mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-28 06:05:58 -04:00
Check for void materials in tracklength tallies
This commit is contained in:
Sterling Harper
parent
bc4683be1c
commit
4895402770
1 changed files with 197 additions and 162 deletions
359
src/tally.F90
359
src/tally.F90
|
|
@ -277,8 +277,8 @@ contains
|
|||
! neutrons exiting a reaction with neutrons in the exit channel
|
||||
if (p % event_MT == ELASTIC .or. p % event_MT == N_LEVEL .or. &
|
||||
(p % event_MT >= N_N1 .and. p % event_MT <= N_NC)) then
|
||||
! Don't waste time on very common reactions we know have multiplicities
|
||||
! of one.
|
||||
! Don't waste time on very common reactions we know have
|
||||
! multiplicities of one.
|
||||
score = p % last_wgt * flux
|
||||
else
|
||||
m = nuclides(p%event_nuclide)%reaction_index% &
|
||||
|
|
@ -304,8 +304,8 @@ contains
|
|||
! neutrons exiting a reaction with neutrons in the exit channel
|
||||
if (p % event_MT == ELASTIC .or. p % event_MT == N_LEVEL .or. &
|
||||
(p % event_MT >= N_N1 .and. p % event_MT <= N_NC)) then
|
||||
! Don't waste time on very common reactions we know have multiplicities
|
||||
! of one.
|
||||
! Don't waste time on very common reactions we know have
|
||||
! multiplicities of one.
|
||||
score = p % last_wgt * flux
|
||||
else
|
||||
m = nuclides(p%event_nuclide)%reaction_index% &
|
||||
|
|
@ -468,18 +468,20 @@ contains
|
|||
score = ZERO
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) % &
|
||||
nu(E, EMISSION_PROMPT) * atom_density_ * flux
|
||||
end do
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) % &
|
||||
nu(E, EMISSION_PROMPT) * atom_density_ * flux
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
||||
|
|
@ -631,6 +633,41 @@ contains
|
|||
type is (DelayedGroupFilter)
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
! Loop over all delayed group bins and tally to them
|
||||
! individually
|
||||
do d_bin = 1, filt % n_bins
|
||||
|
||||
! Get the delayed group for this bin
|
||||
d = filt % groups(d_bin)
|
||||
|
||||
! Get the yield for the desired nuclide and delayed group
|
||||
yield = nuclides(i_nuc) % nu(E, EMISSION_DELAYED, d)
|
||||
|
||||
! Compute the score and tally to bin
|
||||
score = micro_xs(i_nuc) % fission * yield &
|
||||
* atom_density_ * flux
|
||||
call score_fission_delayed_dg(t, d_bin, score, &
|
||||
score_index)
|
||||
end do
|
||||
end do
|
||||
end if
|
||||
cycle SCORE_LOOP
|
||||
end select
|
||||
else
|
||||
|
||||
score = ZERO
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
|
|
@ -639,40 +676,11 @@ contains
|
|||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
! Loop over all delayed group bins and tally to them
|
||||
! individually
|
||||
do d_bin = 1, filt % n_bins
|
||||
|
||||
! Get the delayed group for this bin
|
||||
d = filt % groups(d_bin)
|
||||
|
||||
! Get the yield for the desired nuclide and delayed group
|
||||
yield = nuclides(i_nuc) % nu(E, EMISSION_DELAYED, d)
|
||||
|
||||
! Compute the score and tally to bin
|
||||
score = micro_xs(i_nuc) % fission * yield * atom_density_ * flux
|
||||
call score_fission_delayed_dg(t, d_bin, score, score_index)
|
||||
end do
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) %&
|
||||
nu(E, EMISSION_DELAYED) * atom_density_ * flux
|
||||
end do
|
||||
cycle SCORE_LOOP
|
||||
end select
|
||||
else
|
||||
|
||||
score = ZERO
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) % &
|
||||
nu(E, EMISSION_DELAYED) * atom_density_ * flux
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
|
@ -714,7 +722,8 @@ contains
|
|||
% nu(E, EMISSION_DELAYED, d)
|
||||
|
||||
associate (rxn => nuclides(p % event_nuclide) % &
|
||||
reactions(nuclides(p % event_nuclide) % index_fission(1)))
|
||||
reactions(nuclides(p % event_nuclide) % &
|
||||
index_fission(1)))
|
||||
|
||||
! Compute the score
|
||||
score = p % absorb_wgt * yield * &
|
||||
|
|
@ -740,14 +749,17 @@ contains
|
|||
reactions(nuclides(p % event_nuclide) % index_fission(1)))
|
||||
|
||||
! We need to be careful not to overshoot the number of delayed
|
||||
! groups since this could cause the range of the rxn % products
|
||||
! array to be exceeded. Hence, we use the size of this array
|
||||
! and not the MAX_DELAYED_GROUPS constant for this loop.
|
||||
! groups since this could cause the range of the
|
||||
! rxn % products array to be exceeded. Hence, we use the size
|
||||
! of this array and not the MAX_DELAYED_GROUPS constant for
|
||||
! this loop.
|
||||
do d = 1, size(rxn % products) - 2
|
||||
|
||||
score = score + rxn % products(1 + d) % decay_rate * &
|
||||
p % absorb_wgt * micro_xs(p % event_nuclide) % fission *&
|
||||
nuclides(p % event_nuclide) % nu(E, EMISSION_DELAYED, d)&
|
||||
p % absorb_wgt &
|
||||
* micro_xs(p % event_nuclide) % fission &
|
||||
* nuclides(p % event_nuclide) % &
|
||||
nu(E, EMISSION_DELAYED, d) &
|
||||
/ micro_xs(p % event_nuclide) % absorption * flux
|
||||
end do
|
||||
end associate
|
||||
|
|
@ -794,15 +806,16 @@ contains
|
|||
select type(filt => t % filters(dg_filter) % obj)
|
||||
type is (DelayedGroupFilter)
|
||||
|
||||
! loop over delayed group bins until the corresponding bin is
|
||||
! found
|
||||
! loop over delayed group bins until the corresponding bin
|
||||
! is found
|
||||
do d_bin = 1, filt % n_bins
|
||||
d = filt % groups(d_bin)
|
||||
|
||||
! check whether the delayed group of the particle is equal to
|
||||
! the delayed group of this bin
|
||||
! check whether the delayed group of the particle is equal
|
||||
! to the delayed group of this bin
|
||||
if (d == g) then
|
||||
call score_fission_delayed_dg(t, d_bin, score, score_index)
|
||||
call score_fission_delayed_dg(t, d_bin, score, &
|
||||
score_index)
|
||||
end if
|
||||
end do
|
||||
end select
|
||||
|
|
@ -879,6 +892,52 @@ contains
|
|||
type is (DelayedGroupFilter)
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
if (nuclides(i_nuc) % fissionable) then
|
||||
|
||||
! Loop over all delayed group bins and tally to them
|
||||
! individually
|
||||
do d_bin = 1, filt % n_bins
|
||||
|
||||
! Get the delayed group for this bin
|
||||
d = filt % groups(d_bin)
|
||||
|
||||
! Get the yield for the desired nuclide and delayed
|
||||
! group
|
||||
yield = nuclides(i_nuc) % nu(E, EMISSION_DELAYED, d)
|
||||
|
||||
associate (rxn => nuclides(i_nuc) % &
|
||||
reactions(nuclides(i_nuc) % index_fission(1)))
|
||||
|
||||
! Compute the score
|
||||
score = micro_xs(i_nuc) % fission * yield * flux * &
|
||||
atom_density_ &
|
||||
* rxn % products(1 + d) % decay_rate
|
||||
end associate
|
||||
|
||||
! Tally to bin
|
||||
call score_fission_delayed_dg(t, d_bin, score, &
|
||||
score_index)
|
||||
end do
|
||||
end if
|
||||
end do
|
||||
end if
|
||||
cycle SCORE_LOOP
|
||||
end select
|
||||
else
|
||||
|
||||
score = ZERO
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
|
|
@ -889,63 +948,26 @@ contains
|
|||
|
||||
if (nuclides(i_nuc) % fissionable) then
|
||||
|
||||
! Loop over all delayed group bins and tally to them
|
||||
! individually
|
||||
do d_bin = 1, filt % n_bins
|
||||
associate (rxn => nuclides(i_nuc) % &
|
||||
reactions(nuclides(i_nuc) % index_fission(1)))
|
||||
|
||||
! Get the delayed group for this bin
|
||||
d = filt % groups(d_bin)
|
||||
! We need to be careful not to overshoot the number of
|
||||
! delayed groups since this could cause the range of the
|
||||
! rxn % products array to be exceeded. Hence, we use the
|
||||
! size of this array and not the MAX_DELAYED_GROUPS
|
||||
! constant for this loop.
