diff --git a/src/cross_section.F90 b/src/cross_section.F90
index 82e3c00e8..22598dc42 100644
--- a/src/cross_section.F90
+++ b/src/cross_section.F90
@@ -157,7 +157,7 @@ contains
     real(8) :: sig_t, sig_a, sig_f ! Intermediate multipole variables
 
     ! Initialize cached cross sections to zero
-    micro_xs(i_nuclide) % elastic         = ZERO
+    micro_xs(i_nuclide) % elastic         = -ONE
     micro_xs(i_nuclide) % thermal         = ZERO
     micro_xs(i_nuclide) % thermal_elastic = ZERO
 
@@ -442,16 +442,7 @@ contains
     micro_xs(i_nuclide) % thermal_elastic = sab_frac * elastic
 
     ! Calculate free atom elastic cross section
-    f = micro_xs(i_nuclide) % interp_factor
-    i_grid = micro_xs(i_nuclide) % index_grid
-    i_temp = micro_xs(i_nuclide) % index_temp
-    if (i_temp > 0) then
-      associate (xs => nuclides(i_nuclide) % reactions(1) % xs(i_temp) % value)
-        micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
-      end associate
-    else
-      micro_xs(i_nuclide) % elastic = ZERO
-    end if
+    call calculate_elastic_xs(i_nuclide)
 
     ! Correct total and elastic cross sections
     micro_xs(i_nuclide) % total = micro_xs(i_nuclide) % total &
@@ -581,6 +572,7 @@ contains
 
       ! Multiply by smooth cross-section if needed
       if (urr % multiply_smooth) then
+        call calculate_elastic_xs(i_nuclide)
         elastic = elastic * micro_xs(i_nuclide) % elastic
         capture = capture * (micro_xs(i_nuclide) % absorption - &
              micro_xs(i_nuclide) % fission)
@@ -609,6 +601,36 @@ contains
 
   end subroutine calculate_urr_xs
 
+!===============================================================================
+! CALCULATE_ELASTIC_XS precalculates the free atom elastic scattering cross
+! section. Normally it is not needed until a collision actually occurs in a
+! material. However, in the thermal and unresolved resonance regions, we have to
+! calculate it early to adjust the total cross section correctly.
+!===============================================================================
+
+  subroutine calculate_elastic_xs(i_nuclide)
+    integer, intent(in) :: i_nuclide
+
+    integer :: i_temp
+    integer :: i_grid
+    real(8) :: f
+
+    ! Get temperature index, grid index, and interpolation factor
+    i_temp =  micro_xs(i_nuclide) % index_temp
+    i_grid =  micro_xs(i_nuclide) % index_grid
+    f      =  micro_xs(i_nuclide) % interp_factor
+
+    if (i_temp > 0) then
+      associate (xs => nuclides(i_nuclide) % reactions(1) % xs(i_temp) % value)
+        micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
+      end associate
+    else
+      ! For multipole, elastic is total - absorption
+      micro_xs(i_nuclide) % elastic = micro_xs(i_nuclide) % total - &
+           micro_xs(i_nuclide) % absorption
+    end if
+  end subroutine calculate_elastic_xs
+
 !===============================================================================
 ! MULTIPOLE_EVAL evaluates the windowed multipole equations for cross
 ! sections in the resolved resonance regions
diff --git a/src/nuclide_header.F90 b/src/nuclide_header.F90
index 23e6e5825..328fe123a 100644
--- a/src/nuclide_header.F90
+++ b/src/nuclide_header.F90
@@ -43,9 +43,9 @@ module nuclide_header
        XS_FISSION    = 3, &
        XS_NU_FISSION = 4
 
-  ! The array within SumXS is of shape (6, n_energy) where the first dimension
-  ! corresponds to the following values: 1) total, 2) elastic scattering, 3)
-  ! fission, 4) neutron production, 5) absorption (MT > 100), 6) heating
+  ! The array within SumXS is of shape (4, n_energy) where the first dimension
+  ! corresponds to the following values: 1) total, 2) absorption (MT > 100), 3)
+  ! fission, 4) neutron production
   type SumXS
     real(8), allocatable :: value(:,:)
   end type SumXS
@@ -576,7 +576,7 @@ contains
     do i = 1, n_temperature
       ! Allocate and initialize derived cross sections
       n_grid = size(this % grid(i) % energy)
-      allocate(this % xs(i) % value(6,n_grid))
+      allocate(this % xs(i) % value(4,n_grid))
       this % xs(i) % value(:,:) = ZERO
     end do
 
diff --git a/src/physics.F90 b/src/physics.F90
index 5d31d87d7..cb2877824 100644
--- a/src/physics.F90
+++ b/src/physics.F90
@@ -2,7 +2,7 @@ module physics
 
   use algorithm,              only: binary_search
   use constants
-  use cross_section,          only: elastic_xs_0K
+  use cross_section,          only: elastic_xs_0K, calculate_elastic_xs
   use endf,                   only: reaction_name
   use error,                  only: fatal_error, warning, write_message
   use material_header,        only: Material, materials
@@ -338,16 +338,9 @@ contains
          micro_xs(i_nuclide) % absorption)
     sampled = .false.
 
-    ! Calculate elastic cross section if need be
-    if (micro_xs(i_nuclide) % elastic == ZERO) then
-      if (i_temp > 0) then
-        associate (xs => nuc % reactions(1) % xs(i_temp) % value)
-          micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
-        end associate
-      else
-        micro_xs(i_nuclide) % elastic = micro_xs(i_nuclide) % total - &
-             micro_xs(i_nuclide) % absorption
-      end if
+    ! Calculate elastic cross section if it wasn't precalculated
+    if (micro_xs(i_nuclide) % elastic < ZERO) then
+      call calculate_elastic_xs(i_nuclide)
     end if
 
     prob = micro_xs(i_nuclide) % elastic - micro_xs(i_nuclide) % thermal