Release of version 0.15.0 (#3050)

Author: paulromano

CMakeLists.txt

@@ -3,8 +3,8 @@ project(openmc C CXX)
 
 # Set version numbers
 set(OPENMC_VERSION_MAJOR 0)
-set(OPENMC_VERSION_MINOR 14)
-set(OPENMC_VERSION_RELEASE 1)
+set(OPENMC_VERSION_MINOR 15)
+set(OPENMC_VERSION_RELEASE 0)
 set(OPENMC_VERSION ${OPENMC_VERSION_MAJOR}.${OPENMC_VERSION_MINOR}.${OPENMC_VERSION_RELEASE})
 configure_file(include/openmc/version.h.in "${CMAKE_BINARY_DIR}/include/openmc/version.h" @ONLY)
 

docs/source/conf.py

@@ -69,9 +69,9 @@
 # built documents.
 #
 # The short X.Y version.
-version = "0.14"
+version = "0.15"
 # The full version, including alpha/beta/rc tags.
-release = "0.14.1-dev"
+release = "0.15.0"
 
 # The language for content autogenerated by Sphinx. Refer to documentation
 # for a list of supported languages.

docs/source/io_formats/settings.rst

@@ -284,9 +284,9 @@ then, OpenMC will only use up to the :math:`P_1` data.
   .. note:: This element is not used in the continuous-energy
     :ref:`energy_mode`.
 
-------------------------
+--------------------------------
 ``<max_history_splits>`` Element
-------------------------
+--------------------------------
 
 The ``<max_history_splits>`` element indicates the number of times a particle can split during a history.
 

docs/source/methods/random_ray.rst

@@ -100,11 +100,11 @@ terms on the right hand side.
 .. math::
     :label: transport
 
-    \begin{align*}
+    \begin{aligned}
     \mathbf{\Omega} \cdot \mathbf{\nabla} \psi(\mathbf{r},\mathbf{\Omega},E) & + \Sigma_t(\mathbf{r},E) \psi(\mathbf{r},\mathbf{\Omega},E) = \\
     & \int_0^\infty d E^\prime \int_{4\pi} d \Omega^{\prime} \Sigma_s(\mathbf{r},\mathbf{\Omega}^\prime \rightarrow \mathbf{\Omega}, E^\prime \rightarrow E) \psi(\mathbf{r},\mathbf{\Omega}^\prime, E^\prime) \\
     & + \frac{\chi(\mathbf{r}, E)}{4\pi k_{eff}} \int_0^\infty dE^\prime \nu \Sigma_f(\mathbf{r},E^\prime) \int_{4\pi}d \Omega^\prime \psi(\mathbf{r},\mathbf{\Omega}^\prime,E^\prime)
-    \end{align*}
+    \end{aligned}
 
 In Equation :eq:`transport`, :math:`\psi` is the angular neutron flux. This
 parameter represents the total distance traveled by all neutrons in a particular

docs/source/releasenotes/0.15.0.rst

@@ -7,6 +7,7 @@ Release Notes
 .. toctree::
   :maxdepth: 1
 
+  0.15.0
   0.14.0
   0.13.3
   0.13.2

docs/source/releasenotes/index.rst

@@ -114,8 +114,9 @@
 
 # Create an initial uniform spatial source distribution over fissionable zones
 bounds = [-1, -1, -1, 1, 1, 1]
-uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
-settings_file.source = openmc.IndependentSource(space=uniform_dist)
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
+settings_file.source = openmc.IndependentSource(
+    space=uniform_dist, constraints={'fissionable': True})
 
 settings_file.keff_trigger = {'type' : 'std_dev', 'threshold' : 5E-4}
 settings_file.trigger_active = True

examples/lattice/hexagonal/build_xml.py

@@ -124,8 +124,9 @@
 
 # Create an initial uniform spatial source distribution over fissionable zones
 bounds = [-1, -1, -1, 1, 1, 1]
-uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
-settings_file.source = openmc.IndependentSource(space=uniform_dist)
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
+settings_file.source = openmc.IndependentSource(
+    space=uniform_dist, constraints={'fissionable': True})
 
 settings_file.export_to_xml()
 

examples/lattice/nested/build_xml.py

@@ -115,8 +115,9 @@
 
 # Create an initial uniform spatial source distribution over fissionable zones
 bounds = [-1, -1, -1, 1, 1, 1]
-uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
-settings_file.source = openmc.IndependentSource(space=uniform_dist)
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
+settings_file.source = openmc.IndependentSource(
+    space=uniform_dist, constraints={'fissionable': True})
 
 settings_file.trigger_active = True
 settings_file.trigger_max_batches = 100

examples/lattice/simple/build_xml.py

@@ -67,8 +67,9 @@
 # Create an initial uniform spatial source distribution over fissionable zones
 lower_left = (-pitch/2, -pitch/2, -1)
 upper_right = (pitch/2, pitch/2, 1)
-uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
-settings.source = openmc.IndependentSource(space=uniform_dist)
+uniform_dist = openmc.stats.Box(lower_left, upper_right)
+settings.source = openmc.IndependentSource(
+    space=uniform_dist, constraints={'fissionable': True})
 
 # For source convergence checks, add a mesh that can be used to calculate the
 # Shannon entropy