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Adding methods to automatically apply results to existing Tally objects. (#2671)
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
This commit is contained in:
Patrick Shriwise
GitHub
parent
7089780026
commit
f207d4220a
9 changed files with 223 additions and 84 deletions
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@ -93,9 +93,9 @@ class CrossNuclide:
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Parameters
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----------
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left_nuclide : openmc.Nuclide or CrossNuclide
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left_nuclide : str or CrossNuclide
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The left nuclide in the outer product
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right_nuclide : openmc.Nuclide or CrossNuclide
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right_nuclide : str or CrossNuclide
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The right nuclide in the outer product
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binary_op : str
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The tally arithmetic binary operator (e.g., '+', '-', etc.) used to
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@ -103,9 +103,9 @@ class CrossNuclide:
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Attributes
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----------
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left_nuclide : openmc.Nuclide or CrossNuclide
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left_nuclide : str or CrossNuclide
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The left nuclide in the outer product
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right_nuclide : openmc.Nuclide or CrossNuclide
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right_nuclide : str or CrossNuclide
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The right nuclide in the outer product
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binary_op : str
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The tally arithmetic binary operator (e.g., '+', '-', etc.) used to
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@ -134,7 +134,7 @@ class CrossNuclide:
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@left_nuclide.setter
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def left_nuclide(self, left_nuclide):
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cv.check_type('left_nuclide', left_nuclide,
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(openmc.Nuclide, CrossNuclide, AggregateNuclide))
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(str, CrossNuclide, AggregateNuclide))
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self._left_nuclide = left_nuclide
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@property
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@ -144,7 +144,7 @@ class CrossNuclide:
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@right_nuclide.setter
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def right_nuclide(self, right_nuclide):
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cv.check_type('right_nuclide', right_nuclide,
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(openmc.Nuclide, CrossNuclide, AggregateNuclide))
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(str, CrossNuclide, AggregateNuclide))
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self._right_nuclide = right_nuclide
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@property
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@ -159,26 +159,7 @@ class CrossNuclide:
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@property
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def name(self):
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string = ''
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# If the Summary was linked, the left nuclide is a Nuclide object
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if isinstance(self.left_nuclide, openmc.Nuclide):
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string += '(' + self.left_nuclide.name
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# If the Summary was not linked, the left nuclide is the ZAID
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else:
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string += '(' + str(self.left_nuclide)
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string += ' ' + self.binary_op + ' '
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# If the Summary was linked, the right nuclide is a Nuclide object
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if isinstance(self.right_nuclide, openmc.Nuclide):
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string += self.right_nuclide.name + ')'
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# If the Summary was not linked, the right nuclide is the ZAID
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else:
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string += str(self.right_nuclide) + ')'
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return string
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return f'({self.left_nuclide} {self.binary_op} {self.right_nuclide})'
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class CrossFilter:
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@ -440,7 +421,7 @@ class AggregateNuclide:
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Parameters
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----------
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nuclides : Iterable of str or openmc.Nuclide or CrossNuclide
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nuclides : Iterable of str or CrossNuclide
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The nuclides included in the aggregation
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aggregate_op : str
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The tally aggregation operator (e.g., 'sum', 'avg', etc.) used
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@ -448,7 +429,7 @@ class AggregateNuclide:
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Attributes
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----------
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nuclides : Iterable of str or openmc.Nuclide or CrossNuclide
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nuclides : Iterable of str or CrossNuclide
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The nuclides included in the aggregation
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aggregate_op : str
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The tally aggregation operator (e.g., 'sum', 'avg', etc.) used
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@ -473,13 +454,7 @@ class AggregateNuclide:
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return str(other) == str(self)
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def __repr__(self):
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# Append each nuclide in the aggregate to the string
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string = f'{self.aggregate_op}('
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names = [nuclide.name if isinstance(nuclide, openmc.Nuclide)
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else str(nuclide) for nuclide in self.nuclides]
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string += ', '.join(map(str, names)) + ')'
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return string
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return f'{self.aggregate_op}{self.name}'
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@property
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def nuclides(self):
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@ -502,12 +477,9 @@ class AggregateNuclide:
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@property
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def name(self):
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# Append each nuclide in the aggregate to the string
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names = [nuclide.name if isinstance(nuclide, openmc.Nuclide)
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else str(nuclide) for nuclide in self.nuclides]
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string = '(' + ', '.join(map(str, names)) + ')'
