Store chain in depletion results

openmc/config.py

@@ -20,7 +20,6 @@
 from typing import Any, Dict, Iterator
 
 from openmc.data import DataLibrary
-from openmc.data.decay import _DECAY_ENERGY, _DECAY_PHOTON_ENERGY
 
 __all__ = ["config"]
 
@@ -76,9 +75,6 @@ def __delitem__(self, key: str):
             env_var = self._PATH_KEYS[key]
             if env_var in os.environ:
                 del os.environ[env_var]
-        if key == 'chain_file':
-            _DECAY_PHOTON_ENERGY.clear()
-            _DECAY_ENERGY.clear()
 
     def __setitem__(self, key: str, value: Any):
         """Set a configuration key and its corresponding value.
@@ -102,9 +98,6 @@ def __setitem__(self, key: str, value: Any):
             self._mapping[key] = stored_path
             os.environ[self._PATH_KEYS[key]] = str(stored_path)
 
-            if key == 'chain_file':
-                _DECAY_PHOTON_ENERGY.clear()
-                _DECAY_ENERGY.clear()
 
             if not stored_path.exists():
                 warnings.warn(f"Path '{stored_path}' does not exist.", UserWarning)
@@ -168,7 +161,10 @@ def patch(self, key: str, value: Any):
 
         """
         previous_value = self.get(key)
-        self[key] = value
+        if value is not None:
+            self[key] = value
+        else:
+            del self[key]
         try:
             yield
         finally:

openmc/data/data.py

@@ -6,6 +6,8 @@
 from math import sqrt, log
 from warnings import warn
 
+from ..exceptions import DataError
+
 # Isotopic abundances from Meija J, Coplen T B, et al, "Isotopic compositions
 # of the elements 2013 (IUPAC Technical Report)", Pure. Appl. Chem. 88 (3),
 # pp. 293-306 (2013). The "representative isotopic abundance" values from
@@ -363,7 +365,7 @@ def atomic_weight(element):
         raise ValueError(f"No naturally-occurring isotopes for element '{element}'.")
 
 
-def half_life(isotope):
+def half_life(isotope, chain_file = None):
     """Return half-life of isotope in seconds or None if isotope is stable
 
     Half-life values are from the `ENDF/B-VIII.0 decay sublibrary
@@ -382,6 +384,15 @@ def half_life(isotope):
         Half-life of isotope in [s]
 
     """
+    from openmc.deplete.chain import _get_chain
+
+    try:
+        chain = _get_chain(chain_file)
+        if isotope in chain and chain[isotope].half_life is not None:
+            return chain[isotope].half_life
+    except DataError:
+        pass
+
     global _HALF_LIFE
     if not _HALF_LIFE:
         # Load ENDF/B-VIII.0 data from JSON file
@@ -391,7 +402,7 @@ def half_life(isotope):
     return _HALF_LIFE.get(isotope.lower())
 
 
-def decay_constant(isotope):
+def decay_constant(isotope, chain_file = None):
     """Return decay constant of isotope in [s^-1]
 
     Decay constants are based on half-life values from the
@@ -415,7 +426,7 @@ def decay_constant(isotope):
     openmc.data.half_life
 
     """
-    t = half_life(isotope)
+    t = half_life(isotope, chain_file = chain_file)
     return _LOG_TWO / t if t else 0.0
 
 

openmc/data/decay.py

@@ -569,10 +569,7 @@ def sources(self):
         return merged_sources
 
 
-_DECAY_PHOTON_ENERGY = {}
-
-
-def decay_photon_energy(nuclide: str) -> Univariate | None:
+def decay_photon_energy(nuclide: str, chain_file = None) -> Univariate | None:
     """Get photon energy distribution resulting from the decay of a nuclide
 
     This function relies on data stored in a depletion chain. Before calling it
@@ -585,6 +582,8 @@ def decay_photon_energy(nuclide: str) -> Univariate | None:
     ----------
     nuclide : str
         Name of nuclide, e.g., 'Co58'
+    chain_file : str or path-like
+        Chain file to get decay photon energy from.
 
     Returns
     -------
@@ -593,32 +592,20 @@ def decay_photon_energy(nuclide: str) -> Univariate | None:
         if no photon source exists. Note that the probabilities represent
         intensities, given as [Bq].
     """
-    if not _DECAY_PHOTON_ENERGY:
-        chain_file = openmc.config.get('chain_file')
-        if chain_file is None:
+    from openmc.deplete.chain import _get_chain
+
+    chain = _get_chain(chain_file)
+    if chain is None:
             raise DataError(
                 "A depletion chain file must be specified with "
                 "openmc.config['chain_file'] in order to load decay data."
             )
-
-        from openmc.deplete import Chain
-        chain = Chain.from_xml(chain_file)
-        for nuc in chain.nuclides:
-            if 'photon' in nuc.sources:
-                _DECAY_PHOTON_ENERGY[nuc.name] = nuc.sources['photon']
-
-        # If the chain file contained no sources at all, warn the user
-        if not _DECAY_PHOTON_ENERGY:
-            warn(f"Chain file '{chain_file}' does not have any decay photon "
-                 "sources listed.")
-
-    return _DECAY_PHOTON_ENERGY.get(nuclide)
+    if nuclide in chain:
+        nuclide = chain[nuclide]
+        return nuclide.sources.get('photon')
 
 
-_DECAY_ENERGY = {}
-
-
-def decay_energy(nuclide: str):
+def decay_energy(nuclide: str, chain_file = None):
     """Get decay energy value resulting from the decay of a nuclide
 
     This function relies on data stored in a depletion chain. Before calling it
@@ -631,31 +618,27 @@ def decay_energy(nuclide: str):
     ----------
     nuclide : str
         Name of nuclide, e.g., 'H3'
+    chain_file : str or path-like
+        Chain file to get decay energy from.
 
     Returns
     -------
     float
         Decay energy of nuclide in [eV]. If the nuclide is stable, a value of
         0.0 is returned.
     """
-    if not _DECAY_ENERGY:
-        chain_file = openmc.config.get('chain_file')
-        if chain_file is None:
+    from openmc.deplete.chain import _get_chain
+
+    chain = _get_chain(chain_file)
+    if chain is None:
             raise DataError(
                 "A depletion chain file must be specified with "
                 "openmc.config['chain_file'] in order to load decay data."
             )
+    if nuclide in chain:
+        nuclide = chain[nuclide]
+        return nuclide.decay_energy
 
-        from openmc.deplete import Chain
-        chain = Chain.from_xml(chain_file)
-        for nuc in chain.nuclides:
-            if nuc.decay_energy:
-                _DECAY_ENERGY[nuc.name] = nuc.decay_energy
-
-        # If the chain file contained no decay energy, warn the user
-        if not _DECAY_ENERGY:
-            warn(f"Chain file '{chain_file}' does not have any decay energy.")
-
-    return _DECAY_ENERGY.get(nuclide, 0.0)
+    return 0.0
 
 

openmc/deplete/abc.py

@@ -19,7 +19,7 @@
 import numpy as np
 from uncertainties import ufloat
 
-from openmc.checkvalue import check_type, check_greater_than, PathLike
+from openmc.checkvalue import check_type, check_greater_than, check_value, PathLike
 from openmc.mpi import comm
 from openmc.utility_funcs import change_directory
 from openmc import Material
@@ -155,6 +155,7 @@ def __init__(self, chain_file=None, fission_q=None, prev_results=None):
             self.prev_res = None
         else:
             check_type("previous results", prev_results, Results)
+            check_value("previous results chain file", prev_results.chain, [self.chain])
             self.prev_res = prev_results
 
     @abstractmethod

openmc/deplete/chain.py

@@ -22,6 +22,7 @@
 from openmc.data import gnds_name, zam
 from openmc.exceptions import DataError
 from .nuclide import FissionYieldDistribution, Nuclide
+from ..mixin import EqualityMixin
 from .._xml import get_text
 import openmc.data
 
@@ -229,7 +230,7 @@ def replace_missing_fpy(actinide, fpy_data, decay_data):
     return 'U235'
 
 
-class Chain:
+class Chain(EqualityMixin):
     """Full representation of a depletion chain.
 
     A depletion chain can be created by using the :meth:`from_endf` method which
@@ -580,6 +581,50 @@ def export_to_xml(self, filename):
 
         tree = ET.ElementTree(root_elem)
         tree.write(str(filename), encoding='utf-8', pretty_print=True)
+
+    @classmethod
+    def from_hdf5(cls, group, fission_q=None):
+        """Reads a depletion chain XML file.
+
+        Parameters
+        ----------
+        group : str
+            The hdf5 group to read depletion chain from.
+        fission_q : dict, optional
+            Dictionary of nuclides and their fission Q values [eV].
+            If not given, values will be pulled from ``filename``
+
+        """
+        cgroup = group["depletion_chain"]
+        if fission_q is not None:
+            check_type("fission_q", fission_q, Mapping)
+        else:
+            fission_q = {}
+        if "path" in cgroup.attrs:
+            path = Path(__file__).resolve().parent.joinpath(cgroup.attrs["path"])
+            return Chain.from_xml(path, fission_q)
+        else:
+            chain = cls()
+            nuc_group = cgroup["nuclides"]
+            for name, ngroup in nuc_group.items():
+                this_q = fission_q.get(name)
+                nuc = Nuclide.from_hdf5(nuc_group, name, this_q)
+                chain.add_nuclide(nuc)
+
+    def to_hdf5(self, group):
+        cgroup = group.create_group("depletion_chain")
+        if hasattr(self, "_xml_path"):
+            path = Path(self._xml_path)
+            if openmc.config["resolve_paths"]:
+                path = path.resolve()
+            else:
+                path = path.relative_to(Path(__file__).resolve().parent, walk_up=True)
+            cgroup.attrs["path"] = str(path)
+        else:
+            nuc_group = cgroup.create_group("nuclides")
+            for nuclide in self.nuclides:
+                nuclide.to_hdf5(nuc_group)
+
 
     def get_default_fission_yields(self):
         """Return fission yields at lowest incident neutron energy