Merge branch 'openmc-dev:develop' into Sensitivity-Analysis

Author: Ebiwonjumi

Dockerfile

@@ -24,7 +24,7 @@ ARG compile_cores=1
 ARG build_dagmc=off
 ARG build_libmesh=off
 
-FROM debian:bookworm-slim AS dependencies
+FROM ubuntu:24.04 AS dependencies
 
 ARG compile_cores
 ARG build_dagmc

docs/source/io_formats/settings.rst

@@ -487,7 +487,7 @@ found in the :ref:`random ray user guide <random_ray>`.
 
         :type:
           The type of the domain. Can be ``material``, ``cell``, or ``universe``.
-  
+
   :diagonal_stabilization_rho:
     The rho factor for use with diagonal stabilization. This technique is
     applied when negative diagonal (in-group) elements are detected in
@@ -917,6 +917,16 @@ variable and whose sub-elements/attributes are as follows:
   :dist:
     This sub-element of a ``pair`` element provides information on the corresponding univariate distribution.
 
+---------------------------------------
+``<source_rejection_fraction>`` Element
+---------------------------------------
+
+The ``<source_rejection_fraction>`` element specifies the minimum fraction of
+external source sites that must be accepted when applying rejection sampling
+based on constraints.
+
+   *Default*: 0.05
+
 -------------------------
 ``<state_point>`` Element
 -------------------------

docs/source/methods/random_ray.rst

@@ -109,9 +109,7 @@ terms on the right hand side.
 In Equation :eq:`transport`, :math:`\psi` is the angular neutron flux. This
 parameter represents the total distance traveled by all neutrons in a particular
 direction inside of a control volume per second, and is often given in units of
-:math:`1/(\text{cm}^{2} \text{s})`. As OpenMC does not support time dependence
-in the random ray solver mode, we consider the steady state equation, where the
-units of flux become :math:`1/\text{cm}^{2}`. The angular direction unit vector,
+:math:`1/(\text{cm}^{2} \text{s})`. The angular direction unit vector,
 :math:`\mathbf{\Omega}`, represents the direction of travel for the neutron. The
 spatial position vector, :math:`\mathbf{r}`,  represents the location within the
 simulation. The neutron energy, :math:`E`, or speed in continuous space, is

include/openmc/settings.h

@@ -145,6 +145,8 @@ extern std::unordered_set<int>
   statepoint_batch; //!< Batches when state should be written
 extern std::unordered_set<int>
   source_write_surf_id; //!< Surface ids where sources will be written
+extern double source_rejection_fraction; //!< Minimum fraction of source sites
+                                         //!< that must be accepted
 
 extern int
   max_history_splits; //!< maximum number of particle splits for weight windows

include/openmc/source.h

@@ -21,10 +21,9 @@ namespace openmc {
 // Constants
 //==============================================================================
 
-// Maximum number of external source spatial resamples to encounter before an
-// error is thrown.
+// Minimum number of external source sites rejected before checking againts the
+// source_rejection_fraction
 constexpr int EXTSRC_REJECT_THRESHOLD {10000};
-constexpr double EXTSRC_REJECT_FRACTION {0.05};
 
 //==============================================================================
 // Global variables

openmc/data/njoy.py

@@ -201,12 +201,12 @@
 _THERMAL_TEMPLATE_THERMR = """
 thermr / %%%%%%%%%%%%%%%% Add thermal scattering data (free gas) %%%%%%%%%%%%%%%
 0 {nthermr1_in} {nthermr1}
-0 {mat} 12 {num_temp} 1 0 {iform} 1 221 1/
+0 {mat} {nbin} {num_temp} 1 0 {iform} 1 221 1/
 {temps}
 {error} {energy_max}
 thermr / %%%%%%%%%%%%%%%% Add thermal scattering data (bound) %%%%%%%%%%%%%%%%%%
 {nthermal_endf} {nthermr2_in} {nthermr2}
-{mat_thermal} {mat} 16 {num_temp} {inelastic} {elastic} {iform} {natom} 222 1/
+{mat_thermal} {mat} {nbin} {num_temp} {inelastic} {elastic} {iform} {natom} 222 1/
 {temps}
 {error} {energy_max}
 """
@@ -495,7 +495,7 @@ def make_ace(filename, temperatures=None, acer=True, xsdir=None,
 def make_ace_thermal(filename, filename_thermal, temperatures=None,
                      ace=None, xsdir=None, output_dir=None, error=0.001,
                      iwt=2, evaluation=None, evaluation_thermal=None,
-                     table_name=None, zaids=None, nmix=None, **kwargs):
+                     table_name=None, zaids=None, nmix=None, nbin=16, **kwargs):
     """Generate thermal scattering ACE file from ENDF files
 
     Parameters
@@ -532,6 +532,8 @@ def make_ace_thermal(filename, filename_thermal, temperatures=None,
         ZAIDs that the thermal scattering data applies to
     nmix : int, optional
         Number of atom types in mixed moderator
+    nbin : int, optional
+        Number of equi-probable angles
     **kwargs
         Keyword arguments passed to :func:`openmc.data.njoy.run`
 

openmc/deplete/abc.py

@@ -24,7 +24,7 @@
 from openmc.utility_funcs import change_directory
 from openmc import Material
 from .stepresult import StepResult
-from .chain import Chain
+from .chain import _get_chain
 from .results import Results, _SECONDS_PER_MINUTE, _SECONDS_PER_HOUR, \
     _SECONDS_PER_DAY, _SECONDS_PER_JULIAN_YEAR
 from .pool import deplete
@@ -126,8 +126,8 @@ class TransportOperator(ABC):
 
     Parameters
     ----------
-    chain_file : str
-        Path to the depletion chain XML file
+    chain_file : PathLike or Chain
+        Path to the depletion chain XML file or instance of openmc.deplete.Chain.
     fission_q : dict, optional
         Dictionary of nuclides and their fission Q values [eV]. If not given,
         values will be pulled from the ``chain_file``.
@@ -145,11 +145,12 @@ class TransportOperator(ABC):
         The depletion chain information necessary to form matrices and tallies.
 
     """
-    def __init__(self, chain_file, fission_q=None, prev_results=None):
+    def __init__(self, chain_file=None, fission_q=None, prev_results=None):
         self.output_dir = '.'
 
         # Read depletion chain
-        self.chain = Chain.from_xml(chain_file, fission_q)
+        self.chain = _get_chain(chain_file, fission_q)
+
         if prev_results is None:
             self.prev_res = None
         else:

openmc/deplete/chain.py

@@ -17,8 +17,9 @@
 import lxml.etree as ET
 import scipy.sparse as sp
 
-from openmc.checkvalue import check_type, check_greater_than
+from openmc.checkvalue import check_type, check_greater_than, PathLike
 from openmc.data import gnds_name, zam
+from openmc.exceptions import DataError
 from .nuclide import FissionYieldDistribution, Nuclide
 import openmc.data
 
@@ -1246,3 +1247,38 @@ def _follow(self, isotopes, level):
         found.update(isotopes)
 
         return found
+
+
+def _get_chain(
+    chain_file: PathLike | Chain | None = None,
+    fission_q: dict | None = None
+) -> Chain:
+    """Get a depletion chain from a file or the runtime configuration.
+
+    Parameters
+    ----------
+    chain_file : PathLike or Chain, optional
+        Path to depletion chain XML file, a Chain instance, or None to use
+        the file specified in ``openmc.config['chain_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``.
+
+    Returns
+    -------
+    Chain
+        Depletion chain instance.
+    """
+    if isinstance(chain_file, Chain):
+        return chain_file
+    elif isinstance(chain_file, PathLike | None):
+        if chain_file is None:
+            chain_file = openmc.config.get('chain_file')
+            if 'chain_file' not in openmc.config:
+                raise DataError(
+                    "No depletion chain specified and could not find depletion "
+                    "chain in openmc.config['chain_file']"
+                )
+        return Chain.from_xml(chain_file, fission_q)
+    else:
+        raise TypeError("chain_file must be path-like, a Chain, or None")

openmc/deplete/coupled_operator.py

@@ -102,9 +102,9 @@ class CoupledOperator(OpenMCOperator):
     ----------
     model : openmc.model.Model
         OpenMC model object
-    chain_file : str, optional
-        Path to the depletion chain XML file. Defaults to
-        ``openmc.config['chain_file']``.
+    chain_file : PathLike or Chain, optional
+        Path to the depletion chain XML file or instance of openmc.deplete.Chain.
+        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

openmc/deplete/d1s.py

@@ -14,19 +14,20 @@
 import openmc
 from openmc.data import half_life
 from .abc import _normalize_timesteps
-from .chain import Chain
+from .chain import Chain, _get_chain
+from ..checkvalue import PathLike
 
 
-def get_radionuclides(model: openmc.Model, chain_file: str | None = None) -> list[str]:
+def get_radionuclides(model: openmc.Model, chain_file: PathLike | Chain | None = None) -> list[str]:
     """Determine all radionuclides that can be produced during D1S.
 
     Parameters
     ----------
     model : openmc.Model
         Model that should be used for determining what nuclides are present
-    chain_file : str, optional
-        Which chain file to use for inspecting decay data. If None is passed,
-        defaults to ``openmc.config['chain_file']``
+    chain_file : PathLike | Chain
+        Path to the depletion chain XML file or instance of openmc.deplete.Chain.
+        Used for inspecting decay data. Defaults to ``openmc.config['chain_file']``
 
     Returns
     -------
@@ -39,9 +40,7 @@ def get_radionuclides(model: openmc.Model, chain_file: str | None = None) -> lis
                       for nuc in mat.get_nuclides()}
 
     # Load chain file
-    if chain_file is None:
-        chain_file = openmc.config['chain_file']
-    chain = Chain.from_xml(chain_file)
+    chain = _get_chain(chain_file)
 
     radionuclides = set()
     for nuclide in chain.nuclides: