replaced physics_mg.F90:scatter with the c++ version

CMakeLists.txt

@@ -424,6 +424,7 @@ add_library(libopenmc SHARED
   src/output.cpp
   src/particle.cpp
   src/physics_common.cpp
+  src/physics_mg.cpp
   src/plot.cpp
   src/position.cpp
   src/pugixml/pugixml_c.cpp

src/mgxs_interface.F90

@@ -156,7 +156,7 @@ module mgxs_interface
   real(8), allocatable :: energy_bins(:)
 
   ! Midpoint of the energy group structure
-  real(8), allocatable :: energy_bin_avg(:)
+  real(C_DOUBLE), allocatable :: energy_bin_avg(:)
 
   ! Energy group structure with increasing energy
   real(C_DOUBLE), allocatable, target :: rev_energy_bins(:)

src/physics_mg.F90

@@ -20,6 +20,14 @@ module physics_mg
 
   implicit none
 
+  interface
+    subroutine scatter(p, energy_bin_avg) bind(C)
+      import Particle, C_DOUBLE
+      type(Particle), intent(inout) :: p
+      real(C_DOUBLE), intent(in)    :: energy_bin_avg(*)
+    end subroutine scatter
+  end interface
+
 contains
 
 !===============================================================================
@@ -85,7 +93,7 @@ contains
 
     ! Sample a scattering reaction and determine the secondary energy of the
     ! exiting neutron
-    call scatter(p)
+    call scatter(p, energy_bin_avg)
 
     ! Play russian roulette if survival biasing is turned on
     if (survival_biasing) then
@@ -133,25 +141,6 @@ contains
 
   end subroutine absorption
 
-!===============================================================================
-! SCATTER
-!===============================================================================
-
-  subroutine scatter(p)
-
-    type(Particle), intent(inout)  :: p
-
-    call sample_scatter_c(p % material, p % last_g, p % g, p % mu, &
-         p % wgt, p % coord(1) % uvw)
-
-    ! Update energy value for downstream compatability (in tallying)
-    p % E = energy_bin_avg(p % g)
-
-    ! Set event component
-    p % event = EVENT_SCATTER
-
-  end subroutine scatter
-
 !===============================================================================
 ! CREATE_FISSION_SITES determines the average total, prompt, and delayed
 ! neutrons produced from fission and creates appropriate bank sites.