From 0a05f9dcb4320ae0c7dbd88547ca2bf9b35b68f5 Mon Sep 17 00:00:00 2001 From: yardasol Date: Thu, 14 Jul 2022 12:54:51 -0500 Subject: [PATCH] pep8 fixes --- openmc/deplete/openmc_operator.py | 49 ++++++++++++++++++++++------- openmc/deplete/operator.py | 52 ++++++++++++++++++++++--------- 2 files changed, 75 insertions(+), 26 deletions(-) diff --git a/openmc/deplete/openmc_operator.py b/openmc/deplete/openmc_operator.py index 0a64c6fee5..2f1b932943 100644 --- a/openmc/deplete/openmc_operator.py +++ b/openmc/deplete/openmc_operator.py @@ -43,6 +43,7 @@ def _distribute(items): return items[j:j + chunk_size] j += chunk_size + class OpenMCOperator(TransportOperator): """Abstrct class holding OpenMC-specific functions for running depletion calculations. @@ -136,12 +137,25 @@ class OpenMCOperator(TransportOperator): results are to be used. """ - 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, - reaction_rate_mode=None, reaction_rate_opts=None, - reduce_chain=False, reduce_chain_level=None): + 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, + reaction_rate_mode=None, + reaction_rate_opts=None, + reduce_chain=False, + reduce_chain_level=None): super().__init__(chain_file, fission_q, dilute_initial, prev_results) - self.round_number=False + self.round_number = False self.materials = materials self.cross_sections = cross_sections @@ -172,8 +186,8 @@ class OpenMCOperator(TransportOperator): if self.prev_res is not None: self._load_previous_results() - - self.nuclides_with_data = self._get_nuclides_with_data(self.cross_sections) + self.nuclides_with_data = self._get_nuclides_with_data( + self.cross_sections) # Select nuclides with data that are also in the chain self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides @@ -189,7 +203,12 @@ class OpenMCOperator(TransportOperator): self.reaction_rates = ReactionRates( self.local_mats, self._burnable_nucs, self.chain.reactions) - self._get_helper_classes(reaction_rate_mode, normalization_mode, fission_yield_mode, reaction_rate_opts, fission_yield_opts) + self._get_helper_classes( + reaction_rate_mode, + normalization_mode, + fission_yield_mode, + reaction_rate_opts, + fission_yield_opts) @abstractmethod def _differentiate_burnable_mats(self): @@ -244,7 +263,6 @@ class OpenMCOperator(TransportOperator): return burnable_mats, volume, nuclides - @abstractmethod def _load_previous_results(self): """Load reuslts from a previous depletion simulation""" @@ -268,11 +286,14 @@ class OpenMCOperator(TransportOperator): Results from a previous depletion calculation """ - self.number = AtomNumber(local_mats, all_nuclides, volume, len(self.chain)) + self.number = AtomNumber( + local_mats, all_nuclides, volume, len( + self.chain)) if self.dilute_initial != 0.0: for nuc in self._burnable_nucs: - self.number.set_atom_density(np.s_[:], nuc, self.dilute_initial) + self.number.set_atom_density( + np.s_[:], nuc, self.dilute_initial) # Now extract and store the number densities # From the geometry if no previous depletion results @@ -337,7 +358,13 @@ class OpenMCOperator(TransportOperator): self.number.set_atom_density(mat_id, nuclide, atom_per_cc) @abstractmethod - def _get_helper_classes(self, reaction_rate_mode, normalization_mode, fission_yield_mode, reaction_rate_opts, fission_yield_opts): + def _get_helper_classes( + self, + reaction_rate_mode, + normalization_mode, + fission_yield_mode, + reaction_rate_opts, + fission_yield_opts): """Create the ``_rate_helper``, ``_normalization_helper``, and ``_yield_helper`` objects. diff --git a/openmc/deplete/operator.py b/openmc/deplete/operator.py index 30563623c4..7149cddada 100644 --- a/openmc/deplete/operator.py +++ b/openmc/deplete/operator.py @@ -32,6 +32,7 @@ from .helpers import ( __all__ = ["Operator", "OperatorResult"] + def _find_cross_sections(model): """Determine cross sections to use for depletion""" if model.materials and model.materials.cross_sections is not None: @@ -219,12 +220,21 @@ class Operator(OpenMCOperator): self.cleanup_when_done = True - super().__init__(model.materials, cross_sections, chain_file, prev_results, - diff_burnable_mats, normalization_mode, - fission_q, dilute_initial, - fission_yield_mode, fission_yield_opts, - reaction_rate_mode, reaction_rate_opts, - reduce_chain, reduce_chain_level) + super().__init__( + model.materials, + cross_sections, + chain_file, + prev_results, + diff_burnable_mats, + normalization_mode, + fission_q, + dilute_initial, + fission_yield_mode, + fission_yield_opts, + reaction_rate_mode, + reaction_rate_opts, + reduce_chain, + reduce_chain_level) def _differentiate_burnable_mats(self): """Assign distribmats for each burnable material""" @@ -240,7 +250,7 @@ class Operator(OpenMCOperator): for mat in distribmats: if mat.volume is None: raise RuntimeError("Volume not specified for depletable " - "material with ID={}.".format(mat.id)) + "material with ID={}.".format(mat.id)) mat.volume /= mat.num_instances if distribmats: @@ -252,8 +262,8 @@ class Operator(OpenMCOperator): for i in range(cell.num_instances)] self.materials = openmc.Materials( - model.geometry.get_all_materials().values() - ) + model.geometry.get_all_materials().values() + ) def _load_previous_results(self): """Load results from a previous depletion simulation""" @@ -285,8 +295,13 @@ class Operator(OpenMCOperator): return nuclides - def _get_helper_classes(self, reaction_rate_mode, normalization_mode, fission_yield_mode, - reaction_rate_opts, fission_yield_opts): + def _get_helper_classes( + self, + reaction_rate_mode, + normalization_mode, + fission_yield_mode, + reaction_rate_opts, + fission_yield_opts): """Create the ``_rate_helper``, ``_normalization_helper``, and ``_yield_helper`` objects. @@ -459,17 +474,24 @@ class Operator(OpenMCOperator): densities.append(val) else: # Only output warnings if values are significantly - # negative. CRAM does not guarantee positive values. + # negative. CRAM does not guarantee positive + # values. if val < -1.0e-21: - print("WARNING: nuclide ", nuc, " in material ", mat, - " is negative (density = ", val, " at/barn-cm)") + print( + "WARNING: nuclide ", + nuc, + " in material ", + mat, + " is negative (density = ", + val, + " at/barn-cm)") number_i[mat, nuc] = 0.0 # Update densities on C API side mat_internal = openmc.lib.materials[int(mat)] mat_internal.set_densities(nuclides, densities) - #TODO Update densities on the Python side, otherwise the + # TODO Update densities on the Python side, otherwise the # summary.h5 file contains densities at the first time step @staticmethod