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move reuasble code to new parent class; 1st pass

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- new class OpenMCOperator to contain shared bits of code between the
  currently existing Operator class and the as-of-yet to be created
  FluxDepletionOperator class
- Operator is now a subclass of OpenMCOperator
- move _distribute method to OpenMCOperator
- move _get_burnable_mats, _extract_number, _set_number_from_mat,
  _set_number_from_results to OpenMCOperator
- split initial condition into openmc.lib dependent and non-dependent
  parts; dependent part goes to Operator, non-dependent part goes to
  OpenMCOperator
- move _update_materials to OpenMCOperator
- move _get_tally_nuclides to OpenMCOperator
- rename _unpack_tallies_and_normalize to _calculate_reaction_rates;
  move to OpenMCOperator
- move get_results_info to OpenMCOperator
This commit is contained in:
yardasol 2022-07-12 18:30:29 -05:00
parent b72c4593a3
commit f5db4e3afb
3 changed files with 535 additions and 319 deletions
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1
openmc/deplete/__init__.py
View file

@ -7,6 +7,7 @@ A depletion front-end tool.
from .nuclide import *
from .chain import *
from .openmc_operator import *
from .operator import *
from .reaction_rates import *
from .atom_number import *

522
openmc/deplete/openmc_operator.py Normal file
View file

@ -0,0 +1,522 @@
"""OpenMC transport operator
This module implements a transport operator for OpenMC so that it can be used by
depletion integrators. The implementation makes use of the Python bindings to
OpenMC's C API so that reading tally results and updating material number
densities is all done in-memory instead of through the filesystem.
"""
import copy
from abc import abstractmethod
from collections import OrderedDict
import os
from warnings import warn
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
from .reaction_rates import ReactionRates
from .results import Results
from .helpers import (
DirectReactionRateHelper, ChainFissionHelper, ConstantFissionYieldHelper,
FissionYieldCutoffHelper, AveragedFissionYieldHelper, EnergyScoreHelper,
SourceRateHelper, FluxCollapseHelper)
__all__ = ["OpenMCOperator", "OperatorResult"]
def _distribute(items):
"""Distribute items across MPI communicator
Parameters
----------
items : list
List of items of distribute
Returns
-------
list
Items assigned to process that called
"""
min_size, extra = divmod(len(items), comm.size)
j = 0
for i in range(comm.size):
chunk_size = min_size + int(i < extra)
if comm.rank == i:
return items[j:j + chunk_size]
j += chunk_size
class OpenMCOperator(TransportOperator):
"""OpenMC transport operator for depletion.
Instances of this class can be used to perform depletion using OpenMC as the
transport operator. Normally, a user needn't call methods of this class
directly. Instead, an instance of this class is passed to an integrator
class, such as :class:`openmc.deplete.CECMIntegrator`.
.. versionchanged:: 0.13.0
The geometry and settings parameters have been replaced with a
model parameter that takes a :class:`~openmc.model.Model` object
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.
prev_results : Results, optional
Results from a previous depletion calculation. If this argument is
specified, the depletion calculation will start from the latest state
in the previous results.
fission_q : dict, optional
Dictionary of nuclides and their fission Q values [eV]. If not given,
values will be pulled from the ``chain_file``. Only applicable
if ``"normalization_mode" == "fission-q"``
dilute_initial : float, optional
Initial atom density [atoms/cm^3] to add for nuclides that are zero
in initial condition to ensure they exist in the decay chain.
Only done for nuclides with reaction rates.
Defaults to 1.0e3.
Attributes
----------
model : openmc.model.Model
OpenMC model object
geometry : openmc.Geometry
OpenMC geometry object
settings : openmc.Settings
OpenMC settings object
dilute_initial : float
Initial atom density [atoms/cm^3] to add for nuclides that
are zero in initial condition to ensure they exist in the decay
chain. Only done for nuclides with reaction rates.
output_dir : pathlib.Path
Path to output directory to save results.
round_number : bool
Whether or not to round output to OpenMC to 8 digits.
