add chain_file, dilute_initial parameters to MicroXS.from_model

docs/source/usersguide/depletion.rst

@@ -216,7 +216,7 @@ and a path to a depletion chain file::
     materials = openmc.Materials()
     ...
 
-    micro_xs = openmc.deplete.MicroXS()
+    micro_xs = openmc.deplete.MicroXS
     ...
 
     op = openmc.deplete.IndependentOperator(materials, micro_xs, chain_file)
@@ -267,7 +267,9 @@ Users can generate the one-group microscopic cross sections needed by
 
     model = openmc.Model.from_xml()
 
-    micro_xs = openmc.deplete.MicroXS.from_model(model, model.materials[0])
+    micro_xs = openmc.deplete.MicroXS.from_model(model,
+                                                 model.materials[0],
+                                                 chain_file)
 
 The :meth:`~openmc.deplete.MicroXS.from_model()` method will produce a
 :class:`~openmc.deplete.MicroXS` object with microscopic cross section data in

openmc/deplete/microxs.py

@@ -8,13 +8,16 @@ import tempfile
 from pathlib import Path
 
 from pandas import DataFrame, read_csv, concat
-from numpy import ndarray
+import numpy as np
 
 from openmc.checkvalue import check_type, check_value, check_iterable_type
 from openmc.mgxs import EnergyGroups, ArbitraryXS, FissionXS
-from openmc import Tallies, StatePoint
+from openmc.data import DataLibrary
+from openmc import Tallies, StatePoint, Materials
 
-from .chain import REACTIONS
+
+from .chain import Chain, REACTIONS
+from .coupled_operator import _find_cross_sections
 
 _valid_rxns = list(REACTIONS)
 _valid_rxns.append('fission')
@@ -29,13 +32,8 @@ class MicroXS(DataFrame):
     def from_model(cls,
                    model,
                    reaction_domain,
-                   reactions=['(n,gamma)',
-                              '(n,2n)',
-                              '(n,p)',
-                              '(n,a)',
-                              '(n,3n)',
-                              '(n,4n)',
-                              'fission'],
+                   chain_file,
+                   dilute_initial=1.0e3,
                    energy_bounds=(0, 20e6)):
         """Generate a one-group cross-section dataframe using
         OpenMC. Note that the ``openmc`` executable must be compiled.
@@ -46,6 +44,14 @@ class MicroXS(DataFrame):
             OpenMC model object. Must contain geometry, materials, and settings.
         reaction_domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
             Domain in which to tally reaction rates.
+        chain_file : str
+            Path to the depletion chain XML file that will be used in depletion
+            simulation. Used to determine cross sections for materials not
+            present in the inital composition.
+        dilute_initial : float
+            Initial atom density [atoms/cm^3] to add for nuclides that
+            are zero in initial condition to ensure they exist in the cross
+            section data. Only done for nuclides with reaction rates.
         reactions : list of str, optional
             Reaction names to tally
         energy_bound : 2-tuple of float, optional
@@ -63,6 +69,12 @@ class MicroXS(DataFrame):
         original_tallies = model.tallies
         tallies = Tallies()
         xs = {}
+        reactions, diluted_materials = cls._add_dilute_nuclides(chain_file,
+                                                                model,
+                                                                dilute_initial)
+        diluted_materials.export_to_xml('diluted_materials.xml')
+        model.materials = diluted_materials
+
         for rxn in reactions:
             if rxn == 'fission':
                 xs[rxn] = FissionXS(domain=reaction_domain, groups=groups, by_nuclide=True)
@@ -94,6 +106,56 @@ class MicroXS(DataFrame):
 
         return micro_xs
 
+    @classmethod
+    def _add_dilute_nuclides(cls, chain_file, model, dilute_initial):
+        chain = Chain.from_xml(chain_file)
+        reactions = chain.reactions
+        cross_sections = _find_cross_sections(model)
+        nuclides_with_data = cls._get_nuclides_with_data(cross_sections)
+        burnable_nucs = [nuc.name for nuc in chain.nuclides
+                               if nuc.name in nuclides_with_data]
+        diluted_materials = Materials()
+        for material in model.materials:
+            if material.depletable:
+                nuc_densities = material.get_nuclide_atom_densities()
+                dilute_density = 1.E-24 * dilute_initial
+                material.set_density('sum')
+                for nuc, density in nuc_densities.items():
+                    material.remove_nuclide(nuc)
+                    material.add_nuclide(nuc, density)
+                for burn_nuc in burnable_nucs:
+                    if burn_nuc not in nuc_densities:
+                        material.add_nuclide(burn_nuc,
+                                             dilute_density)
+            diluted_materials.append(material)
+        return reactions, diluted_materials
+
+    @staticmethod
+    def _get_nuclides_with_data(cross_sections):
+        """Loads cross_sections.xml file to find nuclides with neutron data
+
+        Parameters
+        ----------
+        cross_sections : str
+            Path to cross_sections.xml file
+
+        Returns
+        -------
+        nuclides : set of str
+            Set of nuclide names that have cross secton data
+
+        """
+        nuclides = set()
+        data_lib = DataLibrary.from_xml(cross_sections)
+        for library in data_lib.libraries:
+            if library['type'] != 'neutron':
+                continue
+            for name in library['materials']:
+                if name not in nuclides:
+                    nuclides.add(name)
+
+        return nuclides
+
     @classmethod
     def from_array(cls, nuclides, reactions, data):
         """
@@ -162,6 +224,6 @@ class MicroXS(DataFrame):
     def _validate_micro_xs_inputs(nuclides, reactions, data):
         check_iterable_type('nuclides', nuclides, str)
         check_iterable_type('reactions', reactions, str)
-        check_type('data', data, ndarray, expected_iter_type=float)
+        check_type('data', data, np.ndarray, expected_iter_type=float)
         for reaction in reactions:
             check_value('reactions', reaction, _valid_rxns)

