in memory cross-section alteration

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)

[dave-soliton]

Description

This PR adds support for accessing and modifying continuous-energy nuclide reaction cross sections directly from the Python API through the shared library interface.
Currently, perturbation studies involving nuclear data require generating modified HDF5 cross section libraries and reloading them into OpenMC. This workflow introduces additional file generation and I/O overhead and makes iterative sensitivity or uncertainty studies more cumbersome.
The additions in this PR expose reaction data structures needed for in-memory cross section perturbation workflows, including:

• MT reaction identifiers
• Reaction energy grids
• Reaction threshold indices
• Read/write access to reaction cross sections
• Rebuilding nuclide reaction data following modification

These changes enable nuclear data perturbations to be performed directly in memory without requiring regenerated HDF5 libraries.

Motivation
The primary motivation is support for transport sensitivity and uncertainty analysis workflows where reaction cross sections may be repeatedly perturbed during runtime.
Implementation Notes
The implementation extends the shared-library Python bindings to expose reaction-level nuclide data and provide mechanisms for rebuilding internal nuclide data structures after modification.
Testing
Verified retrieval and modification of reaction cross sections through the Python API
Verified updated cross sections are used during transport
Existing regression tests pass

Example
`import openmc
import openmc.lib
import numpy as np
import matplotlib.pyplot as plt

import numpy as np

def build_thresholded_xs(E, xs_reaction, threshold_energy):
"""
Build an XS array aligned to the unionized grid E,
with zeros below threshold_energy.
"""
xs_full = np.zeros_like(E, dtype=xs_reaction.dtype)

# Find the first index in E that is >= threshold
idx_start = np.searchsorted(E, threshold_energy, side='left')
# Fill XS values starting from threshold
n_fill = min(len(xs_reaction), len(E) - idx_start)
xs_full[idx_start : idx_start + n_fill] = xs_reaction[:n_fill]
return xs_full

============================================================

Materials

============================================================

metal

metal = openmc.Material(name="Iron")
metal.add_element("Fe", 1.0)
metal.set_density("g/cm3", 7.87)

materials = openmc.Materials([metal])
materials.export_to_xml()

============================================================

Geometry

============================================================

Sphere

sphere = openmc.Sphere(r=40.0, boundary_type='vacuum')

region = -sphere # inside sphere
cell = openmc.Cell(fill=metal, region=region)

geom = openmc.Geometry([cell])
geom.export_to_xml()

============================================================

Fixed Source

============================================================

monoenergetic source at the origin

source = openmc.Source()
source.space = openmc.stats.Point((0.0, 0.0, 0.0))
source.energy = openmc.stats.Discrete([2.4e6], [1.0])
source.angle = openmc.stats.Isotropic()

============================================================

Settings

============================================================

settings = openmc.Settings()
settings.run_mode = "fixed source"

settings.source = source
settings.batches = 1
settings.particles = int(2e5)

settings.export_to_xml()

============================================================

Tallies

============================================================

tally = openmc.Tally(tally_id=1, name="heating")
tally.scores = ["heating"]
tally.filters = [openmc.CellFilter(cell)]

============================================================

Model

============================================================

model = openmc.Model(geom, materials, settings)
model.tallies.append(tally)
model.export_to_xml()

============================================================

Run normal OpenMC (unperturbed)

============================================================

print("\n--- Running base case ---\n")
openmc.lib.init()
openmc.lib.simulation_init()

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT base case")

============================================================

Access cross section data

============================================================

nuclide_name = 'Fe56'
mt = 2 # elastic scattering
T_index = 0

Get index of nuclide

idx = next(i for i, n in enumerate(openmc.lib.nuclides) if n == nuclide_name)
print(f"{nuclide_name} index = {idx}")
mts = openmc.lib.get_nuclide_mt_numbers(idx)
print(f"Nuclide {nuclide_name} has MT numbers: {mts}")

Retrieve xs grid

xs = openmc.lib.get_nuclide_xs(idx, mt, T_index)

Retrieve energy grid

E = openmc.lib.nuclide_energy_grid(idx, T_index)
threshold = openmc.lib.get_reaction_threshold_energy(idx, mt, T_index)
print(f"For {nuclide_name} and MT = {mt} the threshold is {threshold} eV")

xs_lat = build_thresholded_xs(E, xs, threshold)

============================================================

Apply perturbation

============================================================

xs_pass_to_openmc = xs_lat * 1.2

openmc.lib.new_nuclide_xs_with_threshold(
index=idx,
mt=mt,
energy=E,
xs_perturbed=xs_pass_to_openmc,
threshold_energy=threshold,
temperature=T_index
)

Rebuild derived cross sections to apply change properly (manual call)

openmc.lib.nuclide_rebuild_derived_xs(idx)

print("\nApplied a perturbation to cross section in memory and rebuilt derived XS.\n")

Retrieve xs grid

xs_new = openmc.lib.get_nuclide_xs(idx, mt, T_index)
xs_lat_new = build_thresholded_xs(E, xs_new, threshold)

============================================================

Plot before/after comparison

============================================================

plt.figure(figsize=(6,4))
plt.loglog(E, xs_lat, label='Original')

plt.loglog(E, xs_lat_new, label='From memory', ls='--')
plt.loglog(E, xs_pass_to_openmc, label='passed to memory', ls='none', marker='+')
plt.xlabel('Energy [eV]')
plt.ylabel('σ (barns)')
plt.title(f'{nuclide_name} MT={mt} Perturbation')
plt.grid(True, which='both', ls='--', lw=0.5)
plt.legend()
plt.tight_layout()
plt.show()

============================================================

Run with perturbed data

============================================================

tally_id = next(iter(openmc.lib.tallies))
openmc.lib.tallies[tally_id].reset()
openmc.lib.run()

result = openmc.lib.tallies[tally_id].mean[0][0]

print(f"heating = {result}, for MT {mt}")

openmc.lib.simulation_finalize()
openmc.lib.finalize()
print("\n--- Perturbed case complete ---\n")

`

Checklist

• [✅] I have performed a self-review of my own code
• [✅] I have run clang-format (version 18) on any C++ source files (if applicable)
• [✅] I have followed the style guidelines for Python source files (if applicable)