Useful in testing, as OpenMC is incredibly sensitive to exact values.
number : openmc.deplete.AtomNumber
Total number of atoms in simulation.
nuclides_with_data : set of str
A set listing all unique nuclides available from cross_sections.xml.
chain : openmc.deplete.Chain
The depletion chain information necessary to form matrices and tallies.
reaction_rates : openmc.deplete.ReactionRates
Reaction rates from the last operator step.
burnable_mats : list of str
All burnable material IDs
heavy_metal : float
Initial heavy metal inventory [g]
local_mats : list of str
All burnable material IDs being managed by a single process
prev_res : Results or None
Results from a previous depletion calculation. ``None`` if no
results are to be used.
cleanup_when_done : bool
Whether to finalize and clear the shared library memory when the
depletion operation is complete. Defaults to clearing the library.
"""
## modify
def __init__(self, chain_file=None, fission_q=None, dilute_initial=0.0, prev_results=None):
super().__init__(chain_file, fission_q, dilute_initial, prev_results)
## clear, but we must keep the method
def __call__(self, vec, source_rate):
"""Runs a simulation.
Simulation will abort under the following circumstances:
1) No energy is computed using OpenMC tallies.
Parameters
----------
vec : list of numpy.ndarray
Total atoms to be used in function.
source_rate : float
Power in [W] or source rate in [neutron/sec]
Returns
-------
openmc.deplete.OperatorResult
Eigenvalue and reaction rates resulting from transport operator
"""
## clear, but must keep the method
def write_bos_data(step):
"""Write a state-point file with beginning of step data
Parameters
----------
step : int
Current depletion step including restarts
"""
pass
## keep
def _get_burnable_mats(self):
"""Determine depletable materials, volumes, and nuclides
Returns
-------
burnable_mats : list of str
List of burnable material IDs
volume : OrderedDict of str to float
Volume of each material in [cm^3]
nuclides : list of str
Nuclides in order of how they'll appear in the simulation.
"""
burnable_mats = set()
model_nuclides = set()
volume = OrderedDict()
self.heavy_metal = 0.0
# Iterate once through the geometry to get dictionaries
for mat in self.materials:
for nuclide in mat.get_nuclides():
model_nuclides.add(nuclide)
if mat.depletable:
burnable_mats.add(str(mat.id))
if mat.volume is None:
raise RuntimeError("Volume not specified for depletable "
"material with ID={}.".format(mat.id))
volume[str(mat.id)] = mat.volume
self.heavy_metal += mat.fissionable_mass
# Make sure there are burnable materials
if not burnable_mats:
raise RuntimeError(
"No depletable materials were found in the model.")
# Sort the sets
burnable_mats = sorted(burnable_mats, key=int)
model_nuclides = sorted(model_nuclides)
# Construct a global nuclide dictionary, burned first
nuclides = list(self.chain.nuclide_dict)
for nuc in model_nuclides:
if nuc not in nuclides:
nuclides.append(nuc)
return burnable_mats, volume, nuclides
## keep
def _extract_number(self, local_mats, volume, nuclides, prev_res=None):
"""Construct AtomNumber using geometry
Parameters
----------
local_mats : list of str
Material IDs to be managed by this process
volume : OrderedDict of str to float
Volumes for the above materials in [cm^3]
nuclides : list of str
Nuclides to be used in the simulation.
prev_res : Results, optional
Results from a previous depletion calculation
"""
self.number = AtomNumber(local_mats, nuclides, volume, len(self.chain))
if self.dilute_initial != 0.0:
for nuc in self._burnable_nucs:
self.number.set_atom_density(np.s_[:], nuc, self.dilute_initial)
# Now extract and store the number densities
# From the geometry if no previous depletion results
if prev_res is None:
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_mat(mat)
# Else from previous depletion results
else:
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_results(mat, prev_res)
def _set_number_from_mat(self, mat):
"""Extracts material and number densities from openmc.Material
Parameters
----------
mat : openmc.Material
The material to read from
"""
mat_id = str(mat.id)
for nuclide, atom_per_bcm in mat.get_nuclide_atom_densities().items():
atom_per_cc = atom_per_bcm * 1.0e24
self.number.set_atom_density(mat_id, nuclide, atom_per_cc)
def _set_number_from_results(self, mat, prev_res):
"""Extracts material nuclides and number densities.
