Miscellaneous bug fixes

openmc/data/thermal.py

@@ -615,7 +615,8 @@ def from_ace(cls, ace_or_filename, name=None):
         ace_name, xs = ace.name.split('.')
         if not xs.endswith('t'):
             raise TypeError("{} is not a thermal scattering ACE table.".format(ace))
-        name = get_thermal_name(ace_name)
+        if name is None:
+            name = get_thermal_name(ace_name)
 
         # Assign temperature to the running list
         kTs = [ace.temperature*EV_PER_MEV]
@@ -783,9 +784,10 @@ def from_njoy(cls, filename, filename_thermal, temperatures=None,
 
             # Create instance from ACE tables within library
             lib = Library(kwargs['ace'])
-            data = cls.from_ace(lib.tables[0])
+            name = kwargs.get('table_name')
+            data = cls.from_ace(lib.tables[0], name=name)
             for table in lib.tables[1:]:
-                data.add_temperature_from_ace(table)
+                data.add_temperature_from_ace(table, name=name)
 
             # Load ENDF data to replace incoherent elastic
             if use_endf_data:

openmc/deplete/abc.py

@@ -19,7 +19,6 @@
 from numpy import nonzero, empty, asarray
 from uncertainties import ufloat
 
-from openmc.data import DataLibrary
 from openmc.lib import MaterialFilter, Tally
 from openmc.checkvalue import check_type, check_greater_than
 from openmc.mpi import comm
@@ -70,10 +69,8 @@ class TransportOperator(ABC):
 
     Parameters
     ----------
-    chain_file : str, optional
-        Path to the depletion chain XML file.  Defaults to the file
-        listed under ``depletion_chain`` in
-        :envvar:`OPENMC_CROSS_SECTIONS` environment variable.
+    chain_file : str
+        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 the ``chain_file``.
@@ -95,30 +92,12 @@ class TransportOperator(ABC):
         Results from a previous depletion calculation. ``None`` if no
         results are to be used.
     """
-    def __init__(self, chain_file=None, fission_q=None, dilute_initial=1.0e3,
+    def __init__(self, chain_file, fission_q=None, dilute_initial=1.0e3,
                  prev_results=None):
         self.dilute_initial = dilute_initial
         self.output_dir = '.'
 
         # Read depletion chain
-        if chain_file is None:
-            chain_file = os.environ.get("OPENMC_DEPLETE_CHAIN", None)
-            if chain_file is None:
-                data = DataLibrary.from_xml()
-                # search for depletion_chain path from end of list
-                for lib in reversed(data.libraries):
-                    if lib['type'] == 'depletion_chain':
-                        break
-                else:
-                    raise IOError(
-                        "No chain specified, either manually or "
-                        "under depletion_chain in environment variable "
-                        "OPENMC_CROSS_SECTIONS.")
-                chain_file = lib['path']
-            else:
-                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, fission_q)
         if prev_results is None:
             self.prev_res = None

openmc/deplete/chain.py

@@ -7,14 +7,16 @@
 from io import StringIO
 from itertools import chain
 import math
+import os
 import re
 from collections import OrderedDict, defaultdict, namedtuple
 from collections.abc import Mapping, Iterable
 from numbers import Real, Integral
 from warnings import warn
 
 from openmc.checkvalue import check_type, check_greater_than
-from openmc.data import gnd_name, zam
+from openmc.data import gnd_name, zam, DataLibrary
+from openmc.exceptions import DataError
 from .nuclide import FissionYieldDistribution
 
 # Try to use lxml if it is available. It preserves the order of attributes and
@@ -237,6 +239,24 @@ def replace_missing_fpy(actinide, fpy_data, decay_data):
     return 'U235'
 
 
+def _find_chain_file(cross_sections=None):
+    # First check deprecated OPENMC_DEPLETE_CHAIN environment variable
+    chain_file = os.environ.get("OPENMC_DEPLETE_CHAIN")
+    if chain_file is not None:
+        warn("Use of OPENMC_DEPLETE_CHAIN is deprecated in favor of adding "
+             "depletion_chain to OPENMC_CROSS_SECTIONS", FutureWarning)
+        return chain_file
+
+    # Check for depletion chain in cross_sections.xml
+    data = DataLibrary.from_xml(cross_sections)
+    for lib in reversed(data.libraries):
+        if lib['type'] == 'depletion_chain':
+            return lib['path']
+
+    raise DataError("No depletion chain specified and could not find depletion "
+                    f"chain in {cross_sections}")
+
+
 class Chain:
     """Full representation of a depletion chain.
 

openmc/deplete/operator.py

@@ -10,20 +10,20 @@
 import copy
 from collections import OrderedDict
 import os
-from typing import Type
-import xml.etree.ElementTree as ET
 from warnings import warn
-from pathlib import Path
 
 import numpy as np
 from uncertainties import ufloat
 
 import openmc
 from openmc.checkvalue import check_value
+from openmc.data import DataLibrary
+from openmc.exceptions import DataError
 import openmc.lib
 from openmc.mpi import comm
 from .abc import TransportOperator, OperatorResult
 from .atom_number import AtomNumber
+from .chain import _find_chain_file
 from .reaction_rates import ReactionRates
 from .results_list import ResultsList
 from .helpers import (
@@ -58,6 +58,22 @@ def _distribute(items):
         j += chunk_size
 
 
+def _find_cross_sections(model):
+    """Determine cross sections to use for depletion"""
+    if model.materials and model.materials.cross_sections is not None:
+        # Prefer info from Model class if available
+        return model.materials.cross_sections
+
+    # otherwise fallback to environment variable
+    cross_sections = os.environ.get("OPENMC_CROSS_SECTIONS")
+    if cross_sections is None:
+        raise DataError(
+            "Cross sections were not specified in Model.materials and "
+            "the OPENMC_CROSS_SECTIONS environment variable is not set."
+        )
+    return cross_sections
+
+
 class Operator(TransportOperator):
     """OpenMC transport operator for depletion.
 
@@ -206,6 +222,11 @@ def __init__(self, model, chain_file=None, prev_results=None,
                 "to generate the depletion Operator."
             raise TypeError(msg)
 
+        # Determine cross sections / depletion chain
+        cross_sections = _find_cross_sections(model)
+        if chain_file is None:
+            chain_file = _find_chain_file(cross_sections)
+
         check_value('fission yield mode', fission_yield_mode,
                     self._fission_helpers.keys())
         check_value('normalization mode', normalization_mode,
@@ -270,7 +291,7 @@ def __init__(self, model, chain_file=None, prev_results=None,
                     self.prev_res.append(new_res)
 
         # Determine which nuclides have incident neutron data
-        self.nuclides_with_data = self._get_nuclides_with_data()
+        self.nuclides_with_data = self._get_nuclides_with_data(cross_sections)
 
         # Select nuclides with data that are also in the chain
         self._burnable_nucs = [nuc.name for nuc in self.chain.nuclides
@@ -637,14 +658,6 @@ def _generate_materials_xml(self):
         for mat in self.materials:
             mat._nuclides.sort(key=lambda x: nuclides.index(x[0]))
 
-        # Grab the cross sections tag from the existing file
-        mfile = Path("materials.xml")
-        if mfile.exists():
-            tree = ET.parse(str(mfile))
-            xs = tree.find('cross_sections')
-            if xs is not None and self.materials.cross_sections is None:
-                self.materials.cross_sections = xs.text
-
         self.materials.export_to_xml()
 
     def _get_tally_nuclides(self):
@@ -768,40 +781,14 @@ def _unpack_tallies_and_normalize(self, source_rate):
 
         return OperatorResult(k_combined, rates)
 
-    def _get_nuclides_with_data(self):
-        """Loads a cross_sections.xml file to find participating nuclides.
-
-        This allows for nuclides that are important in the decay chain but not
-        important neutronically, or have no cross section data.
-        """
-
-        # Reads cross_sections.xml to create a dictionary containing
-        # participating (burning and not just decaying) nuclides.
-
-        try:
-            filename = os.environ["OPENMC_CROSS_SECTIONS"]
-        except KeyError:
-            filename = None
-
+    def _get_nuclides_with_data(self, cross_sections):
+        """Loads cross_sections.xml file to find nuclides with neutron data"""
         nuclides = set()
-
-        try:
-            tree = ET.parse(filename)
-        except Exception:
-            if filename is None:
-                msg = "No cross_sections.xml specified in materials."
-            else:
-                msg = 'Cross section file "{}" is invalid.'.format(filename)
-            raise IOError(msg)
-
-        root = tree.getroot()
-        for nuclide_node in root.findall('library'):
-            mats = nuclide_node.get('materials')
-            if not mats:
+        data_lib = DataLibrary.from_xml(cross_sections)
+        for library in data_lib.libraries:
+            if library['type'] != 'neutron':
                 continue
-            for name in mats.split():
-                # Make a burn list of the union of nuclides in cross_sections.xml
-                # and nuclides in depletion chain.
+            for name in library['materials']:
                 if name not in nuclides:
                     nuclides.add(name)
 

openmc/material.py

@@ -700,14 +700,7 @@ def add_s_alpha_beta(self, name, fraction=1.0):
         cv.check_type('S(a,b) fraction', fraction, Real)
         cv.check_greater_than('S(a,b) fraction', fraction, 0.0, True)
         cv.check_less_than('S(a,b) fraction', fraction, 1.0, True)
-
-        new_name = openmc.data.get_thermal_name(name)
-        if new_name != name:
-            msg = 'OpenMC S(a,b) tables follow the GND naming convention. ' \
-                  'Table "{}" is being renamed as "{}".'.format(name, new_name)
-            warnings.warn(msg)
-
-        self._sab.append((new_name, fraction))
+        self._sab.append((name, fraction))
 
     def make_isotropic_in_lab(self):
         self.isotropic = [x.name for x in self._nuclides]

openmc/model/model.py

@@ -1,10 +1,11 @@
 from collections.abc import Iterable
+from contextlib import contextmanager
 from functools import lru_cache
 import os
 from pathlib import Path
 from numbers import Integral
+from tempfile import NamedTemporaryFile
 import time
-from contextlib import contextmanager
 
 import h5py
 
@@ -528,8 +529,11 @@ def run(self, particles=None, threads=None, geometry_debug=False,
         """
 
         # Setting tstart here ensures we don't pick up any pre-existing
-        # statepoint files in the output directory
-        tstart = time.time()
+        # statepoint files in the output directory -- just in case there are
+        # differences between the system clock and the filesystem, we get the
+        # time of a just-created temporary file
+        with NamedTemporaryFile() as fp:
+            tstart = Path(fp.name).stat().st_mtime
         last_statepoint = None
 
         # Operate in the provided working directory

src/source.cpp

@@ -46,12 +46,11 @@ vector<unique_ptr<Source>> external_sources;
 // IndependentSource implementation
 //==============================================================================
 
-IndependentSource::IndependentSource(UPtrSpace space, UPtrAngle angle, UPtrDist energy, UPtrDist time)
-  : space_{std::move(space)},
-    angle_{std::move(angle)},
-    energy_{std::move(energy)},
-    time_{std::move(time)}
-{ }
+IndependentSource::IndependentSource(
+  UPtrSpace space, UPtrAngle angle, UPtrDist energy, UPtrDist time)
+  : space_ {std::move(space)}, angle_ {std::move(angle)},
+    energy_ {std::move(energy)}, time_ {std::move(time)}
+{}
 
 IndependentSource::IndependentSource(pugi::xml_node node)
 {
@@ -150,7 +149,7 @@ IndependentSource::IndependentSource(pugi::xml_node node)
       // Default to a Constant time T=0
       double T[] {0.0};
       double p[] {1.0};
-      time_ = UPtrDist {new Discrete{T, p, 1}};
+      time_ = UPtrDist {new Discrete {T, p, 1}};
     }
   }
 }
@@ -238,7 +237,7 @@ SourceSite IndependentSource::sample(uint64_t* seed) const
 
   // Sample particle creation time
   site.time = time_->sample(seed);
-  
+
   return site;
 }
 

tests/unit_tests/test_deplete_operator.py

@@ -10,7 +10,7 @@
 
 import pytest
 from openmc.deplete.abc import TransportOperator
-from openmc.deplete.chain import Chain
+from openmc.deplete.chain import Chain, _find_chain_file
 
 BARE_XS_FILE = "bare_cross_sections.xml"
 CHAIN_PATH = Path(__file__).parents[1] / "chain_simple.xml"
@@ -31,7 +31,7 @@ def bare_xs(run_in_tmpdir):
     with open(BARE_XS_FILE, "w") as out:
         out.write(bare_xs_contents)
 
-    yield
+    yield BARE_XS_FILE
 
 
 class BareDepleteOperator(TransportOperator):
@@ -54,14 +54,10 @@ def write_bos_data():
         pass
 
 
-@mock.patch.dict(environ, {"OPENMC_CROSS_SECTIONS": BARE_XS_FILE})
 def test_operator_init(bare_xs):
-    """The test will set and unset environment variable OPENMC_CROSS_SECTIONS
-    to point towards a temporary dummy file. This file will be removed
-    at the end of the test, and only contains a
-    depletion_chain node."""
-    # force operator to read from OPENMC_CROSS_SECTIONS
-    bare_op = BareDepleteOperator(chain_file=None)
+    """The test uses a temporary dummy chain. This file will be removed
+    at the end of the test, and only contains a depletion_chain node."""
+    bare_op = BareDepleteOperator(_find_chain_file(bare_xs))
     act_chain = bare_op.chain
     ref_chain = Chain.from_xml(CHAIN_PATH)
     assert len(act_chain) == len(ref_chain)