implement OpenMCDepcode.update_depletable_materials

saltproc/abc.py

@@ -153,5 +153,23 @@ def update_depletable_materials(self, mats, dep_end_time):
 
         """
 
+    def read_plaintext_file(self, file_path):
+        """Reads the content of a plaintext file for use by other methods.
+
+        Parameters
+        ----------
+        file_path : str
+            Path to file.
+
+        Returns
+        -------
+        file_lines : list of str
+            File lines.
+
+        """
+        file_lines = []
+        with open(file_path, 'r') as file:
+            file_lines = file.readlines()
+        return file_lines
 
 

saltproc/app.py

@@ -65,6 +65,28 @@ def run():
         mats = depcode.read_depleted_materials(True)
         simulation.store_mat_data(mats, step_idx, False)
         simulation.store_run_step_info()
+
+        # Preserve depletion and statepoint results files if using OpenMC
+        if depcode.codename == 'openmc':
+            depletion_results_path = \
+                depcode.output_path / f'depletion_results_{step_idx}.h5'
+            (depcode.output_path / 'depletion_results.h5').rename(depletion_results_path)
+
+            if step_idx != 0:
+                statepoint_fname = f'openmc_simulation_n0.h5'
+                statepoint_path = \
+                    depcode.output_path / f'openmc_simulation_n{step_idx}.h5'
+                # rename "base" to "0"
+                if step_idx == len(msr.depletion_timesteps) - 1:
+                    statepoint_path = \
+                        depcode.output_path / f'openmc_simulation_n0.h5'
+                    (depcode.output_path / 'openmc_simulation_base.h5').rename(statepoint_path)
+
+            else:
+                statepoint_path = depletion_results_path = \
+                    depcode.output_path / f'openmc_simulation_base.h5'
+            (depcode.output_path / statepoint_fname).rename(statepoint_path)
+
         # Reprocessing here
         print("\nMass and volume of fuel before reproc: %f g, %f cm3" %
               (mats['fuel'].mass,

saltproc/openmc_depcode.py

@@ -94,7 +94,7 @@ def __init__(self,
              'settings': str((output_path / 'settings.xml').resolve())}
         self.runtime_matfile = str((output_path / 'materials.xml').resolve())
 
-        self._check_for_material_names(self.runtime_matfile)
+        self._check_for_material_names(self.template_input_file_path['materials'])
 
     def _check_for_material_names(self, filename):
         """Checks that all materials in the material file
@@ -186,6 +186,7 @@ def read_depleted_materials(self, read_at_end=False):
             else:
                 burnup = 0
             depleted_materials[name].burnup = burnup
+        del openmc_materials, depleted_openmc_materials, starting_openmc_materials
         return depleted_materials
 
     def _create_mass_percents_dictionary(self, mat):
@@ -311,10 +312,11 @@ def write_runtime_input(self, reactor, depletion_step, restart):
         """
 
         if depletion_step == 0 and not restart:
+            geo_file = self.geo_file_paths.pop(0)
             materials = openmc.Materials.from_xml(
                 self.template_input_file_path['materials'])
             geometry = openmc.Geometry.from_xml(
-                self.geo_file_paths[0], materials=materials)
+                geo_file, materials=materials)
             settings = openmc.Settings.from_xml(
                 self.template_input_file_path['settings'])
             self.npop = settings.particles
@@ -393,6 +395,29 @@ def update_depletable_materials(self, mats, dep_end_time):
             Current time at the end of the depletion step (d).
 
         """
+        runtime_materials = openmc.Materials.from_xml(self.runtime_matfile)
+
+        for material in runtime_materials:
+            # depletable materials only
+            if material.name in mats.keys():
+                components = {}
+                for nuc_code, mass_fraction in mats[material.name].comp.items():
+                    nuc_name = pyname.name(nuc_code)
+                    # Convert nuclide names from PyNE format to OpenMC format
+                    if nuc_name[-1] == 'M':
+                        nuc_name = nuc_name[:-1] + '_m1'
+                    components[nuc_name] = mass_fraction
+
+                material.set_density(mats[material.name].density)
+                material.volume = mats[material.name].vol
+                for element in material.get_elements():
+                    material.remove_element(element)
+                material.add_components(components, percent_type='wo')
+
+        runtime_materials.export_to_xml(path=self.runtime_materials)
+        del runtime_materials
+        del material
+
 
     def write_saltproc_openmc_tallies(self, materials, geometry):
         """

saltproc/serpent_depcode.py

@@ -395,25 +395,6 @@ def read_neutronics_parameters(self):
         self.neutronics_parameters['fission_mass_eds'] = \
             res['TOT_FMASS'][1]
 
-    def read_plaintext_file(self, file_path):
-        """Reads the content of a plaintext file for use by other methods.
-
-        Parameters
-        ----------
-        file_path : str
-            Path to file.
-
-        Returns
-        -------
-        file_lines : list of str
-            File lines.
-
-        """
-        file_lines = []
-        with open(file_path, 'r') as file:
-            file_lines = file.readlines()
-        return file_lines
-
     def set_power_load(self,
                        file_lines,
                        reactor,