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return string
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names = [str(nuclide) for nuclide in self.nuclides]
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return '(' + ', '.join(map(str, names)) + ')'
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class AggregateFilter:
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@ -403,9 +403,8 @@ class Cell(IDManagerMixin):
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if self._atoms is not None:
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volume = self.volume
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for name, atoms in self._atoms.items():
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nuclide = openmc.Nuclide(name)
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density = 1.0e-24 * atoms.n/volume # density in atoms/b-cm
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nuclides[name] = (nuclide, density)
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nuclides[name] = (name, density)
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else:
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raise RuntimeError(
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'Volume information is needed to calculate microscopic '
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@ -956,7 +956,7 @@ class MGXS:
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self.xs_tally._nuclides = []
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nuclides = self.get_nuclides()
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for nuclide in nuclides:
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self.xs_tally.nuclides.append(openmc.Nuclide(nuclide))
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self.xs_tally.nuclides.append(nuclide)
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# Remove NaNs which may have resulted from divide-by-zero operations
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self.xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
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@ -14,7 +14,7 @@ import openmc
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import openmc._xml as xml
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from openmc.dummy_comm import DummyCommunicator
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from openmc.executor import _process_CLI_arguments
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from openmc.checkvalue import check_type, check_value
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from openmc.checkvalue import check_type, check_value, PathLike
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from openmc.exceptions import InvalidIDError
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from openmc.utility_funcs import change_directory
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@ -604,7 +604,8 @@ class Model:
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def run(self, particles=None, threads=None, geometry_debug=False,
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restart_file=None, tracks=False, output=True, cwd='.',
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openmc_exec='openmc', mpi_args=None, event_based=None,
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export_model_xml=True, **export_kwargs):
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export_model_xml=True, apply_tally_results=False,
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**export_kwargs):
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"""Run OpenMC
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If the C API has been initialized, then the C API is used, otherwise,
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@ -654,6 +655,11 @@ class Model:
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to True.
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.. versionadded:: 0.13.3
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apply_tally_results : bool
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Whether to apply results of the final statepoint file to the
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model's tally objects.
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.. versionadded:: 0.15.1
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**export_kwargs
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Keyword arguments passed to either :meth:`Model.export_to_model_xml`
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or :meth:`Model.export_to_xml`.
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@ -725,6 +731,10 @@ class Model:
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if mtime >= tstart: # >= allows for poor clock resolution
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tstart = mtime
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last_statepoint = sp
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if apply_tally_results:
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self.apply_tally_results(last_statepoint)
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return last_statepoint
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def calculate_volumes(self, threads=None, output=True, cwd='.',
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@ -932,6 +942,16 @@ class Model:
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n_samples=n_samples, prn_seed=prn_seed
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)
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def apply_tally_results(self, statepoint: PathLike | openmc.StatePoint):
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"""Apply results from a statepoint to tally objects on the Model
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Parameters
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----------
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statepoint : PathLike or openmc.StatePoint
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Statepoint file used to update tally results
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"""
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self.tallies.add_results(statepoint)
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def plot_geometry(self, output=True, cwd='.', openmc_exec='openmc',
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export_model_xml=True, **export_kwargs):
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"""Creates plot images as specified by the Model.plots attribute
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@ -448,14 +448,11 @@ class StatePoint:
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nuclide_names = group['nuclides'][()]
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# Add all nuclides to the Tally
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for name in nuclide_names:
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nuclide = openmc.Nuclide(name.decode().strip())
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tally.nuclides.append(nuclide)
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tally.nuclides = [name.decode().strip() for name in nuclide_names]
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# Add the scores to the Tally
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scores = group['score_bins'][()]
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for score in scores:
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tally.scores.append(score.decode())
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tally.scores = [score.decode() for score in scores]
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# Add Tally to the global dictionary of all Tallies
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tally.sparse = self.sparse
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@ -526,15 +523,16 @@ class StatePoint:
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def get_tally(self, scores=[], filters=[], nuclides=[],
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name=None, id=None, estimator=None, exact_filters=False,
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exact_nuclides=False, exact_scores=False):
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exact_nuclides=False, exact_scores=False,
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multiply_density=None, derivative=None):
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"""Finds and returns a Tally object with certain properties.
