added testing and working for nuclides

openmc/deplete/coupled_operator.py

@@ -10,6 +10,7 @@
 
 import copy
 from warnings import warn
+from typing import Optional
 
 import numpy as np
 from uncertainties import ufloat
@@ -33,18 +34,19 @@
 __all__ = ["CoupledOperator", "Operator", "OperatorResult"]
 
 
-def _find_cross_sections(model):
+def _find_cross_sections(model: Optional[str] = None):
     """Determine cross sections to use for depletion
 
     Parameters
     ----------
-    model : openmc.model.Model
+    model : openmc.model.Model, optional
         Reactor model
 
     """
-    if model.materials and model.materials.cross_sections is not None:
-        # Prefer info from Model class if available
-        return model.materials.cross_sections
+    if model:
+        if model.materials and model.materials.cross_sections is not None:
+            # Prefer info from Model class if available
+            return model.materials.cross_sections
 
     # otherwise fallback to environment variable
     cross_sections = openmc.config.get("cross_sections")
@@ -67,7 +69,7 @@ def _get_nuclides_with_data(cross_sections):
     Returns
     -------
     nuclides : set of str
-        Set of nuclide names that have cross secton data
+        Set of nuclide names that have cross section data
 
     """
     nuclides = set()

openmc/deplete/microxs.py

@@ -36,16 +36,6 @@ def _resolve_chain_file_path(chain_file:str):
     return chain_file
 
 
-def _resolve_openmc_data_path(openmc_data_path: str=None):
-    if not openmc_data_path:
-        if 'cross_sections' not in openmc.config:
-            raise ValueError("`openmc_data_path` is not defined nor `openmc.config['cross_sections']` is defined.")
-        else:
-            openmc_data_path = openmc.config['cross_sections']
-    return openmc_data_path
-
-
-
 def get_microxs_and_flux(
         model: openmc.Model,
         domains,
@@ -216,8 +206,7 @@ def from_multi_group_flux(
         energies: Union[Iterable[float], str],
         multi_group_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
-        openmc_data_path: Optional[PathLike] = None,
-        temperature: int=294,
+        temperature: float=294.,
         nuclides: Optional[Iterable[str]] = None,
     ):
         """Generated MicroXS object from a known flux and a chain file.
@@ -235,9 +224,6 @@ def from_multi_group_flux(
         chain_file : str, optional
             Path to the depletion chain XML file that will be used in
             depletion simulation.  Defaults to ``openmc.config['chain_file']``.
-        openmc_data_path : str, optional
-            Path to the cross section XML file that contains data library paths.
-            Defaults to ``openmc.config['cross_sections']``.
         temperature : int, optional
             Temperature for cross section evaluation in [K].
         nuclides : list of str
@@ -249,7 +235,7 @@ def from_multi_group_flux(
         MicroXS
         """
 
-        check_type("temperature", temperature, int)
+        check_type("temperature", temperature, (int, float))
         # if energy is string then use group structure of that name
         if isinstance(energies, str):
             energies = openmc.EnergyFilter.from_group_structure(energies).values
@@ -266,15 +252,15 @@ def from_multi_group_flux(
         chain_file_path = _resolve_chain_file_path(chain_file)
         chain = openmc.deplete.Chain.from_xml(chain_file_path)
 
-        openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
-        cross_sections = openmc.data.DataLibrary.from_xml(openmc_data_path)
+        cross_sections = _find_cross_sections(model=None)
+        data_lib = openmc.data.DataLibrary.from_xml(cross_sections)
         nuclides_with_data = _get_nuclides_with_data(cross_sections)
 
         # get reactions and nuclides from chain file
         if not nuclides:
-                nuclides = chain.nuclides
+            nuclides = chain.nuclides
+            nuclides = [nuc.name for nuc in nuclides]
         reactions = chain.reactions
-        nuclides = [nuc.name for nuc in nuclides]
         if 'fission' in reactions:
             # send to back
             reactions.append(reactions.pop(reactions.index('fission')))
@@ -308,6 +294,7 @@ def from_multi_group_flux(
             settings.particles = 1
             settings.batches = 1
             settings.output = {'summary': False}
+
             model = openmc.Model(geometry, materials, settings)
             model.export_to_model_xml()
 

openmc/lib/nuclide.py

@@ -91,7 +91,7 @@ def collapse_rate(self, MT, temperature, energy, flux):
         energy : iterable of float
             Energy group boundaries in [eV]
         flux : iterable of float
-            Flux in each energt group (not normalized per eV)
+            Flux in each energy group (not normalized per eV)
 
         Returns
         -------

tests/unit_tests/test_deplete_microxs.py

@@ -60,6 +60,7 @@ def test_csv():
 
 def test_from_multi_group_flux():
 
+    # test with energy group structure from string
     microxs = MicroXS.from_multi_group_flux(
         energies='CASMO-4',
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
@@ -68,10 +69,22 @@ def test_from_multi_group_flux():
     )
     assert isinstance(microxs, MicroXS)
 
+    # test with energy group structure as floats
     microxs = MicroXS.from_multi_group_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)
+
+    # test with nuclides provided
+    microxs = MicroXS.from_multi_group_flux(
+        energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
+        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        temperature=293,
+        chain_file=Path(__file__).parents[1] / 'chain_simple.xml',
+        nuclides=['Gd157']
+    )
+    assert microxs.nuclides == ['Gd157']
+    assert isinstance(microxs, MicroXS)