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Move openmc_tally_set_nuclides to C++

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This commit is contained in:
Sterling Harper 2019-02-05 17:58:45 -05:00
parent 19e7923d08
commit 53a0dc8d97
3 changed files with 49 additions and 124 deletions
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2
include/openmc/tallies/tally.h
View file

@ -61,7 +61,7 @@ public:
std::vector<int> scores_; //!< Filter integrands (e.g. flux, fission)
//! Index of each nuclide to be tallied. -1 indicates total material.
std::vector<int> nuclides_;
std::vector<int> nuclides_ {-1};
//! True if this tally has a bin for every nuclide in the problem
bool all_nuclides_ {false};

74
src/tallies/tally.cpp
View file

@ -532,6 +532,8 @@ Tally::set_scores(std::vector<std::string> scores)
void
Tally::set_nuclides(pugi::xml_node node)
{
nuclides_.clear();
// By default, we tally just the total material rates.
if (!check_for_node(node, "nuclides")) {
nuclides_.push_back(-1);
@ -1231,6 +1233,52 @@ openmc_tally_set_scores(int32_t index, int n, const char** scores)
return 0;
}
extern "C" int
openmc_tally_get_nuclides(int32_t index, int** nuclides, int* n)
{
// Make sure the index fits in the array bounds.
if (index < 1 || index > model::tallies.size()) {
set_errmsg("Index in tallies array is out of bounds.");
return OPENMC_E_OUT_OF_BOUNDS;
}
//TODO: off-by-one
*n = model::tallies[index-1]->nuclides_.size();
*nuclides = model::tallies[index-1]->nuclides_.data();
return 0;
}
extern "C" int
openmc_tally_set_nuclides(int32_t index, int n, const char** nuclides)
{
// Make sure the index fits in the array bounds.
if (index < 1 || index > model::tallies.size()) {
set_errmsg("Index in tallies array is out of bounds.");
return OPENMC_E_OUT_OF_BOUNDS;
}
std::vector<std::string> words(nuclides, nuclides+n);
std::vector<int> nucs;
for (auto word : words){
if (word == "total") {
nucs.push_back(-1);
} else {
auto search = data::nuclide_map.find(word);
if (search == data::nuclide_map.end()) {
set_errmsg("Nuclide \"" + word + "\" has not been loaded yet");
return OPENMC_E_DATA;
}
nucs.push_back(search->second);
}
}
//TODO: off-by-one
model::tallies[index-1]->nuclides_ = nucs;
return 0;
}
extern "C" int
openmc_tally_get_filters(int32_t index, const int32_t** indices, int* n)
{
@ -1266,22 +1314,6 @@ openmc_tally_set_filters(int32_t index, int n, const int32_t* indices)
return 0;
}
extern "C" int
openmc_tally_get_nuclides(int32_t index, int** nuclides, int* n)
{
// Make sure the index fits in the array bounds.
if (index < 1 || index > model::tallies.size()) {
set_errmsg("Index in tallies array is out of bounds.");
return OPENMC_E_OUT_OF_BOUNDS;
}
//TODO: off-by-one
*n = model::tallies[index-1]->nuclides_.size();
*nuclides = model::tallies[index-1]->nuclides_.data();
return 0;
}
//==============================================================================
// Fortran compatibility functions
//==============================================================================
@ -1353,16 +1385,6 @@ extern "C" {
int tally_get_nuclide_bins_c(Tally* tally, int i)
{return tally->nuclides_[i-1];}
void
tally_set_nuclide_bins_c(Tally* tally, int n, int bins[], bool all_nuclides)
{
tally->nuclides_.clear();
tally->nuclides_.assign(bins, bins + n);
tally->all_nuclides_ = all_nuclides;
}
bool tally_get_all_nuclides_c(Tally* tally) {return tally->all_nuclides_;}
int tally_get_energyout_filter_c(Tally* tally)
{return tally->energyout_filter_;}

97
src/tallies/tally_header.F90
View file

@ -98,10 +98,6 @@ module tally_header
character(len=104) :: name = "" ! user-defined name
real(8) :: volume ! volume of region
! Values to score, e.g. flux, absorption, etc.
!integer :: n_score_bins = 0
!integer, allocatable :: score_bins_(:)
! Results for each bin -- the first dimension of the array is for scores
! (e.g. flux, total reaction rate, fission reaction rate, etc.) and the
! second dimension of the array is for the combination of filters
@ -133,8 +129,6 @@ module tally_header
procedure :: n_filter_bins => tally_get_n_filter_bins
procedure :: n_nuclide_bins => tally_get_n_nuclide_bins
procedure :: nuclide_bins => tally_get_nuclide_bins
procedure :: set_nuclide_bins => tally_set_nuclide_bins
procedure :: all_nuclides => tally_get_all_nuclides
procedure :: energyout_filter => tally_get_energyout_filter
procedure :: delayedgroup_filter => tally_get_delayedgroup_filter
procedure :: deriv => tally_get_deriv
@ -279,12 +273,6 @@ contains
subroutine tally_allocate_results(this)
class(TallyObject), intent(inout) :: this
integer, parameter :: default_nuclide_bins(1) = (/-1/)
! If no nuclides were specified, add a single bin for total material
if (this % n_nuclide_bins() == 0) then
call this % set_nuclide_bins(default_nuclide_bins)
end if
! Set total number of filter and scoring bins
this % total_score_bins = this % n_score_bins() * this % n_nuclide_bins()
@ -505,41 +493,6 @@ contains
nuclide = tally_get_nuclide_bins_c(this % ptr, i)
end function
subroutine tally_set_nuclide_bins(this, bins, all_nuclides)
class(TallyObject) :: this
integer :: bins(:)
logical, optional :: all_nuclides
logical(C_BOOL) :: all_
interface
subroutine tally_set_nuclide_bins_c(this, n, bins, all_nuclides) bind(C)
import C_PTR, C_INT, C_BOOL
type(C_PTR), value :: this
integer(C_INT), value :: n
integer(C_INT) :: bins(n)
logical(C_BOOL), value :: all_nuclides
end subroutine
end interface
if (present(all_nuclides)) then
all_ = logical(all_nuclides, kind=C_BOOL)
else
all_ = .false._C_BOOL
end if
call tally_set_nuclide_bins_c(this % ptr, size(bins), bins, all_)
end subroutine
function tally_get_all_nuclides(this) result(all_nuc)
class(TallyObject) :: this
logical(C_BOOL) :: all_nuc
interface
function tally_get_all_nuclides_c(tally) result(all_nuc) bind(C)
import C_PTR, C_BOOL
type(C_PTR), value :: tally
logical(C_BOOL) :: all_nuc
end function
end interface
all_nuc = tally_get_all_nuclides_c(this % ptr)
end function
function tally_get_energyout_filter(this) result(filt)
class(TallyObject) :: this
integer(C_INT) :: filt
@ -874,56 +827,6 @@ contains
end function openmc_tally_set_id
function openmc_tally_set_nuclides(index, n, nuclides) result(err) bind(C)
! Sets the nuclides in the tally which results should be scored for
integer(C_INT32_T), value :: index
integer(C_INT), value :: n
type(C_PTR), intent(in) :: nuclides(n)
integer(C_INT) :: err
integer, allocatable :: bins(:)
integer :: i
integer :: idx
character(C_CHAR), pointer :: string(:)
character(len=:, kind=C_CHAR), allocatable :: nuclide_
err = E_UNASSIGNED
if (index >= 1 .and. index <= size(tallies)) then
associate (t => tallies(index) % obj)
allocate(bins(n))
do i = 1, n
! Convert C string to Fortran string
call c_f_pointer(nuclides(i), string, [10])
nuclide_ = to_f_string(string)
select case (nuclide_)
case ('total')
bins(i) = -1
case default
idx = nuclide_map_get(to_c_string(nuclide_))
if (idx /= -1) then
bins(i) = idx - 1
else
err = E_DATA
call set_errmsg("Nuclide '" // trim(to_f_string(string)) // &
"' has not been loaded yet.")
return
end if
end select
end do
call t % set_nuclide_bins(bins)
err = 0
end associate
else
err = E_OUT_OF_BOUNDS
call set_errmsg('Index in tallies array is out of bounds.')
end if
end function openmc_tally_set_nuclides
subroutine openmc_get_tally_next_id(id) bind(C)
! Returns an ID number that has not been used by any other tallies.
integer(C_INT32_T), intent(out) :: id