crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-29 06:35:48 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Remove elastic from sum_xs and caches

Browse source
This commit is contained in:
Paul Romano 2018-01-12 15:44:30 -06:00
parent df40af793b
commit 21585e937c
4 changed files with 33 additions and 38 deletions
Download patch file Download diff file Expand all files Collapse all files

29
src/cross_section.F90
View file

@ -42,7 +42,6 @@ contains
! Set all material macroscopic cross sections to zero
material_xs % total = ZERO
material_xs % elastic = ZERO
material_xs % absorption = ZERO
material_xs % fission = ZERO
material_xs % nu_fission = ZERO
@ -115,10 +114,6 @@ contains
material_xs % total = material_xs % total + &
atom_density * micro_xs(i_nuclide) % total
! Add contributions to material macroscopic scattering cross section
material_xs % elastic = material_xs % elastic + &
atom_density * micro_xs(i_nuclide) % elastic
! Add contributions to material macroscopic absorption cross section
material_xs % absorption = material_xs % absorption + &
atom_density * micro_xs(i_nuclide) % absorption
@ -162,6 +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) % thermal = ZERO
micro_xs(i_nuclide) % thermal_elastic = ZERO
@ -181,7 +177,6 @@ contains
call multipole_eval(nuc % multipole, E, sqrtkT, sig_t, sig_a, sig_f)
micro_xs(i_nuclide) % total = sig_t
micro_xs(i_nuclide) % elastic = sig_t - sig_a
micro_xs(i_nuclide) % absorption = sig_a
micro_xs(i_nuclide) % fission = sig_f
@ -268,9 +263,9 @@ contains
micro_xs(i_nuclide) % total = (ONE - f) * xs % value(XS_TOTAL,i_grid) &
+ f * xs % value(XS_TOTAL,i_grid + 1)
! Calculate microscopic nuclide elastic cross section
micro_xs(i_nuclide) % elastic = (ONE - f) * xs % value(XS_ELASTIC,i_grid) &
+ f * xs % value(XS_ELASTIC,i_grid + 1)
! Calculate microscopic nuclide absorption cross section
micro_xs(i_nuclide) % absorption = (ONE - f) * xs % value(XS_ABSORPTION, &
i_grid) + f * xs % value(XS_ABSORPTION,i_grid + 1)
if (nuc % fissionable) then
! Calculate microscopic nuclide total cross section
@ -284,10 +279,6 @@ contains
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(XS_ABSORPTION, &
i_grid) + f * xs % value(XS_ABSORPTION,i_grid + 1)
end associate
! Depletion-related reactions
@ -450,6 +441,18 @@ contains
micro_xs(i_nuclide) % thermal = sab_frac * (elastic + inelastic)
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
! Correct total and elastic cross sections
micro_xs(i_nuclide) % total = micro_xs(i_nuclide) % total &
+ micro_xs(i_nuclide) % thermal &

16
src/nuclide_header.F90
View file

@ -39,11 +39,9 @@ module nuclide_header
! Positions for first dimension of Nuclide % xs
integer, parameter :: &
XS_TOTAL = 1, &
XS_ELASTIC = 2, &
XS_ABSORPTION = 2, &
XS_FISSION = 3, &
XS_NU_FISSION = 4, &
XS_ABSORPTION = 5, &
XS_HEATING = 6
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)
@ -120,11 +118,12 @@ module nuclide_header
type NuclideMicroXS
! Microscopic cross sections in barns
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 ! absorption (disappearance)
real(8) :: fission ! fission
real(8) :: nu_fission ! neutron production from fission
real(8) :: elastic ! If sab_frac is not 1 or 0, then this value is
! averaged over bound and non-bound nuclei
real(8) :: thermal ! Bound thermal elastic & inelastic scattering
real(8) :: thermal_elastic ! Bound thermal elastic scattering
@ -155,7 +154,6 @@ module nuclide_header
type MaterialMacroXS
real(8) :: total ! macroscopic total xs
real(8) :: elastic ! macroscopic elastic scattering xs
real(8) :: absorption ! macroscopic absorption xs
real(8) :: fission ! macroscopic fission xs
real(8) :: nu_fission ! macroscopic production xs
@ -606,10 +604,6 @@ contains
j = rx % xs(t) % threshold
n = size(rx % xs(t) % value)
! Copy elastic
if (rx % MT == ELASTIC) this % xs(t) % value(XS_ELASTIC,:) = &
rx % xs(t) % value
! Add contribution to total cross section
this % xs(t) % value(XS_TOTAL,j:j+n-1) = this % xs(t) % &
value(XS_TOTAL,j:j+n-1) + rx % xs(t) % value

12
src/physics.F90
View file

@ -338,6 +338,18 @@ 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
end if
prob = micro_xs(i_nuclide) % elastic - micro_xs(i_nuclide) % thermal
if (prob > cutoff) then
! =======================================================================

14
src/tallies/tally.F90
View file

@ -1009,20 +1009,6 @@ contains
! Simply count number of scoring events
score = ONE
case (ELASTIC)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= ELASTIC) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % elastic * atom_density * flux
else
score = material_xs % elastic * flux
end if
end if
case (SCORE_FISS_Q_PROMPT)
score = ZERO