address comments from @yardasol

openmcyclus/DepleteReactor.py

@@ -282,9 +282,9 @@ def enter_notify(self):
         '''
         Calls the enter_notify method of the parent class.
         Also defines a list for the input commodity preferences if
-        not are provided by the user.
+        none are provided by the user.
 
-        Establish the openmc.deplete.MicroXs and openmc.Materials
+        Establish the openmc.deplete.MicroXSxz and openmc.Materials
         objects for use in simulation.
         '''
         super().enter_notify()

openmcyclus/depletion.py

@@ -68,7 +68,7 @@ def update_materials(self, comp_list, materials):
         --------
         material_ids: list of strs
             material id numbers for the OpenMC model
-        mats: openmc.Materials()
+        materials: openmc.Materials()
             updated material object
         '''
 
@@ -83,45 +83,8 @@ def update_materials(self, comp_list, materials):
                     m = nuclide - Z * int(1e7) - A * int(1e4)
                     nucname = openmc.data.gnds_name(Z, A, m)
                     material.add_nuclide(nucname, percent, percent_type='wo')
-        mats = materials
-
-        return material_ids, mats
-
-    def run_depletion(self, flux, materials, microxs):
-        '''
-        Run the IndependentOperator class in OpenMC to perform
-        transport-independent depletion. This method is only
-        used in the test suite, and not in the OpenMCyclus
-        archetype
-
-        Parameters:
-        -----------
-        flux: float
-            flux through nuclear fuel
-        materials: openmc.Materials()
-            material definitions to deplete
-        microxs: openmc.deplete.MicroXS
-            microscopic cross section data
-
-        Outputs:
-        --------
-        depletion_results.h5: database
-            HDF5 data base with the results of the depletion simulation
-        '''
-        ind_op = od.IndependentOperator(materials,
-                                        [np.array([flux])] * len(materials),
-                                        [microxs] * len(materials),
-                                        str(self.path + self.chain_file))
-        ind_op.output_dir = self.path
-        integrator = od.PredictorIntegrator(
-            ind_op,
-            np.ones(self.timesteps) * 30,
-            power=self.power *
-            1e6,
-            timestep_units='d')
-        integrator.integrate()
 
-        return
+        return material_ids, materials
 
     def get_spent_comps(self, material_ids, microxs):
         '''

tests/unit_tests/test_depletion.py

@@ -22,6 +22,40 @@ def setUp(self):
         self.materials = openmc.Materials().from_xml("./examples/materials.xml")
         self.micro_xs = od.MicroXS.from_csv("./examples/micro_xs.csv")
 
+    def run_depletion(self, flux):
+        '''
+        Run the IndependentOperator class in OpenMC to perform
+        transport-independent depletion.
+
+        Parameters:
+        -----------
+        flux: float
+            flux through nuclear fuel
+        materials: openmc.Materials()
+            material definitions to deplete
+        microxs: openmc.deplete.MicroXS
+            microscopic cross section data
+
+        Outputs:
+        --------
+        depletion_results.h5: database
+            HDF5 data base with the results of the depletion simulation
+        '''
+        ind_op = od.IndependentOperator(self.materials,
+                                        [np.array([flux])] * len(self.materials),
+                                        [self.micro_xs] * len(self.materials),
+                                        str(self.deplete.path + self.deplete.chain_file))
+        ind_op.output_dir = self.path
+        integrator = od.PredictorIntegrator(
+            ind_op,
+            np.ones(self.deplete.timesteps) * 30,
+            power=self.deplete.power *
+            1e6,
+            timestep_units='d')
+        integrator.integrate()
+
+        return
+
     def test_update_materials(self):
         '''
         Test that the provided compositions get written to the
@@ -33,18 +67,15 @@ def test_update_materials(self):
         material_ids, materials = self.deplete.update_materials(
             comps, self.materials)
         assert materials[0].nuclides == [
-            openmc.material.NuclideTuple(
-                'U235', 0.05, 'wo'), openmc.material.NuclideTuple(
-                'U238', 0.95, 'wo')]
+            openmc.material.NuclideTuple('U235', 0.05, 'wo'),
+            openmc.material.NuclideTuple('U238', 0.95, 'wo')]
         assert materials[1].nuclides == [
-            openmc.material.NuclideTuple(
-                'Cs137', 0.1, 'wo'), openmc.material.NuclideTuple(
-                'Kr85', 0.80, 'wo'), openmc.material.NuclideTuple(
-                'Xe135', 0.10, 'wo')]
+            openmc.material.NuclideTuple('Cs137', 0.1, 'wo'), 
+            openmc.material.NuclideTuple('Kr85', 0.80, 'wo'), 
+            openmc.material.NuclideTuple('Xe135', 0.10, 'wo')]
         assert materials[2].nuclides == [
-            openmc.material.NuclideTuple(
-                'Pu239', 0.10, 'wo'), openmc.material.NuclideTuple(
-                'Pu241', 0.90, 'wo')]
+            openmc.material.NuclideTuple('Pu239', 0.10, 'wo'), 
+            openmc.material.NuclideTuple('Pu241', 0.90, 'wo')]
         assert material_ids == [5, 6, 7]
 
     def test_run_depletion(self):
@@ -53,7 +84,7 @@ def test_run_depletion(self):
         This test makes sure that the depletion runs with the correct
         output file created.
         '''
-        self.deplete.run_depletion(10.3, self.materials, self.micro_xs)
+        self.run_depletion(10.3, self.materials, self.micro_xs)
         assert os.path.isfile('examples/depletion_results.h5')
         os.system('rm examples/depletion_results.h5')