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Outgoing energy filter can now be used with nu-fission reactions. This means it is now possible to make production matrices.

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This commit is contained in:
Paul Romano 2011-10-18 13:12:43 -04:00
parent c6d85b082d
commit d92e22983c
4 changed files with 118 additions and 58 deletions
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6
src/input_xml.f90
View file

@ -790,12 +790,6 @@ contains
end if
case ('nu-fission')
t % macro_bins(j) % scalar = MACRO_NU_FISSION
! TODO: Add fission energy transfer and remove this
if (t % n_bins(T_ENERGYOUT) > 0) then
msg = "Cannot tally nu-fission rate with an outgoing " &
// "energy filter."
call fatal_error(msg)
end if
case default
msg = "Unknown macro reaction: " // trim(words(j))
call fatal_error(msg)

4
src/particle_header.f90
View file

@ -27,6 +27,9 @@ module particle_header
real(8) :: last_wgt ! last particle weight
real(8) :: last_E ! last energy
! Post-collision physical data
integer :: n_bank ! number of fission sites banked
! Energy grid data
integer :: IE ! index on energy grid
real(8) :: interp ! interpolation factor for energy grid
@ -65,6 +68,7 @@ contains
p % type = NEUTRON
p % wgt = ONE
p % alive = .true.
p % n_bank = 0
p % cell = 0
p % cell_born = 0
p % universe = 0

12
src/physics.f90
View file

@ -285,13 +285,15 @@ contains
real(8) :: scatter ! microscopic scattering cross-section
real(8) :: inelastic ! microscopic inelastic scattering cross-section
logical :: scattered ! was this a scattering reaction?
logical :: fissioned ! was this a fission reaction?
character(MAX_LINE_LEN) :: msg ! output/error message
type(Material), pointer :: mat
type(Nuclide), pointer :: nuc
type(Reaction), pointer :: rxn
! Set scatter to false by default
! Set scattered/fissioned to false by default
scattered = .false.
fissioned = .false.
! Store pre-collision particle properties
p % last_wgt = p % wgt
@ -433,7 +435,6 @@ contains
call inelastic_scatter(p, nuc, rxn)
end if
! With survival biasing, the particle will always scatter
scattered = .true.
@ -472,6 +473,7 @@ contains
scattered = .true.
case (N_FISSION, N_F, N_NF, N_2NF, N_3NF)
call create_fission_sites(p, index_nuclide, rxn, .true.)
fissioned = .true.
p % alive = .false.
case (N_GAMMA : N_DA)
call n_absorption(p)
@ -500,9 +502,12 @@ contains
! filter
if (tallies_on) then
call score_tally(p, scattered)
call score_tally(p, scattered, fissioned)
end if
! Reset number of particles banked during collision
p % n_bank = 0
! find energy index, interpolation factor
call find_energy_index(p)
@ -767,6 +772,7 @@ contains
! increment number of bank sites
n_bank = min(n_bank + nu, 3*n_particles)
p % n_bank = nu
end subroutine create_fission_sites

