Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable #3980

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a694b3c
Add group_xs C API for group-averaged cross sections
yrrepy Jun 9, 2026
81e033d
Add Nuclide.group_xs Python binding
yrrepy Jun 9, 2026
d92af26
Build the group cross section table once for flux collapse
yrrepy Jun 9, 2026
5e0f004
Add unit tests for the group_xs flux-collapse path
yrrepy Jun 9, 2026
1911ba4
Make from_multigroup_flux the multi-flux collapse engine
yrrepy Jun 10, 2026
0b5b807
Group Materials.deplete by energy and temperature
yrrepy Jun 10, 2026
9aad1cc
Streamline the flux-collapse additions
yrrepy Jun 11, 2026
302e301
Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper
yrrepy Jun 10, 2026
30f5d1a
Avoid copying the whole flux batch in from_multigroup_flux
yrrepy Jun 18, 2026
0c92039
Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…
GuySten Jul 25, 2026
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9 changes: 9 additions & 0 deletions include/openmc/nuclide.h
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Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
double collapse_rate(int MT, double temperature, span<const double> energy,
span<const double> flux) const;

//! Calculate group-averaged cross sections for a reaction
//
//! \param[in] MT ENDF MT value for desired reaction
//! \param[in] temperature Temperature in [K]
//! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
//! \param[out] xs Group-averaged cross section in each energy group
void group_xs(int MT, double temperature, span<const double> energy,
span<double> xs) const;

//! Return a ParticleType object representing this nuclide
ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down
10 changes: 10 additions & 0 deletions include/openmc/reaction.h
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Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
double collapse_rate(int64_t i_temp, span<const double> energy,
span<const double> flux, const vector<double>& grid) const;

//! Calculate group-averaged cross sections over a group structure
//
//! \param[in] i_temp Temperature index
//! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
//! \param[in] grid Nuclide energy grid at i_temp
//! \param[out] xs Group-averaged cross section in each energy group;
//! must be zeroed on entry
void group_xs(int64_t i_temp, span<const double> energy,
const vector<double>& grid, span<double> xs) const;

//! Cross section at a single temperature
struct TemperatureXS {
int threshold;
Expand Down
45 changes: 32 additions & 13 deletions openmc/deplete/helpers.py
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Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

.. versionadded:: 0.12.1

.. versionchanged:: 0.15.4
Flux-collapsed rates reuse a group cross section table built once per
material temperature instead of collapsing each pair every step.

Parameters
----------
n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
nuclides : list of str
All nuclides with desired reaction rates.

Notes
-----
One group cross section table is cached per distinct material temperature,
which suits a handful of temperatures but not hundreds of distinct ones.

"""
def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
super().__init__(n_nucs, n_reacts)
self._energies = asarray(energies)
self._xs_tables = {} # group XS tables cached by temperature
self._reactions_direct = list(reactions) if reactions is not None else []
self._nuclides_direct = list(nuclides) if nuclides is not None else None

@ReactionRateHelper.nuclides.setter
def nuclides(self, nuclides):
# Invalidate cached tables when the nuclide set changes (rare)
if nuclides != self._nuclides:
self._xs_tables = {}
ReactionRateHelper.nuclides.fset(self, nuclides)
if self._reactions_direct and self._nuclides_direct is None:
self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
if self._reactions_direct:
self._rate_tally_means_cache = None

def _table(self, temperature):
"""Return the cached group cross section table for a temperature."""
if temperature not in self._xs_tables:
# Lazy import avoids the microxs -> coupled_operator -> helpers cycle
from .microxs import _build_xs_table_ce
self._xs_tables[temperature] = _build_xs_table_ce(
self.nuclides, self._scores, self._energies, temperature,
set(self.nuclides))
return self._xs_tables[temperature]

def get_material_rates(self, mat_index, nuc_index, react_index):
"""Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

mat = self._materials[mat_index]

# One matvec gives all flux-mode rates, in (nuclides, scores) order
self._results_cache[ix_(nuc_index, react_index)] = \
self._table(mat.temperature).collapse(flux)