If the nuclide concentration's evolution is tracked, the densities come
from depletion results. Else, densities are extracted from the geometry
in the summary.
Parameters
----------
mat : openmc.Material
The material to read from
prev_res : Results
Results from a previous depletion calculation
"""
mat_id = str(mat.id)
# Get nuclide lists from geometry and depletion results
depl_nuc = prev_res[-1].nuc_to_ind
geom_nuc_densities = mat.get_nuclide_atom_densities()
# Merge lists of nuclides, with the same order for every calculation
geom_nuc_densities.update(depl_nuc)
for nuclide, atom_per_bcm in geom_nuc_densities.items():
if nuclide in depl_nuc:
concentration = prev_res.get_atoms(mat_id, nuclide)[1][-1]
volume = prev_res[-1].volume[mat_id]
atom_per_cc = concentration / volume
else:
atom_per_cc = atom_per_bcm * 1.0e24
self.number.set_atom_density(mat_id, nuclide, atom_per_cc)
## keep, will need to modify
def initial_condition(self, materials):
"""Performs final setup and returns initial condition.
Parameters
----------
materials : ???
Returns
-------
list of numpy.ndarray
Total density for initial conditions.
"""
self._rate_helper.generate_tallies(materials, self.chain.reactions)
self._normalization_helper.prepare(
self.chain.nuclides, self.reaction_rates.index_nuc)
# Tell fission yield helper what materials this process is
# responsible for
self._yield_helper.generate_tallies(
materials, tuple(sorted(self._mat_index_map.values())))
# Return number density vector
return list(self.number.get_mat_slice(np.s_[:]))
## keep? there is a non-removable part of this
## that needs the openmc c executable
def _update_materials(self):
"""Updates material compositions in OpenMC on all processes."""
for rank in range(comm.size):
number_i = comm.bcast(self.number, root=rank)
for mat in number_i.materials:
nuclides = []
densities = []
for nuc in number_i.nuclides:
if nuc in self.nuclides_with_data:
val = 1.0e-24 * number_i.get_atom_density(mat, nuc)
# If nuclide is zero, do not add to the problem.
if val > 0.0:
if self.round_number:
val_magnitude = np.floor(np.log10(val))
val_scaled = val / 10**val_magnitude
val_round = round(val_scaled, 8)
val = val_round * 10**val_magnitude
nuclides.append(nuc)
densities.append(val)
else:
# Only output warnings if values are significantly
# negative. CRAM does not guarantee positive values.
if val < -1.0e-21:
print("WARNING: nuclide ", nuc, " in material ", mat,
" is negative (density = ", val, " at/barn-cm)")
number_i[mat, nuc] = 0.0
# Update densities on C API side
mat_internal = openmc.lib.materials[int(mat)]
mat_internal.set_densities(nuclides, densities)
#TODO Update densities on the Python side, otherwise the
# summary.h5 file contains densities at the first time step
## keep, but rename and modify docstring
## get_tally_nuclides
def _get_tally_nuclides(self):
"""Determine nuclides that should be tallied for reaction rates.
This method returns a list of all nuclides that have neutron data and
are listed in the depletion chain. Technically, we should tally nuclides
that may not appear in the depletion chain because we still need to get
the fission reaction rate for these nuclides in order to normalize
power, but that is left as a future exercise.
