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https://github.com/openmc-dev/openmc.git
synced 2026-07-27 21:55:41 -04:00
Fix last mesh bugs (all tests pass now)
This commit is contained in:
Paul Romano
parent
7a6d36c279
commit
f1fb400f92
6 changed files with 8 additions and 131 deletions
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@ -312,7 +312,6 @@ add_library(libopenmc SHARED
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src/material_header.F90
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src/math.F90
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src/matrix_header.F90
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src/mesh.F90
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src/mesh_header.F90
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src/message_passing.F90
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src/mgxs_data.F90
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@ -41,6 +41,8 @@ _dll.openmc_mesh_set_params.errcheck = _error_handler
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_dll.openmc_get_mesh_index.argtypes = [c_int32, POINTER(c_int32)]
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_dll.openmc_get_mesh_index.restype = c_int
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_dll.openmc_get_mesh_index.errcheck = _error_handler
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_dll.n_meshes.argtypes = []
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_dll.n_meshes.restype = c_int
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class Mesh(_FortranObjectWithID):
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@ -172,10 +174,10 @@ class _MeshMapping(Mapping):
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def __iter__(self):
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for i in range(len(self)):
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yield Mesh(index=i + 1).id
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yield Mesh(index=i).id
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def __len__(self):
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return c_int32.in_dll(_dll, 'n_meshes').value
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return _dll.n_meshes()
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def __repr__(self):
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return repr(dict(self))
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@ -214,7 +214,6 @@ contains
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use bank_header, only: source_bank
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use constants, only: ZERO, ONE
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use error, only: warning, fatal_error
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use mesh, only: count_bank_sites
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use message_passing
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use string, only: to_str
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106
src/mesh.F90
106
src/mesh.F90
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@ -1,106 +0,0 @@
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module mesh
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use algorithm, only: binary_search
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use bank_header, only: bank
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use constants
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use mesh_header
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use message_passing
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implicit none
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contains
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!===============================================================================
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! COUNT_BANK_SITES determines the number of fission bank sites in each cell of a
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! given mesh as well as an optional energy group structure. This can be used for
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! a variety of purposes (Shannon entropy, CMFD, uniform fission source
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! weighting)
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!===============================================================================
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subroutine count_bank_sites(m, bank_array, cnt, energies, size_bank, &
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sites_outside)
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type(RegularMesh), intent(in) :: m ! mesh to count sites
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type(Bank), intent(in) :: bank_array(:) ! fission or source bank
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real(8), intent(out) :: cnt(:,:) ! weight of sites in each
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! cell and energy group
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real(8), intent(in), optional :: energies(:) ! energy grid to search
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integer(8), intent(in), optional :: size_bank ! # of bank sites (on each proc)
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logical, intent(inout), optional :: sites_outside ! were there sites outside mesh?
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real(8), allocatable :: cnt_(:,:)
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integer :: i ! loop index for local fission sites
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integer :: n_sites ! size of bank array
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integer :: n ! number of energy groups / size
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integer :: mesh_bin ! mesh bin
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integer :: e_bin ! energy bin
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#ifdef OPENMC_MPI
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integer :: mpi_err ! MPI error code
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#endif
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logical :: outside ! was any site outside mesh?
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! initialize variables
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allocate(cnt_(size(cnt,1), size(cnt,2)))
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cnt_ = ZERO
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outside = .false.
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! Set size of bank
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if (present(size_bank)) then
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n_sites = int(size_bank,4)
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else
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n_sites = size(bank_array)
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end if
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! Determine number of energies in group structure
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if (present(energies)) then
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n = size(energies) - 1
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else
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n = 1
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end if
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! loop over fission sites and count how many are in each mesh box
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FISSION_SITES: do i = 1, n_sites
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! determine scoring bin for entropy mesh
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call m % get_bin(bank_array(i) % xyz, mesh_bin)
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! if outside mesh, skip particle
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if (mesh_bin == NO_BIN_FOUND) then
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outside = .true.
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cycle
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end if
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! determine energy bin
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if (present(energies)) then
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if (bank_array(i) % E < energies(1)) then
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e_bin = 1
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elseif (bank_array(i) % E > energies(n + 1)) then
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e_bin = n
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else
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e_bin = binary_search(energies, n + 1, bank_array(i) % E)
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end if
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else
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e_bin = 1
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end if
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! add to appropriate mesh box
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cnt_(e_bin, mesh_bin) = cnt_(e_bin, mesh_bin) + bank_array(i) % wgt
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end do FISSION_SITES
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#ifdef OPENMC_MPI
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! collect values from all processors
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n = size(cnt_)
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call MPI_REDUCE(cnt_, cnt, n, MPI_REAL8, MPI_SUM, 0, mpi_intracomm, mpi_err)
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! Check if there were sites outside the mesh for any processor
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if (present(sites_outside)) then
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call MPI_REDUCE(outside, sites_outside, 1, MPI_LOGICAL, MPI_LOR, 0, &
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mpi_intracomm, mpi_err)
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end if
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#else
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sites_outside = outside
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cnt = cnt_
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#endif
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end subroutine count_bank_sites
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end module mesh
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@ -443,7 +443,7 @@ void RegularMesh::bins_crossed(const Particle* p, std::vector<int>& bins,
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}
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}
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int j = xt::argmin(d)(0);
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j = xt::argmin(d)(0);
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if (u[j] > 0.0) {
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++ijk0(j);
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} else {
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@ -729,11 +729,11 @@ xt::xarray<double> RegularMesh::count_sites(int64_t n, const Bank* bank,
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extern "C" int
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openmc_extend_meshes(int32_t n, int32_t* index_start, int32_t* index_end)
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{
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if (!index_start) *index_start = meshes.size();
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if (index_start) *index_start = meshes.size();
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for (int i = 0; i < n; ++i) {
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meshes.emplace_back(new RegularMesh{});
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}
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if (!index_end) *index_end = meshes.size() - 1;
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if (index_end) *index_end = meshes.size() - 1;
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return 0;
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}
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@ -824,6 +824,7 @@ openmc_mesh_get_params(int32_t index, double** ll, double** ur, double** width,
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*ll = m->lower_left_.data();
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*ur = m->upper_right_.data();
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*width = m->width_.data();
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*n = m->n_dimension_;
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return 0;
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}
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@ -73,25 +73,7 @@ contains
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integer, intent(in) :: estimator
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type(TallyFilterMatch), intent(inout) :: match
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integer, parameter :: MAX_SEARCH_ITER = 100 ! Maximum number of times we can
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! can loop while trying to find
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! the first intersection.
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integer :: j ! loop index for direction
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integer :: n
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integer :: ijk0(3) ! indices of starting coordinates
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integer :: ijk1(3) ! indices of ending coordinates
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integer :: search_iter ! loop count for intersection search
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integer :: bin
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real(8) :: uvw(3) ! cosine of angle of particle
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real(8) :: xyz0(3) ! starting/intermediate coordinates
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real(8) :: xyz1(3) ! ending coordinates of particle
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real(8) :: xyz_cross ! coordinates of next boundary
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real(8) :: d(3) ! distance to each bounding surface
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real(8) :: total_distance ! distance of entire particle track
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real(8) :: distance ! distance traveled in mesh cell
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logical :: start_in_mesh ! starting coordinates inside mesh?
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logical :: end_in_mesh ! ending coordinates inside mesh?
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type(RegularMesh) :: m
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type(C_PTR) :: ptr_bins, ptr_weights
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