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Move nuclide and S(a,b) table arrays to their respective modules

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This commit is contained in:
Paul Romano 2017-08-07 12:10:10 -05:00
parent 56021e1e42
commit 2864915409
12 changed files with 62 additions and 77 deletions
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10
src/api.F90
View file

@ -424,7 +424,7 @@ contains
if (.not. nuclide_dict % has_key(to_lower(name_))) then
if (library_dict % has_key(to_lower(name_))) then
! allocate extra space in nuclides array
n = n_nuclides_total
n = n_nuclides
allocate(new_nuclides(n + 1))
new_nuclides(1:n) = nuclides(:)
call move_alloc(FROM=new_nuclides, TO=nuclides)
@ -446,7 +446,7 @@ contains
! Add entry to nuclide dictionary
call nuclide_dict % add_key(to_lower(name_), n)
n_nuclides_total = n
n_nuclides = n
! Assign resonant scattering data
if (res_scat_on) call assign_0K_elastic_scattering(nuclides(n))
@ -585,7 +585,7 @@ contains
err = E_UNASSIGNED
if (index >= 1 .and. index <= size(materials)) then
associate (m => materials(index))
err = m % set_density(density, nuclides)
err = m % set_density(density)
end associate
else
err = E_OUT_OF_BOUNDS
@ -636,10 +636,10 @@ contains
m % p0(:) = .false.
! Set total density to the sum of the vector
err = m % set_density(sum(density), nuclides)
err = m % set_density(sum(density))
! Assign S(a,b) tables
call m % assign_sab_tables(nuclides, sab_tables)
call m % assign_sab_tables()
end associate
else
err = E_OUT_OF_BOUNDS

17
src/geometry_header.F90
View file

@ -4,7 +4,9 @@ module geometry_header
use constants, only: HALF, TWO, THREE, INFINITY, K_BOLTZMANN, &
MATERIAL_VOID, NONE
use dict_header, only: DictCharInt, DictIntInt
use nuclide_header
use material_header, only: Material
use sab_header
use stl_vector, only: VectorReal
use string, only: to_lower
@ -330,16 +332,12 @@ contains
! temperatures to read (which may be different if interpolation is used)
!===============================================================================
subroutine get_temperatures(cells, materials, material_dict, nuclide_dict, &
n_nucs, nuc_temps, sab_dict, n_sabs, sab_temps)
subroutine get_temperatures(cells, materials, material_dict, &
nuc_temps, sab_temps)
type(Cell), allocatable, intent(in) :: cells(:)
type(Material), allocatable, intent(in) :: materials(:)
type(DictIntInt), intent(in) :: material_dict
type(DictCharInt), intent(in) :: nuclide_dict
integer, intent(in) :: n_nucs
type(VectorReal), allocatable, intent(out) :: nuc_temps(:)
type(DictCharInt), optional, intent(in) :: sab_dict
integer, optional, intent(in) :: n_sabs
type(VectorReal), optional, allocatable, intent(out) :: sab_temps(:)
integer :: i, j, k
@ -348,8 +346,8 @@ contains
integer :: i_material
real(8) :: temperature ! temperature in Kelvin
allocate(nuc_temps(n_nucs))
if (present(n_sabs) .and. present(sab_temps)) allocate(sab_temps(n_sabs))
allocate(nuc_temps(n_nuclides))
if (present(sab_temps)) allocate(sab_temps(n_sab_tables))
do i = 1, size(cells)
do j = 1, size(cells(i) % material)
@ -376,8 +374,7 @@ contains
end if
end do NUC_NAMES_LOOP
if (present(sab_temps) .and. present(sab_dict) .and. &
mat % n_sab > 0) then
if (present(sab_temps) .and. mat % n_sab > 0) then
SAB_NAMES_LOOP: do k = 1, size(mat % sab_names)
! Get index in nuc_temps array
i_sab = sab_dict % get_key(to_lower(mat % sab_names(k)))