|
||||
do d = 1, size(rxn % products) - 2
|
||||
|
||||
! Get the yield for the desired nuclide and delayed group
|
||||
yield = nuclides(i_nuc) % nu(E, EMISSION_DELAYED, d)
|
||||
|
||||
associate (rxn => nuclides(i_nuc) % &
|
||||
reactions(nuclides(i_nuc) % index_fission(1)))
|
||||
|
||||
! Compute the score
|
||||
score = micro_xs(i_nuc) % fission * yield * flux * &
|
||||
atom_density_ * rxn % products(1 + d) % decay_rate
|
||||
end associate
|
||||
|
||||
! Tally to bin
|
||||
call score_fission_delayed_dg(t, d_bin, score, score_index)
|
||||
end do
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission &
|
||||
* nuclides(i_nuc) % nu(E, EMISSION_DELAYED) &
|
||||
* atom_density_ * flux &
|
||||
* rxn % products(1 + d) % decay_rate
|
||||
end do
|
||||
end associate
|
||||
end if
|
||||
end do
|
||||
cycle SCORE_LOOP
|
||||
end select
|
||||
else
|
||||
|
||||
score = ZERO
|
||||
|
||||
! Loop over all nuclides in the current material
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
if (nuclides(i_nuc) % fissionable) then
|
||||
|
||||
associate (rxn => nuclides(i_nuc) % &
|
||||
reactions(nuclides(i_nuc) % index_fission(1)))
|
||||
|
||||
! We need to be careful not to overshoot the number of delayed
|
||||
! groups since this could cause the range of the rxn % products
|
||||
! array to be exceeded. Hence, we use the size of this array
|
||||
! and not the MAX_DELAYED_GROUPS constant for this loop.
|
||||
do d = 1, size(rxn % products) - 2
|
||||
|
||||
! Accumulate the contribution from each nuclide
|
||||
score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) %&
|
||||
nu(E, EMISSION_DELAYED) * atom_density_ * flux * &
|
||||
rxn % products(1 + d) % decay_rate
|
||||
end do
|
||||
end associate
|
||||
end if
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
|
@ -996,20 +1018,24 @@ contains
|
|||
end if
|
||||
end associate
|
||||
else
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
! Determine atom density and index of nuclide
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
if (p % material == MATERIAL_VOID) then
|
||||
score = ZERO
|
||||
else
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
! Determine atom density and index of nuclide
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
! If nuclide is fissionable, accumulate kappa fission
|
||||
associate(nuc => nuclides(i_nuc))
|
||||
if (nuc % fissionable) then
|
||||
score = score + &
|
||||
nuc % reactions(nuc % index_fission(1)) % Q_value * &
|
||||
micro_xs(i_nuc) % fission * atom_density_ * flux
|
||||
end if
|
||||
end associate
|
||||
end do
|
||||
! If nuclide is fissionable, accumulate kappa fission
|
||||
associate(nuc => nuclides(i_nuc))
|
||||
if (nuc % fissionable) then
|
||||
score = score + &
|
||||
nuc % reactions(nuc % index_fission(1)) % Q_value * &
|
||||
micro_xs(i_nuc) % fission * atom_density_ * flux
|
||||
end if
|
||||
end associate
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
||||
|
|
@ -1077,15 +1103,17 @@ contains
|
|||
* nuclides(i_nuclide) % fission_q_prompt % evaluate(E)
|
||||
end if
|
||||
else
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
if (allocated(nuclides(i_nuc) % fission_q_prompt)) then
|
||||
score = score + micro_xs(i_nuc) % fission * atom_density_ &
|
||||
* flux &
|
||||
* nuclides(i_nuc) % fission_q_prompt % evaluate(E)
|
||||
end if
|
||||
end do
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
if (allocated(nuclides(i_nuc) % fission_q_prompt)) then
|
||||
score = score + micro_xs(i_nuc) % fission * atom_density_ &
|
||||
* flux &
|
||||
* nuclides(i_nuc) % fission_q_prompt % evaluate(E)
|
||||
end if
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
||||
|
|
@ -1135,14 +1163,17 @@ contains
|
|||
* nuclides(i_nuclide) % fission_q_recov % evaluate(E)
|
||||
end if
|
||||
else
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
if (allocated(nuclides(i_nuc) % fission_q_recov)) then
|
||||
score = score + micro_xs(i_nuc) % fission * atom_density_ &
|
||||
* flux * nuclides(i_nuc) % fission_q_recov % evaluate(E)
|
||||
end if
|
||||
end do
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
if (allocated(nuclides(i_nuc) % fission_q_recov)) then
|
||||
score = score + micro_xs(i_nuc) % fission * atom_density_ &
|
||||
* flux &
|
||||
* nuclides(i_nuc) % fission_q_recov % evaluate(E)
|
||||
end if
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
end if
|
||||
|
||||
|
|
@ -1172,7 +1203,8 @@ contains
|
|||
i_energy = micro_xs(i_nuclide) % index_grid
|
||||
f = micro_xs(i_nuclide) % interp_factor
|
||||
|
||||
associate (xs => nuclides(i_nuclide) % reactions(m) % xs(i_temp))
|
||||
associate (xs => nuclides(i_nuclide) % reactions(m) &
|
||||
% xs(i_temp))
|
||||
if (i_energy >= xs % threshold) then
|
||||
score = ((ONE - f) * xs % value(i_energy - &
|
||||
xs % threshold + 1) + f * xs % value(i_energy - &
|
||||
|
|
@ -1182,31 +1214,34 @@ contains
|
|||
end if
|
||||
|
||||
else
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
if (p % material /= MATERIAL_VOID) then
|
||||
do l = 1, materials(p % material) % n_nuclides
|
||||
! Get atom density
|
||||
atom_density_ = materials(p % material) % atom_density(l)
|
||||
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
! Get index in nuclides array
|
||||
i_nuc = materials(p % material) % nuclide(l)
|
||||
|
||||
if (nuclides(i_nuc)%reaction_index%has_key(score_bin)) then
|
||||
m = nuclides(i_nuc)%reaction_index%get_key(score_bin)
|
||||
if (nuclides(i_nuc)%reaction_index%has_key(score_bin)) then
|
||||
m = nuclides(i_nuc)%reaction_index%get_key(score_bin)
|
||||
|
||||
! Retrieve temperature and energy grid index and interpolation
|
||||
! factor
|
||||
i_temp = micro_xs(i_nuc) % index_temp
|
||||
i_energy = micro_xs(i_nuc) % index_grid
|
||||
f = micro_xs(i_nuc) % interp_factor
|
||||
! Retrieve temperature and energy grid index and
|
||||
! interpolation factor
|
||||
i_temp = micro_xs(i_nuc) % index_temp
|
||||
i_energy = micro_xs(i_nuc) % index_grid
|
||||
f = micro_xs(i_nuc) % interp_factor
|
||||
|
||||
associate (xs => nuclides(i_nuc) % reactions(m) % xs(i_temp))
|
||||
if (i_energy >= xs % threshold) then
|
||||
score = score + ((ONE - f) * xs % value(i_energy - &
|
||||
xs % threshold + 1) + f * xs % value(i_energy - &
|
||||
xs % threshold + 2)) * atom_density_ * flux
|
||||
end if
|
||||
end associate
|
||||
end if
|
||||
end do
|
||||
associate (xs => nuclides(i_nuc) % reactions(m) &
|
||||
% xs(i_temp))
|
||||
if (i_energy >= xs % threshold) then
|
||||
score = score + ((ONE - f) * xs % value(i_energy - &
|
||||
xs % threshold + 1) + f * xs % value(i_energy - &
|
||||
xs % threshold + 2)) * atom_density_ * flux
|
||||
end if
|
||||
end associate
|
||||
end if
|
||||
end do
|
||||
end if
|
||||
end if
|
||||
|
||||
else
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue