diff --git a/openmc/deplete/openmc_wrapper.py b/openmc/deplete/openmc_wrapper.py index f3afedb29..3c3b974c9 100644 --- a/openmc/deplete/openmc_wrapper.py +++ b/openmc/deplete/openmc_wrapper.py @@ -123,10 +123,6 @@ class OpenMCOperator(Operator): Reaction rates from the last operator step. burn_mat_to_ind : OrderedDict of str to int Dictionary mapping material ID (as a string) to an index in reaction_rates. - burn_nuc_to_ind : OrderedDict of str to int - Dictionary mapping nuclide name (as a string) to an index in - reaction_rates. Consists of all nuclides with neutron data and appearing - in the depletion chain. burnable_mats : list of str All burnable material IDs @@ -144,7 +140,9 @@ class OpenMCOperator(Operator): local_mats = _distribute(self.burnable_mats) # Determine which nuclides have incident neutron data - self._load_participating() + self.nuclides_with_data = self._get_nuclides_with_data() + self._burnable_nucs = [nuc for nuc in self.nuclides_with_data + if nuc in self.chain] # Extract number densities from the geometry self._extract_number(local_mats, volume, nuc_dict) @@ -152,7 +150,7 @@ class OpenMCOperator(Operator): # Create reaction rates array index_rx = {rx: i for i, rx in enumerate(self.chain.reactions)} self.reaction_rates = ReactionRates( - self.burn_mat_to_ind, self.burn_nuc_to_ind, index_rx) + self.burn_mat_to_ind, self._burnable_nucs, index_rx) def __call__(self, vec, print_out=True): """Runs a simulation. @@ -264,7 +262,7 @@ class OpenMCOperator(Operator): len(self.chain)) if self.settings.dilute_initial != 0.0: - for nuc in self.burn_nuc_to_ind: + for nuc in self._burnable_nucs: self.number.set_atom_density(np.s_[:], nuc, self.settings.dilute_initial) @@ -545,7 +543,7 @@ class OpenMCOperator(Operator): return OperatorResult(k_combined, rates) - def _load_participating(self): + def _get_nuclides_with_data(self): """Loads a cross_sections.xml file to find participating nuclides. This allows for nuclides that are important in the decay chain but not @@ -560,7 +558,7 @@ class OpenMCOperator(Operator): except KeyError: filename = None - self.nuclides_with_data = set() + nuclides = set() try: tree = ET.parse(filename) @@ -572,9 +570,6 @@ class OpenMCOperator(Operator): raise IOError(msg) root = tree.getroot() - self.burn_nuc_to_ind = OrderedDict() - nuc_ind = 0 - for nuclide_node in root.findall('library'): mats = nuclide_node.get('materials') if not mats: @@ -582,11 +577,10 @@ class OpenMCOperator(Operator): for name in mats.split(): # Make a burn list of the union of nuclides in cross_sections.xml # and nuclides in depletion chain. - if name not in self.nuclides_with_data: - self.nuclides_with_data.add(name) - if name in self.chain: - self.burn_nuc_to_ind[name] = nuc_ind - nuc_ind += 1 + if name not in nuclides: + nuclides.add(name) + + return nuclides def get_results_info(self): """Returns volume list, cell lists, and nuc lists. diff --git a/openmc/deplete/reaction_rates.py b/openmc/deplete/reaction_rates.py index a47908517..d1e1b0f1e 100644 --- a/openmc/deplete/reaction_rates.py +++ b/openmc/deplete/reaction_rates.py @@ -15,8 +15,8 @@ class ReactionRates(np.ndarray): ---------- index_mat : OrderedDict of str to int A dictionary mapping material ID as string to index. - index_nuc : OrderedDict of str to int - A dictionary mapping nuclide name as string to index. + nuclides : list of str + Depletable nuclides index_rx : OrderedDict of str to int A dictionary mapping reaction name as string to index. @@ -36,15 +36,15 @@ class ReactionRates(np.ndarray): Number of reactions. """ - def __new__(cls, index_mat, index_nuc, index_rx): + def __new__(cls, index_mat, nuclides, index_rx): # Create appropriately-sized zeroed-out ndarray - shape = (len(index_mat), len(index_nuc), len(index_rx)) + shape = (len(index_mat), len(nuclides), len(index_rx)) obj = super().__new__(cls, shape) obj[:] = 0.0 # Add mapping attributes obj.index_mat = index_mat - obj.index_nuc = index_nuc + obj.index_nuc = {nuc: i for i, nuc in enumerate(nuclides)} obj.index_rx = index_rx return obj