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Improvements to depletion reaction caching

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This commit is contained in:
Paul Romano 2017-12-22 14:47:42 +07:00
parent f774c29254
commit 9ff4e620b6
7 changed files with 162 additions and 396 deletions
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11
src/constants.F90
View file

@ -308,7 +308,7 @@ module constants
! Tally score type -- if you change these, make sure you also update the
! _SCORES dictionary in openmc/capi/tally.py
integer, parameter :: N_SCORE_TYPES = 30
integer, parameter :: N_SCORE_TYPES = 24
integer, parameter :: &
SCORE_FLUX = -1, & ! flux
SCORE_TOTAL = -2, & ! total reaction rate
@ -333,13 +333,8 @@ module constants
SCORE_INVERSE_VELOCITY = -21, & ! flux-weighted inverse velocity
SCORE_FISS_Q_PROMPT = -22, & ! prompt fission Q-value
SCORE_FISS_Q_RECOV = -23, & ! recoverable fission Q-value
SCORE_DECAY_RATE = -24, & ! delayed neutron precursor decay rate
SCORE_N2N = -25, &
SCORE_NGAMMA = -26, &
SCORE_N3N = -27, &
SCORE_N4N = -28, &
SCORE_NP = -29, &
SCORE_NALPHA = -30
SCORE_DECAY_RATE = -24 ! delayed neutron precursor decay rate
! Maximum scattering order supported
integer, parameter :: MAX_ANG_ORDER = 10

