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Fix up material nuclide mapping

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This commit is contained in:
Paul Romano 2017-12-21 17:16:26 +07:00
parent 7f83000812
commit 4e40788128
3 changed files with 41 additions and 51 deletions
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13
src/input_xml.F90
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@ -2450,13 +2450,14 @@ contains
n_nuclides = index_nuclide
n_sab_tables = index_sab
do i=1, n_materials
! Create direct address tables for tally optimization purposes
do i = 1, n_materials
mat => materials(i)
allocate(mat % mat_nuclide_list(n_nuclides_total))
mat % mat_nuclide_list(:) = 0
do j=1, mat % n_nuclides
mat % mat_nuclide_list(mat % nuclide(j))=j
allocate(mat % mat_nuclide_index(n_nuclides))
mat % mat_nuclide_index(:) = 0
do j = 1, mat % n_nuclides
mat % mat_nuclide_index(mat % nuclide(j)) = j
end do
end do

7
src/material_header.F90
View file

@ -32,11 +32,16 @@ module material_header
character(len=104) :: name = "" ! User-defined name
integer :: n_nuclides = 0 ! number of nuclides
integer, allocatable :: nuclide(:) ! index in nuclides array
integer, allocatable :: mat_nuclide_list(:)
real(8) :: density ! total atom density in atom/b-cm
real(8), allocatable :: atom_density(:) ! nuclide atom density in atom/b-cm
real(8) :: density_gpcc ! total density in g/cm^3
! To improve performance of tallying, we store an array (direct address
! table) that indicates for each nuclide in the global nuclides(:) array the
! index of the corresponding nuclide in the Material % nuclide(:) array. If
! it is not present in the material, the entry is set to zero.
integer, allocatable :: mat_nuclide_index(:)
! Energy grid information
integer :: n_grid ! # of union material grid points
real(8), allocatable :: e_grid(:) ! union material grid energies

72
src/tallies/tally.F90
View file

@ -2377,10 +2377,6 @@ contains
i_tally = active_analog_tallies % data(i)
associate (t => tallies(i_tally) % obj)
! Get pointer to current material. We need this in order to determine what
! nuclides are in the material
mat => materials(p % material)
! Find all valid bins in each filter if they have not already been found
! for a previous tally.
do j = 1, size(t % filter)
@ -2423,33 +2419,28 @@ contains
! Nuclide logic
! Check for nuclide bins
k = 0
NUCLIDE_LOOP: do while (k < t % n_nuclide_bins)
! Increment the index in the list of nuclide bins
k = k + 1
NUCLIDE_LOOP: do k = 1, t % n_nuclide_bins
! Get index of nuclide in nuclides array
i_nuclide = t % nuclide_bins(k)
if (i_nuclide > 0) then
atom_density = -ONE
! Check to see if this nuclide was in the material of our collision
do m = 1, mat % n_nuclides
if (mat % nuclide(m) == i_nuclide) then
atom_density = mat % atom_density(m)
exit
end if
end do
if (p % material /= MATERIAL_VOID) then
! Get pointer to current material
mat => materials(p % material)
! Determine index of nuclide in Material % atom_density array
j = mat % mat_nuclide_index(i_nuclide)
if (j == 0) cycle NUCLIDE_LOOP
! Copy corresponding atom density
atom_density = mat % atom_density(j)
end if
else
atom_density = ZERO
end if
! If we found the nuclide, determine the score for each bin
if (atom_density >= ZERO) then
call score_general(p, t, (k-1)*t % n_score_bins, filter_index, &
i_nuclide, atom_density, filter_weight)
end if
call score_general(p, t, (k-1)*t % n_score_bins, filter_index, &
i_nuclide, atom_density, filter_weight)
end do NUCLIDE_LOOP
! ======================================================================
@ -2730,7 +2721,7 @@ contains
type(Particle), intent(in) :: p
real(8), intent(in) :: distance
integer :: check_value
integer :: i
integer :: i_tally
integer :: i_filt
@ -2744,7 +2735,7 @@ contains
real(8) :: filter_weight ! combined weight of all filters
logical :: finished ! found all valid bin combinations
type(Material), pointer :: mat
! Determine track-length estimate of flux
flux = p % wgt * distance
@ -2811,17 +2802,18 @@ contains
if (p % material /= MATERIAL_VOID) then
! Get pointer to current material
mat => materials(p % material)
check_value= mat % mat_nuclide_list(i_nuclide)
if (check_value == 0) then
cycle NUCLIDE_BIN_LOOP
end if
atom_density = mat % atom_density(check_value)
! Determine index of nuclide in Material % atom_density array
j = mat % mat_nuclide_index(i_nuclide)
if (j == 0) cycle NUCLIDE_BIN_LOOP
! Copy corresponding atom density
atom_density = mat % atom_density(j)
else
atom_density = ZERO
end if
end if
! Determine score for each bin
call score_general(p, t, (k-1)*t % n_score_bins, filter_index, &
i_nuclide, atom_density, flux * filter_weight)
@ -2968,19 +2960,11 @@ contains
! Get pointer to current material
mat => materials(p % material)
! Determine if nuclide is actually in material
NUCLIDE_MAT_LOOP: do j = 1, mat % n_nuclides
! If index of nuclide matches the j-th nuclide listed in the
! material, break out of the loop
if (i_nuclide == mat % nuclide(j)) exit
! If we've reached the last nuclide in the material, it means
! the specified nuclide to be tallied is not in this material
if (j == mat % n_nuclides) then
cycle NUCLIDE_BIN_LOOP
end if
end do NUCLIDE_MAT_LOOP
! Determine index of nuclide in Material % atom_density array
j = mat % mat_nuclide_index(i_nuclide)
if (j == 0) cycle NUCLIDE_BIN_LOOP
! Copy corresponding atom density
atom_density = mat % atom_density(j)
else
atom_density = ZERO