diff --git a/src/material.cpp b/src/material.cpp
index ca29ce7404..30dfa5ed50 100644
--- a/src/material.cpp
+++ b/src/material.cpp
@@ -349,24 +349,26 @@ Material::~Material()
 void Material::finalize()
 {
   // Set fissionable if any nuclide is fissionable
-  for (const auto& i_nuc : nuclide_) {
-    if (data::nuclides[i_nuc]->fissionable_) {
-      fissionable_ = true;
-      break;
+  if (settings::run_CE) {
+    for (const auto& i_nuc : nuclide_) {
+      if (data::nuclides[i_nuc]->fissionable_) {
+        fissionable_ = true;
+        break;
+      }
     }
+
+    // Generate material bremsstrahlung data for electrons and positrons
+    if (settings::photon_transport &&
+        settings::electron_treatment == ElectronTreatment::TTB) {
+      this->init_bremsstrahlung();
+    }
+
+    // Assign thermal scattering tables
+    this->init_thermal();
   }
 
-  // Generate material bremsstrahlung data for electrons and positrons
-  if (settings::photon_transport &&
-      settings::electron_treatment == ElectronTreatment::TTB) {
-    this->init_bremsstrahlung();
-  }
-
-  // Assign thermal scattering tables
-  this->init_thermal();
-
-  // Normalize density
-  this->normalize_density();
+// Normalize density
+this->normalize_density();
 }
 
 void Material::normalize_density()
diff --git a/src/mgxs_interface.cpp b/src/mgxs_interface.cpp
index 1a6dab5c6a..77085d57c8 100644
--- a/src/mgxs_interface.cpp
+++ b/src/mgxs_interface.cpp
@@ -123,10 +123,14 @@ void MgxsInterface::create_macro_xs()
   // Therefore type(nuclides[mat->nuclide_[0]]) dictates type(macroxs).
   // At the same time, we will find the scattering type, as that will dictate
   // how we allocate the scatter object within macroxs.
+
   for (int i = 0; i < model::materials.size(); ++i) {
+    // First we have to normalize the densities as it has not been called yet
+    // for MG mode
+    auto& mat {model::materials[i]};
+    mat->finalize();
     if (kTs[i].size() > 0) {
       // Convert atom_densities to a vector
-      auto& mat {model::materials[i]};
       vector<double> atom_densities(
         mat->atom_density_.begin(), mat->atom_density_.end());
 
diff --git a/src/volume_calc.cpp b/src/volume_calc.cpp
index ae738d6bb1..3ae98fdc38 100644
--- a/src/volume_calc.cpp
+++ b/src/volume_calc.cpp
@@ -8,6 +8,7 @@
 #include "openmc/hdf5_interface.h"
 #include "openmc/material.h"
 #include "openmc/message_passing.h"
+#include "openmc/mgxs_interface.h"
 #include "openmc/nuclide.h"
 #include "openmc/output.h"
 #include "openmc/random_lcg.h"
@@ -237,7 +238,8 @@ vector<VolumeCalculation::Result> VolumeCalculation::execute() const
       // Create 2D array to store atoms/uncertainty for each nuclide. Later this
       // is compressed into vectors storing only those nuclides that are
       // non-zero
-      auto n_nuc = data::nuclides.size();
+      auto n_nuc = settings::run_CE ? data::nuclides.size()
+                                    : data::mg.nuclides_.size();
       xt::xtensor<double, 2> atoms({n_nuc, 2}, 0.0);
 
 #ifdef OPENMC_MPI
@@ -442,7 +444,8 @@ void VolumeCalculation::to_hdf5(
 
     vector<std::string> nucnames;
     for (int i_nuc : result.nuclides) {
-      nucnames.push_back(data::nuclides[i_nuc]->name_);
+      nucnames.push_back(settings::run_CE ? data::nuclides[i_nuc]->name_
+                                          : data::mg.nuclides_[i_nuc].name);
     }
 
     // Create array of total # of atoms with uncertainty for each nuclide
diff --git a/tests/regression_tests/volume_calc/test.py b/tests/regression_tests/volume_calc/test.py
index fc96e4a551..ddca3ceaa7 100644
--- a/tests/regression_tests/volume_calc/test.py
+++ b/tests/regression_tests/volume_calc/test.py
@@ -100,6 +100,8 @@ class VolumeTest(PyAPITestHarness):
             for iso in ['H1', 'O16', 'B10', 'Mo99', 'U235']:
                 mat = openmc.XSdata(iso, groups)
                 mat.order = 0
+                mat.atomic_weight_ratio = \
+                    openmc.data.atomic_mass(iso) / openmc.data.NEUTRON_MASS
                 mat.set_scatter_matrix(scatter)
                 if iso == 'U235':
                     mat.set_nu_fission(np.multiply(nu, fiss))