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https://github.com/openmc-dev/openmc.git
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Add material_add_nuclide function to C API (required a little restructuring)
This commit is contained in:
Paul Romano
parent
75af32e019
commit
76f99c2a00
10 changed files with 252 additions and 139 deletions
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@ -317,7 +317,6 @@ set(LIBOPENMC_FORTRAN_SRC
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src/endf.F90
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src/endf_header.F90
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src/energy_distribution.F90
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src/energy_grid.F90
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src/error.F90
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src/finalize.F90
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src/geometry.F90
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@ -28,6 +28,9 @@ class _OpenMCLibrary(object):
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self._dll.openmc_init.restype = None
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self._dll.openmc_load_nuclide.argtypes = [c_char_p]
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self._dll.openmc_load_nuclide.restype = c_int
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self._dll.openmc_material_add_nuclide.argtypes = [
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c_int, c_char_p, c_double]
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self._dll.openmc_material_add_nuclide.restype = c_int
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self._dll.openmc_material_get_densities.argtypes = [
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c_int, _double_array]
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self._dll.openmc_material_get_densities.restype = c_int
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@ -126,6 +129,27 @@ class _OpenMCLibrary(object):
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"""
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return self._dll.openmc_load_nuclide(name.encode())
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def material_add_nuclide(self, mat_id, name, density):
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"""Add a nuclide to a material.
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Parameters
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----------
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mat_id : int
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ID of the material
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name : str
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Name of nuclide, e.g. 'U235'
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density : float
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Density in atom/b-cm
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Returns
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-------
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int
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Return status (negative if an error occurs).
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"""
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return self._dll.openmc_material_add_nuclide(
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mat_id, name.encode(), density)
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def material_get_densities(self, mat_id):
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"""Get atom densities in a material.
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196
src/api.F90
196
src/api.F90
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@ -24,6 +24,7 @@ module openmc_api
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public :: openmc_finalize
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public :: openmc_find
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public :: openmc_init
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public :: openmc_material_add_nuclide
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public :: openmc_material_get_densities
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public :: openmc_material_set_density
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public :: openmc_load_nuclide
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@ -36,44 +37,6 @@ module openmc_api
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contains
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function openmc_material_set_density(id, density) result(err) bind(C)
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integer(C_INT), value, intent(in) :: id
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real(C_DOUBLE), value, intent(in) :: density
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integer(C_INT) :: err
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integer :: i
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err = -1
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if (allocated(materials)) then
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if (material_dict % has_key(id)) then
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i = material_dict % get_key(id)
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associate (m => materials(i))
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err = m % set_density(density, nuclides)
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end associate
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end if
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end if
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end function openmc_material_set_density
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function openmc_material_get_densities(id, ptr) result(n) bind(C)
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integer(C_INT), intent(in), value :: id
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type(C_PTR), intent(out) :: ptr
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integer(C_INT) :: n
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ptr = C_NULL_PTR
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n = 0
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if (allocated(materials)) then
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if (material_dict % has_key(id)) then
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i = material_dict % get_key(id)
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associate (m => materials(i))
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if (allocated(m % atom_density)) then
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ptr = C_LOC(m % atom_density(1))
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n = size(m % atom_density)
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end if
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end associate
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end if
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end if
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end function openmc_material_get_densities
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!===============================================================================
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! OPENMC_CELL_SET_TEMPERATURE sets the temperature of a cell
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!===============================================================================
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@ -138,20 +101,16 @@ contains
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character(kind=C_CHAR) :: name(*)
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integer(C_INT) :: err
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integer :: i, n
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integer(HID_T) :: file_id, group_id
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character(10) :: name_
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integer :: n
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integer(HID_T) :: file_id
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integer(HID_T) :: group_id
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character(:), allocatable :: name_
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real(8) :: minmax(2) = [ZERO, INFINITY]
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type(VectorReal) :: temperature
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type(Nuclide), allocatable :: new_nuclides(:)
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! Copy array of C_CHARs to normal Fortran string
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name_ = ''
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i = 1
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do while (name(i) /= C_NULL_CHAR)
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name_(i:i) = name(i)
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i = i + 1
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end do
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name_ = to_f_string(name)
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err = -1
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if (.not. nuclide_dict % has_key(to_lower(name_))) then
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@ -184,12 +143,136 @@ contains
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! Assign resonant scattering data
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if (res_scat_on) call assign_0K_elastic_scattering(nuclides(n))
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! Initialize nuclide grid
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call nuclides(n) % init_grid(energy_min_neutron, &
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energy_max_neutron, n_log_bins)
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err = 0
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else
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err = -2
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end if
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end if
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end function openmc_load_nuclide
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!===============================================================================
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! OPENMC_MATERIAL_ADD_NUCLIDE
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!===============================================================================
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function openmc_material_add_nuclide(id, name, density) result(err) bind(C)
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integer(C_INT), value, intent(in) :: id
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character(kind=C_CHAR) :: name(*)
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real(C_DOUBLE), value, intent(in) :: density
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integer(C_INT) :: err
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integer :: i, j, k, n
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integer :: err2
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real(8) :: awr
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integer, allocatable :: new_nuclide(:)
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real(8), allocatable :: new_density(:)
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character(:), allocatable :: name_
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name_ = to_f_string(name)
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err = -1
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if (allocated(materials)) then
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if (material_dict % has_key(id)) then
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i = material_dict % get_key(id)
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associate (m => materials(i))
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! Check if nuclide is already in material
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do j = 1, size(m % nuclide)
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k = m % nuclide(j)
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if (nuclides(k) % name == name_) then
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awr = nuclides(k) % awr
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m % density = m % density + density - m % atom_density(j)
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m % density_gpcc = m % density_gpcc + (density - &
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m % atom_density(j)) * awr * MASS_NEUTRON / N_AVOGADRO
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m % atom_density(j) = density
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err = 0
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end if
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end do
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! If nuclide wasn't found, extend nuclide/density arrays
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if (err /= 0) then
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! If nuclide hasn't been loaded, load it now
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err2 = openmc_load_nuclide(name)
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if (err2 /= -2) then
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! Extend arrays
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n = size(m % nuclide)
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allocate(new_nuclide(n + 1))
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new_nuclide(1:n) = m % nuclide
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call move_alloc(FROM=new_nuclide, TO=m % nuclide)
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allocate(new_density(n + 1))
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new_density(1:n) = m % atom_density
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call move_alloc(FROM=new_density, TO=m % atom_density)
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! Append new nuclide/density
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k = nuclide_dict % get_key(to_lower(name_))
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m % nuclide(n + 1) = k
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m % atom_density(n + 1) = density
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m % density = m % density + density
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m % density_gpcc = m % density_gpcc + &
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density * nuclides(k) % awr * MASS_NEUTRON / N_AVOGADRO
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m % n_nuclides = n + 1
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err = 0
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end if
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end if
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end associate
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end if
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end if
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end function openmc_material_add_nuclide
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!===============================================================================
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! OPENMC_MATERIAL_GET_DENSITIES returns an array of nuclide densities in a
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! material
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!===============================================================================
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function openmc_material_get_densities(id, ptr) result(n) bind(C)
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integer(C_INT), intent(in), value :: id
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type(C_PTR), intent(out) :: ptr
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integer(C_INT) :: n
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ptr = C_NULL_PTR
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n = 0
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if (allocated(materials)) then
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if (material_dict % has_key(id)) then
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i = material_dict % get_key(id)
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associate (m => materials(i))
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if (allocated(m % atom_density)) then
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ptr = C_LOC(m % atom_density(1))
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n = size(m % atom_density)
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end if
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end associate
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end if
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end if
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end function openmc_material_get_densities
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!===============================================================================
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! OPENMC_MATERIAL_SET_DENSITY sets the total density of a material in atom/b-cm
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!===============================================================================
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function openmc_material_set_density(id, density) result(err) bind(C)
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integer(C_INT), value, intent(in) :: id
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real(C_DOUBLE), value, intent(in) :: density
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integer(C_INT) :: err
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integer :: i
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err = -1
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if (allocated(materials)) then
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if (material_dict % has_key(id)) then
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i = material_dict % get_key(id)
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associate (m => materials(i))
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err = m % set_density(density, nuclides)
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end associate
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end if
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end if
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end function openmc_material_set_density
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!===============================================================================
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! OPENMC_RESET resets all tallies
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!===============================================================================
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@ -306,4 +389,23 @@ contains
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end if
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end subroutine openmc_tally_results
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function to_f_string(c_string) result(f_string)
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character(kind=C_CHAR), intent(in) :: c_string(*)
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character(:), allocatable :: f_string
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integer :: i, n
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! Determine length of original string
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n = 0
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do while (c_string(n + 1) /= C_NULL_CHAR)
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n = n + 1
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end do
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! Copy C string character by character
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allocate(character(len=n) :: f_string)
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do i = 1, n
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f_string(i:i) = c_string(i)
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end do
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end function to_f_string
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end module openmc_api
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@ -2,7 +2,6 @@ module cross_section
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use algorithm, only: binary_search
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use constants
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use energy_grid, only: log_spacing
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use error, only: fatal_error
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use global
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use list_header, only: ListElemInt
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@ -1,67 +0,0 @@
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module energy_grid
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use global
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use list_header, only: ListReal
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use output, only: write_message
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implicit none
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integer :: grid_method ! how to treat the energy grid
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integer :: n_log_bins ! number of bins for logarithmic grid
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real(8) :: log_spacing ! spacing on logarithmic grid
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contains
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!===============================================================================
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! LOGARITHMIC_GRID determines a logarithmic mapping for energies to bounding
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! indices on a nuclide energy grid
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!===============================================================================
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subroutine logarithmic_grid()
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integer :: i, j, k ! Loop indices
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integer :: t ! temperature index
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integer :: M ! Number of equally log-spaced bins
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real(8) :: E_max ! Maximum energy in MeV
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real(8) :: E_min ! Minimum energy in MeV
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real(8), allocatable :: umesh(:) ! Equally log-spaced energy grid
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! Set minimum/maximum energies
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E_max = energy_max_neutron
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E_min = energy_min_neutron
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! Determine equal-logarithmic energy spacing
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M = n_log_bins
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log_spacing = log(E_max/E_min)/M
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! Create equally log-spaced energy grid
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allocate(umesh(0:M))
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umesh(:) = [(i*log_spacing, i=0, M)]
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do i = 1, n_nuclides_total
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associate (nuc => nuclides(i))
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do t = 1, size(nuc % grid)
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! Allocate logarithmic mapping for nuclide
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allocate(nuc % grid(t) % grid_index(0:M))
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! Determine corresponding indices in nuclide grid to energies on
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! equal-logarithmic grid
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j = 1
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do k = 0, M
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do while (log(nuc % grid(t) % energy(j + 1)/E_min) <= umesh(k))
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! Ensure that for isotopes where maxval(nuc % energy) << E_max
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! that there are no out-of-bounds issues.
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if (j + 1 == size(nuc % grid(t) % energy)) exit
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j = j + 1
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end do
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nuc % grid(t) % grid_index(k) = j
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end do
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end do
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end associate
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end do
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deallocate(umesh)
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end subroutine logarithmic_grid
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end module energy_grid
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@ -112,6 +112,9 @@ module global
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real(8) :: temperature_default = 293.6_8
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real(8) :: temperature_range(2) = [ZERO, ZERO]
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integer :: n_log_bins ! number of bins for logarithmic grid
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real(8) :: log_spacing ! spacing on logarithmic grid
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! ============================================================================
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! MULTI-GROUP CROSS SECTION RELATED VARIABLES
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@ -11,7 +11,6 @@ module initialize
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use constants
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use dict_header, only: DictIntInt, ElemKeyValueII
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use set_header, only: SetInt
