Merge pull request #1273 from drewejohnson/feat-op-qvalues

Option to pass fission q values to Chain, Operator, Model

openmc/deplete/abc.py

@@ -49,6 +49,9 @@ class TransportOperator(metaclass=ABCMeta):
         Path to the depletion chain XML file.  Defaults to the file
         listed under ``depletion_chain`` in
         :envvar:`OPENMC_CROSS_SECTIONS` environment variable.
+    fission_q : dict, optional
+        Dictionary of nuclides and their fission Q values [eV]. If not given,
+        values will be pulled from the ``chain_file``.
 
     Attributes
     ----------
@@ -58,7 +61,7 @@ class TransportOperator(metaclass=ABCMeta):
         nuclides with reaction rates. Defaults to 1.0e3.
 
     """
-    def __init__(self, chain_file=None):
+    def __init__(self, chain_file=None, fission_q=None):
         self.dilute_initial = 1.0e3
         self.output_dir = '.'
 
@@ -81,7 +84,7 @@ class TransportOperator(metaclass=ABCMeta):
                 warn("Use of OPENMC_DEPLETE_CHAIN is deprecated in favor "
                      "of adding depletion_chain to OPENMC_CROSS_SECTIONS",
                      FutureWarning)
-        self.chain = Chain.from_xml(chain_file)
+        self.chain = Chain.from_xml(chain_file, fission_q)
 
     @abstractmethod
     def __call__(self, vec, print_out=True):

openmc/deplete/chain.py

@@ -4,16 +4,18 @@ This module contains information about a depletion chain.  A depletion chain is
 loaded from an .xml file and all the nuclides are linked together.
 """
 
-from collections import OrderedDict, defaultdict
 from io import StringIO
 from itertools import chain
 import math
 import re
-import os
+from collections import OrderedDict, defaultdict
+from collections.abc import Mapping
+
+from openmc.checkvalue import check_type
 
 # Try to use lxml if it is available. It preserves the order of attributes and
-# provides a pretty-printer by default. If not available, use OpenMC function to
-# pretty print.
+# provides a pretty-printer by default. If not available,
+# use OpenMC function to pretty print.
 try:
     import lxml.etree as ET
     _have_lxml = True
@@ -312,22 +314,32 @@ class Chain(object):
         return chain
 
     @classmethod
-    def from_xml(cls, filename):
+    def from_xml(cls, filename, fission_q=None):
         """Reads a depletion chain XML file.
 
         Parameters
         ----------
         filename : str
             The path to the depletion chain XML file.
+        fission_q : dict, optional
+            Dictionary of nuclides and their fission Q values [eV].
+            If not given, values will be pulled from ``filename``
 
         """
         chain = cls()
 
+        if fission_q is not None:
+            check_type("fission_q", fission_q, Mapping)
+        else:
+            fission_q = {}
+
         # Load XML tree
         root = ET.parse(str(filename))
 
         for i, nuclide_elem in enumerate(root.findall('nuclide')):
-            nuc = Nuclide.from_xml(nuclide_elem)
+            this_q = fission_q.get(nuclide_elem.get("name"))
+
+            nuc = Nuclide.from_xml(nuclide_elem, this_q)
             chain.nuclide_dict[nuc.name] = i
 
             # Check for reaction paths

openmc/deplete/nuclide.py

@@ -112,13 +112,16 @@ class Nuclide(object):
         return len(self.reactions)
 
     @classmethod
-    def from_xml(cls, element):
+    def from_xml(cls, element, fission_q=None):
         """Read nuclide from an XML element.
 
         Parameters
         ----------
         element : xml.etree.ElementTree.Element
             XML element to write nuclide data to
+        fission_q : None or float
+            User-supplied fission Q value [eV].
+            Will be read from the element if not given
 
         Returns
         -------
@@ -153,6 +156,8 @@ class Nuclide(object):
                 target = reaction_elem.get('target')
             else:
                 target = None
+                if fission_q is not None:
+                    Q = fission_q
 
             # Append reaction
             nuc.reactions.append(ReactionTuple(

openmc/deplete/operator.py

@@ -73,6 +73,9 @@ class Operator(TransportOperator):
         in the previous results.
     diff_burnable_mats : bool, optional
         Whether to differentiate burnable materials with multiple instances
+    fission_q : dict, optional
+        Dictionary of nuclides and their fission Q values [eV]. If not given,
+        values will be pulled from the ``chain_file``.
 
     Attributes
     ----------
@@ -110,14 +113,14 @@ class Operator(TransportOperator):
 
     """
     def __init__(self, geometry, settings, chain_file=None, prev_results=None,
-                 diff_burnable_mats=False):
-        super().__init__(chain_file)
+                 diff_burnable_mats=False, fission_q=None):
+        super().__init__(chain_file, fission_q)
         self.round_number = False
         self.settings = settings
         self.geometry = geometry
         self.diff_burnable_mats = diff_burnable_mats
 
-        if prev_results != None:
+        if prev_results is not None:
             # Reload volumes into geometry
             prev_results[-1].transfer_volumes(geometry)
 
@@ -149,7 +152,6 @@ class Operator(TransportOperator):
         self.reaction_rates = ReactionRates(
             self.local_mats, self._burnable_nucs, self.chain.reactions)
 
-
     def __call__(self, vec, power, print_out=True):
         """Runs a simulation.
 

openmc/model/model.py

@@ -124,7 +124,7 @@ class Model(object):
                 self._plots.append(plot)
 
     def deplete(self, timesteps, chain_file=None, method='cecm',
-                **kwargs):
+                fission_q=None, **kwargs):
         """Deplete model using specified timesteps/power
 
         Parameters
@@ -138,6 +138,9 @@ class Model(object):
             :envvar:`OPENMC_CROSS_SECTIONS` environment variable if it exists.
         method : str
              Integration method used for depletion (e.g., 'cecm', 'predictor')
+        fission_q : dict, optional
+            Dictionary of nuclides and their fission Q values [eV].
+            If not given, values will be pulled from the ``chain_file``.
         **kwargs
             Keyword arguments passed to integration function (e.g.,
             :func:`openmc.deplete.integrator.cecm`)
@@ -149,7 +152,10 @@ class Model(object):
         import openmc.deplete as dep
 
         # Create OpenMC transport operator
-        op = dep.Operator(self.geometry, self.settings, chain_file)
+        op = dep.Operator(
+            self.geometry, self.settings, chain_file,
+            fission_q=fission_q,
+        )
 
         # Perform depletion
         check_value('method', method, ('cecm', 'predictor', 'cf4', 'epc_rk4',

tests/unit_tests/test_deplete_chain.py

@@ -231,3 +231,15 @@ def test_getitem():
     assert "NucA" == chain["NucA"]
     assert "NucB" == chain["NucB"]
     assert "NucC" == chain["NucC"]
+
+
+def test_set_fiss_q():
+    """Make sure new fission q values can be set on the chain"""
+    new_q = {"U235": 2.0E8, "U238": 2.0E8, "U234": 5.0E7}
+    chain_file = Path(__file__).parents[1] / "chain_simple.xml"
+    mod_chain = Chain.from_xml(chain_file, new_q)
+    for name, q in new_q.items():
+        chain_nuc = mod_chain[name]
+        for rx in chain_nuc.reactions:
+            if rx.type == 'fission':
+                assert rx.Q == q

tests/unit_tests/test_deplete_operator.py

@@ -68,3 +68,16 @@ def test_operator_init(bare_xs):
                 'decay_modes', 'yield_data', 'yield_energies',
                 ]:
             assert getattr(act_nuc, prop) == getattr(ref_nuc, prop), prop
+
+
+def test_operator_fiss_q():
+    """Make sure fission q values can be set"""
+    new_q = {"U235": 2.0E8, "U238": 2.0E8, "U234": 5.0E7}
+    chain_file = Path(__file__).parents[1] / "chain_simple.xml"
+    operator = BareDepleteOperator(chain_file=chain_file, fission_q=new_q)
+    mod_chain = operator.chain
+    for name, q in new_q.items():
+        chain_nuc = mod_chain[name]
+        for rx in chain_nuc.reactions:
+            if rx.type == 'fission':
+                assert rx.Q == q