32
src/global.F90
View file

@ -14,9 +14,7 @@ module global
use material_header, only: Material
use mesh_header, only: RegularMesh
use mgxs_header, only: Mgxs, MgxsContainer
use nuclide_header
use plot_header, only: ObjectPlot
use sab_header, only: SAlphaBeta
use set_header, only: SetInt
use stl_vector, only: VectorInt
use surface_header, only: SurfaceContainer
@ -27,6 +25,10 @@ module global
use timer_header, only: Timer
use volume_header, only: VolumeCalculation
! Inherit module variables from nuclide/S(a,b) modules
use nuclide_header
use sab_header
implicit none
! ============================================================================
@ -70,39 +72,13 @@ module global
! ENERGY TREATMENT RELATED VARIABLES
logical :: run_CE = .true. ! Run in CE mode?
! ============================================================================
! CROSS SECTION RELATED VARIABLES NEEDED REGARDLESS OF CE OR MG
! Number of nuclide cross section tables
integer(C_INT), bind(C, name='n_nuclides') :: n_nuclides_total
! Cross section caches
type(NuclideMicroXS), allocatable :: micro_xs(:) ! Cache for each nuclide
type(MaterialMacroXS) :: material_xs ! Cache for current material
! Dictionaries to look up cross sections and listings
type(DictCharInt) :: nuclide_dict
type(DictCharInt) :: library_dict
! Cross section libraries
type(Library), allocatable :: libraries(:)
! ============================================================================
! CONTINUOUS-ENERGY CROSS SECTION RELATED VARIABLES
! Cross section arrays
type(Nuclide), allocatable, target :: nuclides(:) ! Nuclide cross-sections
type(SAlphaBeta), allocatable, target :: sab_tables(:) ! S(a,b) tables
integer :: n_sab_tables ! Number of S(a,b) thermal scattering tables
! Minimum/maximum energies
real(8) :: energy_min_neutron = ZERO
real(8) :: energy_max_neutron = INFINITY
! Dictionaries to look up cross sections and listings
type(DictCharInt) :: sab_dict
! Unreoslved resonance probablity tables
logical :: urr_ptables_on = .true.

24
src/input_xml.F90
View file

@ -2187,9 +2187,7 @@ contains
call assign_temperatures(material_temps)
! Determine desired temperatures for each nuclide and S(a,b) table
call get_temperatures(cells, materials, material_dict, nuclide_dict, &
n_nuclides_total, nuc_temps, sab_dict, &
n_sab_tables, sab_temps)
call get_temperatures(cells, materials, material_dict, nuc_temps, sab_temps)
! Read continuous-energy cross sections
if (run_CE .and. run_mode /= MODE_PLOTTING) then
@ -2617,8 +2615,8 @@ contains
end do
! Set total number of nuclides and S(a,b) tables
n_nuclides_total = index_nuclide
n_sab_tables = index_sab
n_nuclides = index_nuclide
n_sab_tables = index_sab
! Close materials XML file
call doc % clear()
@ -3464,15 +3462,15 @@ contains
if (trim(sarray(1)) == 'all') then
! Handle special case <nuclides>all</nuclides>
allocate(t % nuclide_bins(n_nuclides_total + 1))
allocate(t % nuclide_bins(n_nuclides + 1))
! Set bins to 1, 2, 3, ..., n_nuclides_total, -1
t % nuclide_bins(1:n_nuclides_total) = &
(/ (j, j=1, n_nuclides_total) /)
t % nuclide_bins(n_nuclides_total + 1) = -1
! Set bins to 1, 2, 3, ..., n_nuclides, -1
t % nuclide_bins(1:n_nuclides) = &
(/ (j, j=1, n_nuclides) /)
t % nuclide_bins(n_nuclides + 1) = -1
! Set number of nuclide bins
t % n_nuclide_bins = n_nuclides_total + 1
t % n_nuclide_bins = n_nuclides + 1
! Set flag so we can treat this case specially
t % all_nuclides = .true.
@ -5183,7 +5181,7 @@ contains
character(MAX_WORD_LEN) :: name
type(SetChar) :: already_read
allocate(nuclides(n_nuclides_total))
allocate(nuclides(n_nuclides))
allocate(sab_tables(n_sab_tables))
! Read cross sections
@ -5275,7 +5273,7 @@ contains
end do
! Associate S(a,b) tables with specific nuclides
call materials(i) % assign_sab_tables(nuclides, sab_tables)
call materials(i) % assign_sab_tables()
end do
! Show which nuclide results in lowest energy for neutron transport

11
src/material_header.F90
View file

@ -2,8 +2,8 @@ module material_header
use constants
use error
use nuclide_header, only: Nuclide
use sab_header, only: SAlphaBeta
use nuclide_header
use sab_header
use stl_vector, only: VectorReal, VectorInt
use string, only: to_str
@ -57,10 +57,9 @@ contains
! MATERIAL_SET_DENSITY sets the total density of a material in atom/b-cm.
!===============================================================================
function material_set_density(m, density, nuclides) result(err)
function material_set_density(m, density) result(err)
class(Material), intent(inout) :: m
real(8), intent(in) :: density
type(Nuclide), intent(in) :: nuclides(:)
integer :: err
integer :: i
@ -95,10 +94,8 @@ contains
! materials so the code knows when to apply bound thermal scattering data
!===============================================================================
subroutine material_assign_sab_tables(this, nuclides, sab_tables)
subroutine material_assign_sab_tables(this)
class(Material), intent(inout) :: this
type(Nuclide), intent(in) :: nuclides(:)
type(SAlphaBeta), intent(in) :: sab_tables(:)
integer :: j ! index over nuclides in material
integer :: k ! index over S(a,b) tables in material

