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Make cross section arrays in sum_xs contiguous

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Paul Romano 2018-01-12 12:36:53 -06:00
parent b9286c0322
commit ac29b7c719
2 changed files with 42 additions and 42 deletions
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24
src/cross_section.F90
View file

@ -266,29 +266,29 @@ contains
micro_xs(i_nuclide) % interp_factor = f
! Calculate microscopic nuclide total cross section
micro_xs(i_nuclide) % total = (ONE - f) * xs % total(i_grid) &
+ f * xs % total(i_grid + 1)
micro_xs(i_nuclide) % total = (ONE - f) * xs % value(SUM_XS_TOTAL,i_grid) &
+ f * xs % value(SUM_XS_TOTAL,i_grid + 1)
! Calculate microscopic nuclide elastic cross section
micro_xs(i_nuclide) % elastic = (ONE - f) * xs % elastic(i_grid) &
+ f * xs % elastic(i_grid + 1)
! Calculate microscopic nuclide absorption cross section
micro_xs(i_nuclide) % absorption = (ONE - f) * xs % absorption( &
i_grid) + f * xs % absorption(i_grid + 1)
micro_xs(i_nuclide) % elastic = (ONE - f) * xs % value(SUM_XS_ELASTIC,i_grid) &
+ f * xs % value(SUM_XS_ELASTIC,i_grid + 1)
if (nuc % fissionable) then
! Calculate microscopic nuclide total cross section
micro_xs(i_nuclide) % fission = (ONE - f) * xs % fission(i_grid) &
+ f * xs % fission(i_grid + 1)
micro_xs(i_nuclide) % fission = (ONE - f) * xs % value(SUM_XS_FISSION,i_grid) &
+ f * xs % value(SUM_XS_FISSION,i_grid + 1)
! Calculate microscopic nuclide nu-fission cross section
micro_xs(i_nuclide) % nu_fission = (ONE - f) * xs % nu_fission( &
i_grid) + f * xs % nu_fission(i_grid + 1)
micro_xs(i_nuclide) % nu_fission = (ONE - f) * xs % value(SUM_XS_NU_FISSION, &
i_grid) + f * xs % value(SUM_XS_NU_FISSION,i_grid + 1)
else
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_fission = ZERO
end if
! Calculate microscopic nuclide absorption cross section
micro_xs(i_nuclide) % absorption = (ONE - f) * xs % value(SUM_XS_ABSORPTION, &
i_grid) + f * xs % value(SUM_XS_ABSORPTION,i_grid + 1)
end associate
! Depletion-related reactions

60
src/nuclide_header.F90
View file

@ -36,13 +36,22 @@ module nuclide_header
real(8), allocatable :: energy(:) ! energy values corresponding to xs
end type EnergyGrid
integer, parameter :: &
SUM_XS_TOTAL = 1, &
SUM_XS_ELASTIC = 2, &
SUM_XS_FISSION = 3, &
SUM_XS_NU_FISSION = 4, &
SUM_XS_ABSORPTION = 5, &
SUM_XS_HEATING = 6
type SumXS
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
real(8), allocatable :: value(:,:)
!!$ 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
end type SumXS
type :: Nuclide
@ -571,16 +580,8 @@ contains
do i = 1, n_temperature
! Allocate and initialize derived cross sections
n_grid = size(this % grid(i) % energy)
allocate(this % sum_xs(i) % total(n_grid))
allocate(this % sum_xs(i) % elastic(n_grid))
allocate(this % sum_xs(i) % fission(n_grid))
allocate(this % sum_xs(i) % nu_fission(n_grid))
allocate(this % sum_xs(i) % absorption(n_grid))
this % sum_xs(i) % total(:) = ZERO
this % sum_xs(i) % elastic(:) = ZERO
this % sum_xs(i) % fission(:) = ZERO
this % sum_xs(i) % nu_fission(:) = ZERO
this % sum_xs(i) % absorption(:) = ZERO
allocate(this % sum_xs(i) % value(6,n_grid))
this % sum_xs(i) % value(:,:) = ZERO
end do
i_fission = 0
@ -608,16 +609,17 @@ contains
n = size(rx % xs(t) % value)
! Copy elastic
if (rx % MT == ELASTIC) this % sum_xs(t) % elastic(:) = rx % xs(t) % value
if (rx % MT == ELASTIC) this % sum_xs(t) % value(SUM_XS_ELASTIC,:) = &
rx % xs(t) % value
! Add contribution to total cross section
this % sum_xs(t) % total(j:j+n-1) = this % sum_xs(t) % total(j:j+n-1) + &
rx % xs(t) % value
this % sum_xs(t) % value(SUM_XS_TOTAL,j:j+n-1) = this % sum_xs(t) % &
value(SUM_XS_TOTAL,j:j+n-1) + rx % xs(t) % value
! Add contribution to absorption cross section
if (is_disappearance(rx % MT)) then
this % sum_xs(t) % absorption(j:j+n-1) = this % sum_xs(t) % &
absorption(j:j+n-1) + rx % xs(t) % value
this % sum_xs(t) % value(SUM_XS_ABSORPTION,j:j+n-1) = this % sum_xs(t) % &
value(SUM_XS_ABSORPTION,j:j+n-1) + rx % xs(t) % value
end if
! Information about fission reactions
@ -633,12 +635,12 @@ contains
! Add contribution to fission cross section
if (is_fission(rx % MT)) then
this % fissionable = .true.
this % sum_xs(t) % fission(j:j+n-1) = this % sum_xs(t) % &
fission(j:j+n-1) + rx % xs(t) % value
this % sum_xs(t) % value(SUM_XS_FISSION,j:j+n-1) = this % sum_xs(t) % &
value(SUM_XS_FISSION,j:j+n-1) + rx % xs(t) % value
! Also need to add fission cross sections to absorption
this % sum_xs(t) % absorption(j:j+n-1) = this % sum_xs(t) % &
absorption(j:j+n-1) + rx % xs(t) % value
this % sum_xs(t) % value(SUM_XS_ABSORPTION,j:j+n-1) = this % sum_xs(t) % &
value(SUM_XS_ABSORPTION,j:j+n-1) + rx % xs(t) % value
! If total fission reaction is present, there's no need to store the
! reaction cross-section since it was copied to this % fission
@ -689,12 +691,10 @@ contains
! Calculate nu-fission cross section
do t = 1, n_temperature
if (this % fissionable) then
do i = 1, size(this % sum_xs(t) % fission)
this % sum_xs(t) % nu_fission(i) = this % nu(this % grid(t) % energy(i), &
EMISSION_TOTAL) * this % sum_xs(t) % fission(i)
do i = 1, n_grid
this % sum_xs(t) % value(SUM_XS_NU_FISSION,i) = this % nu(this % grid(t) % energy(i), &
EMISSION_TOTAL) * this % sum_xs(t) % value(SUM_XS_FISSION,i)
end do
else
this % sum_xs(t) % nu_fission(:) = ZERO
end if
end do
end subroutine nuclide_create_derived