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Add the ability to tally microscopic cross sections in void materials #3771

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Add the ability to tally microscopic cross sections in void materials #3771

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e518aa8
add option to score micro xs in void material
GuySten Feb 4, 2026
f3f18e7
add unit test
GuySten Feb 4, 2026
5828551
fix error
GuySten Feb 4, 2026
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56 changes: 27 additions & 29 deletions src/tallies/tally_scoring.cpp
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Original file line number Diff line number Diff line change
Expand Up @@ -2437,24 +2437,22 @@ void score_tracklength_tally_general(
if (p.material() != MATERIAL_VOID) {
const auto& mat = model::materials[p.material()];
auto j = mat->mat_nuclide_index_[i_nuclide];
if (j == C_NONE) {
// Determine log union grid index
if (i_log_union == C_NONE) {
int neutron = ParticleType::neutron().transport_index();
i_log_union = std::log(p.E() / data::energy_min[neutron]) /
simulation::log_spacing;
}

// Update micro xs cache
if (!tally.multiply_density()) {
p.update_neutron_xs(i_nuclide, i_log_union);
atom_density = 1.0;
}
} else {
atom_density = tally.multiply_density()
? mat->atom_density(j, p.density_mult())
: 1.0;
if (j != C_NONE)
atom_density = mat->atom_density(j, p.density_mult());
}
if (atom_density > 0) {
if (!tally.multiply_density())
atom_density = 1.0;
} else if (!tally.multiply_density()) {
// Determine log union grid index
if (i_log_union == C_NONE) {
int neutron = ParticleType::neutron().transport_index();
i_log_union = std::log(p.E() / data::energy_min[neutron]) /
simulation::log_spacing;
}
// Update micro xs cache
p.update_neutron_xs(i_nuclide, i_log_union);
atom_density = 1.0;
}
}

Expand Down Expand Up @@ -2566,25 +2564,25 @@ void score_collision_tally(Particle& p)

double atom_density = 0.;
if (i_nuclide >= 0) {
const auto& mat = model::materials[p.material()];
auto j = mat->mat_nuclide_index_[i_nuclide];
if (j == C_NONE) {
if (p.material() != MATERIAL_VOID) {
const auto& mat = model::materials[p.material()];
auto j = mat->mat_nuclide_index_[i_nuclide];
if (j != C_NONE)
atom_density = mat->atom_density(j, p.density_mult());
}
if (atom_density > 0) {
if (!tally.multiply_density())
atom_density = 1.0;
} else if (!tally.multiply_density()) {
// Determine log union grid index
if (i_log_union == C_NONE) {
int neutron = ParticleType::neutron().transport_index();
i_log_union = std::log(p.E() / data::energy_min[neutron]) /
simulation::log_spacing;
}

// Update micro xs cache
if (!tally.multiply_density()) {
p.update_neutron_xs(i_nuclide, i_log_union);
atom_density = 1.0;
}
} else {
atom_density = tally.multiply_density()
? mat->atom_density(j, p.density_mult())
: 1.0;
p.update_neutron_xs(i_nuclide, i_log_union);
atom_density = 1.0;
}
}

Expand Down
42 changes: 42 additions & 0 deletions tests/unit_tests/test_void.py
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Original file line number Diff line number Diff line change
@@ -0,0 +1,42 @@
import numpy as np
import openmc
import pytest


@pytest.fixture
def empty_sphere():
openmc.reset_auto_ids()
model = openmc.model.Model()
surf = openmc.Sphere(r=10, boundary_type = 'vacuum')
cell = openmc.Cell(region=-surf)
model.geometry = openmc.Geometry([cell])

model.settings.run_mode = 'fixed source'
model.settings.batches = 3
model.settings.particles = 1000
model.settings.source = openmc.IndependentSource(space=openmc.stats.Point())

tally = openmc.Tally()
tally.scores = ['fission']
tally.nuclides = ['U235']

model.tallies.append(tally)

return model

def test_equivalent_microxs(empty_sphere, run_in_tmpdir):
sp_file = empty_sphere.run()
with openmc.StatePoint(sp_file) as sp:
tally1 = sp.tallies[1]

mat = openmc.Material()
mat.add_nuclide('H1', 1e-16)

empty_sphere.geometry.get_all_cells()[1].fill = mat
empty_sphere.materials.append(mat)

sp_file = empty_sphere.run()
with openmc.StatePoint(sp_file) as sp:
tally2 = sp.tallies[1]

assert np.isclose(tally1.mean.sum(),tally2.mean.sum(), rtol=1e-10, atol=0)
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