From feb0054ded2f753eea1f0fa1021f50d3a260a510 Mon Sep 17 00:00:00 2001
From: Paul Romano <romano7@gmail.com>
Date: Mon, 31 Oct 2011 16:35:15 -0400
Subject: [PATCH] Removed old get_macro_xs subroutine.

---
 src/DEPENDENCIES      |  1 -
 src/cross_section.f90 | 63 -------------------------------------------
 src/tally.f90         |  1 -
 3 files changed, 65 deletions(-)

diff --git a/src/DEPENDENCIES b/src/DEPENDENCIES
index acc83a4a3b..7b0d4d71a4 100644
--- a/src/DEPENDENCIES
+++ b/src/DEPENDENCIES
@@ -174,7 +174,6 @@ string.o: error.o
 string.o: global.o
 
 tally.o: constants.o
-tally.o: cross_section.o
 tally.o: error.o
 tally.o: global.o
 tally.o: mesh.o
diff --git a/src/cross_section.f90 b/src/cross_section.f90
index 419fbea811..de240e6663 100644
--- a/src/cross_section.f90
+++ b/src/cross_section.f90
@@ -1319,69 +1319,6 @@ contains
 
   end function get_real
 
-!===============================================================================
-! GET_MACRO_XS
-!===============================================================================
-
-  function get_macro_xs(p, mat, MT) result(xs)
-
-    type(Particle), pointer    :: p
-    type(Material), pointer    :: mat
-    integer,        intent(in) :: MT
-    real(8)                    :: xs
-
-    integer :: i, j
-    integer :: n_nuclides
-    integer :: IE
-    real(8) :: density_i
-    real(8) :: sigma_i
-    real(8) :: f
-    type(Nuclide),  pointer :: nuc => null()
-    type(Reaction), pointer :: rxn => null()
-
-    ! initialize xs
-    xs = ZERO
-
-    ! loop over all nuclides in material
-    n_nuclides = mat % n_nuclides
-    do i = 1, n_nuclides
-       nuc => nuclides(mat % nuclide(i))
-
-       ! determine nuclide atom density
-       density_i = mat % atom_density(i)
-
-       ! search nuclide energy grid
-       IE = nuc%grid_index(p % IE)
-       f = (p%E - nuc%energy(IE))/(nuc%energy(IE+1) - nuc%energy(IE))
-       
-       ! handle special case of total cross section
-       if (MT == 1) then
-          xs = xs + mat % density * (ONE-f) * nuc%total(IE) + & 
-               & f * (nuc%total(IE+1))
-          cycle
-       end if
-
-       ! loop over reactions in isotope
-       do j = 1, nuc % n_reaction
-          rxn => nuc % reactions(i)
-
-          ! check for matching MT
-          if (MT /= rxn % MT) cycle
-
-          ! if energy is below threshold for this reaction, skip it
-          if (IE < rxn % IE) cycle
-
-          ! add to cumulative probability
-          sigma_i = (ONE-f) * rxn%sigma(IE-rxn%IE+1) + & 
-               & f * (rxn%sigma(IE-rxn%IE+2))
-       end do
-
-       ! calculate nuclide macroscopic cross-section
-       xs = xs + density_i * sigma_i
-    end do
-
-  end function get_macro_xs
-
 !===============================================================================
 ! READ_XSDATA reads the data in a SERPENT xsdata file and builds a dictionary to
 ! find cross-section information later on.
diff --git a/src/tally.f90 b/src/tally.f90
index f3da1f0038..be09cd69f3 100644
--- a/src/tally.f90
+++ b/src/tally.f90
@@ -1,7 +1,6 @@
 module tally
 
   use constants
-  use cross_section, only: get_macro_xs
   use error,         only: fatal_error
   use global
   use mesh,          only: get_mesh_bin, bin_to_mesh_indices, get_mesh_indices