crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 06:05:58 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

added comments to delayed-nu-fission tally

Browse source
This commit is contained in:
Sam Shaner 2015-10-16 08:41:32 -04:00
parent 0b7ee32a88
commit 133bbb510b
2 changed files with 56 additions and 15 deletions
Download patch file Download diff file Expand all files Collapse all files

17
src/fission.F90
View file

@ -132,25 +132,28 @@ contains
! since no prompt or delayed data is present, this means all neutron
! emission is prompt -- WARNING: This currently returns zero. The calling
! routine needs to know this situation is occurring since we don't want
! to call yield unnecessarily if it has already been called.
! to call yield_delayed unnecessarily if it has already been called.
yield = ZERO
else if (nuc % nu_d_type == NU_TABULAR) then
lc = 1
! determine the yield for this group
! loop over delayed groups and determine the yield for the desired group
do d = 1, nuc % n_precursor
! determine number of interpolation regions and energies
NR = int(nuc % nu_d_precursor_data(lc + 1))
NE = int(nuc % nu_d_precursor_data(lc + 2 + 2*NR))
! determine delayed neutron precursor yield for group d
yield = interpolate_tab1(nuc % nu_d_precursor_data( &
lc+1:lc+2+2*NR+2*NE), E)
! check if this is the desired group
if (d == g) then
! Check if this group is the requested group
if (d == g) exit
! determine delayed neutron precursor yield for group g
yield = interpolate_tab1(nuc % nu_d_precursor_data( &
lc+1:lc+2+2*NR+2*NE), E)
exit
end if
! advance pointer
lc = lc + 2 + 2*NR + 2*NE + 1

54
src/tally.F90
View file

@ -57,7 +57,6 @@ contains
integer :: score_index ! scoring bin index
integer :: d ! delayed neutron index
integer :: d_bin ! delayed group bin index
integer :: n_bins ! number of delayed group bins
integer :: i_filter ! filter index
integer :: dg_filter ! index of delayed group filter
real(8) :: yield ! delayed neutron yield
@ -408,11 +407,6 @@ contains
! Set the delayedgroup filter index and the number of delayed group bins
dg_filter = t % find_filter(FILTER_DELAYEDGROUP)
n_bins = 0
if (dg_filter > 0) then
n_bins = t % filters(dg_filter) % n_bins
end if
if (t % estimator == ESTIMATOR_ANALOG) then
if (survival_biasing .or. p % fission) then
@ -429,7 +423,7 @@ contains
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! nu-fission
! delayed-nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO) then
! Get the event nuclide
@ -440,9 +434,15 @@ contains
! Loop over all delayed group bins and tally to them
! individually
do d_bin = 1, n_bins
do d_bin = 1, t % filters(dg_filter) % n_bins
! Get the delayed group for this bin
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the yield for this delayed group
yield = yield_delayed(nuc, p % E, d)
! Compute the score and tally to bin
score = p % absorb_wgt * yield * micro_xs(p % event_nuclide) &
% fission * nu_delayed(nuc, p % E) / &
micro_xs(p % event_nuclide) % absorption
@ -450,6 +450,9 @@ contains
end do
cycle SCORE_LOOP
else
! If the delayed group filter is not present, compute the score
! by multiplying the absorbed weight by the fraction of the
! delayed-nu-fission xs to the absorption xs
score = p % absorb_wgt * micro_xs(p % event_nuclide) &
% fission * nu_delayed(nuc, p % E) / &
micro_xs(p % event_nuclide) % absorption
@ -472,13 +475,21 @@ contains
! Loop over all delayed group bins and tally to them individually
do d_bin = 1, t % filters(dg_filter) % n_bins
! Get the delayed group for this bin
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the score and tally to bin
score = keff * p % wgt_bank / p % n_bank * p % n_delayed_bank(d)
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
else
score = ZERO
! Loop over all delayed groups and accumulate the contribution
! from each group
do d = 1, nuclides(p % event_nuclide) % n_precursor
score = score + keff * p % wgt_bank / p % n_bank * &
p % n_delayed_bank(d)
@ -498,14 +509,23 @@ contains
! Loop over all delayed group bins and tally to them individually
do d_bin = 1, t % filters(dg_filter) % n_bins
! Get the delayed group for this bin
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the yield for this delayed group
yield = yield_delayed(nuc, p % E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuclide) % fission * yield &
* nu_delayed(nuc, p % E) * atom_density * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
else
! If the delayed group filter is not present, compute the score
! by multiplying the delayed-nu-fission macro xs by the flux
score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, p % E)&
* atom_density * flux
end if
@ -519,17 +539,28 @@ contains
! Check if the delayed group filter is present
if (dg_filter > 0) then
! Loop over all nuclides in the current material
do l = 1, mat % n_nuclides
! Get atom density
atom_density_ = mat % atom_density(l)
! Get index in nuclides array
i_nuc = mat % nuclide(l)
! Loop over all delayed group bins and tally to them individually
do d_bin = 1, t % filters(dg_filter) % n_bins
! Get the delayed group for this bin
d = t % filters(dg_filter) % int_bins(d_bin)
! Get the current nuclide
nuc => nuclides(i_nuc)
! Get the yield for the desired nuclide and delayed group
yield = yield_delayed(nuc, p % E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuc) % fission * yield &
* nu_delayed(nuc, p % E) * atom_density_ * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
@ -540,12 +571,19 @@ contains
score = ZERO
! Loop over all nuclides in the current material
do l = 1, mat % n_nuclides
! Get atom density
atom_density_ = mat % atom_density(l)
! Get index in nuclides array
i_nuc = mat % nuclide(l)
! Get the current nuclide
nuc => nuclides(i_nuc)
! Accumulate the contribution from each nuclide
score = score + micro_xs(i_nuc) % fission &
* nu_delayed(nuc, p % E) * atom_density_ * flux
end do