diff --git a/src/tallies/tally_scoring.cpp b/src/tallies/tally_scoring.cpp
index 51b5d9ffcc5..2f4f9bfb14a 100644
--- a/src/tallies/tally_scoring.cpp
+++ b/src/tallies/tally_scoring.cpp
@@ -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;
           }
         }
 
@@ -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;
           }
         }
 
diff --git a/tests/unit_tests/test_void.py b/tests/unit_tests/test_void.py
new file mode 100644
index 00000000000..be7a09d75e0
--- /dev/null
+++ b/tests/unit_tests/test_void.py
@@ -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)