141
src/cross_section.F90
View file

@ -2,7 +2,6 @@ module cross_section
use algorithm, only: binary_search
use constants
use endf
use error, only: fatal_error
use list_header, only: ListElemInt
use material_header, only: Material, materials
@ -52,6 +51,7 @@ contains
material_xs % ngamma = ZERO
material_xs % np = ZERO
material_xs % nalpha = ZERO
! Exit subroutine if material is void
if (p % material == MATERIAL_VOID) return
@ -136,17 +136,11 @@ contains
material_xs % nu_fission = material_xs % nu_fission + &
atom_density * micro_xs(i_nuclide) % nu_fission
! Add contributions to material macroscopic n2n cross section
! Add contributions to material macroscopic n2n cross section
material_xs % n2n = material_xs % n2n + &
atom_density * micro_xs(i_nuclide) % n2n
! Add contributions to material macroscopic ngamma cross section
material_xs % ngamma = material_xs % ngamma + &
atom_density * micro_xs(i_nuclide) % ngamma
! Add contributions to material macroscopic n3n cross section
! Add contributions to material macroscopic n3n cross section
material_xs % n3n = material_xs % n3n + &
atom_density * micro_xs(i_nuclide) % n3n
@ -154,16 +148,17 @@ contains
material_xs % n4n = material_xs % n4n + &
atom_density * micro_xs(i_nuclide) % n4n
! Add contributions to material macroscopic np cross section
! Add contributions to material macroscopic ngamma cross section
material_xs % ngamma = material_xs % ngamma + &
atom_density * micro_xs(i_nuclide) % ngamma
! Add contributions to material macroscopic np cross section
material_xs % np = material_xs % np + &
atom_density * micro_xs(i_nuclide) % np
! Add contributions to material macroscopic nalpha cross section
! Add contributions to material macroscopic nalpha cross section
material_xs % nalpha = material_xs % nalpha + &
atom_density * micro_xs(i_nuclide) % nalpha
end do
end associate
@ -189,12 +184,23 @@ contains
integer :: i_grid ! index on nuclide energy grid
integer :: i_low ! lower logarithmic mapping index
integer :: i_high ! upper logarithmic mapping index
integer :: i_rxn ! reaction index
integer :: j
real(8) :: val
real(8) :: f ! interp factor on nuclide energy grid
real(8) :: kT ! temperature in eV
real(8) :: sigT, sigA, sigF ! Intermediate multipole variables
integer :: rxn
type(TallyObject) :: t
character(15) :: threshold_rxn_type
integer, parameter :: DEPLETION_RX(6) = [N_2N, N_3N, N_4N, N_GAMMA, N_P, N_A]
! Initialize cached cross sections to zero
micro_xs(i_nuclide) % n2n = ZERO
micro_xs(i_nuclide) % n3n = ZERO
micro_xs(i_nuclide) % n4n = ZERO
micro_xs(i_nuclide) % np = ZERO
micro_xs(i_nuclide) % nalpha = ZERO
micro_xs(i_nuclide) % thermal = ZERO
micro_xs(i_nuclide) % thermal_elastic = ZERO
associate (nuc => nuclides(i_nuclide))
! Check to see if there is multipole data present at this energy
use_mp = .false.
@ -221,6 +227,7 @@ contains
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_fission = ZERO
end if
micro_xs(i_nuclide) % ngamma = sigF - sigA
! Ensure these values are set
! Note, the only time either is used is in one of 4 places:
@ -239,8 +246,6 @@ contains
else
! Find the appropriate temperature index.
kT = sqrtkT**2
!rxn = nuc % rxn_index_MT(p % event_MT)
select case (temperature_method)
case (TEMPERATURE_NEAREST)
i_temp = minloc(abs(nuclides(i_nuclide) % kTs - kT), dim=1)
@ -286,22 +291,9 @@ contains
(grid % energy(i_grid + 1) - grid % energy(i_grid))
micro_xs(i_nuclide) % index_temp = i_temp
micro_xs(i_nuclide) % index_grid = i_grid
micro_xs(i_nuclide) % interp_factor = f
! Initialize nuclide cross-sections to zero
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_fission = ZERO
micro_xs(i_nuclide) % n2n = ZERO
micro_xs(i_nuclide) % n3n = ZERO
micro_xs(i_nuclide) % n4n = ZERO
micro_xs(i_nuclide) % np = ZERO
micro_xs(i_nuclide) % nalpha = ZERO
micro_xs(i_nuclide) % thermal = ZERO
micro_xs(i_nuclide) % thermal_elastic = ZERO
! Calculate microscopic nuclide total cross section
micro_xs(i_nuclide) % total = (ONE - f) * xs % total(i_grid) &
+ f * xs % total(i_grid + 1)
@ -320,66 +312,41 @@ contains
+ f * xs % fission(i_grid + 1)
! Calculate microscopic nuclide nu-fission cross section
micro_xs(i_nuclide) % nu_fission = (ONE - f) * xs % nu_fission(&
micro_xs(i_nuclide) % nu_fission = (ONE - f) * xs % nu_fission( &
i_grid) + f * xs % nu_fission(i_grid + 1)
else
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_fission = ZERO
end if
end associate
micro_xs(i_nuclide) % ngamma = (micro_xs(i_nuclide) % absorption - &
micro_xs(i_nuclide) % fission)
if (t % has_threshold_rxn) then
!if (nuclides(i_nuclide)%reaction_index%has_key(p % event_MT)) the
rxn = nuclides(i_nuclide) % rxn_index_MT(p % event_MT)
associate(xs_threshold => nuc % reactions(rxn) % xs(i_temp))
!end if
threshold_rxn_type = reaction_name(p % event_MT)
if (threshold_rxn_type == 'n2n') then
if (i_grid >= xs_threshold % threshold) then
micro_xs(i_nuclide) % n2n = (ONE - f) * xs_threshold % value(i_grid - xs_threshold &
% threshold + 1) + f * xs_threshold % value(i_grid - xs_threshold &
% threshold + 2)
end if
end if
if (threshold_rxn_type == 'n3n') then
if (i_grid >= xs_threshold % threshold) then
micro_xs(i_nuclide) % n3n = (ONE - f) * xs_threshold % value(i_grid - xs_threshold &
% threshold + 1) + f * xs_threshold % value(i_grid - xs_threshold &
% threshold + 2)
end if
end if
if (threshold_rxn_type == 'n4n') then
if (i_grid >= xs_threshold % threshold) then
micro_xs(i_nuclide) % n4n = (ONE - f) * xs_threshold % value(i_grid - xs_threshold &
% threshold + 1) + f * xs_threshold % value(i_grid - xs_threshold &
% threshold + 2)
end if
end if
if (threshold_rxn_type == 'np') then
if (i_grid >= xs_threshold % threshold) then
micro_xs(i_nuclide) % np = (ONE - f) * xs_threshold % value(i_grid - xs_threshold &
% threshold + 1) + f * xs_threshold % value(i_grid - xs_threshold &
% threshold + 2)
end if
end if
if (threshold_rxn_type == 'nalpha') then
if (i_grid >= xs_threshold % threshold) then
micro_xs(i_nuclide) % nalpha = (ONE - f) * xs_threshold % value(i_grid - xs_threshold &
% threshold + 1) + f * xs_threshold % value(i_grid - xs_threshold &
% threshold + 2)
end if
end if
! Depletion-related reactions
do j = 1, 6
i_rxn = nuc % rxn_index_MT(DEPLETION_RX(j))
if (i_rxn > 0) then
associate (xs => nuc % reactions(i_rxn) % xs(i_temp))
if (i_grid >= xs % threshold) then
val = (ONE - f) * xs % value(i_grid - xs % threshold + 1) + &
f * xs % value(i_grid - xs % threshold + 2)
select case (DEPLETION_RX(j))
case (N_2N)
micro_xs(i_nuclide) % n2n = val
case (N_3N)
micro_xs(i_nuclide) % n3n = val
case (N_4N)
micro_xs(i_nuclide) % n4n = val
case (N_GAMMA)
micro_xs(i_nuclide) % ngamma = val
case (N_P)
micro_xs(i_nuclide) % np = val
case (N_A)
micro_xs(i_nuclide) % nalpha = val
end select
end if
end associate
end if
end associate
end if
end do
end if
! Initialize sab treatment to false
micro_xs(i_nuclide) % index_sab = NONE

36
src/endf.F90
View file

@ -66,18 +66,6 @@ contains
string = "fission-q-prompt"
case (SCORE_FISS_Q_RECOV)
string = "fission-q-recoverable"
case (SCORE_N2N)
string = "n2n"
case (SCORE_N3N)
string = "n3n"
case (SCORE_N4N)
string = "n4n"
case (SCORE_NGAMMA)
string = "ngamma"
case (SCORE_NP)
string = "np"
case (SCORE_NALPHA)
string = "nalpha"
! Normal ENDF-based reactions
case (TOTAL_XS)
@ -88,12 +76,12 @@ contains
string = '(n,level)'
case (N_2ND)
string = '(n,2nd)'
!case (N_2N)
!string = '(n,2n)'
!case (N_3N)
!string = '(n,3n)'
!case (N_FISSION)
! string = '(n,fission)'
case (N_2N)
string = '(n,2n)'
case (N_3N)
string = '(n,3n)'
case (N_FISSION)
string = '(n,fission)'
case (N_F)
string = '(n,f)'
case (N_NF)
@ -142,18 +130,18 @@ contains
string = '(n,nc)'
case (N_DISAPPEAR)
string = '(n,disappear)'
!case (N_GAMMA)
! string = '(n,gamma)'
!case (N_P)
!string = '(n,p)'
case (N_GAMMA)
string = '(n,gamma)'
case (N_P)
string = '(n,p)'
case (N_D)
string = '(n,d)'
case (N_T)
string = '(n,t)'
case (N_3HE)
string = '(n,3He)'
!case (N_A)
!string = '(n,a)'
case (N_A)
string = '(n,a)'
case (N_2A)
string = '(n,2a)'
case (N_3A)

33
src/input_xml.F90
View file

@ -2450,17 +2450,6 @@ contains
n_nuclides = index_nuclide
n_sab_tables = index_sab
do i=1, n_materials
mat => materials(i)
allocate(mat % mat_nuclide_list(n_nuclides_total))
mat % mat_nuclide_list(:) = 0
do j=1, mat % n_nuclides
mat % mat_nuclide_list(mat % nuclide(j))=j
end do
end do
! Close materials XML file
call doc % clear()
@ -3058,14 +3047,14 @@ contains
call fatal_error("n1n score no longer supported for tallies, &
&please remove")
case ('n2n', '(n,2n)')
t % score_bins(j) = SCORE_N2N
t % has_threshold_rxn = .true.
t % score_bins(j) = N_2N
case ('n3n', '(n,3n)')
t % score_bins(j) = SCORE_N3N
t % has_threshold_rxn = .true.
t % score_bins(j) = N_3N
case ('n4n', '(n,4n)')
t % score_bins(j) = SCORE_N4N
t % has_threshold_rxn = .true.
t % score_bins(j) = N_4N
case ('absorption')
t % score_bins(j) = SCORE_ABSORPTION
if (t % find_filter(FILTER_ENERGYOUT) > 0) then
@ -3249,12 +3238,9 @@ contains
case ('(n,nc)')
t % score_bins(j) = N_NC
case ('(n,gamma)')
t % score_bins(j) = SCORE_NGAMMA
t % score_bins(j) = N_GAMMA
case ('(n,p)')
t % score_bins(j) = SCORE_NP
t % has_threshold_rxn = .true.
t % score_bins(j) = N_P
case ('(n,d)')
t % score_bins(j) = N_D
case ('(n,t)')
@ -3262,8 +3248,7 @@ contains
case ('(n,3He)')
t % score_bins(j) = N_3HE
case ('(n,a)')
t % score_bins(j) = SCORE_NALPHA
t % has_threshold_rxn = .true.
t % score_bins(j) = N_A
case ('(n,2a)')
t % score_bins(j) = N_2A
case ('(n,3a)')

23
src/nuclide_header.F90
View file

@ -43,14 +43,6 @@ module nuclide_header
real(8), allocatable :: nu_fission(:) ! neutron production
real(8), allocatable :: absorption(:) ! absorption (MT > 100)
real(8), allocatable :: heating(:) ! heating
real(8), allocatable :: ngamma(:)
real(8), allocatable :: n2n(:)
real(8), allocatable :: n3n(:)
real(8), allocatable :: n4n(:)
real(8), allocatable :: np(:)
real(8), allocatable :: nalpha(:)
end type SumXS
type :: Nuclide
@ -127,9 +119,9 @@ module nuclide_header
real(8) :: fission
real(8) :: nu_fission
real(8) :: n2n
real(8) :: ngamma
real(8) :: n3n
real(8) :: n4n
real(8) :: ngamma
real(8) :: np
real(8) :: nalpha
real(8) :: thermal ! Bound thermal elastic & inelastic scattering
@ -162,13 +154,12 @@ module nuclide_header
real(8) :: absorption ! macroscopic absorption xs
real(8) :: fission ! macroscopic fission xs
real(8) :: nu_fission ! macroscopic production xs
real(8) :: n2n
real(8) :: n3n
real(8) :: n4n
real(8) :: ngamma
real(8) :: np
real(8) :: nalpha
real(8) :: n2n ! macroscopic (n,2n) xs
real(8) :: n3n ! macroscopic (n,3n) xs
real(8) :: n4n ! macroscopic (n,4n) xs
real(8) :: ngamma ! macroscopic (n,gamma) xs
real(8) :: np ! macroscopic (n,p) xs
real(8) :: nalpha ! macroscopic (n,alpha) xs
end type MaterialMacroXS
!===============================================================================