Returns
-------
list of str
Tally nuclides
"""
nuc_set = set()
# Create the set of all nuclides in the decay chain in materials marked
# for burning in which the number density is greater than zero.
for nuc in self.number.nuclides:
if nuc in self.nuclides_with_data:
if np.sum(self.number[:, nuc]) > 0.0:
nuc_set.add(nuc)
# Communicate which nuclides have nonzeros to rank 0
if comm.rank == 0:
for i in range(1, comm.size):
nuc_newset = comm.recv(source=i, tag=i)
nuc_set |= nuc_newset
else:
comm.send(nuc_set, dest=0, tag=comm.rank)
if comm.rank == 0:
# Sort nuclides in the same order as self.number
nuc_list = [nuc for nuc in self.number.nuclides
if nuc in nuc_set]
else:
nuc_list = None
# Store list of tally nuclides on each process
nuc_list = comm.bcast(nuc_list)
return [nuc for nuc in nuc_list if nuc in self.chain]
## modify
## the tally-specific methods
## for the normalization and fission helper classes
## will already work with the current implementation.
## For transport-less depletion, we'll need to write
## a new ReactionRateHelper
def _calculate_reaction_rates(self, source_rate):
"""Unpack tallies from OpenMC and return an operator result
This method uses OpenMC's C API bindings to determine the k-effective
value and reaction rates from the simulation. The reaction rates are
normalized by a helper class depending on the method being used.
Parameters
----------
source_rate : float
Power in [W] or source rate in [neutron/sec]
Returns
-------
rates : openmc.deplete.ReactionRates
Reaction rates for nuclides
"""
rates = self.reaction_rates
rates.fill(0.0)
# Extract tally bins
nuclides = self._rate_helper.nuclides
# Form fast map
nuc_ind = [rates.index_nuc[nuc] for nuc in nuclides]
react_ind = [rates.index_rx[react] for react in self.chain.reactions]
# Keep track of energy produced from all reactions in eV per source
# particle
self._normalization_helper.reset()
self._yield_helper.unpack()
# Store fission yield dictionaries
fission_yields = []
# Create arrays to store fission Q values, reaction rates, and nuclide
# numbers, zeroed out in material iteration
number = np.empty(rates.n_nuc)
fission_ind = rates.index_rx.get("fission")
# Extract results
for i, mat in enumerate(self.local_mats):
# Get tally index
mat_index = self._mat_index_map[mat]
# Zero out reaction rates and nuclide numbers
number.fill(0.0)
# Get new number densities
for nuc, i_nuc_results in zip(nuclides, nuc_ind):
number[i_nuc_results] = self.number[mat, nuc]
tally_rates = self._rate_helper.get_material_rates(
mat_index, nuc_ind, react_ind)
# Compute fission yields for this material
fission_yields.append(self._yield_helper.weighted_yields(i))
# Accumulate energy from fission
if fission_ind is not None:
self._normalization_helper.update(tally_rates[:, fission_ind])
# Divide by total number and store
rates[i] = self._rate_helper.divide_by_adens(number)
# Scale reaction rates to obtain units of reactions/sec
rates *= self._normalization_helper.factor(source_rate)
# Store new fission yields on the chain
self.chain.fission_yields = fission_yields
return rates
@abstractmethod
def _get_nuclides_with_data(self):
"""Find nuclides with cross section data"""
## Keep
def get_results_info(self):
"""Returns volume list, material lists, and nuc lists.
Returns
-------
volume : dict of str float
Volumes corresponding to materials in full_burn_dict
nuc_list : list of str
A list of all nuclide names. Used for sorting the simulation.
burn_list : list of int
A list of all material IDs to be burned. Used for sorting the simulation.