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This routine searches the list of Tallies and returns the first Tally
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found which satisfies all of the input parameters.
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NOTE: If any of the "exact" parameters are False (default), the input
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parameters do not need to match the complete Tally specification and
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may only represent a subset of the Tally's properties. If an "exact"
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parameters do not need to match the complete Tally specification and may
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only represent a subset of the Tally's properties. If an "exact"
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parameter is True then number of scores, filters, or nuclides in the
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parameters must precisely match those of any matching Tally.
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@ -561,9 +559,15 @@ class StatePoint:
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to those in the matching Tally. If False (default), the nuclides in
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the parameters may be a subset of those in the matching Tally.
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exact_scores : bool
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If True, the number of scores in the parameters must be identical
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to those in the matching Tally. If False (default), the scores
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in the parameters may be a subset of those in the matching Tally.
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If True, the number of scores in the parameters must be identical to
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those in the matching Tally. If False (default), the scores in the
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parameters may be a subset of those in the matching Tally. Default
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is None (no check).
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multiply_density : bool, optional
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If not None, the Tally must have the multiply_density attribute set
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to the same value as this parameter.
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derivative : openmc.TallyDerivative, optional
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TallyDerivative object to match.
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Returns
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-------
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@ -591,17 +595,25 @@ class StatePoint:
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if id and id != test_tally.id:
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continue
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# Determine if Tally has queried estimator
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if estimator and estimator != test_tally.estimator:
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# Determine if Tally has queried estimator, only move on to next tally
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# if the estimator is both specified and the tally estimtor does not
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# match
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if estimator is not None and estimator != test_tally.estimator:
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continue
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# The number of filters, nuclides and scores must exactly match
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if exact_scores and len(scores) != test_tally.num_scores:
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continue
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if exact_nuclides and len(nuclides) != test_tally.num_nuclides:
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if exact_nuclides and nuclides and len(nuclides) != test_tally.num_nuclides:
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continue
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if exact_nuclides and not nuclides and test_tally.nuclides != ['total']:
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continue
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if exact_filters and len(filters) != test_tally.num_filters:
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continue
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if derivative is not None and derivative != test_tally.derivative:
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continue
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if multiply_density is not None and multiply_density != test_tally.multiply_density:
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continue
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# Determine if Tally has the queried score(s)
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if scores:
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@ -1,3 +1,4 @@
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from __future__ import annotations
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from collections.abc import Iterable, MutableSequence
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import copy
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from functools import partial, reduce
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@ -33,7 +34,7 @@ _NUCLIDE_CLASSES = (str, openmc.CrossNuclide, openmc.AggregateNuclide)
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_FILTER_CLASSES = (openmc.Filter, openmc.CrossFilter, openmc.AggregateFilter)
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# Valid types of estimators
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ESTIMATOR_TYPES = ['tracklength', 'collision', 'analog']
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ESTIMATOR_TYPES = {'tracklength', 'collision', 'analog'}
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class Tally(IDManagerMixin):
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@ -65,7 +66,9 @@ class Tally(IDManagerMixin):
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scores : list of str
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List of defined scores, e.g. 'flux', 'fission', etc.
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estimator : {'analog', 'tracklength', 'collision'}
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Type of estimator for the tally
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Type of estimator for the tally. If unset (None), OpenMC will automatically
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select an appropriate estimator based on the tally filters and scores
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with a preference for 'tracklength'.
