added openmc.Material.mean_free_path function

openmc/material.py

@@ -1718,6 +1718,36 @@ def from_xml_element(cls, elem: ET.Element) -> Material:
         return mat
 
 
+    def mean_free_path(self, energy: float) -> float:
+        """Calculate the mean free path of neutrons in the material at a given
+        energy.
+
+        .. versionadded:: 0.15.3
+
+        Parameters
+        ----------
+        energy : float
+            Neutron energy in eV
+
+        Returns
+        -------
+        float
+            Mean free path in cm
+
+        """
+        from openmc.plotter import _calculate_cexs_elem_mat
+
+        energy_grid, cexs = _calculate_cexs_elem_mat(
+            this=self,
+            types=["total"],
+        )
+        total_cexs = cexs[0]
+
+        interpolated_cexs = float(np.interp(energy, energy_grid, total_cexs))
+
+        return 1.0 / interpolated_cexs
+
+
 class Materials(cv.CheckedList):
     """Collection of Materials used for an OpenMC simulation.
 

tests/unit_tests/test_material.py

@@ -710,3 +710,16 @@ def test_avoid_subnormal(run_in_tmpdir):
     # When read back in, the density should be zero
     mats = openmc.Materials.from_xml()
     assert mats[0].get_nuclide_atom_densities()['H2'] == 0.0
+
+
+def test_mean_free_path():
+
+    mat1 = openmc.Material()
+    mat1.add_nuclide('Si28', 1.0)
+    mat1.set_density('g/cm3', 2.32)
+    assert mat1.mean_free_path(energy=14e6) == pytest.approx(11.41, abs=1e-2)
+
+    mat2 = openmc.Material()
+    mat2.add_nuclide('Pb208', 1.0)
+    mat2.set_density('g/cm3', 11.34)
+    assert mat2.mean_free_path(energy=14e6) == pytest.approx(5.65, abs=1e-2)