crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-26 05:05:30 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Use openmc.config['cross_sections'] instead of OPENMC_CROSS_SECTIONS

Browse source
This commit is contained in:
Paul Romano 2022-08-31 22:13:54 -05:00
parent c29fb407d5
commit a074e0a5cf
8 changed files with 51 additions and 69 deletions
Download patch file Download diff file Expand all files Collapse all files

17
openmc/data/library.py
View file

@ -4,6 +4,7 @@ import pathlib
import h5py
import openmc
from openmc.mixin import EqualityMixin
from openmc._xml import clean_indentation, reorder_attributes
@ -124,8 +125,8 @@ class DataLibrary(EqualityMixin):
Parameters
----------
path : str, optional
Path to XML file to read. If not provided, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used.
Path to XML file to read. If not provided,
openmc.config['cross_sections'] will be used.
Returns
-------
@ -136,15 +137,14 @@ class DataLibrary(EqualityMixin):
data = cls()
# If path is None, get the cross sections from the
# OPENMC_CROSS_SECTIONS environment variable
# If path is None, get the cross sections from the global configuration
if path is None:
path = os.environ.get('OPENMC_CROSS_SECTIONS')
path = openmc.config.get('cross_sections')
# Check to make sure there was an environmental variable.
# Check to make sure we picked up cross sections
if path is None:
raise ValueError("Either path or OPENMC_CROSS_SECTIONS "
"environmental variable must be set")
raise ValueError("Either path or openmc.config['cross_sections'] "
"must be set")
tree = ET.parse(path)
root = tree.getroot()
@ -162,7 +162,6 @@ class DataLibrary(EqualityMixin):
data.libraries.append(library)
# get depletion chain data
dep_node = root.find("depletion_chain")
if dep_node is not None:
filename = os.path.join(directory, dep_node.attrib['path'])

23
openmc/deplete/chain.py
View file

@ -239,24 +239,6 @@ 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.
@ -265,9 +247,8 @@ class Chain:
yield sublibrary files. The depletion chain used during a depletion
simulation is indicated by either an argument to
:class:`openmc.deplete.CoupledOperator` or
:class:`openmc.deplete.IndependentOperator`, or through the
``depletion_chain`` item in the :envvar:`OPENMC_CROSS_SECTIONS`
environment variable.
:class:`openmc.deplete.IndependentOperator`, or through
openmc.config['chain_file'].
Attributes
----------

15
openmc/deplete/coupled_operator.py
View file

@ -9,7 +9,6 @@ filesystem.
"""
import copy
import os
from warnings import warn
import numpy as np
@ -22,7 +21,6 @@ from openmc.exceptions import DataError
import openmc.lib
from openmc.mpi import comm
from .abc import OperatorResult
from .chain import _find_chain_file
from .openmc_operator import OpenMCOperator, _distribute
from .results import Results
from .helpers import (
@ -48,11 +46,11 @@ def _find_cross_sections(model):
return model.materials.cross_sections
# otherwise fallback to environment variable
cross_sections = os.environ.get("OPENMC_CROSS_SECTIONS")
cross_sections = openmc.config.get("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."
"openmc.config['cross_sections'] is not set."
)
return cross_sections
@ -103,9 +101,8 @@ class CoupledOperator(OpenMCOperator):
model : openmc.model.Model
OpenMC model object
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.
Path to the depletion chain XML file. Defaults to
``openmc.config['chain_file']``.
prev_results : Results, optional
Results from a previous depletion calculation. If this argument is
specified, the depletion calculation will start from the latest state
@ -231,10 +228,8 @@ class CoupledOperator(OpenMCOperator):
" model with which to generate the transport Operator."
raise TypeError(msg)
# Determine cross sections / depletion chain
# Determine cross sections
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())

16
openmc/deplete/openmc_operator.py
View file

@ -11,6 +11,7 @@ from collections import OrderedDict
import numpy as np
import openmc
from openmc.exceptions import DataError
from openmc.mpi import comm
from .abc import TransportOperator, OperatorResult
from .atom_number import AtomNumber
@ -57,9 +58,8 @@ class OpenMCOperator(TransportOperator):
Path to continuous energy cross section library, or object containing
one-group cross-sections.
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.
Path to the depletion chain XML file. Defaults to
openmc.config['chain_file'].
prev_results : Results, optional
Results from a previous depletion calculation. If this argument is
specified, the depletion calculation will start from the latest state
@ -83,7 +83,6 @@ class OpenMCOperator(TransportOperator):
if ``reduce_chain`` evaluates to true. The default value of
``None`` implies no limit on the depth.
Attributes
----------
materials : openmc.Materials
@ -133,6 +132,15 @@ class OpenMCOperator(TransportOperator):
reduce_chain=False,
reduce_chain_level=None):
# If chain file was not specified, try to get it from global config
if chain_file is None:
chain_file = openmc.config.get('chain_file')
if chain_file is None:
raise DataError(
"No depletion chain specified and could not find depletion "
"chain in openmc.config['chain_file']"
)
super().__init__(chain_file, fission_q, dilute_initial, prev_results)
self.round_number = False
self.materials = materials