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use energy_grid, only: logarithmic_grid, grid_method
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use error, only: fatal_error, warning
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use geometry, only: neighbor_lists, count_instance, calc_offsets, &
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maximum_levels
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@ -109,12 +108,6 @@ contains
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end if
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if (run_mode /= MODE_PLOTTING) then
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! Construct information needed for nuclear data
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if (run_CE) then
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! Construct log energy grid for cross-sections
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call logarithmic_grid()
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end if
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! Allocate and setup tally stride, filter_matches, and tally maps
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call configure_tallies()
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@ -7,7 +7,6 @@ module input_xml
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use distribution_multivariate
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use distribution_univariate
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use endf, only: reaction_name
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use energy_grid, only: grid_method, n_log_bins
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use error, only: fatal_error, warning
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use geometry_header, only: Cell, Lattice, RectLattice, HexLattice, &
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get_temperatures, root_universe
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@ -5240,9 +5239,6 @@ contains
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allocate(nuclides(n_nuclides_total))
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allocate(sab_tables(n_sab_tables))
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!$omp parallel
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allocate(micro_xs(n_nuclides_total))
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!$omp end parallel
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! Read cross sections
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do i = 1, size(materials)
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@ -5294,6 +5290,13 @@ contains
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end do
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end do
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! Set up logarithmic grid for nuclides
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do i = 1, size(nuclides)
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call nuclides(i) % init_grid(energy_min_neutron, &
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energy_max_neutron, n_log_bins)
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end do
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log_spacing = log(energy_max_neutron/energy_min_neutron) / n_log_bins
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do i = 1, size(materials)
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! Skip materials with no S(a,b) tables
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if (.not. allocated(materials(i) % sab_names)) cycle
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@ -97,6 +97,7 @@ module nuclide_header
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contains
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procedure :: clear => nuclide_clear
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procedure :: from_hdf5 => nuclide_from_hdf5
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procedure :: init_grid => nuclide_init_grid
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procedure :: nu => nuclide_nu
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procedure, private :: create_derived => nuclide_create_derived
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end type Nuclide
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@ -683,4 +684,42 @@ module nuclide_header
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end function nuclide_nu
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subroutine nuclide_init_grid(this, E_min, E_max, M)
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class(Nuclide), intent(inout) :: this
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real(8), intent(in) :: E_min ! Minimum energy in MeV
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real(8), intent(in) :: E_max ! Maximum energy in MeV
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integer, intent(in) :: M ! Number of equally log-spaced bins
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integer :: i, j, k ! Loop indices
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integer :: t ! temperature index
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real(8) :: spacing
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real(8), allocatable :: umesh(:) ! Equally log-spaced energy grid
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! Determine equal-logarithmic energy spacing
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spacing = log(E_max/E_min)/M
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! Create equally log-spaced energy grid
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allocate(umesh(0:M))
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umesh(:) = [(i*spacing, i=0, M)]
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do t = 1, size(this % grid)
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! Allocate logarithmic mapping for nuclide
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allocate(this % grid(t) % grid_index(0:M))
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! Determine corresponding indices in nuclide grid to energies on
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! equal-logarithmic grid
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j = 1
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do k = 0, M
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do while (log(this % grid(t) % energy(j + 1)/E_min) <= umesh(k))
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||||
! Ensure that for isotopes where maxval(this % energy) << E_max
|
||||
! that there are no out-of-bounds issues.
|
||||
if (j + 1 == size(this % grid(t) % energy)) exit
|
||||
j = j + 1
|
||||
end do
|
||||
this % grid(t) % grid_index(k) = j
|
||||
end do
|
||||
end do
|
||||
|
||||
end subroutine nuclide_init_grid
|
||||
|
||||
end module nuclide_header
|
||||
|
|
|
|||
|
|
@ -42,17 +42,7 @@ contains
|
|||
type(Particle) :: p
|
||||
integer(8) :: i_work
|
||||
|
||||
if (.not. restart_run) call initialize_source()
|
||||
|
||||
! Display header
|
||||
if (master) then
|
||||
if (run_mode == MODE_FIXEDSOURCE) then
|
||||
call header("FIXED SOURCE TRANSPORT SIMULATION", 3)
|
||||
elseif (run_mode == MODE_EIGENVALUE) then
|
||||
call header("K EIGENVALUE SIMULATION", 3)
|
||||
if (verbosity >= 7) call print_columns()
|
||||
end if
|
||||
end if
|
||||
call initialize_simulation()
|
||||
|
||||
! Turn on inactive timer
|
||||
call time_inactive % start()
|
||||
|
|
@ -380,12 +370,40 @@ contains
|
|||
|
||||
end subroutine replay_batch_history
|
||||
|
||||
!===============================================================================
|
||||
! INITIALIZE_SIMULATION
|
||||
!===============================================================================
|
||||
|
||||
subroutine initialize_simulation()
|
||||
|
||||
!$omp parallel
|
||||
allocate(micro_xs(n_nuclides_total))
|
||||
!$omp end parallel
|
||||
|
||||
if (.not. restart_run) call initialize_source()
|
||||
|
||||
! Display header
|
||||
if (master) then
|
||||
if (run_mode == MODE_FIXEDSOURCE) then
|
||||
call header("FIXED SOURCE TRANSPORT SIMULATION", 3)
|
||||
elseif (run_mode == MODE_EIGENVALUE) then
|
||||
call header("K EIGENVALUE SIMULATION", 3)
|
||||
if (verbosity >= 7) call print_columns()
|
||||
end if
|
||||
end if
|
||||
|
||||
end subroutine initialize_simulation
|
||||
|
||||
!===============================================================================
|
||||
! FINALIZE_SIMULATION calculates tally statistics, writes tallies, and displays
|
||||
! execution time and results
|
||||
!===============================================================================
|
||||
|
||||
subroutine finalize_simulation
|
||||
subroutine finalize_simulation()
|
||||
|
||||
!$omp parallel
|
||||
deallocate(micro_xs)
|
||||
!$omp end parallel
|
||||
|
||||
! Increment total number of generations
|
||||
total_gen = total_gen + n_batches*gen_per_batch
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue