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Release of version 0.15.0 (#3050)
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lines changed

‎CMakeLists.txt

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Original file line numberDiff line numberDiff line change
@@ -3,8 +3,8 @@ project(openmc C CXX)
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44
# Set version numbers
55
set(OPENMC_VERSION_MAJOR 0)
6-
set(OPENMC_VERSION_MINOR 14)
7-
set(OPENMC_VERSION_RELEASE 1)
6+
set(OPENMC_VERSION_MINOR 15)
7+
set(OPENMC_VERSION_RELEASE 0)
88
set(OPENMC_VERSION ${OPENMC_VERSION_MAJOR}.${OPENMC_VERSION_MINOR}.${OPENMC_VERSION_RELEASE})
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configure_file(include/openmc/version.h.in "${CMAKE_BINARY_DIR}/include/openmc/version.h" @ONLY)
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‎docs/source/conf.py

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@@ -69,9 +69,9 @@
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# built documents.
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#
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# The short X.Y version.
72-
version = "0.14"
72+
version = "0.15"
7373
# The full version, including alpha/beta/rc tags.
74-
release = "0.14.1-dev"
74+
release = "0.15.0"
7575

7676
# The language for content autogenerated by Sphinx. Refer to documentation
7777
# for a list of supported languages.

‎docs/source/io_formats/settings.rst

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@@ -284,9 +284,9 @@ then, OpenMC will only use up to the :math:`P_1` data.
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.. note:: This element is not used in the continuous-energy
285285
:ref:`energy_mode`.
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287-
------------------------
287+
--------------------------------
288288
``<max_history_splits>`` Element
289-
------------------------
289+
--------------------------------
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291291
The ``<max_history_splits>`` element indicates the number of times a particle can split during a history.
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‎docs/source/methods/random_ray.rst

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@@ -100,11 +100,11 @@ terms on the right hand side.
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.. math::
101101
:label: transport
102102
103-
\begin{align*}
103+
\begin{aligned}
104104
\mathbf{\Omega} \cdot \mathbf{\nabla} \psi(\mathbf{r},\mathbf{\Omega},E) & + \Sigma_t(\mathbf{r},E) \psi(\mathbf{r},\mathbf{\Omega},E) = \\
105105
& \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) \\
106106
& + \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)
107-
\end{align*}
107+
\end{aligned}
108108
109109
In Equation :eq:`transport`, :math:`\psi` is the angular neutron flux. This
110110
parameter represents the total distance traveled by all neutrons in a particular

‎docs/source/releasenotes/0.15.0.rst

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‎docs/source/releasenotes/index.rst

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@@ -7,6 +7,7 @@ Release Notes
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.. toctree::
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:maxdepth: 1
99

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0.15.0
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0.14.0
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0.13.3
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0.13.2

‎examples/lattice/hexagonal/build_xml.py

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@@ -114,8 +114,9 @@
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# Create an initial uniform spatial source distribution over fissionable zones
116116
bounds = [-1, -1, -1, 1, 1, 1]
117-
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
118-
settings_file.source = openmc.IndependentSource(space=uniform_dist)
117+
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
118+
settings_file.source = openmc.IndependentSource(
119+
space=uniform_dist, constraints={'fissionable': True})
119120

120121
settings_file.keff_trigger = {'type' : 'std_dev', 'threshold' : 5E-4}
121122
settings_file.trigger_active = True

‎examples/lattice/nested/build_xml.py

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@@ -124,8 +124,9 @@
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125125
# Create an initial uniform spatial source distribution over fissionable zones
126126
bounds = [-1, -1, -1, 1, 1, 1]
127-
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
128-
settings_file.source = openmc.IndependentSource(space=uniform_dist)
127+
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
128+
settings_file.source = openmc.IndependentSource(
129+
space=uniform_dist, constraints={'fissionable': True})
129130

130131
settings_file.export_to_xml()
131132

‎examples/lattice/simple/build_xml.py

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@@ -115,8 +115,9 @@
115115

116116
# Create an initial uniform spatial source distribution over fissionable zones
117117
bounds = [-1, -1, -1, 1, 1, 1]
118-
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
119-
settings_file.source = openmc.IndependentSource(space=uniform_dist)
118+
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
119+
settings_file.source = openmc.IndependentSource(
120+
space=uniform_dist, constraints={'fissionable': True})
120121

121122
settings_file.trigger_active = True
122123
settings_file.trigger_max_batches = 100

‎examples/pincell/build_xml.py

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@@ -67,8 +67,9 @@
6767
# Create an initial uniform spatial source distribution over fissionable zones
6868
lower_left = (-pitch/2, -pitch/2, -1)
6969
upper_right = (pitch/2, pitch/2, 1)
70-
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
71-
settings.source = openmc.IndependentSource(space=uniform_dist)
70+
uniform_dist = openmc.stats.Box(lower_left, upper_right)
71+
settings.source = openmc.IndependentSource(
72+
space=uniform_dist, constraints={'fissionable': True})
7273

7374
# For source convergence checks, add a mesh that can be used to calculate the
7475
# Shannon entropy

‎examples/pincell_multigroup/build_xml.py

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@@ -113,8 +113,9 @@
113113
# Create an initial uniform spatial source distribution over fissionable zones
114114
lower_left = (-pitch/2, -pitch/2, -1)
115115
upper_right = (pitch/2, pitch/2, 1)
116-
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
117-
settings.source = openmc.IndependentSource(space=uniform_dist)
116+
uniform_dist = openmc.stats.Box(lower_left, upper_right)
117+
settings.source = openmc.IndependentSource(
118+
space=uniform_dist, constraints={'fissionable': True})
118119
settings.export_to_xml()
119120

120121
###############################################################################

‎include/openmc/version.h.in

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@@ -10,7 +10,7 @@ namespace openmc {
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constexpr int VERSION_MAJOR {@OPENMC_VERSION_MAJOR@};
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constexpr int VERSION_MINOR {@OPENMC_VERSION_MINOR@};
1212
constexpr int VERSION_RELEASE {@OPENMC_VERSION_RELEASE@};
13-
constexpr bool VERSION_DEV {true};
13+
constexpr bool VERSION_DEV {false};
1414
constexpr std::array<int, 3> VERSION {VERSION_MAJOR, VERSION_MINOR, VERSION_RELEASE};
1515
// clang-format on
1616

‎openmc/__init__.py

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@@ -39,4 +39,4 @@
3939
from openmc.model import Model
4040

4141

42-
__version__ = '0.14.1-dev'
42+
__version__ = '0.15.0'

‎openmc/deplete/abc.py

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@@ -786,7 +786,7 @@ def integrate(
786786
path : PathLike
787787
Path to file to write. Defaults to 'depletion_results.h5'.
788788
789-
.. versionadded:: 0.14.1
789+
.. versionadded:: 0.15.0
790790
"""
791791
with change_directory(self.operator.output_dir):
792792
n = self.operator.initial_condition()
@@ -993,7 +993,7 @@ def integrate(
993993
path : PathLike
994994
Path to file to write. Defaults to 'depletion_results.h5'.
995995
996-
.. versionadded:: 0.14.1
996+
.. versionadded:: 0.15.0
997997
"""
998998
with change_directory(self.operator.output_dir):
999999
n = self.operator.initial_condition()

‎openmc/deplete/microxs.py

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@@ -219,7 +219,7 @@ def from_multigroup_flux(
219219
sections available. MicroXS entry will be 0 if the nuclide cross section
220220
is not found.
221221
222-
.. versionadded:: 0.14.1
222+
.. versionadded:: 0.15.0
223223
224224
Parameters
225225
----------

‎openmc/lib/mesh.py

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@@ -174,7 +174,7 @@ def material_volumes(
174174
) -> List[List[Tuple[Material, float]]]:
175175
"""Determine volume of materials in each mesh element
176176
177-
.. versionadded:: 0.14.1
177+
.. versionadded:: 0.15.0
178178
179179
Parameters
180180
----------
@@ -231,7 +231,7 @@ def get_plot_bins(
231231
) -> np.ndarray:
232232
"""Get mesh bin indices for a rasterized plot.
233233
234-
.. versionadded:: 0.14.1
234+
.. versionadded:: 0.15.0
235235
236236
Parameters
237237
----------

‎openmc/mesh.py

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@@ -157,7 +157,7 @@ def get_homogenized_materials(
157157
) -> List[openmc.Material]:
158158
"""Generate homogenized materials over each element in a mesh.
159159
160-
.. versionadded:: 0.14.1
160+
.. versionadded:: 0.15.0
161161
162162
Parameters
163163
----------
@@ -1487,6 +1487,8 @@ def get_indices_at_coords(
14871487
) -> Tuple[int, int, int]:
14881488
"""Finds the index of the mesh voxel at the specified x,y,z coordinates.
14891489
1490+
.. versionadded:: 0.15.0
1491+
14901492
Parameters
14911493
----------
14921494
coords : Sequence[float]

‎openmc/plotter.py

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@@ -169,7 +169,7 @@ def plot_xs(reactions, divisor_types=None, temperature=294., axis=None,
169169
energy_axis_units : {'eV', 'keV', 'MeV'}
170170
Units used on the plot energy axis
171171
172-
.. versionadded:: 0.14.1
172+
.. versionadded:: 0.15.0
173173
174174
Returns
175175
-------

‎openmc/region.py

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@@ -344,7 +344,7 @@ def rotate(self, rotation, pivot=(0., 0., 0.), order='xyz', inplace=False,
344344
def plot(self, *args, **kwargs):
345345
"""Display a slice plot of the region.
346346
347-
.. versionadded:: 0.14.1
347+
.. versionadded:: 0.15.0
348348
349349
Parameters
350350
----------

‎openmc/settings.py

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@@ -113,7 +113,7 @@ class Settings:
113113
max_particle_events : int
114114
Maximum number of allowed particle events per source particle.
115115
116-
.. versionadded:: 0.14.1
116+
.. versionadded:: 0.15.0
117117
max_order : None or int
118118
Maximum scattering order to apply globally when in multi-group mode.
119119
max_history_splits : int
@@ -158,7 +158,7 @@ class Settings:
158158
Starting ray distribution (must be uniform in space and angle) as
159159
specified by a :class:`openmc.SourceBase` object.
160160
161-
.. versionadded:: 0.14.1
161+
.. versionadded:: 0.15.0
162162
resonance_scattering : dict
163163
Settings for resonance elastic scattering. Accepted keys are 'enable'
164164
(bool), 'method' (str), 'energy_min' (float), 'energy_max' (float), and

‎openmc/source.py

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@@ -264,7 +264,7 @@ class IndependentSource(SourceBase):
264264
Domains to reject based on, i.e., if a sampled spatial location is not
265265
within one of these domains, it will be rejected.
266266
267-
.. deprecated:: 0.14.1
267+
.. deprecated:: 0.15.0
268268
Use the `constraints` argument instead.
269269
constraints : dict
270270
Constraints on sampled source particles. Valid keys include 'domains',
@@ -482,7 +482,7 @@ class MeshSource(SourceBase):
482482
strength of the mesh source as a whole is the sum of all source strengths
483483
applied to the elements.
484484
485-
.. versionadded:: 0.14.1
485+
.. versionadded:: 0.15.0
486486
487487
Parameters
488488
----------
@@ -1049,7 +1049,7 @@ def write_source_file(
10491049
def read_source_file(filename: PathLike) -> typing.List[SourceParticle]:
10501050
"""Read a source file and return a list of source particles.
10511051
1052-
.. versionadded:: 0.14.1
1052+
.. versionadded:: 0.15.0
10531053
10541054
Parameters
10551055
----------

‎openmc/stats/multivariate.py

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769769
Whether spatial sites should only be accepted if they occur in
770770
fissionable materials
771771
772-
.. deprecated:: 0.14.1
772+
.. deprecated:: 0.15.0
773773
Use the `constraints` argument when defining a source object instead.
774774
775775
Attributes
@@ -782,7 +782,7 @@ class Box(Spatial):
782782
Whether spatial sites should only be accepted if they occur in
783783
fissionable materials
784784
785-
.. deprecated:: 0.14.1
785+
.. deprecated:: 0.15.0
786786
Use the `constraints` argument when defining a source object instead.
787787
788788
"""

‎openmc/trigger.py

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2222
can cause the trigger to fire prematurely if there are zero scores in
2323
any bin at the first evaluation.
2424
25-
.. versionadded:: 0.14.1
25+
.. versionadded:: 0.15.0
2626
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Attributes
2828
----------

‎tests/regression_tests/filter_cellfrom/inputs_true.dat

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4949
<batches>15</batches>
5050
<inactive>5</inactive>
5151
<source particle="neutron" strength="1.0" type="independent">
52-
<space type="fission">
52+
<space type="box">
5353
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
5454
</space>
55+
<constraints>
56+
<fissionable>true</fissionable>
57+
</constraints>
5558
</source>
5659
<seed>1</seed>
5760
</settings>

‎tests/regression_tests/filter_cellfrom/test.py

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@@ -142,8 +142,9 @@ def model():
142142
core_radius,
143143
core_height / 2.0,
144144
]
145-
distribution = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
146-
model.settings.source = openmc.IndependentSource(space=distribution)
145+
distribution = openmc.stats.Box(bounds[:3], bounds[3:])
146+
model.settings.source = openmc.IndependentSource(
147+
space=distribution, constraints={'fissionable': True})
147148

148149
# =============================================================================
149150
# Tallies

‎tests/regression_tests/surface_source_write/case-01/inputs_true.dat

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4242
<batches>5</batches>
4343
<inactive>1</inactive>
4444
<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
5053
<max_particles>300</max_particles>

‎tests/regression_tests/surface_source_write/case-02/inputs_true.dat

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4242
<batches>5</batches>
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<inactive>1</inactive>
4444
<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
5053
<surface_ids>8</surface_ids>

‎tests/regression_tests/surface_source_write/case-03/inputs_true.dat

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4242
<batches>5</batches>
4343
<inactive>1</inactive>
4444
<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
5053
<surface_ids>4 5 6 7 8 9</surface_ids>

‎tests/regression_tests/surface_source_write/case-04/inputs_true.dat

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<batches>5</batches>
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<inactive>1</inactive>
4444
<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
5053
<surface_ids>4 5 6 7 8 9</surface_ids>

‎tests/regression_tests/surface_source_write/case-05/inputs_true.dat

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<batches>5</batches>
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<inactive>1</inactive>
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<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
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<surface_ids>4 5 6 7 8 9</surface_ids>

‎tests/regression_tests/surface_source_write/case-06/inputs_true.dat

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<batches>5</batches>
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<inactive>1</inactive>
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<source particle="neutron" strength="1.0" type="independent">
45-
<space type="fission">
45+
<space type="box">
4646
<parameters>-2.0 -2.0 -2.0 2.0 2.0 2.0</parameters>
4747
</space>
48+
<constraints>
49+
<fissionable>true</fissionable>
50+
</constraints>
4851
</source>
4952
<surf_source_write>
5053
<max_particles>300</max_particles>

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