Merge branch 'develop' into mg_docs

docs/source/usersguide/input.rst

@@ -897,11 +897,19 @@ Each ``<surface>`` element can have the following attributes or sub-elements:
     *Default*: None
 
   :boundary:
-    The boundary condition for the surface. This can be "transmission",
-    "vacuum", or "reflective".
+     The boundary condition for the surface. This can be "transmission",
+     "vacuum", "reflective", or "periodic". Periodic boundary conditions can
+     only be applied to x-, y-, and z-planes. Only axis-aligned periodicity is
+     supported, i.e., x-planes can only be paired with x-planes. Specify which
+     planes are periodic and the code will automatically identify which planes
+     are paired together.
 
     *Default*: "transmission"
 
+  :periodic_surface_id:
+     If a periodic boundary condition is applied, this attribute identifies the
+     ``id`` of the corresponding periodic sufrace.
+
 The following quadratic surfaces can be modeled:
 
   :x-plane:
@@ -1033,6 +1041,20 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
 
         <cell fill="..." rotation="0 0 90" />
 
+    The rotation applied is an intrinsic rotation whose Tait-Bryan angles are
+    given as those specified about the x, y, and z axes respectively. That is to
+    say, if the angles are :math:`(\phi, \theta, \psi)`, then the rotation
+    matrix applied is :math:`R_z(\psi) R_y(\theta) R_x(\phi)` or
+
+    .. math::
+
+       \left [ \begin{array}{ccc} \cos\theta \cos\psi & -\cos\theta \sin\psi +
+       \sin\phi \sin\theta \cos\psi & \sin\phi \sin\psi + \cos\phi \sin\theta
+       \cos\psi \\ \cos\theta \sin\psi & \cos\phi \cos\psi + \sin\phi \sin\theta
+       \sin\psi & -\sin\phi \cos\psi + \cos\phi \sin\theta \sin\psi \\
+       -\sin\theta & \sin\phi \cos\theta & \cos\phi \cos\theta \end{array}
+       \right ]
+
     *Default*: None
 
   :translation:
@@ -1215,11 +1237,10 @@ Each ``material`` element can have the following attributes or sub-elements:
     An element with attributes/sub-elements called ``value`` and ``units``. The
     ``value`` attribute is the numeric value of the density while the ``units``
     can be "g/cm3", "kg/m3", "atom/b-cm", "atom/cm3", or "sum". The "sum" unit
-    indicates that values appearing in ``ao`` attributes for ``<nuclide>`` and
-    ``<element>`` sub-elements are to be interpreted as nuclide/element
-    densities in atom/b-cm, and the total density of the material is taken as
-    the sum of all nuclides/elements. The "sum" option cannot be used in
-    conjunction with weight percents.  The "macro" unit is used with
+    indicates that values appearing in ``ao`` or ``wo`` attributes for ``<nuclide>``
+    and ``<element>`` sub-elements are to be interpreted as absolute nuclide/element
+    densities in atom/b-cm or g/cm3, and the total density of the material is
+    taken as the sum of all nuclides/elements. The "macro" unit is used with
     a ``macroscopic`` quantity to indicate that the density is already included
     in the library and thus not needed here.  However, if a value is provided
     for the ``value``, then this is treated as a number density multiplier on
@@ -1258,6 +1279,9 @@ Each ``material`` element can have the following attributes or sub-elements:
 
     *Default*: None
 
+    .. note:: The ``scattering`` attribute/sub-element is not used in the
+              multi-group :ref:`energy_mode`.
+
   :element:
 
     Specifies that a natural element is present in the material. The natural
@@ -1293,6 +1317,9 @@ Each ``material`` element can have the following attributes or sub-elements:
 
     *Default*: None
 
+    .. note:: The ``scattering`` attribute/sub-element is not used in the
+              multi-group :ref:`energy_mode`.
+
   :sab:
     Associates an S(a,b) table with the material. This element has
     attributes/sub-elements called ``name`` and ``xs``. The ``name`` attribute
@@ -1301,6 +1328,8 @@ Each ``material`` element can have the following attributes or sub-elements:
 
     *Default*: None
 
+    .. note:: This element is not used in the multi-group :ref:`energy_mode`.
+
   :macroscopic:
     The ``macroscopic`` element is similar to the ``nuclide`` element, but,
     recognizes that some multi-group libraries may be providing material
@@ -1569,7 +1598,8 @@ The ``<tally>`` element accepts the following sub-elements:
         |Score                 | Description                                       |
         +======================+===================================================+
         |absorption            |Total absorption rate. This accounts for all       |
-        |                      |reactions which do not produce secondary neutrons. |
+        |                      |reactions which do not produce secondary neutrons  |
+        |                      |as well as fission.                                |
         +----------------------+---------------------------------------------------+
         |elastic               |Elastic scattering reaction rate.                  |
         +----------------------+---------------------------------------------------+

docs/source/usersguide/mgxs_library.rst

@@ -8,10 +8,10 @@ OpenMC can be run in continuous-energy mode or multi-group mode, provided the
 nuclear data is available.  In continuous-energy mode, the
 ``cross_sections.xml`` file contains necessary meta-data for each data set,
 including the name and a file system location where the complete library
-can be found.  In multi-group mode, this ``cross_sections.xml`` file contains
+can be found.  In multi-group mode, this ``mgxs.xml`` file contains
 this same meta-data describing the nuclide or material, but also contains the
 group-wise nuclear data.  This portion of the manual describes the format of
-the multi-group data library required to be used in the ``cross_sections.xml``
+the multi-group data library required to be used in the ``mgxs.xml``
 file.
 
 Similar to the other input file types, the multi-group library is provided in
@@ -22,9 +22,9 @@ materials.
 
 .. _XML: http://www.w3.org/XML/
 
-------------------------------------------------
-MGXS Library Specification -- cross_sections.xml
-------------------------------------------------
+--------------------------------------
+MGXS Library Specification -- mgxs.xml
+--------------------------------------
 
 The multi-group library meta-data is contained within the groups_,
 group_structure_, and inverse_velocities_ elements.
@@ -33,7 +33,7 @@ The actual multi-group data itself is contained within the xsdata_ element.
 .. _groups:
 
 ``<groups>`` Element
-----------------------------------
+--------------------
 
 The ``<groups>`` element has no attributes and simply provides the number of
 energy groups contained within the library.
@@ -172,7 +172,7 @@ attributes/sub-elements required to describe the meta-data:
     during the scattering process.  Specifically, the options are to either
     convert the Legendre expansion to a tabular representation or leave it as
     a set of Legendre coefficients.  Converting to a tabular representation will
-    cost memory but is likely to decrease runtime compared to leaving as a
+    cost memory but can allow for a decrease in runtime compared to leaving as a
     set of Legendre coefficients.  This element has the following
     attributes/sub-elements:
 
@@ -181,7 +181,7 @@ attributes/sub-elements required to describe the meta-data:
       tabular format should be performed or not.  A value of "true" means
       the conversion should be performed, "false" means it should not.
 
-      *Default*: "true"
+      *Default*: "false"
 
     :num_points:
       If the conversion is to take place the number of tabular points is

docs/source/usersguide/output/statepoint.rst

@@ -248,7 +248,7 @@ if run_mode == 'k-eigenvalue':
     Accumulated sum and sum-of-squares for each global tally. The compound type
     has fields named ``sum`` and ``sum_sq``.
 
-**tallies_present** (*int*)
+**/tallies_present** (*int*)
 
     Flag indicated if tallies are present in the file.
 
@@ -260,3 +260,69 @@ if (run_mode == 'k-eigenvalue' and source_present > 0)
         ``wgt``, ``xyz``, ``uvw``, ``E``, ``g``, and ``delayed_group``, which
         represent the weight, position, direction, energy, energy group, and
         delayed_group of the source particle, respectively.
+
+**/runtime/total initialization** (*double*)
+
+    Time (in seconds on the master process) spent reading inputs, allocating
+    arrays, etc.
+
+**/runtime/reading cross sections** (*double*)
+
+    Time (in seconds on the master process) spent loading cross section
+    libraries (this is a subset of initialization).
+
+**/runtime/simulation** (*double*)
+
+    Time (in seconds on the master process) spent between initialization and
+    finalization.
+
+**/runtime/transport** (*double*)
+
+    Time (in seconds on the master process) spent transporting particles.
+
+**/runtime/inactive batches** (*double*)
+
+    Time (in seconds on the master process) spent in the inactive batches
+    (including non-transport activities like communcating sites).
+
+**/runtime/active batches** (*double*)
+
+    Time (in seconds on the master process) spent in the active batches
+    (including non-transport activities like communicating sites).
+
+**/runtime/synchronizing fission bank** (*double*)
+
+    Time (in seconds on the master process) spent sampling source particles
+    from fission sites and communicating them to other processes for load
+    balancing.
+
+**/runtime/sampling source sites** (*double*)
+
+    Time (in seconds on the master process) spent sampling source particles
+    from fission sites.
+
+**/runtime/SEND-RECV source sites** (*double*)
+
+    Time (in seconds on the master process) spent communicating source sites
+    between processes for load balancing.
+
+**/runtime/accumulating tallies** (*double*)
+
+    Time (in seconds on the master process) spent communicating tally results
+    and evaluating their statistics.
+
+**/runtime/CMFD** (*double*)
+
+    Time (in seconds on the master process) spent evaluating CMFD.
+
+**/runtime/CMFD building matrices** (*double*)
+
+    Time (in seconds on the master process) spent buliding CMFD matrices.
+
+**/runtime/CMFD solving matrices** (*double*)
+
+    Time (in seconds on the master process) spent solving CMFD matrices.
+
+**/runtime/total** (*double*)
+
+    Total time spent (in seconds on the master process) in the program.

docs/source/usersguide/output/summary.rst

@@ -293,6 +293,13 @@ The current revision of the summary file format is 1.
 
     Filter offset (used for distribcell filter).
 
+**/tallies/tally <uid>/filter <j>/paths** (*char[][]*)
+
+    The paths traversed through the CSG tree to reach each distribcell
+    instance (for 'distribcell' filters only). This consists of the integer
+    IDs for each universe, cell and lattice delimited by '->'. Each lattice
+    cell is specified by its (x,y) or (x,y,z) indices.
+
 **/tallies/tally <uid>/filter <j>/n_bins** (*int*)
 
     Number of bins for the j-th filter.

docs/source/usersguide/processing.rst

@@ -196,10 +196,10 @@ Data Extraction
 
 A great deal of information is available in statepoint files (See
 :ref:`usersguide_statepoint`), all of which is accessible through the Python
-API. The ``openmc.statepoint`` module (see :ref:`pythonapi_statepoint`) provides
-a class to load statepoints and access data as requested; it is used in many of
-the provided plotting utilities, OpenMC's regression test suite, and can be used
-in user-created scripts to carry out manipulations of the data.
+API. The :class:`openmc.StatePoint` class can load statepoints and access data
+as requested; it is used in many of the provided plotting utilities, OpenMC's
+regression test suite, and can be used in user-created scripts to carry out
+manipulations of the data.
 
 An :ref:`example IPython notebook <notebook_post_processing>` demonstrates how
 to extract data from a statepoint using the Python API.