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Removed temporary index i in calculate_nuclide_xs.

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This commit is contained in:
Paul Romano 2011-12-06 15:03:22 -05:00
parent 8bf7b3f77e
commit 73ae7cb2bb
1 changed files with 18 additions and 22 deletions
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40
src/physics.f90
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@ -225,7 +225,6 @@ contains
integer, intent(in) :: index_nuclide ! index into nuclides array
integer, intent(in) :: index_sab ! index into sab_tables array
integer :: i ! index into nuclides array
integer :: IE ! index on nuclide energy grid
integer :: IE_sab ! index on S(a,b) energy grid
real(8) :: f ! interp factor on nuclide energy grid
@ -236,11 +235,8 @@ contains
type(Nuclide), pointer :: nuc => null()
type(SAB_Table), pointer :: sab => null()
! Copy index of nuclide
i = index_nuclide
! Set pointer to nuclide
nuc => nuclides(i)
nuc => nuclides(index_nuclide)
! TODO: If not using unionized energy grid, we need to find the index on the
! nuclide energy grid using lethargy mapping or whatever other technique
@ -249,37 +245,37 @@ contains
IE = nuc % grid_index(p % IE)
f = (p%E - nuc%energy(IE))/(nuc%energy(IE+1) - nuc%energy(IE))
micro_xs(i) % index_grid = IE
micro_xs(i) % interp_factor = f
micro_xs(index_nuclide) % index_grid = IE
micro_xs(index_nuclide) % interp_factor = f
! Initialize sab treatment to false
micro_xs(i) % use_sab = .false.
micro_xs(i) % elastic_sab = ZERO
micro_xs(index_nuclide) % use_sab = .false.
micro_xs(index_nuclide) % elastic_sab = ZERO
! Initialize nuclide cross-sections to zero
micro_xs(i) % fission = ZERO
micro_xs(i) % nu_fission = ZERO
micro_xs(index_nuclide) % fission = ZERO
micro_xs(index_nuclide) % nu_fission = ZERO
! Calculate microscopic nuclide total cross section
micro_xs(i) % total = &
micro_xs(index_nuclide) % total = &
(ONE-f) * nuc % total(IE) + f * nuc % total(IE+1)
! Calculate microscopic nuclide total cross section
micro_xs(i) % elastic = &
micro_xs(index_nuclide) % elastic = &
(ONE-f) * nuc % elastic(IE) + f * nuc % elastic(IE+1)
! Calculate microscopic nuclide absorption cross section
micro_xs(i) % absorption = &
micro_xs(index_nuclide) % absorption = &
(ONE-f) * nuc % absorption(IE) + f * nuc % absorption(IE+1)
if (nuc % fissionable) then
! Calculate microscopic nuclide total cross section
micro_xs(i) % fission = &
micro_xs(index_nuclide) % fission = &
(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(i) % nu_fission = nu * micro_xs(i) % fission
micro_xs(index_nuclide) % nu_fission = nu * micro_xs(index_nuclide) % fission
end if
! If there is S(a,b) data for this nuclide, we need to do a few
@ -288,7 +284,7 @@ contains
! then add back in the calculated S(a,b) elastic+inelastic cross section.
if (index_sab > 0) then
micro_xs(i) % use_sab = .true.
micro_xs(index_nuclide) % use_sab = .true.
! Get pointer to S(a,b) table
sab => sab_tables(index_sab)
@ -344,16 +340,16 @@ contains
end if
! Correct total and elastic cross sections
micro_xs(i) % total = micro_xs(i) % total - micro_xs(i) % elastic &
+ inelastic + elastic
micro_xs(i) % elastic = inelastic + elastic
micro_xs(index_nuclide) % total = micro_xs(index_nuclide) % total - &
micro_xs(index_nuclide) % elastic + inelastic + elastic
micro_xs(index_nuclide) % elastic = inelastic + elastic
! Store S(a,b) elastic cross section for sampling later
micro_xs(i) % elastic_sab = elastic
micro_xs(index_nuclide) % elastic_sab = elastic
end if
! Set last evaluated energy
micro_xs(i) % last_E = p % E
micro_xs(index_nuclide) % last_E = p % E
end subroutine calculate_nuclide_xs