5
src/mgxs_data.F90
View file

@ -51,8 +51,7 @@ contains
call write_message("Loading cross section data...", 5)
! Get temperatures
call get_temperatures(cells, materials, material_dict, nuclide_dict, &
n_nuclides_total, temps)
call get_temperatures(cells, materials, material_dict, temps)
! Open file for reading
file_id = file_open(path_cross_sections, 'r', parallel=.true.)
@ -72,7 +71,7 @@ contains
end if
! allocate arrays for MGXS storage and cross section cache
allocate(nuclides_MG(n_nuclides_total))
allocate(nuclides_MG(n_nuclides))
! ==========================================================================
! READ ALL MGXS CROSS SECTION TABLES

17
src/nuclide_header.F90
View file

@ -7,7 +7,7 @@ module nuclide_header
use algorithm, only: sort, find
use constants
use dict_header, only: DictIntInt
use dict_header, only: DictIntInt, DictCharInt
use endf, only: reaction_name, is_fission, is_disappearance
use endf_header, only: Function1D, Polynomial, Tabulated1D
use error, only: fatal_error, warning
@ -157,7 +157,20 @@ module nuclide_header
character(MAX_FILE_LEN) :: path
end type Library
contains
! Cross section libraries
type(Library), allocatable :: libraries(:)
type(DictCharInt) :: library_dict
! Nuclear data for each nuclide
type(Nuclide), allocatable, target :: nuclides(:)
integer(C_INT), bind(C) :: n_nuclides
type(DictCharInt) :: nuclide_dict
! Minimum/maximum energies
real(8) :: energy_min_neutron = ZERO
real(8) :: energy_max_neutron = INFINITY
contains
!===============================================================================
! NUCLIDE_CLEAR resets and deallocates data in Nuclide

2
src/particle_restart.F90
View file

@ -32,7 +32,7 @@ contains
! Set verbosity high
verbosity = 10
allocate(micro_xs(n_nuclides_total))
allocate(micro_xs(n_nuclides))
! Initialize the particle to be tracked
call p % initialize()

7
src/sab_header.F90
View file

@ -4,7 +4,7 @@ module sab_header
use algorithm, only: find, sort
use constants
use dict_header, only: DictIntInt
use dict_header, only: DictIntInt, DictCharInt
use distribution_univariate, only: Tabular
use error, only: warning, fatal_error
use hdf5, only: HID_T, HSIZE_T, SIZE_T
@ -78,6 +78,11 @@ module sab_header
procedure :: from_hdf5 => salphabeta_from_hdf5
end type SAlphaBeta
! S(a,b) tables
type(SAlphaBeta), allocatable, target :: sab_tables(:)
integer :: n_sab_tables
type(DictCharInt) :: sab_dict
contains
subroutine salphabeta_from_hdf5(this, group_id, temperature, method, &

2
src/simulation.F90
View file

@ -370,7 +370,7 @@ contains
subroutine initialize_simulation()
!$omp parallel
allocate(micro_xs(n_nuclides_total))
allocate(micro_xs(n_nuclides))
! Allocate array for matching filter bins
allocate(filter_matches(n_filters))

8
src/summary.F90
View file

@ -79,12 +79,12 @@ contains
! Write useful data from nuclide objects
nuclide_group = create_group(file_id, "nuclides")
call write_attribute(nuclide_group, "n_nuclides", n_nuclides_total)
call write_attribute(nuclide_group, "n_nuclides", n_nuclides)
! Build array of nuclide names and awrs
allocate(nucnames(n_nuclides_total))
allocate(awrs(n_nuclides_total))
do i = 1, n_nuclides_total
allocate(nucnames(n_nuclides))
allocate(awrs(n_nuclides))
do i = 1, n_nuclides
if (run_CE) then
nucnames(i) = nuclides(i) % name
awrs(i) = nuclides(i) % awr

4
src/tally.F90
View file

@ -2205,7 +2205,7 @@ contains
atom_density = ZERO
! Determine score for each bin
call score_general(p, t, n_nuclides_total*t % n_score_bins, filter_index, &
call score_general(p, t, n_nuclides*t % n_score_bins, filter_index, &
i_nuclide, atom_density, flux)
end subroutine score_all_nuclides
@ -2300,7 +2300,7 @@ contains
elseif (k == p % event_nuclide + 1) then
! After we've tallied the individual nuclide bin, we also need
! to contribute to the total material bin which is the last bin
k = n_nuclides_total + 1
k = n_nuclides + 1
else
! After we've tallied in both the individual nuclide bin and the
! total material bin, we're done