Pre-calculate microscopic nu-fission cross sections.

src/DEPENDENCIES

@@ -5,6 +5,7 @@ cross_section.o: datatypes_header.o
 cross_section.o: endf.o
 cross_section.o: error.o
 cross_section.o: fileio.o
+cross_section.o: fission.o
 cross_section.o: global.o 
 cross_section.o: material_header.o
 cross_section.o: output.o 

src/cross_section.F90

@@ -9,6 +9,7 @@ module cross_section
   use endf,                 only: reaction_name
   use error,                only: fatal_error
   use fileio,               only: read_line, skip_lines
+  use fission,              only: nu_total
   use global
   use material_header,      only: Material
   use output,               only: write_message
@@ -322,12 +323,20 @@ contains
        nuc % kT   = kT
        nuc % zaid = NXS(2)
 
+       ! read all blocks
        call read_esz(nuc)
        call read_nu_data(nuc)
        call read_reactions(nuc)
        call read_angular_dist(nuc)
        call read_energy_dist(nuc)
        call read_unr_res(nuc)
+
+       ! for fissionable nuclides, precalculate microscopic nu-fission cross
+       ! sections so that we don't need to call the nu_total function during
+       ! cross section lookups
+
+       if (nuc % fissionable) call generate_nu_fission(nuc)
+
     case (ACE_THERMAL)
        sab => sab_tables(index_table)
        sab % name = name
@@ -366,6 +375,7 @@ contains
     allocate(nuc % total(NE))
     allocate(nuc % elastic(NE))
     allocate(nuc % fission(NE))
+    allocate(nuc % nu_fission(NE))
     allocate(nuc % absorption(NE))
     allocate(nuc % heating(NE))
 
@@ -373,6 +383,7 @@ contains
     nuc % total      = ZERO
     nuc % elastic    = ZERO
     nuc % fission    = ZERO
+    nuc % nu_fission = ZERO
     nuc % absorption = ZERO
     nuc % heating    = ZERO
 
@@ -1106,6 +1117,33 @@ contains
 
   end subroutine read_unr_res
 
+!===============================================================================
+! GENERATE_NU_FISSION precalculates the microscopic nu-fission cross section for
+! a given nuclide. This is done so that the nu_total function does not need to
+! be called during cross section lookups.
+!===============================================================================
+
+  subroutine generate_nu_fission(nuc)
+
+    type(Nuclide), pointer :: nuc
+
+    integer :: i  ! index on nuclide energy grid
+    real(8) :: E  ! energy
+    real(8) :: nu ! # of neutrons per fission
+
+    do i = 1, nuc % n_grid
+       ! determine energy
+       E = nuc % energy(i)
+
+       ! determine total nu at given energy
+       nu = nu_total(nuc, E)
+
+       ! determine nu-fission microscopic cross section
+       nuc % nu_fission(i) = nu * nuc % fission(i)
+    end do
+
+  end subroutine generate_nu_fission
+
 !===============================================================================
 ! READ_THERMAL_DATA reads elastic and inelastic cross sections and corresponding
 ! secondary energy/angle distributions derived from experimental S(a,b)

src/cross_section_header.F90

@@ -81,6 +81,7 @@ module cross_section_header
      real(8), allocatable :: total(:)      ! total cross section
      real(8), allocatable :: elastic(:)    ! elastic scattering
      real(8), allocatable :: fission(:)    ! fission
+     real(8), allocatable :: nu_fission(:) ! neutron production
      real(8), allocatable :: absorption(:) ! absorption (MT > 100)
      real(8), allocatable :: heating(:)    ! heating
 

src/physics.F90

@@ -231,7 +231,6 @@ contains
     real(8) :: f_sab     ! interp factor on S(a,b) energy grid
     real(8) :: inelastic ! S(a,b) inelastic cross section
     real(8) :: elastic   ! S(a,b) elastic cross section
-    real(8) :: nu        ! total # of neutrons emitted per fission
     type(Nuclide),   pointer :: nuc => null()
     type(SAB_Table), pointer :: sab => null()
 
@@ -274,8 +273,8 @@ contains
             (ONE-f) * nuc % fission(IE) + f * nuc % fission(IE+1)
 
        ! Calculate microscopic nuclide nu-fission cross section
-       nu = nu_total(nuc, p % E)
-       micro_xs(index_nuclide) % nu_fission = nu * micro_xs(index_nuclide) % fission
+       micro_xs(index_nuclide) % nu_fission = &
+            (ONE-f) * nuc % nu_fission(IE) + f * nuc % nu_fission(IE+1)
     end if
 
     ! If there is S(a,b) data for this nuclide, we need to do a few