# Overwrite the (nuclide, reaction) pairs tallied directly
for name, i_nuc in zip(self.nuclides, nuc_index):
for mt, score, i_rx in zip(self._mts, self._scores, react_index):
for score, i_rx in zip(self._scores, react_index):
if score in self._reactions_direct and name in nuclides_direct:
# Get reaction rate from tally
i_rx_direct = direct_rx_index[score]
i_nuc_direct = direct_nuc_index[name]
self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
else:
# Use flux to collapse reaction rate (per N)
nuc = openmc.lib.nuclides[name]
rate_per_nuc = nuc.collapse_rate(
mt, mat.temperature, self._energies, flux)

self._results_cache[i_nuc, i_rx] = rate_per_nuc
self._results_cache[i_nuc, i_rx] = rx_rates[
direct_nuc_index[name], direct_rx_index[score]]

return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

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Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

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Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

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Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
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a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

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Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable#3980

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 Open

yrrepy
wants to merge
10
commits into

base:develop
Choose a base branch

from

 Open


Deplete: Flux mode (mg-flux); faster collapse via SparseXSTable
#3980
Show file tree
Hide file tree
 Changes from all commits
Commits

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10 commits


Select commit
Hold shift + click to select a range
a694b3c

Add group_xs C API for group-averaged cross sections


yrrepy Jun 9, 2026
81e033d

Add Nuclide.group_xs Python binding


yrrepy Jun 9, 2026
d92af26

Build the group cross section table once for flux collapse


yrrepy Jun 9, 2026
5e0f004

Add unit tests for the group_xs flux-collapse path


yrrepy Jun 9, 2026
1911ba4

Make from_multigroup_flux the multi-flux collapse engine


yrrepy Jun 10, 2026
0b5b807

Group Materials.deplete by energy and temperature


yrrepy Jun 10, 2026
9aad1cc

Streamline the flux-collapse additions


yrrepy Jun 11, 2026
302e301

Optional/Bonus: Cache per-temperature XS table in FluxCollapseHelper


yrrepy Jun 10, 2026
30f5d1a

Avoid copying the whole flux batch in from_multigroup_flux


yrrepy Jun 18, 2026
0c92039

Merge branch 'develop' into microxs_mg-flux_faster-collapse_SparseXST…


GuySten Jul 25, 2026

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9 changes: 9 additions & 0 deletions
include/openmc/nuclide.h

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -84,6 +84,15 @@ class Nuclide {
 double collapse_rate(int MT, double temperature, span<const double> energy,
 span<const double> flux) const;

 //! Calculate group-averaged cross sections for a reaction
 //
 //! \param[in] MT ENDF MT value for desired reaction
 //! \param[in] temperature Temperature in [K]
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[out] xs Group-averaged cross section in each energy group
 void group_xs(int MT, double temperature, span<const double> energy,
 span<double> xs) const;

 //! Return a ParticleType object representing this nuclide
 ParticleType particle_type() const { return {Z_, A_, metastable_}; }

Expand Down

10 changes: 10 additions & 0 deletions
include/openmc/reaction.h

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Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -52,6 +52,16 @@ class Reaction {
 double collapse_rate(int64_t i_temp, span<const double> energy,
 span<const double> flux, const vector<double>& grid) const;

 //! Calculate group-averaged cross sections over a group structure
 //
 //! \param[in] i_temp Temperature index
 //! \param[in] energy Energy group boundaries in [eV] (size n_groups+1)
 //! \param[in] grid Nuclide energy grid at i_temp
 //! \param[out] xs Group-averaged cross section in each energy group;
 //! must be zeroed on entry
 void group_xs(int64_t i_temp, span<const double> energy,
 const vector<double>& grid, span<double> xs) const;

 //! Cross section at a single temperature
 struct TemperatureXS {
 int threshold;
Expand Down

45 changes: 32 additions & 13 deletions
openmc/deplete/helpers.py

View file


Open in desktop

Original file line numberDiff line numberDiff line change
Expand Up@@ -9,7 +9,7 @@
from numbers import Real
import sys

from numpy import dot, zeros, newaxis, asarray
from numpy import dot, zeros, newaxis, asarray, ix_

from openmc.mpi import comm
from openmc.checkvalue import check_type, check_greater_than
Expand DownExpand Up@@ -235,6 +235,10 @@ class FluxCollapseHelper(ReactionRateHelper):

 .. versionadded:: 0.12.1

 .. versionchanged:: 0.15.4
 Flux-collapsed rates reuse a group cross section table built once per
 material temperature instead of collapsing each pair every step.

 Parameters
 ----------
 n_nucs : int
Expand All@@ -255,15 +259,24 @@ class FluxCollapseHelper(ReactionRateHelper):
 nuclides : list of str
 All nuclides with desired reaction rates.

 Notes
 -----
 One group cross section table is cached per distinct material temperature,
 which suits a handful of temperatures but not hundreds of distinct ones.

 """
 def __init__(self, n_nucs, n_reacts, energies, reactions=None, nuclides=None):
 super().__init__(n_nucs, n_reacts)
 self._energies = asarray(energies)
 self._xs_tables = {} # group XS tables cached by temperature
 self._reactions_direct = list(reactions) if reactions is not None else []
 self._nuclides_direct = list(nuclides) if nuclides is not None else None

 @ReactionRateHelper.nuclides.setter
 def nuclides(self, nuclides):
 # Invalidate cached tables when the nuclide set changes (rare)
 if nuclides != self._nuclides:
 self._xs_tables = {}
 ReactionRateHelper.nuclides.fset(self, nuclides)
 if self._reactions_direct and self._nuclides_direct is None:
 self._rate_tally.nuclides = nuclides
Expand DownExpand Up@@ -347,6 +360,16 @@ def reset_tally_means(self):
 if self._reactions_direct:
 self._rate_tally_means_cache = None

 def _table(self, temperature):
 """Return the cached group cross section table for a temperature."""
 if temperature not in self._xs_tables:
 # Lazy import avoids the microxs -> coupled_operator -> helpers cycle
 from .microxs import _build_xs_table_ce
 self._xs_tables[temperature] = _build_xs_table_ce(
 self.nuclides, self._scores, self._energies, temperature,
 set(self.nuclides))
 return self._xs_tables[temperature]

 def get_material_rates(self, mat_index, nuc_index, react_index):
 """Return an array of reaction rates for a material

Expand DownExpand Up@@ -384,20 +407,16 @@ def get_material_rates(self, mat_index, nuc_index, react_index):

 mat = self._materials[mat_index]

 # One matvec gives all flux-mode rates, in (nuclides, scores) order
 self._results_cache[ix_(nuc_index, react_index)] = \
 self._table(mat.temperature).collapse(flux)

 # Overwrite the (nuclide, reaction) pairs tallied directly
 for name, i_nuc in zip(self.nuclides, nuc_index):
 for mt, score, i_rx in zip(self._mts, self._scores, react_index):
 for score, i_rx in zip(self._scores, react_index):
 if score in self._reactions_direct and name in nuclides_direct:
 # Get reaction rate from tally
 i_rx_direct = direct_rx_index[score]
 i_nuc_direct = direct_nuc_index[name]
 self._results_cache[i_nuc, i_rx] = rx_rates[i_nuc_direct, i_rx_direct]
 else:
 # Use flux to collapse reaction rate (per N)
 nuc = openmc.lib.nuclides[name]
 rate_per_nuc = nuc.collapse_rate(
 mt, mat.temperature, self._energies, flux)

 self._results_cache[i_nuc, i_rx] = rate_per_nuc
 self._results_cache[i_nuc, i_rx] = rx_rates[
 direct_nuc_index[name], direct_rx_index[score]]

 return self._results_cache

Expand Down
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