docs.scipy.org/doc/scipy/reference', None),
+    'scipy': ('https://docs.scipy.org/doc/scipy/', None),
     'pandas': ('https://pandas.pydata.org/pandas-docs/stable/', None),
-    'matplotlib': ('https://matplotlib.org/', None)
+    'matplotlib': ('https://matplotlib.org/stable/', None)
 }

docs/source/devguide/tests.rst

@@ -31,9 +31,13 @@ Prerequisites
 
 - The test suite requires a specific set of cross section data in order for
   tests to pass. A download URL for the data that OpenMC expects can be found
-  within ``tools/ci/download-xs.sh``.
+  within ``tools/ci/download-xs.sh``. Once the tarball is downloaded and
+  unpacked, set the :envvar:`OPENMC_CROSS_SECTIONS` environment variable to the
+  path of the ``cross_sections.xml`` file within the unpacked data.
 - In addition to the HDF5 data, some tests rely on ENDF files. A download URL
-  for those can also be found in ``tools/ci/download-xs.sh``.
+  for those can also be found in ``tools/ci/download-xs.sh``. Once the tarball
+  is downloaded and unpacked, set the :envvar:`OPENMC_ENDF_DATA` environment
+  variable to the top-level directory of the unpacked tarball.
 - Some tests require `NJOY <https://www.njoy21.io/NJOY2016>`_ to preprocess
   cross section data. The test suite assumes that you have an ``njoy``
   executable available on your :envvar:`PATH`.

docs/source/io_formats/settings.rst

@@ -732,6 +732,14 @@ attributes/sub-elements:
 
     *Default*: false
 
+  :mcpl:
+    If this element is set to "true", the source point file containing the
+    source bank will be written as an MCPL_ file name ``source.mcpl`` instead of
+    an HDF5 file. This option is only applicable if the ``<separate>`` element
+    is set to true.
+
+    *Default*: false
+
 ------------------------------
 ``<surf_source_read>`` Element
 ------------------------------
@@ -767,6 +775,16 @@ certain surfaces and write out the source bank in a separate file called
 
     *Default*: None
 
+  :mcpl:
+    An optional boolean which indicates if the banked particles should be
+    written to a file in the MCPL_-format instead of the native HDF5-based
+    format. If activated the output file name is changed to
+    ``surface_source.mcpl``.
+
+    *Default*: false
+
+    .. _MCPL: https://mctools.github.io/mcpl/mcpl.pdf
+
 ------------------------------
 ``<survival_biasing>`` Element
 ------------------------------
@@ -832,7 +850,9 @@ cell, the nearest temperature at which cross sections are given is to be
 applied, within a given tolerance (see :ref:`temperature_tolerance`). A value of
 "interpolation" indicates that cross sections are to be linear-linear
 interpolated between temperatures at which nuclear data are present (see
-:ref:`temperature_treatment`).
+:ref:`temperature_treatment`). With the "interpolation" method, temperatures
+outside of the bounds of the nuclear data may be accepted, provided they still
+fall within the tolerance (see :ref:`temperature_tolerance`).
 
   *Default*: "nearest"
 
@@ -871,7 +891,12 @@ The ``<temperature_tolerance>`` element specifies a tolerance in Kelvin that is
 to be applied when the "nearest" temperature method is used. For example, if a
 cell temperature is 340 K and the tolerance is 15 K, then the closest
 temperature in the range of 325 K to 355 K will be used to evaluate cross
-sections.
+sections. If the ``<temperature_method>`` is "interpolation", the tolerance
+specified applies to cell temperatures outside of the data bounds. For example,
+if a cell is specified at 695K, a tolerance of 15K and data is only available
+at 700K and 1000K, the cell's cross sections will be evaluated at 700K, since
+the desired temperature of 695K is within the tolerance of the actual data
+despite not being bounded on both sides.
 
   *Default*: 10 K
 

docs/source/io_formats/statepoint.rst

@@ -109,7 +109,8 @@ The current version of the statepoint file format is 17.0.
            - **y** (*double[]*) -- Interpolant values for energyfunction
              interpolation. Only used for 'energyfunction' filters.
 
-             :Attributes: - **interpolation** (*int*) -- Interpolation type. Only used for
+             :Attributes:
+                          - **interpolation** (*int*) -- Interpolation type. Only used for
                             'energyfunction' filters.
 
 **/tallies/derivatives/derivative <id>/**

docs/source/pythonapi/base.rst

@@ -120,7 +120,7 @@ Constructing Tallies
    openmc.MaterialFilter
    openmc.CellFilter
    openmc.CellFromFilter
-   openmc.CellbornFilter
+   openmc.CellBornFilter
    openmc.CellInstanceFilter
    openmc.CollisionFilter
    openmc.SurfaceFilter

docs/source/pythonapi/data.rst

@@ -63,9 +63,10 @@ Core Functions
     atomic_weight
     combine_distributions
     decay_constant
+    decay_energy
     decay_photon_energy
     dose_coefficients
-    gnd_name
+    gnds_name
     half_life
     isotopes
     kalbach_slope

docs/source/pythonapi/model.rst

@@ -31,6 +31,7 @@ Composite Surfaces
    openmc.model.XConeOneSided
    openmc.model.YConeOneSided
    openmc.model.ZConeOneSided
+   openmc.model.Polygon
 
 TRISO Fuel Modeling
 -------------------

docs/source/pythonapi/stats.rst

@@ -22,6 +22,12 @@ Univariate Probability Distributions
    openmc.stats.Legendre
    openmc.stats.Mixture
    openmc.stats.Normal
+
+.. autosummary::
+   :toctree: generated
+   :nosignatures:
+   :template: myfunction.rst
+
    openmc.stats.muir
 
 Angular Distributions

docs/source/releasenotes/0.13.2.rst

@@ -0,0 +1,99 @@
+====================
+What's New in 0.13.2
+====================
+
+.. currentmodule:: openmc
+
+-------
+Summary
+-------
+
+This release of OpenMC includes several bug fixes, performance improvements for
+complex geometries and depletion simulations, and other general enhancements.
+Notably, a capability has been added to compute the photon spectra from decay of
+unstable nuclides. Alongside that, a new :data:`openmc.config` configuration
+variable has been introduced that allows easier configuration of data sources.
+Additionally, users can now perform cell or material rejection when sampling
+external source distributions. Finally,
+
+------------------------------------
+Compatibility Notes and Deprecations
+------------------------------------
+
+- If you are building against libMesh for unstructured mesh tally support,
+  version 1.6 or higher is now required.
+- The ``openmc.stats.Muir`` class has been replaced by a
+  :func:`openmc.stats.muir` function that returns an instance of
+  :class:`openmc.stats.Normal`.
+
+------------
+New Features
+------------
+
+- The :meth:`openmc.Material.get_nuclide_atom_densities` method now takes an
+  optional ``nuclide`` argument.
+- Functions/methods in the :mod:`openmc.deplete` module now accept times in
+  Julian years (``'a'``).
+- The :meth:`openmc.Universe.plot` method now allows a pre-existing axes object
+  to be passed in.
+- Performance optimization for geometries with many complex regions.
+- Performance optimization for depletion by avoiding deepcopies and caching
+  reaction rates.
+- The :class:`openmc.RegularMesh` class now has a
+  :meth:`~openmc.RegularMesh.from_domain` classmethod.
+- The :class:`openmc.CylindricalMesh` class now has a
+  :meth:`~openmc.CylindricalMesh.from_domain` classmethod.
+- Improved method to condense diffusion coefficients from the :mod:`openmc.mgxs`
+  module.
+- A new :data:`openmc.config` configuration variable has been introduced that
+  allows data sources to be specified at runtime or via environment variables.
+- The :class:`openmc.EnergyFunctionFilter` class now supports multiple
+  interpolation schemes, not just linear-linear interpolation.
+- The :class:`openmc.DAGMCUniverse` class now has ``material_names``,
+  ``n_cells``, and ``n_surfaces`` attributes.
+- A new :func:`openmc.data.decay_photon_energy` function has been added that
+  returns the energy spectrum of photons emitted from the decay of an unstable
+  nuclide.
+- The :class:`openmc.Material` class also has a new
+  :attr:`~openmc.Material.decay_photon_energy` attribute that gives the decay
+  photon energy spectrum from the material based on its constituent nuclides.
+- The :class:`openmc.deplete.StepResult` now has a
+  :meth:`~openmc.deplete.StepResult.get_material` method.
+- The :class:`openmc.Source` class now takes a ``domains`` argument that
+  specifies a list of cells, materials, or universes that is used to reject
+  source sites (i.e., if the sampled sites are not within the specified domain,
+  they are rejected).
+
+---------
+Bug Fixes
+---------
+
+- `Delay call to Tally::set_strides <https://github.com/openmc-dev/openmc/pull/2183>`_
+- `Fix reading reference direction from XML for angular distributions <https://github.com/openmc-dev/openmc/pull/2204>`_
+- `Fix erroneous behavior in Material.add_components <https://github.com/openmc-dev/openmc/pull/2205>`_
+- `Fix reading thermal elastic data from ACE <https://github.com/openmc-dev/openmc/pull/2226>`_
+- `Fix reading source file with time attribute <https://github.com/openmc-dev/openmc/pull/2228>`_
+- `Fix conversion of multiple thermal scattering data files from ACE <https://github.com/openmc-dev/openmc/pull/2232>`_
+- `Fix reading values from wwinp file <https://github.com/openmc-dev/openmc/pull/2242>`_
+- `Handle possibility of .ppm file in Universe.plot <https://github.com/openmc-dev/openmc/pull/2251>`_
+- `Update volume calc types to mitigate overflow issues <https://github.com/openmc-dev/openmc/pull/2270>`_
+
+------------
+Contributors
+------------
+
+- `Lewis Gross <https://github.com/lewisgross1296>`_
+- `Andrew Johnson <https://github.com/drewejohnson>`_
+- `Miriam Kreher <https://github.com/mkreher13>`_
+- `James Logan <https://github.com/jlogan03>`_
+- `Jose Ignacio Marquez Damien <https://github.com/marquezj>`_
+- `Josh May <https://github.com/joshmay1>`_
+- `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>`_
+- `Patrick Shriwise <https://github.com/pshriwise>`_
+- `Jonathan Shimwell <https://github.com/Shimwell>`_
+- `Olek Yardas <https://github.com/yardasol>`_