313
src/tallies/tally.F90
View file

@ -1131,223 +1131,81 @@ contains
end if
end if
case (SCORE_N2N)
score = ZERO
if (i_nuclide > 0) then
!if (nuclides(i_nuclide)%reaction_index%has_key(score_bin)) then
!m = nuclides(i_nuclide)%reaction_index%get_key(score_bin)
! Retrieve temperature and energy grid index and interpolation
! factor
!i_temp = micro_xs(i_nuclide) % index_temp
!if (i_temp > 0) then
!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))
!if (i_energy >= xs % threshold) then
score = micro_xs(i_nuclide) % n2n * atom_density * flux
!else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
! score = ZERO
else
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)
score = score + micro_xs(i_nuc) % n2n * atom_density_ * flux
end do
end if
case (SCORE_N3N)
score = ZERO
if (i_nuclide > 0) then
!if (nuclides(i_nuclide)%reaction_index%has_key(score_bin)) then
!m = nuclides(i_nuclide)%reaction_index%get_key(score_bin)
! Retrieve temperature and energy grid index and interpolation
! factor
!i_temp = micro_xs(i_nuclide) % index_temp
!if (i_temp > 0) then
!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))
!if (i_energy >= xs % threshold) then
score = micro_xs(i_nuclide) % n3n * atom_density * flux
!else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
! score = ZERO
else
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)
score = score + micro_xs(i_nuc) % n3n * atom_density_ * flux
end do
end if
case (SCORE_N4N)
score = ZERO
if (i_nuclide > 0) then
!if (nuclides(i_nuclide)%reaction_index%has_key(score_bin)) then
!m = nuclides(i_nuclide)%reaction_index%get_key(score_bin)
! Retrieve temperature and energy grid index and interpolation
! factor
!i_temp = micro_xs(i_nuclide) % index_temp
!if (i_temp > 0) then
!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))
!if (i_energy >= xs % threshold) then
score = micro_xs(i_nuclide) % n4n * atom_density * flux
!else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
! score = ZERO
else
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)
score = score + micro_xs(i_nuc) % n4n * atom_density_ * flux
end do
end if
case (SCORE_NP)
score = ZERO
if (i_nuclide > 0) then
!if (nuclides(i_nuclide)%reaction_index%has_key(score_bin)) then
!m = nuclides(i_nuclide)%reaction_index%get_key(score_bin)
! Retrieve temperature and energy grid index and interpolation
! factor
!i_temp = micro_xs(i_nuclide) % index_temp
!if (i_temp > 0) then
!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))
!if (i_energy >= xs % threshold) then
score = micro_xs(i_nuclide) % np * atom_density * flux
!else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
! score = ZERO
else
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)
score = score + micro_xs(i_nuc) % np * atom_density_ * flux
end do
end if
case (SCORE_NALPHA)
score = ZERO
if (i_nuclide > 0) then
!if (nuclides(i_nuclide)%reaction_index%has_key(score_bin)) then
!m = nuclides(i_nuclide)%reaction_index%get_key(score_bin)
! Retrieve temperature and energy grid index and interpolation
! factor
!i_temp = micro_xs(i_nuclide) % index_temp
!if (i_temp > 0) then
!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))
!if (i_energy >= xs % threshold) then
score = micro_xs(i_nuclide) % nalpha * atom_density * flux
!else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
! score = ZERO
else
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)
score = score + micro_xs(i_nuc) % nalpha * atom_density_ * flux
end do
end if
case (SCORE_NGAMMA)
case (N_2N)
if (t % estimator == ESTIMATOR_ANALOG) then
if (survival_biasing) then
! No absorption events actually occur if survival biasing is on --
! just use weight absorbed in survival biasing
score = p % absorb_wgt * flux
! Check if event MT matches
if (p % event_MT /= N_2N) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n2n * atom_density * flux
else
! Skip any event where the particle wasn't absorbed
if (p % event == EVENT_SCATTER) cycle SCORE_LOOP
! All fission and absorption events will contribute here, so we
! can just use the particle's weight entering the collision
score = p % last_wgt * flux
score = material_xs % n2n * flux
end if
end if
case (N_3N)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= N_3N) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n3n * atom_density * flux
else
score = material_xs % n3n * flux
end if
end if
case (N_4N)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= N_4N) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % n4n * atom_density * flux
else
score = material_xs % n4n * flux
end if
end if
case (N_P)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= N_P) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % np * atom_density * flux
else
score = material_xs % np * flux
end if
end if
case (N_A)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= N_A) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % nalpha * atom_density * flux
else
score = material_xs % nalpha * flux
end if
end if
case (N_GAMMA)
if (t % estimator == ESTIMATOR_ANALOG) then
! Check if event MT matches
if (p % event_MT /= N_GAMMA) cycle SCORE_LOOP
score = p % last_wgt * flux
else
if (i_nuclide > 0) then
@ -1357,8 +1215,6 @@ contains
end if
end if
case default
if (t % estimator == ESTIMATOR_ANALOG) then
! Any other score is assumed to be a MT number. Thus, we just need
@ -1395,14 +1251,8 @@ contains
end associate
else
! This block is reached if multipole is turned on and we're in
! the resolved range. For (n,gamma), use absorption -
! fission. For everything else, assume it's zero.
if (score_bin == N_GAMMA) then
score = (micro_xs(i_nuclide) % absorption - &
micro_xs(i_nuclide) % fission) * atom_density * flux
else
score = ZERO
end if
! the resolved range. Assume xs is zero.
score = ZERO
end if
end if
@ -1434,16 +1284,8 @@ contains
end associate
else
! This block is reached if multipole is turned on and
! we're in the resolved range. For (n,gamma), use
! absorption - fission. For everything else, assume it's
! zero.
if (score_bin == N_GAMMA) then
score = (micro_xs(i_nuc) % absorption &
- micro_xs(i_nuc) % fission) &
* atom_density_ * flux
else
score = ZERO
end if
! we're in the resolved range. Assume xs is zero.
score = ZERO
end if
end if
end do
@ -2949,7 +2791,7 @@ contains
type(Particle), intent(in) :: p
real(8), intent(in) :: distance
integer :: check_value
integer :: i
integer :: i_tally
integer :: i_filt
@ -3030,7 +2872,6 @@ contains
if (p % material /= MATERIAL_VOID) then
! Get pointer to current material
mat => materials(p % material)
check_value= mat % mat_nuclide_list(i_nuclide)
! Determine index of nuclide in Material % atom_density array
j = mat % mat_nuclide_index(i_nuclide)

1
src/tallies/tally_header.F90
View file

@ -88,7 +88,6 @@ module tally_header
! reset property - allows a tally to be reset after every batch
logical :: reset = .false.
logical :: has_threshold_rxn = .false.
! Number of realizations of tally random variables
integer :: n_realizations = 0