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triggers : list of openmc.Trigger
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List of tally triggers
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num_scores : int
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@ -131,6 +134,36 @@ class Tally(IDManagerMixin):
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self._sp_filename = None
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self._results_read = False
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def __eq__(self, other):
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if other.id != self.id:
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return False
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if other.name != self.name:
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return False
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# estimators are automatically set based on the tally filters and scores
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# during OpenMC initialization if this value is None, so it is not
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# considered a requirement for equivalence if it is unset on either
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# tally as it implies that the user is allowing OpenMC to select an
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# appropriate estimator. If the value is explicitly set on both tallies,
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# then the values must match for the tallies to be considered equivalent.
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if self.estimator is not None and other.estimator is not None and other.estimator != self.estimator:
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return False
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if other.filters != self.filters:
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return False
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# for tallies are loaded from statpoint files
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# an empty nuclide list is equivalent to a list with 'total'
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other_nuclides = other.nuclides.copy()
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self_nuclides = self.nuclides.copy()
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if 'total' in other_nuclides:
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other_nuclides.remove('total')
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if 'total' in self_nuclides:
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self_nuclides.remove('total')
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if other_nuclides != self_nuclides:
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return False
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for attr in {'scores', 'triggers', 'derivative', 'multiply_density'}:
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if getattr(other, attr) != getattr(self, attr):
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return False
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return True
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def __repr__(self):
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parts = ['Tally']
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parts.append('{: <15}=\t{}'.format('ID', self.id))
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@ -266,7 +299,8 @@ class Tally(IDManagerMixin):
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@estimator.setter
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def estimator(self, estimator):
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cv.check_value('estimator', estimator, ESTIMATOR_TYPES)
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# allow the estimator to be set to None (let OpenMC choose the estimator at runtime)
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cv.check_value('estimator', estimator, ESTIMATOR_TYPES | {None})
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self._estimator = estimator
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@property
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@ -304,8 +338,16 @@ class Tally(IDManagerMixin):
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# Open the HDF5 statepoint file
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with h5py.File(self._sp_filename, 'r') as f:
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# Set number of realizations
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group = f[f'tallies/tally {self.id}']
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self.num_realizations = int(group['n_realizations'][()])
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# Update nuclides
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nuclide_names = group['nuclides'][()]
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self.nuclides = [name.decode().strip() for name in nuclide_names]
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# Extract Tally data from the file
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data = f[f'tallies/tally {self.id}/results']
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data = group['results']
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sum_ = data[:, :, 0]
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sum_sq = data[:, :, 1]
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@ -511,7 +553,7 @@ class Tally(IDManagerMixin):
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Parameters
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----------
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nuclide : openmc.Nuclide
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nuclide : str
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Nuclide to remove
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"""
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@ -883,6 +925,21 @@ class Tally(IDManagerMixin):
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return element
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def add_results(self, statepoint: cv.PathLike | openmc.StatePoint):
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"""Add results from the provided statepoint file to this tally instance
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.. versionadded:: 0.15.1
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Parameters
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----------
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statepoint : openmc.PathLike or openmc.StatePoint
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Statepoint used to update tally results
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"""
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if isinstance(statepoint, openmc.StatePoint):
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self._sp_filename = statepoint._f.filename
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else:
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self._sp_filename = str(statepoint)
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@classmethod
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def from_xml_element(cls, elem, **kwargs):
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"""Generate tally object from an XML element
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@ -1020,17 +1077,10 @@ class Tally(IDManagerMixin):
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in the Tally.
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"""
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# Look for the user-requested nuclide in all of the Tally's Nuclides
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# Look for the user-requested nuclide in all of the Tally's nuclides
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for i, test_nuclide in enumerate(self.nuclides):
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# If the Summary was linked, then values are Nuclide objects
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if isinstance(test_nuclide, openmc.Nuclide):
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if test_nuclide.name == nuclide:
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return i
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# If the Summary has not been linked, then values are ZAIDs
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else:
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if test_nuclide == nuclide:
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return i
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if test_nuclide == nuclide:
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return i
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msg = (f'Unable to get the nuclide index for Tally since "{nuclide}" '
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'is not one of the nuclides')
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||||
|
|
@ -1357,9 +1407,7 @@ class Tally(IDManagerMixin):
|
|||
column_name = 'nuclide'
|
||||
|
||||
for nuclide in self.nuclides:
|
||||
if isinstance(nuclide, openmc.Nuclide):
|
||||
nuclides.append(nuclide.name)
|
||||
elif isinstance(nuclide, openmc.AggregateNuclide):
|
||||
if isinstance(nuclide, openmc.AggregateNuclide):
|
||||
nuclides.append(nuclide.name)
|
||||
column_name = f'{nuclide.aggregate_op}(nuclide)'
|
||||
else:
|
||||
|
|
@ -2350,7 +2398,8 @@ class Tally(IDManagerMixin):
|
|||
new_tally.name = self.name
|
||||
new_tally._mean = self.mean / other
|
||||
new_tally._std_dev = self.std_dev * np.abs(1. / other)
|
||||
new_tally.estimator = self.estimator
|
||||
if self.estimator is not None:
|
||||
new_tally.estimator = self.estimator
|
||||
new_tally.with_summary = self.with_summary
|
||||
new_tally.num_realizations = self.num_realizations
|
||||
|
||||
|
|
@ -2643,8 +2692,8 @@ class Tally(IDManagerMixin):
|
|||
|
||||
# Determine the nuclide indices from any of the requested nuclides
|
||||
for nuclide in self.nuclides:
|
||||
if nuclide.name not in nuclides:
|
||||
nuclide_index = self.get_nuclide_index(nuclide.name)
|
||||
if nuclide not in nuclides:
|
||||
nuclide_index = self.get_nuclide_index(nuclide)
|
||||
nuclide_indices.append(nuclide_index)
|
||||
|
||||
# Loop over indices in reverse to remove excluded Nuclides
|
||||
|
|
@ -3166,6 +3215,19 @@ class Tallies(cv.CheckedList):
|
|||
# Continue iterating from the first loop
|
||||
break
|
||||
|
||||
def add_results(self, statepoint: cv.PathLike | openmc.StatePoint):
|
||||
"""Add results from the provided statepoint file
|
||||
|
||||
.. versionadded:: 0.15.1
|
||||
|
||||
Parameters
|
||||
----------
|
||||
statepoint : openmc.PathLike or openmc.StatePoint
|
||||
Statepoint used to update tally results
|
||||
"""
|
||||
for tally in self:
|
||||
tally.add_results(statepoint)
|
||||
|
||||
def _create_tally_subelements(self, root_element):
|
||||
for tally in self:
|
||||
root_element.append(tally.to_xml_element())
|
||||
|
|
|
|||
|
|
@ -598,9 +598,8 @@ class UniverseBase(ABC, IDManagerMixin):
|
|||
if self._atoms:
|
||||
volume = self.volume
|
||||
for name, atoms in self._atoms.items():
|
||||
nuclide = openmc.Nuclide(name)
|
||||
density = 1.0e-24 * atoms.n/volume # density in atoms/b-cm
|
||||
nuclides[name] = (nuclide, density)
|
||||
nuclides[name] = (name, density)
|
||||
else:
|
||||
raise RuntimeError(
|
||||
'Volume information is needed to calculate microscopic cross '
|
||||
|
|
|
|||
|
|
@ -11,7 +11,7 @@ class SourceTestHarness(PyAPITestHarness):
|
|||
super().__init__(*args, **kwargs)
|
||||
mat1 = openmc.Material(material_id=1, temperature=294)
|
||||
mat1.set_density('g/cm3', 4.5)
|
||||
mat1.add_nuclide(openmc.Nuclide('U235'), 1.0)
|
||||
mat1.add_nuclide('U235', 1.0)
|
||||
self._model.materials = openmc.Materials([mat1])
|
||||
|
||||
sphere = openmc.Sphere(surface_id=1, r=10.0, boundary_type='vacuum')
|
||||
|
|
|
|||
|
|
@ -41,4 +41,79 @@ def test_xml_roundtrip(run_in_tmpdir):
|
|||
assert len(new_tally.triggers) == 1
|
||||
assert new_tally.triggers[0].trigger_type == tally.triggers[0].trigger_type
|
||||
assert new_tally.triggers[0].threshold == tally.triggers[0].threshold
|
||||
assert new_tally.triggers[0].scores == tally.triggers[0].scores
|
||||
assert new_tally.triggers[0].scores == tally.triggers[0].scores
|
||||
assert new_tally.multiply_density == tally.multiply_density
|
||||
|
||||
|
||||
def test_tally_equivalence():
|
||||
tally_a = openmc.Tally()
|
||||
tally_b = openmc.Tally(tally_id=tally_a.id)
|
||||
|
||||
tally_a.name = 'new name'
|
||||
assert tally_a != tally_b
|
||||
tally_b.name = tally_a.name
|
||||
assert tally_a == tally_b
|
||||
|
||||
assert tally_a == tally_b
|
||||
ef_a = openmc.EnergyFilter([0.0, 0.1, 1.0, 10.0e6])
|
||||
ef_b = openmc.EnergyFilter([0.0, 0.1, 1.0, 10.0e6])
|
||||
|
||||
tally_a.filters = [ef_a]
|
||||
assert tally_a != tally_b
|
||||
tally_b.filters = [ef_b]
|
||||
assert tally_a == tally_b
|
||||
|
||||
tally_a.scores = ['flux', 'absorption', 'fission', 'scatter']
|
||||
assert tally_a != tally_b
|
||||
tally_b.scores = ['flux', 'absorption', 'fission', 'scatter']
|
||||
assert tally_a == tally_b
|
||||
|
||||
tally_a.nuclides = []
|
||||
tally_b.nuclides = []
|
||||
assert tally_a == tally_b
|
||||
|
||||
tally_a.nuclides = ['total']
|
||||
assert tally_a == tally_b
|
||||
|
||||
# a tally with an estimator set to None is equal to
|
||||
# a tally with an estimator specified
|
||||
tally_a.estimator = 'collision'
|
||||
assert tally_a == tally_b
|
||||
tally_b.estimator = 'collision'
|
||||
assert tally_a == tally_b
|
||||
|
||||
tally_a.multiply_density = False
|
||||
assert tally_a != tally_b
|
||||
tally_b.multiply_density = False
|
||||
assert tally_a == tally_b
|
||||
|
||||
trigger_a = openmc.Trigger('rel_err', 0.025)
|
||||
trigger_b = openmc.Trigger('rel_err', 0.025)
|
||||
|
||||
tally_a.triggers = [trigger_a]
|
||||
assert tally_a != tally_b
|
||||
tally_b.triggers = [trigger_b]
|
||||
assert tally_a == tally_b
|
||||
|
||||
|
||||
def test_tally_application(sphere_model, run_in_tmpdir):
|
||||
# Create a tally with most possible gizmos
|
||||
tally = openmc.Tally(name='test tally')
|
||||
ef = openmc.EnergyFilter([0.0, 0.1, 1.0, 10.0e6])
|
||||
mesh = openmc.RegularMesh.from_domain(sphere_model.geometry, (2, 2, 2))
|
||||
mf = openmc.MeshFilter(mesh)
|
||||
tally.filters = [ef, mf]
|
||||
tally.scores = ['flux', 'absorption', 'fission', 'scatter']
|
||||
sphere_model.tallies = [tally]
|
||||
|
||||
# run the simulation and apply results
|
||||
sp_file = sphere_model.run(apply_tally_results=True)
|
||||
with openmc.StatePoint(sp_file) as sp:
|
||||
assert tally in sp.tallies.values()
|
||||
sp_tally = sp.tallies[tally.id]
|
||||
|
||||
# at this point the tally information regarding results should be the same
|
||||
assert (sp_tally.mean == tally.mean).all()
|
||||
assert (sp_tally.std_dev == tally.std_dev).all()
|
||||
assert sp_tally.nuclides == tally.nuclides
|
||||
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue