Respond to @nelsonag comments on #852

docs/source/devguide/workflow.rst

@@ -105,15 +105,20 @@ in the tests directory. We recommend to developers to test their branches
 before submitting a formal pull request using gfortran and Intel compilers
 if available.
 
-The test suite is designed to integrate with cmake using ctest_.
-It is configured to run with cross sections from NNDC_. To
-download these cross sections please do the following:
+The test suite is designed to integrate with cmake using ctest_.  It is
+configured to run with cross sections from NNDC_ augmented with 0 K elastic
+scattering data for select nuclides as well as multipole data. To download the
+proper data, run the following commands:
 
 .. code-block:: sh
 
-    cd ../scripts
-    ./openmc-get-nndc-data
-    export OPENMC_CROSS_SECTIONS=<path_to_data_folder>/nndc_hdf5/cross_sections.xml
+    wget -O nndc_hdf5.tar.xz $(cat <openmc_root>/.travis.yml | grep anl.box | awk '{print $2}')
+    tar xJvf nndc_hdf5.tar.xz
+    export OPENMC_CROSS_SECTIONS=$(pwd)/nndc_hdf5/cross_sections.xml
+
+    git clone --branch=master git://github.com/smharper/windowed_multipole_library.git wmp_lib
+    tar xzvf wmp_lib/multipole_lib.tar.gz
+    export OPENMC_MULTIPOLE_LIBRARY=$(pwd)/multipole_lib
 
 The test suite can be run on an already existing build using:
 

docs/source/usersguide/cross_sections.rst

@@ -199,6 +199,30 @@ OpenMC.
           etc. For a more thorough overview of the capabilities of this class,
           see the :ref:`notebook_nuclear_data` example notebook.
 
+Enabling Resonance Scattering Treatments
+----------------------------------------
+
+In order for OpenMC to correctly treat elastic scattering in heavy nuclides
+where low-lying resonances might be present (see
+:ref:`energy_dependent_xs_model`), the elastic scattering cross section at 0 K
+must be present. To add the 0 K elastic scattering cross section to existing
+:class:`IncidentNeutron` instance, you can use the
+:meth:`IncidentNeutron.add_elastic_0K_from_endf` method which requires an ENDF
+file for the nuclide you are modifying::
+
+  u238 = openmc.data.IncidentNeutron.from_hdf5('U238.h5')
+  u238.add_elastic_0K_from_endf('n-092_U_238.endf')
+  u238.export_to_hdf5('U238_with_0K.h5')
+
+With 0 K elastic scattering data present, you can turn on a resonance scattering
+method using :attr:`Settings.resonance_scattering`.
+
+.. note:: The process of reconstructing resonances and generating tabulated 0 K
+          cross sections can be computationally expensive, especially for
+          nuclides like U-238 where thousands of resonances are present. Thus,
+          running the :meth:`IncidentNeutron.add_elastic_0K_from_endf` method
+          may take several minutes to complete.
+
 -----------------------
 Windowed Multipole Data
 -----------------------

openmc/examples.py

@@ -7,6 +7,11 @@ import openmc.model
 def pwr_pin_cell():
     """Create a PWR pin-cell model.
 
+    This model is a single fuel pin with 2.4 w/o enriched UO2 corresponding to a
+    beginning-of-cycle condition and borated water. The specifications are from
+    the `BEAVRS <http://crpg.mit.edu/research/beavrs>`_ benchmark. Note that the
+    number of particles/batches is initially set very low for testing purposes.
+
     Returns
     -------
     model : openmc.model.Model
@@ -85,7 +90,8 @@ def pwr_core():
 
     This model is the OECD/NEA Monte Carlo Performance benchmark which is a
     grossly simplified pressurized water reactor (PWR) with 241 fuel
-    assemblies.
+    assemblies. Note that the number of particles/batches is initially set very
+    low for testing purposes.
 
     Returns
     -------
@@ -428,6 +434,11 @@ def pwr_core():
 def pwr_assembly():
     """Create a PWR assembly model.
 
+    This model is a reflected 17x17 fuel assembly from the the `BEAVRS
+    <http://crpg.mit.edu/research/beavrs>`_ benchmark. The fuel is 2.4 w/o
+    enriched UO2 corresponding to a beginning-of-cycle condition. Note that the
+    number of particles/batches is initially set very low for testing purposes.
+
     Returns
     -------
     model : openmc.model.Model
@@ -538,8 +549,12 @@ def slab_mg(reps=None, as_macro=True):
     Parameters
     ----------
     reps : list, optional
-        List of angular representations. Items can be 'ang', 'ang_mu', 'iso', or
-        'iso_mu'.
+        List of angular representations. Each item corresponds to materials and
+        dictates the angular representation of the multi-group cross
+        sections---isotropic ('iso') or angle-dependent ('ang'), and if Legendre
+        scattering or tabular scattering ('mu') is used. Thus, items can be
+        'ang', 'ang_mu', 'iso', or 'iso_mu'.
+
     as_macro : bool, optional
         Whether :class:`openmc.Macroscopic` is used
 

openmc/model/model.py

@@ -10,7 +10,10 @@ class Model(object):
     This class can be used to store instances of :class:`openmc.Geometry`,
     :class:`openmc.Materials`, :class:`openmc.Settings`,
     :class:`openmc.Tallies`, :class:`openmc.Plots`, and :class:`openmc.CMFD`,
-    thus making a complete model.
+    thus making a complete model. The :meth:`Model.export_to_xml` method will
+    export XML files for all attributes that have been set. If the
+    :meth:`Model.materials` attribute is not set, it will attempt to create a
+    ``materials.xml`` file based on all materials appearing in the geometry.
 
     Parameters
     ----------
@@ -50,9 +53,8 @@ class Model(object):
         self.materials = openmc.Materials()
         self.settings = openmc.Settings()
         self.cmfd = cmfd
-
-        self._tallies = openmc.Tallies()
-        self._plots = openmc.Plots()
+        self.tallies = openmc.Tallies()
+        self.plots = openmc.Plots()
 
         if geometry is not None:
             self.geometry = geometry