docs/source/releasenotes/index.rst

@@ -7,6 +7,7 @@ Release Notes
 .. toctree::
   :maxdepth: 1
 
+  0.13.2
   0.13.1
   0.13.0
   0.12.2

docs/source/usersguide/scripts.rst

@@ -157,9 +157,9 @@ geometry.xml
   added. Any 'surfaces' attributes/elements on a cell will be renamed 'region'.
 
 materials.xml
-  Nuclide names will be changed from ACE aliases (e.g., Am-242m) to HDF5/GND
+  Nuclide names will be changed from ACE aliases (e.g., Am-242m) to HDF5/GNDS
   names (e.g., Am242_m1). Thermal scattering table names will be changed from
-  ACE aliases (e.g., HH2O) to HDF5/GND names (e.g., c_H_in_H2O).
+  ACE aliases (e.g., HH2O) to HDF5/GNDS names (e.g., c_H_in_H2O).
 
 ----------------------
 ``openmc-update-mgxs``

docs/source/usersguide/settings.rst

@@ -474,7 +474,6 @@ selected::
    Some features related to photon transport are not currently implemented,
    including:
 
-     * Tallying photon energy deposition.
      * Generating a photon source from a neutron calculation that can be used
        for a later fixed source photon calculation.
      * Photoneutron reactions.

docs/source/usersguide/tallies.rst

@@ -269,9 +269,8 @@ The following tables show all valid scores:
     |heating               |Total nuclear heating in units of eV per source    |
     |                      |particle. For neutrons, this corresponds to MT=301 |
     |                      |produced by NJOY's HEATR module while for photons, |
-    |                      |this is tallied from either direct photon energy   |
-    |                      |deposition (analog estimator) or pre-generated     |
-    |                      |photon heating number. See :ref:`methods_heating`  |
+    |                      |this is tallied from direct photon energy          |
+    |                      |deposition. See :ref:`methods_heating`.            |
     +----------------------+---------------------------------------------------+
     |heating-local         |Total nuclear heating in units of eV per source    |
     |                      |particle assuming energy from secondary photons is |