154
src/tally.f90
View file

@ -259,24 +259,30 @@ contains
! SCORE_TALLY contains the main logic for scoring user-specified tallies
!===============================================================================
subroutine score_tally(p, scattered)
subroutine score_tally(p, scattered, fissioned)
type(Particle), pointer :: p
logical, intent(in) :: scattered
logical, intent(in) :: fissioned
integer :: i
integer :: j
integer :: k
integer :: n
integer :: bins(TALLY_TYPES)
integer :: bin_energyout
integer :: score_index
integer :: score_index0
integer :: macro_bin
real(8) :: score
real(8) :: last_wgt
real(8) :: wgt
real(8) :: mu
real(8) :: E_out
logical :: in_mesh
logical :: has_outgoing_energy
logical :: has_outgoing_angle
logical :: has_energyout_bin
logical :: analog
character(MAX_LINE_LEN) :: msg
type(TallyObject), pointer :: t
type(StructuredMesh), pointer :: m
@ -364,10 +370,10 @@ contains
! search to find incoming energy bin
bins(T_ENERGYOUT) = binary_search(t % energy_out, n + 1, p % E)
has_outgoing_energy = .true.
has_energyout_bin = .true.
else
bins(T_ENERGYOUT) = 1
has_outgoing_energy = .false.
has_energyout_bin = .false.
end if
! =======================================================================
@ -386,60 +392,110 @@ contains
macro_bin = t % macro_bins(j) % scalar
! determine if we need outgoing angle
has_outgoing_angle = (macro_bin == MACRO_NU_SCATTER .or. &
analog = (macro_bin == MACRO_NU_SCATTER .or. &
macro_bin == MACRO_SCATTER_1 .or. macro_bin == MACRO_SCATTER_2 .or. &
macro_bin == MACRO_SCATTER_3 .or. macro_bin == MACRO_N_1N .or. &
macro_bin == MACRO_N_2N .or. macro_bin == MACRO_N_3N .or. &
macro_bin == MACRO_N_4N)
if (has_outgoing_energy .or. has_outgoing_angle) then
if (has_energyout_bin .or. analog) then
! If this tally has an outgoing energy filter, the only supported
! reaction is scattering. For all other reactions, about
! reaction is scattering or nu-fission. Note that some macro
! quantities can only be scored in analog such as scattering
! moments and (n,xn)
if (.not. scattered) cycle
! Make sure bin is scattering -- since scattering has already
! occured, we do not need to multiply by the scattering cross
! section
if ((.not. scattered) .and. (.not. fissioned)) cycle
select case (macro_bin)
case (MACRO_SCATTER)
score = last_wgt
case (MACRO_NU_SCATTER)
score = wgt
case (MACRO_SCATTER_1)
score = last_wgt * mu
case (MACRO_SCATTER_2)
score = last_wgt * 0.5*(3.0*mu*mu - ONE)
case (MACRO_SCATTER_3)
score = last_wgt * 0.5*(5.0*mu*mu*mu - 3.0*mu)
case (MACRO_N_1N)
if (wgt == last_wgt) then
if (scattered) then
! since scattering has already occured, we do not need to
! multiply by the scattering cross section
select case (macro_bin)
case (MACRO_SCATTER)
score = last_wgt
else
case (MACRO_NU_SCATTER)
score = wgt
case (MACRO_SCATTER_1)
score = last_wgt * mu
case (MACRO_SCATTER_2)
score = last_wgt * 0.5*(3.0*mu*mu - ONE)
case (MACRO_SCATTER_3)
score = last_wgt * 0.5*(5.0*mu*mu*mu - 3.0*mu)
case (MACRO_N_1N)
if (wgt == last_wgt) then
score = last_wgt
else
cycle
end if
case (MACRO_N_2N)
if (int(wgt/last_wgt) == 2) then
score = last_wgt
else
cycle
end if
case (MACRO_N_3N)
if (int(wgt/last_wgt) == 3) then
score = last_wgt
else
cycle
end if
case (MACRO_N_4N)
if (int(wgt/last_wgt) == 4) then
score = last_wgt
else
cycle
end if
case (MACRO_NU_FISSION)
cycle
end if
case (MACRO_N_2N)
if (int(wgt/last_wgt) == 2) then
score = last_wgt
else
cycle
end if
case (MACRO_N_3N)
if (int(wgt/last_wgt) == 3) then
score = last_wgt
else
cycle
end if
case (MACRO_N_4N)
if (int(wgt/last_wgt) == 4) then
score = last_wgt
else
cycle
end if
case default
! call fatal_error
end select
case default
msg = "Invalid macro reaction on analog tally."
call fatal_error(msg)
end select
end if
if (fissioned) then
if (macro_bin /= MACRO_NU_FISSION) cycle
! Normally, we only need to make contributions to one scoring
! bin. However, in the case of fission, since multiple fission
! neutrons were emitted with different energies, multiple
! outgoing energy bins may have been scored to. The following
! logic treats this special case and scores to multiple bins
! save original outgoing energy bin and score index
bin_energyout = bins(T_ENERGYOUT)
score_index0 = score_index
! loop over number of particles banked
do k = 1, p % n_bank
! determine outgoing energy from fission bank
E_out = fission_bank(n_bank - p % n_bank + k) % E
! change outgoing energy bin
bins(T_ENERGYOUT) = binary_search(t % energy_out, n + 1, E_out)
! determine scoring index
score_index = sum((bins - 1) * t % stride) + 1
! Since the creation of fission sites is weighted such that
! it is expected to create n_particles sites, we need to
! multiply the score by keff to get the true nu-fission
! rate. Otherwise, the sum of all nu-fission rates would be
! ~1.0.
score = keff
! Add score to tally
call add_to_score(t % scores(score_index, j), score)
end do
! reset outgoing energy bin and score index
bins(T_ENERGYOUT) = bin_energyout
score_index = score_index0
cycle
end if
else
! For tallies with no outgoing energy filter, the score is
! calculated normally depending on the quantity specified