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pep8 fixes
This commit is contained in:
yardasol
parent
6ff164131c
commit
0a05f9dcb4
2 changed files with 75 additions and 26 deletions
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@ -43,6 +43,7 @@ def _distribute(items):
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return items[j:j + chunk_size]
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j += chunk_size
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class OpenMCOperator(TransportOperator):
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"""Abstrct class holding OpenMC-specific functions for running
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depletion calculations.
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@ -136,12 +137,25 @@ class OpenMCOperator(TransportOperator):
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results are to be used.
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"""
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def __init__(self, materials=None, cross_sections=None, chain_file=None, prev_results=None, diff_burnable_mats=False, normalization_mode=None, fission_q=None, dilute_initial=0.0, fission_yield_mode=None, fission_yield_opts=None,
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reaction_rate_mode=None, reaction_rate_opts=None,
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reduce_chain=False, reduce_chain_level=None):
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def __init__(
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self,
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materials=None,
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cross_sections=None,
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chain_file=None,
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prev_results=None,
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diff_burnable_mats=False,
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normalization_mode=None,
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fission_q=None,
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dilute_initial=0.0,
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fission_yield_mode=None,
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fission_yield_opts=None,
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reaction_rate_mode=None,
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reaction_rate_opts=None,
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reduce_chain=False,
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reduce_chain_level=None):
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super().__init__(chain_file, fission_q, dilute_initial, prev_results)
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self.round_number=False
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self.round_number = False
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self.materials = materials
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self.cross_sections = cross_sections
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@ -172,8 +186,8 @@ class OpenMCOperator(TransportOperator):
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if self.prev_res is not None:
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self._load_previous_results()
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self.nuclides_with_data = self._get_nuclides_with_data(self.cross_sections)
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self.nuclides_with_data = self._get_nuclides_with_data(
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self.cross_sections)
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# Select nuclides with data that are also in the chain
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self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides
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@ -189,7 +203,12 @@ class OpenMCOperator(TransportOperator):
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self.reaction_rates = ReactionRates(
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self.local_mats, self._burnable_nucs, self.chain.reactions)
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self._get_helper_classes(reaction_rate_mode, normalization_mode, fission_yield_mode, reaction_rate_opts, fission_yield_opts)
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self._get_helper_classes(
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reaction_rate_mode,
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normalization_mode,
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fission_yield_mode,
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reaction_rate_opts,
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fission_yield_opts)
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@abstractmethod
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def _differentiate_burnable_mats(self):
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@ -244,7 +263,6 @@ class OpenMCOperator(TransportOperator):
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return burnable_mats, volume, nuclides
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@abstractmethod
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def _load_previous_results(self):
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"""Load reuslts from a previous depletion simulation"""
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@ -268,11 +286,14 @@ class OpenMCOperator(TransportOperator):
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Results from a previous depletion calculation
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"""
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self.number = AtomNumber(local_mats, all_nuclides, volume, len(self.chain))
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self.number = AtomNumber(
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local_mats, all_nuclides, volume, len(
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self.chain))
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if self.dilute_initial != 0.0:
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for nuc in self._burnable_nucs:
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self.number.set_atom_density(np.s_[:], nuc, self.dilute_initial)
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self.number.set_atom_density(
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np.s_[:], nuc, self.dilute_initial)
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# Now extract and store the number densities
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# From the geometry if no previous depletion results
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@ -337,7 +358,13 @@ class OpenMCOperator(TransportOperator):
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self.number.set_atom_density(mat_id, nuclide, atom_per_cc)
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@abstractmethod
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def _get_helper_classes(self, reaction_rate_mode, normalization_mode, fission_yield_mode, reaction_rate_opts, fission_yield_opts):
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def _get_helper_classes(
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self,
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reaction_rate_mode,
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normalization_mode,
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fission_yield_mode,
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reaction_rate_opts,
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fission_yield_opts):
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"""Create the ``_rate_helper``, ``_normalization_helper``, and
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``_yield_helper`` objects.
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@ -32,6 +32,7 @@ from .helpers import (
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__all__ = ["Operator", "OperatorResult"]
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def _find_cross_sections(model):
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"""Determine cross sections to use for depletion"""
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if model.materials and model.materials.cross_sections is not None:
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@ -219,12 +220,21 @@ class Operator(OpenMCOperator):
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self.cleanup_when_done = True
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super().__init__(model.materials, cross_sections, chain_file, prev_results,
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diff_burnable_mats, normalization_mode,
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fission_q, dilute_initial,
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fission_yield_mode, fission_yield_opts,
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reaction_rate_mode, reaction_rate_opts,
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reduce_chain, reduce_chain_level)
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super().__init__(
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model.materials,
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cross_sections,
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chain_file,
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prev_results,
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diff_burnable_mats,
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normalization_mode,
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fission_q,
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dilute_initial,
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fission_yield_mode,
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fission_yield_opts,
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reaction_rate_mode,
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reaction_rate_opts,
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reduce_chain,
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reduce_chain_level)
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def _differentiate_burnable_mats(self):
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"""Assign distribmats for each burnable material"""
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@ -240,7 +250,7 @@ class Operator(OpenMCOperator):
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for mat in distribmats:
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if mat.volume is None:
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raise RuntimeError("Volume not specified for depletable "
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"material with ID={}.".format(mat.id))
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"material with ID={}.".format(mat.id))
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mat.volume /= mat.num_instances
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if distribmats:
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@ -252,8 +262,8 @@ class Operator(OpenMCOperator):
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for i in range(cell.num_instances)]
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self.materials = openmc.Materials(
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model.geometry.get_all_materials().values()
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)
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model.geometry.get_all_materials().values()
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)
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def _load_previous_results(self):
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"""Load results from a previous depletion simulation"""
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@ -285,8 +295,13 @@ class Operator(OpenMCOperator):
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return nuclides
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def _get_helper_classes(self, reaction_rate_mode, normalization_mode, fission_yield_mode,
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reaction_rate_opts, fission_yield_opts):
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def _get_helper_classes(
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self,
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reaction_rate_mode,
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normalization_mode,
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fission_yield_mode,
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reaction_rate_opts,
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fission_yield_opts):
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"""Create the ``_rate_helper``, ``_normalization_helper``, and
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``_yield_helper`` objects.
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@ -459,17 +474,24 @@ class Operator(OpenMCOperator):
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densities.append(val)
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else:
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# Only output warnings if values are significantly
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# negative. CRAM does not guarantee positive values.
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# negative. CRAM does not guarantee positive
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# values.
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if val < -1.0e-21:
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print("WARNING: nuclide ", nuc, " in material ", mat,
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" is negative (density = ", val, " at/barn-cm)")
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print(
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"WARNING: nuclide ",
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nuc,
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" in material ",
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mat,
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" is negative (density = ",
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val,
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" at/barn-cm)")
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number_i[mat, nuc] = 0.0
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# Update densities on C API side
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mat_internal = openmc.lib.materials[int(mat)]
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mat_internal.set_densities(nuclides, densities)
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#TODO Update densities on the Python side, otherwise the
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# TODO Update densities on the Python side, otherwise the
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# summary.h5 file contains densities at the first time step
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@staticmethod
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