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Make sure elastic scattering cross section is precalculated for use in URR

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This commit is contained in:
Paul Romano 2018-01-15 17:25:21 -06:00
parent 0cf1f918f5
commit 7d7453b8ff
3 changed files with 41 additions and 26 deletions
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44
src/cross_section.F90
View file

@ -157,7 +157,7 @@ contains
real(8) :: sig_t, sig_a, sig_f ! Intermediate multipole variables
! Initialize cached cross sections to zero
micro_xs(i_nuclide) % elastic = ZERO
micro_xs(i_nuclide) % elastic = -ONE
micro_xs(i_nuclide) % thermal = ZERO
micro_xs(i_nuclide) % thermal_elastic = ZERO
@ -442,16 +442,7 @@ contains
micro_xs(i_nuclide) % thermal_elastic = sab_frac * elastic
! Calculate free atom elastic cross section
f = micro_xs(i_nuclide) % interp_factor
i_grid = micro_xs(i_nuclide) % index_grid
i_temp = micro_xs(i_nuclide) % index_temp
if (i_temp > 0) then
associate (xs => nuclides(i_nuclide) % reactions(1) % xs(i_temp) % value)
micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
end associate
else
micro_xs(i_nuclide) % elastic = ZERO
end if
call calculate_elastic_xs(i_nuclide)
! Correct total and elastic cross sections
micro_xs(i_nuclide) % total = micro_xs(i_nuclide) % total &
@ -581,6 +572,7 @@ contains
! Multiply by smooth cross-section if needed
if (urr % multiply_smooth) then
call calculate_elastic_xs(i_nuclide)
elastic = elastic * micro_xs(i_nuclide) % elastic
capture = capture * (micro_xs(i_nuclide) % absorption - &
micro_xs(i_nuclide) % fission)
@ -609,6 +601,36 @@ contains
end subroutine calculate_urr_xs
!===============================================================================
! CALCULATE_ELASTIC_XS precalculates the free atom elastic scattering cross
! section. Normally it is not needed until a collision actually occurs in a
! material. However, in the thermal and unresolved resonance regions, we have to
! calculate it early to adjust the total cross section correctly.
!===============================================================================
subroutine calculate_elastic_xs(i_nuclide)
integer, intent(in) :: i_nuclide
integer :: i_temp
integer :: i_grid
real(8) :: f
! Get temperature index, grid index, and interpolation factor
i_temp = micro_xs(i_nuclide) % index_temp
i_grid = micro_xs(i_nuclide) % index_grid
f = micro_xs(i_nuclide) % interp_factor
if (i_temp > 0) then
associate (xs => nuclides(i_nuclide) % reactions(1) % xs(i_temp) % value)
micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
end associate
else
! For multipole, elastic is total - absorption
micro_xs(i_nuclide) % elastic = micro_xs(i_nuclide) % total - &
micro_xs(i_nuclide) % absorption
end if
end subroutine calculate_elastic_xs
!===============================================================================
! MULTIPOLE_EVAL evaluates the windowed multipole equations for cross
! sections in the resolved resonance regions

8
src/nuclide_header.F90
View file

@ -43,9 +43,9 @@ module nuclide_header
XS_FISSION = 3, &
XS_NU_FISSION = 4
! The array within SumXS is of shape (6, n_energy) where the first dimension
! corresponds to the following values: 1) total, 2) elastic scattering, 3)
! fission, 4) neutron production, 5) absorption (MT > 100), 6) heating
! The array within SumXS is of shape (4, n_energy) where the first dimension
! corresponds to the following values: 1) total, 2) absorption (MT > 100), 3)
! fission, 4) neutron production
type SumXS
real(8), allocatable :: value(:,:)
end type SumXS
@ -576,7 +576,7 @@ contains
do i = 1, n_temperature
! Allocate and initialize derived cross sections
n_grid = size(this % grid(i) % energy)
allocate(this % xs(i) % value(6,n_grid))
allocate(this % xs(i) % value(4,n_grid))
this % xs(i) % value(:,:) = ZERO
end do

15
src/physics.F90
View file

@ -2,7 +2,7 @@ module physics
use algorithm, only: binary_search
use constants
use cross_section, only: elastic_xs_0K
use cross_section, only: elastic_xs_0K, calculate_elastic_xs
use endf, only: reaction_name
use error, only: fatal_error, warning, write_message
use material_header, only: Material, materials
@ -338,16 +338,9 @@ contains
micro_xs(i_nuclide) % absorption)
sampled = .false.
! Calculate elastic cross section if need be
if (micro_xs(i_nuclide) % elastic == ZERO) then
if (i_temp > 0) then
associate (xs => nuc % reactions(1) % xs(i_temp) % value)
micro_xs(i_nuclide) % elastic = (ONE - f)*xs(i_grid) + f*xs(i_grid + 1)
end associate
else
micro_xs(i_nuclide) % elastic = micro_xs(i_nuclide) % total - &
micro_xs(i_nuclide) % absorption
end if
! Calculate elastic cross section if it wasn't precalculated
if (micro_xs(i_nuclide) % elastic < ZERO) then
call calculate_elastic_xs(i_nuclide)
end if
prob = micro_xs(i_nuclide) % elastic - micro_xs(i_nuclide) % thermal