full_burn_list : list
List of all burnable material IDs
"""
nuc_list = self.number.burnable_nuclides
burn_list = self.local_mats
volume = {}
for i, mat in enumerate(burn_list):
volume[mat] = self.number.volume[i]
# Combine volume dictionaries across processes
volume_list = comm.allgather(volume)
volume = {k: v for d in volume_list for k, v in d.items()}
return volume, nuc_list, burn_list, self.burnable_mats

331
openmc/deplete/operator.py
View file

@ -24,6 +24,7 @@ from openmc.mpi import comm
from .abc import TransportOperator, OperatorResult
from .atom_number import AtomNumber
from .chain import _find_chain_file
from .openmc_operator import OpenMCOperator, _distribute
from .reaction_rates import ReactionRates
from .results import Results
from .helpers import (
@ -34,30 +35,6 @@ from .helpers import (
__all__ = ["Operator", "OperatorResult"]
def _distribute(items):
"""Distribute items across MPI communicator
Parameters
----------
items : list
List of items of distribute
Returns
-------
list
Items assigned to process that called
"""
min_size, extra = divmod(len(items), comm.size)
j = 0
for i in range(comm.size):
chunk_size = min_size + int(i < extra)
if comm.rank == i:
return items[j:j + chunk_size]
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:
@ -74,7 +51,7 @@ def _find_cross_sections(model):
return cross_sections
class Operator(TransportOperator):
class Operator(OpenMCOperator):
"""OpenMC transport operator for depletion.
Instances of this class can be used to perform depletion using OpenMC as the
@ -267,7 +244,7 @@ class Operator(TransportOperator):
# Clear out OpenMC, create task lists, distribute
openmc.reset_auto_ids()
self.burnable_mats, volume, nuclides = self._get_burnable_mats()
self.burnable_mats, volume, nuclides = super()._get_burnable_mats()
self.local_mats = _distribute(self.burnable_mats)
# Generate map from local materials => material index
@ -298,7 +275,7 @@ class Operator(TransportOperator):
if nuc.name in self.nuclides_with_data]
# Extract number densities from the geometry / previous depletion run
self._extract_number(self.local_mats, volume, nuclides, self.prev_res)
super()._extract_number(self.local_mats, volume, nuclides, self.prev_res)
# Create reaction rates array
self.reaction_rates = ReactionRates(
@ -380,7 +357,7 @@ class Operator(TransportOperator):
openmc.reset_auto_ids()
# Update tally nuclides data in preparation for transport solve
nuclides = self._get_tally_nuclides()
nuclides = super()._get_tally_nuclides()
self._rate_helper.nuclides = nuclides
self._normalization_helper.nuclides = nuclides
self._yield_helper.update_tally_nuclides(nuclides)
@ -390,7 +367,12 @@ class Operator(TransportOperator):
openmc.lib.reset_timers()
# Extract results
op_result = self._unpack_tallies_and_normalize(source_rate)
rates = self._calculate_reaction_rates(source_rate)
# Get k and uncertainty
keff = ufloat(*openmc.lib.keff())
op_result = OperatorResult(keff, rates)
return copy.deepcopy(op_result)
@ -434,138 +416,6 @@ class Operator(TransportOperator):
cell.fill = [mat.clone()
for i in range(cell.num_instances)]
def _get_burnable_mats(self):
"""Determine depletable materials, volumes, and nuclides
Returns
-------
burnable_mats : list of str
List of burnable material IDs
volume : OrderedDict of str to float
Volume of each material in [cm^3]
nuclides : list of str
Nuclides in order of how they'll appear in the simulation.
"""
burnable_mats = set()
model_nuclides = set()
volume = OrderedDict()
self.heavy_metal = 0.0
# Iterate once through the geometry to get dictionaries
for mat in self.materials:
for nuclide in mat.get_nuclides():
model_nuclides.add(nuclide)
if mat.depletable:
burnable_mats.add(str(mat.id))
if mat.volume is None:
raise RuntimeError("Volume not specified for depletable "
"material with ID={}.".format(mat.id))
volume[str(mat.id)] = mat.volume
self.heavy_metal += mat.fissionable_mass
# Make sure there are burnable materials
if not burnable_mats:
raise RuntimeError(
"No depletable materials were found in the model.")
