crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-27 21:55:41 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Changes to get tallying on better footing, moved C5G7 problem to higher subdir in examples

Browse source
This commit is contained in:
Adam Nelson 2015-11-08 14:12:33 -05:00
parent c571a4b126
commit 2262045e63
25 changed files with 123 additions and 64 deletions
Download patch file Download diff file Expand all files Collapse all files

5
examples/xml/multigroup/c5g7/2d/cmfd.xml → examples/xml/c5g7/2d/cmfd.xml
View file

@ -2,12 +2,12 @@
<cmfd>
<begin>2</begin>
<feedback>false</feedback>
<solver>power</solver>
<feedback>true</feedback>
<mesh>
<lower_left> 0.0 0.0 -100.0</lower_left>
<upper_right> 64.26 64.26 100.0</upper_right>
<dimension>6 6 1</dimension>
<!-- <energy>1.0E-11 0.0635E-6 10.0E-6 1.0E-4 1.0E-3 0.5 1.0 20.0</energy> -->
<map>
2 2 2 2 1 1
2 2 2 2 1 1
@ -16,5 +16,6 @@
1 1 1 1 1 1
1 1 1 1 1 1
</map>
<albedo>1 0 0 1 1 1</albedo>
</mesh>
</cmfd>

0
examples/xml/multigroup/c5g7/2d/geometry.xml → examples/xml/c5g7/2d/geometry.xml
View file

0
examples/xml/multigroup/c5g7/2d/materials.xml → examples/xml/c5g7/2d/materials.xml
View file

0
examples/xml/multigroup/c5g7/2d/plots.xml → examples/xml/c5g7/2d/plots.xml
View file

12
examples/xml/multigroup/c5g7/2d/settings.xml → examples/xml/c5g7/2d/settings.xml
View file

@ -6,17 +6,11 @@
in an eigenvalue calculation mode
-->
<eigenvalue>
<batches>2500</batches>
<batches>2000</batches>
<inactive>500</inactive>
<particles> 1000 </particles>
<particles>1000</particles>
</eigenvalue>
<uniform_fs>
<lower_left> 0.0 21.42 -100 </lower_left>
<upper_right> 42.84 64.26 100 </upper_right>
<dimension> 34 34 1 </dimension>
</uniform_fs>
<!--
Start with uniformally distributed neutron source
with the default energy spectrum of a Maxwellian
@ -33,7 +27,7 @@
<!-- Establish statepoints to aid in examining convergence -->
<state_point>
<interval> 2500 </interval>
<interval> 2000 </interval>
<source_write>true</source_write>
</state_point>

0
examples/xml/multigroup/c5g7/2d/tallies.xml → examples/xml/c5g7/2d/tallies.xml
View file

0
examples/xml/multigroup/c5g7/3d/geometry.xml → examples/xml/c5g7/3d/geometry.xml
View file

0
examples/xml/multigroup/c5g7/3d/materials.xml → examples/xml/c5g7/3d/materials.xml
View file

0
examples/xml/multigroup/c5g7/3d/plots.xml → examples/xml/c5g7/3d/plots.xml
View file

2
examples/xml/multigroup/c5g7/3d/settings.xml → examples/xml/c5g7/3d/settings.xml
View file

@ -42,8 +42,6 @@
<tallies>true</tallies>
</output>
<run_cmfd>false</run_cmfd>
<cross_sections>../data.xml</cross_sections>
</settings>

0
examples/xml/multigroup/c5g7/3d/tallies.xml → examples/xml/c5g7/3d/tallies.xml
View file

0
examples/xml/multigroup/c5g7/data.xml → examples/xml/c5g7/data.xml
View file

0
examples/xml/multigroup/c5g7/materials.xml → examples/xml/c5g7/materials.xml
View file

0
examples/xml/multigroup/c5g7/pin/geometry.xml → examples/xml/c5g7/pin/geometry.xml
View file

0
examples/xml/multigroup/c5g7/pin/materials.xml → examples/xml/c5g7/pin/materials.xml
View file

0
examples/xml/multigroup/c5g7/pin/plots.xml → examples/xml/c5g7/pin/plots.xml
View file

0
examples/xml/multigroup/c5g7/pin/settings.xml → examples/xml/c5g7/pin/settings.xml
View file

0
examples/xml/multigroup/c5g7/pin/tallies.xml → examples/xml/c5g7/pin/tallies.xml
View file

19
src/cmfd_input.F90
View file

@ -52,7 +52,7 @@ contains
use xml_interface
use, intrinsic :: ISO_FORTRAN_ENV
integer :: i
integer :: i, g
integer :: ng
integer :: n_params
integer, allocatable :: iarray(:)
@ -102,6 +102,23 @@ contains
if(.not.allocated(cmfd%egrid)) allocate(cmfd%egrid(ng))
call get_node_array(node_mesh, "energy", cmfd%egrid)
cmfd % indices(4) = ng - 1 ! sets energy group dimension
! If using MG mode, check to see if these egrid points at least match
! the MG Data breakpoints
if (.not. run_CE) then
do i = 1, ng
found = .false.
do g = 1, energy_groups + 1
if (cmfd%egrid(i) == energy_bins(g)) then
found = .true.
exit
end if
end do
if (.not. found) then
call fatal_error("CMFD energy mesh boundaries must align with&
& boundaries of multi-group data!")
end if
end do
end if
else
if(.not.allocated(cmfd % egrid)) allocate(cmfd % egrid(2))
cmfd % egrid = [ ZERO, 20.0_8 ]

4
src/geometry.F90
View file

@ -199,6 +199,7 @@ contains
type(Cell), pointer :: c ! pointer to cell
class(Lattice), pointer :: lat ! pointer to lattice
type(Universe), pointer :: univ ! universe to search in
real(8) :: new_xyz(3) ! perturbed location used to look for cell
do j = p % n_coord + 1, MAX_COORD
call p % coord(j) % reset()
@ -285,7 +286,8 @@ contains
lat => lattices(c % fill) % obj
! Determine lattice indices
i_xyz = lat % get_indices(p % coord(j) % xyz + TINY_BIT * p % coord(j) % uvw)
new_xyz = p % coord(j) % xyz + TINY_BIT * p % coord(j) % uvw
i_xyz = lat % get_indices(new_xyz)
! Store lower level coordinates
p % coord(j + 1) % xyz = lat % get_local_xyz(p % coord(j) % xyz, i_xyz)

8
src/input_xml.F90
View file

@ -2705,6 +2705,8 @@ contains
! Allocate and store bins
allocate(t % filters(j) % real_bins(n_words))
call get_node_array(node_filt, "bins", t % filters(j) % real_bins)
if (.not. run_CE) t % energy_matches_groups = .false.
else if (n_words == -1) then
! Set number of bins
t % filters(j) % n_bins = energy_groups
@ -2712,6 +2714,8 @@ contains
! Allocate and store bins
allocate(t % filters(j) % real_bins(energy_groups))
t % filters(j) % real_bins = energy_bins
if (.not. run_CE) t % energy_matches_groups = .true.
end if
case ('energyout')
@ -2725,6 +2729,8 @@ contains
! Allocate and store bins
allocate(t % filters(j) % real_bins(n_words))
call get_node_array(node_filt, "bins", t % filters(j) % real_bins)
if (.not. run_CE) t % energyout_matches_groups = .false.
else if (n_words == -1) then
! Set number of bins
t % filters(j) % n_bins = energy_groups
@ -2732,6 +2738,8 @@ contains
! Allocate and store bins
allocate(t % filters(j) % real_bins(energy_groups))
t % filters(j) % real_bins = energy_bins
if (.not. run_CE) t % energyout_matches_groups = .true.
end if
! Set to analog estimator

4
src/particle_header.F90
View file

@ -198,8 +198,8 @@ contains
this % E = src % E
this % last_E = src % E
if (.not. run_CE) then
this % g = src % g
this % last_g = src % g
this % g = src % g
this % last_g = src % g
end if
end subroutine initialize_from_source

1
src/physics_mg.F90
View file

@ -241,6 +241,7 @@ contains
! Sample secondary energy distribution for fission reaction and set energy
! in fission bank
fission_bank(i) % g = sample_fission_energy(xs, p % g, fission_bank(i) % uvw)
fission_bank(i) % E = energy_bin_avg(fission_bank(i) % g)
end do
! increment number of bank sites

128
src/tally.F90
View file

@ -799,8 +799,7 @@ contains
real(8) :: micro_abs ! nuclidic microscopic abs
class(Nuclide_MG), pointer :: nuc
if (i_nuclide > 0) &
nuc => nuclides_MG(i_nuclide) % obj
if (i_nuclide > 0) nuc => nuclides_MG(i_nuclide) % obj
i = 0
SCORE_LOOP: do q = 1, t % n_user_score_bins
@ -1047,7 +1046,7 @@ contains
else
if (i_nuclide > 0) then
score = nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw) &
score = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw) &
* atom_density * flux
else
score = material_xs % absorption * flux
@ -1061,10 +1060,10 @@ contains
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! fission
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'fission', UVW=p % coord(1) % uvw) &
nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) &
/ micro_abs
else
score = ZERO
@ -1076,13 +1075,13 @@ contains
! particle's weight entering the collision as the estimate for the
! fission reaction rate
score = p % last_wgt &
* nuc % get_xs(p % g, 'fission', UVW=p % coord(1) % uvw) &
/ nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw)
* nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) &
/ nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
end if
else
if (i_nuclide > 0) then
score = nuc % get_xs(p % g, 'fission', UVW=p % coord(1) % uvw) * &
score = nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) * &
atom_density * flux
else
score = material_xs % fission * flux
@ -1107,10 +1106,11 @@ contains
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO) then
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'fission', UVW=p % coord(1) % uvw) / &
nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw)
nuc % get_xs(p % g, 'fission', UVW=p % coord(i) % uvw) / &
micro_abs
else
score = ZERO
end if
@ -1127,7 +1127,7 @@ contains
else
if (i_nuclide > 0) then
score = nuc % get_xs(p % g, 'nu_fission', UVW=p % coord(1) % uvw) &
score = nuc % get_xs(p % g, 'nu_fission', UVW=p % coord(i) % uvw) &
* atom_density * flux
else
score = material_xs % nu_fission * flux
@ -1141,10 +1141,10 @@ contains
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! fission scale by kappa-fission
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw)
micro_abs = nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
if (micro_abs > ZERO) then
score = p % absorb_wgt * &
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(1) % uvw) / &
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) / &
micro_abs
else
score = ZERO
@ -1156,13 +1156,13 @@ contains
! particle's weight entering the collision as the estimate for
! the fission energy production rate
score = p % last_wgt * &
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(1) % uvw) / &
nuc % get_xs(p % g, 'absorption', UVW=p % coord(1) % uvw)
nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) / &
nuc % get_xs(p % g, 'absorption', UVW=p % coord(i) % uvw)
end if
else
if (i_nuclide > 0) then
score = nuc % get_xs(p % g, 'k_fission', UVW=p % coord(1) % uvw) &
score = nuc % get_xs(p % g, 'k_fission', UVW=p % coord(i) % uvw) &
* atom_density * flux
else
score = material_xs % kappa_fission * flux
@ -1534,13 +1534,14 @@ contains
integer :: i_filter ! index for matching filter bin combination
real(8) :: score ! actual score
integer :: gout ! energy group of fission bank site
real(8) :: E_out
! save original outgoing energy bin and score index
i = t % find_filter(FILTER_ENERGYOUT)
bin_energyout = matching_bins(i)
! Get number of energies on filter
n = size(t % filters(i) % int_bins)
n = size(t % filters(i) % real_bins)
! Since the creation of fission sites is weighted such that it is
! expected to create n_particles sites, we need to multiply the
@ -1552,11 +1553,23 @@ contains
! determine score based on bank site weight and keff
score = keff * fission_bank(n_bank - p % n_bank + k) % wgt
! determine outgoing energy from fission bank
gout = fission_bank(n_bank - p % n_bank + k) % g
if (t % energyout_matches_groups) then
! determine outgoing energy from fission bank
gout = fission_bank(n_bank - p % n_bank + k) % g
! change outgoing energy bin
matching_bins(i) = gout
! change outgoing energy bin
matching_bins(i) = gout
else
! determine outgoing energy from fission bank
E_out = fission_bank(n_bank - p % n_bank + k) % E
! check if outgoing energy is within specified range on filter
if (E_out < t % filters(i) % real_bins(1) .or. &
E_out > t % filters(i) % real_bins(n)) cycle
! change outgoing energy bin
matching_bins(i) = binary_search(t % filters(i) % real_bins, n, E_out)
end if
! determine scoring index
i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
@ -2427,8 +2440,8 @@ contains
integer :: j
integer :: n ! number of bins for single filter
integer :: offset ! offset for distribcell
integer :: g ! particle energy group
real(8) :: theta, phi ! Polar and Azimuthal Angles, respectively
real(8) :: E
type(TallyObject), pointer :: t
type(RegularMesh), pointer :: m
@ -2507,34 +2520,55 @@ contains
p % surface, i_tally)
case (FILTER_ENERGYIN)
! make sure the correct energy is used
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
g = p % g
else
g = p % last_g
end if
! Since all groups are filters, the filter bin is the group
matching_bins(i) = g
case (FILTER_ENERGYOUT)
! Since all groups are filters, the filter bin is the group
matching_bins(i) = p % g
case (FILTER_DELAYEDGROUP)
if (survival_biasing .and. t % find_filter(FILTER_ENERGYOUT) <= 0) then
matching_bins(i) = 1
elseif (active_tracklength_tallies % size() > 0) then
matching_bins(i) = 1
else
if (p % delayed_group == 0) then
matching_bins = NO_BIN_FOUND
if (t % energy_matches_groups) then
! make sure the correct energy group is used
! Since all groups are filters, the filter bin is the group
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
matching_bins(i) = p % g
else
matching_bins(i) = p % delayed_group
matching_bins(i) = p % last_g
end if
else
! make sure the correct energy is used
if (t % estimator == ESTIMATOR_TRACKLENGTH) then
E = p % E
else
E = p % last_E
end if
n = t % filters(i) % n_bins
! check if energy of the particle is within energy bins
if (E < t % filters(i) % real_bins(1) .or. &
E > t % filters(i) % real_bins(n + 1)) then
matching_bins(i) = NO_BIN_FOUND
else
! search to find incoming energy bin
matching_bins(i) = binary_search(t % filters(i) % real_bins, &
n + 1, E)
end if
end if
case (FILTER_ENERGYOUT)
if (t % energyout_matches_groups) then
! Since all groups are filters, the filter bin is the group
matching_bins(i) = p % g
else
! determine outgoing energy bin
n = t % filters(i) % n_bins
! check if energy of the particle is within energy bins
if (p % E < t % filters(i) % real_bins(1) .or. &
p % E > t % filters(i) % real_bins(n + 1)) then
matching_bins(i) = NO_BIN_FOUND
else
! search to find incoming energy bin
matching_bins(i) = binary_search(t % filters(i) % real_bins, &
n + 1, p % E)
end if
end if
case (FILTER_MU)
! determine mu bin
n = t % filters(i) % n_bins

4
src/tally_header.F90
View file

@ -130,6 +130,10 @@ module tally_header
integer :: n_triggers = 0 ! # of triggers
type(TriggerObject), allocatable :: triggers(:) ! Array of triggers
! Multi-Group Specific Information To Enable Rapid Tallying
logical :: energy_matches_groups = .false.
logical :: energyout_matches_groups = .false.
! Type-Bound procedures
contains
procedure :: clear => tallyobject_clear ! Deallocates TallyObject