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Don't cache macroscopic cross sections for depletion reactions

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This commit is contained in:
Paul Romano 2017-12-23 12:53:54 +07:00
parent 26eedca662
commit e137187510
3 changed files with 69 additions and 53 deletions
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32
src/cross_section.F90
View file

@ -49,12 +49,6 @@ contains
material_xs % absorption = ZERO
material_xs % fission = ZERO
material_xs % nu_fission = ZERO
material_xs % n2n = ZERO
material_xs % n3n = ZERO
material_xs % n4n = ZERO
material_xs % ngamma = ZERO
material_xs % np = ZERO
material_xs % nalpha = ZERO
! Exit subroutine if material is void
if (p % material == MATERIAL_VOID) return
@ -139,32 +133,6 @@ contains
! Add contributions to material macroscopic nu-fission cross section
material_xs % nu_fission = material_xs % nu_fission + &
atom_density * micro_xs(i_nuclide) % nu_fission
if (in_active) then
! Add contributions to material macroscopic n2n cross section
material_xs % n2n = material_xs % n2n + &
atom_density * micro_xs(i_nuclide) % n2n
! Add contributions to material macroscopic n3n cross section
material_xs % n3n = material_xs % n3n + &
atom_density * micro_xs(i_nuclide) % n3n
! Add contributions to material macroscopic n4n cross section
material_xs % n4n = material_xs % n4n + &
atom_density * micro_xs(i_nuclide) % n4n
! Add contributions to material macroscopic ngamma cross section
material_xs % ngamma = material_xs % ngamma + &
atom_density * micro_xs(i_nuclide) % ngamma
! Add contributions to material macroscopic np cross section
material_xs % np = material_xs % np + &
atom_density * micro_xs(i_nuclide) % np
! Add contributions to material macroscopic nalpha cross section
material_xs % nalpha = material_xs % nalpha + &
atom_density * micro_xs(i_nuclide) % nalpha
end if
end do
end associate

24
src/nuclide_header.F90
View file

@ -115,15 +115,15 @@ module nuclide_header
real(8) :: total
real(8) :: elastic ! If sab_frac is not 1 or 0, then this value is
! averaged over bound and non-bound nuclei
real(8) :: absorption
real(8) :: fission
real(8) :: nu_fission
real(8) :: n2n
real(8) :: n3n
real(8) :: n4n
real(8) :: ngamma
real(8) :: np
real(8) :: nalpha
real(8) :: absorption ! absorption (disappearance)
real(8) :: fission ! fission
real(8) :: nu_fission ! neutron production from fission
real(8) :: n2n ! (n,2n)
real(8) :: n3n ! (n,3n)
real(8) :: n4n ! (n,4n)
real(8) :: ngamma ! (n,gamma)
real(8) :: np ! (n,p)
real(8) :: nalpha ! (n,alpha)
real(8) :: thermal ! Bound thermal elastic & inelastic scattering
real(8) :: thermal_elastic ! Bound thermal elastic scattering
@ -154,12 +154,6 @@ module nuclide_header
real(8) :: absorption ! macroscopic absorption xs
real(8) :: fission ! macroscopic fission xs
real(8) :: nu_fission ! macroscopic production xs
real(8) :: n2n ! macroscopic (n,2n) xs
real(8) :: n3n ! macroscopic (n,3n) xs
real(8) :: n4n ! macroscopic (n,4n) xs
real(8) :: ngamma ! macroscopic (n,gamma) xs
real(8) :: np ! macroscopic (n,p) xs
real(8) :: nalpha ! macroscopic (n,alpha) xs
end type MaterialMacroXS
!===============================================================================

66
src/tallies/tally.F90
View file

@ -1141,7 +1141,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n2n * atom_density * flux
else
score = material_xs % n2n * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % n2n * atom_density_ * flux
end do
end associate
end if
end if
end if
@ -1155,7 +1164,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n3n * atom_density * flux
else
score = material_xs % n3n * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % n3n * atom_density_ * flux
end do
end associate
end if
end if
end if
@ -1169,7 +1187,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n4n * atom_density * flux
else
score = material_xs % n4n * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % n4n * atom_density_ * flux
end do
end associate
end if
end if
end if
@ -1183,7 +1210,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % np * atom_density * flux
else
score = material_xs % np * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % np * atom_density_ * flux
end do
end associate
end if
end if
end if
@ -1197,7 +1233,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % nalpha * atom_density * flux
else
score = material_xs % nalpha * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % nalpha * atom_density_ * flux
end do
end associate
end if
end if
end if
@ -1211,7 +1256,16 @@ contains
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % ngamma * atom_density * flux
else
score = material_xs % ngamma * flux
score = ZERO
if (p % material /= MATERIAL_VOID) then
associate (mat => materials(p % material))
do l = 1, materials(p % material) % n_nuclides
i_nuc = mat % nuclide(l)
atom_density_ = mat % atom_density(l)
score = score + micro_xs(i_nuc) % ngamma * atom_density_ * flux
end do
end associate
end if
end if
end if