# Sort the sets
burnable_mats = sorted(burnable_mats, key=int)
model_nuclides = sorted(model_nuclides)
# Construct a global nuclide dictionary, burned first
nuclides = list(self.chain.nuclide_dict)
for nuc in model_nuclides:
if nuc not in nuclides:
nuclides.append(nuc)
return burnable_mats, volume, nuclides
def _extract_number(self, local_mats, volume, nuclides, prev_res=None):
"""Construct AtomNumber using geometry
Parameters
----------
local_mats : list of str
Material IDs to be managed by this process
volume : OrderedDict of str to float
Volumes for the above materials in [cm^3]
nuclides : list of str
Nuclides to be used in the simulation.
prev_res : Results, optional
Results from a previous depletion calculation
"""
self.number = AtomNumber(local_mats, nuclides, volume, len(self.chain))
if self.dilute_initial != 0.0:
for nuc in self._burnable_nucs:
self.number.set_atom_density(np.s_[:], nuc, self.dilute_initial)
# Now extract and store the number densities
# From the geometry if no previous depletion results
if prev_res is None:
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_mat(mat)
# Else from previous depletion results
else:
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_results(mat, prev_res)
def _set_number_from_mat(self, mat):
"""Extracts material and number densities from openmc.Material
Parameters
----------
mat : openmc.Material
The material to read from
"""
mat_id = str(mat.id)
for nuclide, atom_per_bcm in mat.get_nuclide_atom_densities().items():
atom_per_cc = atom_per_bcm * 1.0e24
self.number.set_atom_density(mat_id, nuclide, atom_per_cc)
def _set_number_from_results(self, mat, prev_res):
"""Extracts material nuclides and number densities.
If the nuclide concentration's evolution is tracked, the densities come
from depletion results. Else, densities are extracted from the geometry
in the summary.
Parameters
----------
mat : openmc.Material
The material to read from
prev_res : Results
Results from a previous depletion calculation
"""
mat_id = str(mat.id)
# Get nuclide lists from geometry and depletion results
depl_nuc = prev_res[-1].nuc_to_ind
geom_nuc_densities = mat.get_nuclide_atom_densities()
# Merge lists of nuclides, with the same order for every calculation
geom_nuc_densities.update(depl_nuc)
for nuclide, atom_per_bcm in geom_nuc_densities.items():
if nuclide in depl_nuc:
concentration = prev_res.get_atoms(mat_id, nuclide)[1][-1]
volume = prev_res[-1].volume[mat_id]
atom_per_cc = concentration / volume
else:
atom_per_cc = atom_per_bcm * 1.0e24
self.number.set_atom_density(mat_id, nuclide, atom_per_cc)
def initial_condition(self):
"""Performs final setup and returns initial condition.
@ -589,16 +439,8 @@ class Operator(TransportOperator):
# Generate tallies in memory
materials = [openmc.lib.materials[int(i)]
for i in self.burnable_mats]
self._rate_helper.generate_tallies(materials, self.chain.reactions)
self._normalization_helper.prepare(
self.chain.nuclides, self.reaction_rates.index_nuc)
# Tell fission yield helper what materials this process is
# responsible for
self._yield_helper.generate_tallies(
materials, tuple(sorted(self._mat_index_map.values())))
# Return number density vector
return list(self.number.get_mat_slice(np.s_[:]))
return super().initial_condition(materials)
def finalize(self):
"""Finalize a depletion simulation and release resources."""
@ -659,127 +501,6 @@ class Operator(TransportOperator):
self.materials.export_to_xml()
def _get_tally_nuclides(self):
"""Determine nuclides that should be tallied for reaction rates.
This method returns a list of all nuclides that have neutron data and
are listed in the depletion chain. Technically, we should tally nuclides
that may not appear in the depletion chain because we still need to get
the fission reaction rate for these nuclides in order to normalize
power, but that is left as a future exercise.
Returns
-------
list of str
Tally nuclides
"""
nuc_set = set()
# Create the set of all nuclides in the decay chain in materials marked
# for burning in which the number density is greater than zero.
for nuc in self.number.nuclides:
if nuc in self.nuclides_with_data:
if np.sum(self.number[:, nuc]) > 0.0:
nuc_set.add(nuc)
# Communicate which nuclides have nonzeros to rank 0
if comm.rank == 0:
for i in range(1, comm.size):
nuc_newset = comm.recv(source=i, tag=i)
nuc_set |= nuc_newset
else:
comm.send(nuc_set, dest=0, tag=comm.rank)
if comm.rank == 0:
# Sort nuclides in the same order as self.number
nuc_list = [nuc for nuc in self.number.nuclides
if nuc in nuc_set]
else:
nuc_list = None
# Store list of tally nuclides on each process
nuc_list = comm.bcast(nuc_list)
return [nuc for nuc in nuc_list if nuc in self.chain]
def _unpack_tallies_and_normalize(self, source_rate):
"""Unpack tallies from OpenMC and return an operator result
This method uses OpenMC's C API bindings to determine the k-effective
value and reaction rates from the simulation. The reaction rates are
normalized by a helper class depending on the method being used.
Parameters
----------
source_rate : float
Power in [W] or source rate in [neutron/sec]
Returns
-------
openmc.deplete.OperatorResult
Eigenvalue and reaction rates resulting from transport operator
"""
rates = self.reaction_rates
rates.fill(0.0)
# Get k and uncertainty
keff = ufloat(*openmc.lib.keff())
# Extract tally bins
nuclides = self._rate_helper.nuclides
# Form fast map
nuc_ind = [rates.index_nuc[nuc] for nuc in nuclides]
react_ind = [rates.index_rx[react] for react in self.chain.reactions]
# Keep track of energy produced from all reactions in eV per source
# particle
self._normalization_helper.reset()
self._yield_helper.unpack()
# Store fission yield dictionaries
fission_yields = []
# Create arrays to store fission Q values, reaction rates, and nuclide
# numbers, zeroed out in material iteration
number = np.empty(rates.n_nuc)
fission_ind = rates.index_rx.get("fission")
# Extract results
for i, mat in enumerate(self.local_mats):
# Get tally index
mat_index = self._mat_index_map[mat]
# Zero out reaction rates and nuclide numbers
number.fill(0.0)
# Get new number densities
for nuc, i_nuc_results in zip(nuclides, nuc_ind):
number[i_nuc_results] = self.number[mat, nuc]
tally_rates = self._rate_helper.get_material_rates(
mat_index, nuc_ind, react_ind)
# Compute fission yields for this material
fission_yields.append(self._yield_helper.weighted_yields(i))
# Accumulate energy from fission
if fission_ind is not None:
self._normalization_helper.update(tally_rates[:, fission_ind])
# Divide by total number and store
rates[i] = self._rate_helper.divide_by_adens(number)
# Scale reaction rates to obtain units of reactions/sec
rates *= self._normalization_helper.factor(source_rate)
# Store new fission yields on the chain
self.chain.fission_yields = fission_yields
return OperatorResult(keff, rates)
def _get_nuclides_with_data(self, cross_sections):
"""Loads cross_sections.xml file to find nuclides with neutron data"""
nuclides = set()
@ -792,31 +513,3 @@ class Operator(TransportOperator):
nuclides.add(name)
return nuclides
def get_results_info(self):
"""Returns volume list, material lists, and nuc lists.
Returns
-------
volume : dict of str float
Volumes corresponding to materials in full_burn_dict
nuc_list : list of str
A list of all nuclide names. Used for sorting the simulation.
burn_list : list of int
A list of all material IDs to be burned. Used for sorting the simulation.
full_burn_list : list
List of all burnable material IDs
"""
nuc_list = self.number.burnable_nuclides
burn_list = self.local_mats
volume = {}
for i, mat in enumerate(burn_list):
volume[mat] = self.number.volume[i]
# Combine volume dictionaries across processes
volume_list = comm.allgather(volume)
volume = {k: v for d in volume_list for k, v in d.items()}
return volume, nuc_list, burn_list, self.burnable_mats