4
openmc/deplete/results.py
View file

@ -392,7 +392,7 @@ class Results(list):
as such cannot be used in subsequent transport calculations.
If not provided, nuclides from the cross_sections element of
materials.xml will be used. If that element is not present,
nuclides from OPENMC_CROSS_SECTIONS will be used.
nuclides from openmc.config['cross_sections'] will be used.
Returns
-------
@ -412,7 +412,7 @@ class Results(list):
# the new materials XML file. The precedence of nuclides to select
# is first ones provided as a kwarg here, then ones specified
# in the materials.xml file if provided, then finally from
# the environment variable OPENMC_CROSS_SECTIONS.
# openmc.config['cross_sections'].
if nuc_with_data:
cv.check_iterable_type('nuclide names', nuc_with_data, str)
available_cross_sections = nuc_with_data

33
openmc/element.py
View file

@ -1,9 +1,9 @@
from collections import OrderedDict
import os
import re
from xml.etree import ElementTree as ET
import openmc.checkvalue as cv
import openmc
from openmc.data import NATURAL_ABUNDANCE, atomic_mass, \
isotopes as natural_isotopes
@ -40,10 +40,10 @@ class Element(str):
cross_sections=None):
"""Expand natural element into its naturally-occurring isotopes.
An optional cross_sections argument or the :envvar:`OPENMC_CROSS_SECTIONS`
environment variable is used to specify a cross_sections.xml file.
If the cross_sections.xml file is found, the element is expanded only
into the isotopes/nuclides present in cross_sections.xml. If no
An optional cross_sections argument or the ``cross_sections``
configuration value is used to specify a cross_sections.xml file. If the
cross_sections.xml file is found, the element is expanded only into the
isotopes/nuclides present in cross_sections.xml. If no
cross_sections.xml file is found, the element is expanded based on its
naturally occurring isotopes.
@ -54,12 +54,13 @@ class Element(str):
percent_type : {'ao', 'wo'}
'ao' for atom percent and 'wo' for weight percent
enrichment : float, optional
Enrichment of an enrichment_target nuclide in percent (ao or wo).
If enrichment_target is not supplied then it is enrichment for U235
in weight percent. For example, input 4.95 for 4.95 weight percent
Enrichment of an enrichment_target nuclide in percent (ao or wo). If
enrichment_target is not supplied then it is enrichment for U235 in
weight percent. For example, input 4.95 for 4.95 weight percent
enriched U. Default is None (natural composition).
enrichment_target: str, optional
Single nuclide name to enrich from a natural composition (e.g., 'O16')
Single nuclide name to enrich from a natural composition (e.g.,
'O16')
.. versionadded:: 0.12
enrichment_type: {'ao', 'wo'}, optional
@ -82,8 +83,8 @@ class Element(str):
ValueError
No data is available for any of natural isotopes of the element
ValueError
If only some natural isotopes are available in the cross-section data
library and the element is not O, W, or Ta
If only some natural isotopes are available in the cross-section
data library and the element is not O, W, or Ta
ValueError
If a non-naturally-occurring isotope is requested
ValueError
@ -101,8 +102,8 @@ class Element(str):
`ORNL/CSD/TM-244 <https://doi.org/10.2172/5561567>`_ is used to
calculate the weight fractions of U234, U235, U236, and U238. Namely,
the weight fraction of U234 and U236 are taken to be 0.89% and 0.46%,
respectively, of the U235 weight fraction. The remainder of the
isotopic weight is assigned to U238.
respectively, of the U235 weight fraction. The remainder of the isotopic
weight is assigned to U238.
When the `enrichment` argument is specified with `enrichment_target`, a
general enrichment procedure is used for elements composed of exactly
@ -125,10 +126,10 @@ class Element(str):
# Create dict to store the expanded nuclides and abundances
abundances = OrderedDict()
# If cross_sections is None, get the cross sections from the
# OPENMC_CROSS_SECTIONS environment variable
# If cross_sections is None, get the cross sections from the global
# configuration
if cross_sections is None:
cross_sections = os.environ.get('OPENMC_CROSS_SECTIONS')
cross_sections = openmc.config.get('cross_sections')
# If a cross_sections library is present, check natural nuclides
# against the nuclides in the library

10
openmc/material.py
View file

@ -1310,9 +1310,8 @@ class Materials(cv.CheckedList):
"""Collection of Materials used for an OpenMC simulation.
This class corresponds directly to the materials.xml input file. It can be
thought of as a normal Python list where each member is a
:class:`Material`. It behaves like a list as the following example
demonstrates:
thought of as a normal Python list where each member is a :class:`Material`.
It behaves like a list as the following example demonstrates:
>>> fuel = openmc.Material()
>>> clad = openmc.Material()
@ -1330,9 +1329,8 @@ class Materials(cv.CheckedList):
----------
cross_sections : str or path-like
Indicates the path to an XML cross section listing file (usually named
cross_sections.xml). If it is not set, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used for
continuous-energy calculations and
cross_sections.xml). If it is not set, openmc.config['cross_sections']
will be used for continuous-energy calculations and
:envvar:`OPENMC_MG_CROSS_SECTIONS` will be used for multi-group
calculations to find the path to the HDF5 cross section file.

2
openmc/mgxs_library.py
View file

@ -2546,7 +2546,7 @@ class MGXSLibrary:
"""
# If filename is None, get the cross sections from the
# OPENMC_CROSS_SECTIONS environment variable
# OPENMC_MG_CROSS_SECTIONS environment variable
if filename is None:
filename = os.environ.get('OPENMC_MG_CROSS_SECTIONS')