tests/micro_xs_simple.csv

@@ -1,7 +1,13 @@
 nuclide,"(n,gamma)",fission
-U234,23.518634203050645,0.49531930067650976
-U235,10.621118186344665,49.10955932965905
-U238,0.8652742788116043,0.10579281644765708
-U236,9.095623870006154,0.32315392339237936
-O16,0.0029535689693132015,0.0
-O17,0.05980775766572907,0.0
+U234,22.231989822002465,0.49620744663749855
+U235,10.479008971197121,48.41787337164604
+U238,0.8673334105437324,0.10467880588762352
+U236,8.65171044607122,0.31948392400019293
+O16,7.497851000107524e-05,0.0
+O17,0.0004079227797153371,0.0
+I135,6.842395323713927,0.0
+Xe135,227463.86426990604,0.0
+Xe136,0.02317896034753588,0.0
+Cs135,2.1721665580713623,0.0
+Gd157,12786.09939237018,0.0
+Gd156,3.4006085445846983,0.0

tests/regression_tests/microxs/test.py

@@ -1,10 +1,14 @@
 """Test one-group cross section generation"""
+from pathlib import Path
+
 import numpy as np
 import pytest
 import openmc
 
 from openmc.deplete import MicroXS
 
+CHAIN_FILE = Path(__file__).parents[2] / "chain_simple.xml"
+
 @pytest.fixture(scope="module")
 def model():
     fuel = openmc.Material(name="uo2")
@@ -24,8 +28,6 @@ def model():
 
     radii = [0.42, 0.45]
     fuel.volume = np.pi * radii[0] ** 2
-    clad.volume = np.pi * (radii[1]**2 - radii[0]**2)
-    water.volume = 1.24**2 - (np.pi * radii[1]**2)
 
     materials = openmc.Materials([fuel, clad, water])
 
@@ -45,6 +47,6 @@ def model():
 
 def test_from_model(model):
     ref_xs = MicroXS.from_csv('test_reference.csv')
-    test_xs = MicroXS.from_model(model, model.materials[0])
+    test_xs = MicroXS.from_model(model, model.materials[0], CHAIN_FILE)
 
     np.testing.assert_allclose(test_xs, ref_xs, rtol=1e-11)

tests/regression_tests/microxs/test_reference.csv

@@ -1,7 +1,13 @@
-nuclide,"(n,gamma)","(n,2n)","(n,p)","(n,a)","(n,3n)","(n,4n)",fission
-U234,23.518634203050674,0.0008255330840536984,0.0,0.0,9.404554397521455e-07,0.0,0.49531930067650964
-U235,10.621118186344795,0.004359401013759254,0.0,0.0,7.2974901306692395e-06,0.0,49.10955932965902
-U238,0.8652742788116055,0.005661917442209096,0.0,0.0,4.92273921631416e-05,0.0,0.10579281644765708
-U236,9.095623870006163,0.0024373322002926834,0.0,0.0,1.966889146690413e-05,0.0,0.3231539233923791
-O16,7.511380881289377e-05,0.0,1.3764104470470622e-05,0.002862620940027927,0.0,0.0,0.0
-O17,0.00041221042693945085,1.9826700699084533e-05,7.764409141772239e-06,0.05938517754333328,0.0,0.0,0.0
+nuclide,"(n,gamma)",fission
+U234,22.231989822002465,0.49620744663749855
+U235,10.479008971197121,48.41787337164604
+U238,0.8673334105437324,0.10467880588762352
+U236,8.65171044607122,0.31948392400019293
+O16,7.497851000107524e-05,0.0
+O17,0.0004079227797153371,0.0
+I135,6.842395323713927,0.0
+Xe135,227463.86426990604,0.0
+Xe136,0.02317896034753588,0.0
+Cs135,2.1721665580713623,0.0
+Gd157,12786.09939237018,0.0
+Gd156,3.4006085445846983,0.0