From 20331519beb2e940fff582f46423e4635bc61684 Mon Sep 17 00:00:00 2001
From: Paul Romano <paul.k.romano@gmail.com>
Date: Sat, 29 Jul 2017 15:15:39 -0500
Subject: [PATCH] Add TallyView.nuclides

---
 openmc/capi/tally.py | 12 ++++++++++++
 src/api.F90          | 27 ++++++++++++++++++++++++++-
 2 files changed, 38 insertions(+), 1 deletion(-)

diff --git a/openmc/capi/tally.py b/openmc/capi/tally.py
index f60d94e92..b7139bcb6 100644
--- a/openmc/capi/tally.py
+++ b/openmc/capi/tally.py
@@ -16,6 +16,10 @@ _dll.openmc_get_tally.errcheck = _error_handler
 _dll.openmc_tally_id.argtypes = [c_int32, POINTER(c_int32)]
 _dll.openmc_tally_id.restype = c_int
 _dll.openmc_tally_id.errcheck = _error_handler
+_dll.openmc_tally_nuclides.argtypes = [
+    c_int32, POINTER(POINTER(c_int)), POINTER(c_int)]
+_dll.openmc_tally_nuclides.restype = c_int
+_dll.openmc_tally_nuclides.errcheck = _error_handler
 _dll.openmc_tally_results.argtypes = [
     c_int32, POINTER(POINTER(c_double)), POINTER(c_int*3)]
 _dll.openmc_tally_results.restype = c_int
@@ -39,6 +43,14 @@ class TallyView(object):
         _dll.openmc_tally_id(self._index, tally_id)
         return tally_id.value
 
+    @property
+    def nuclides(self):
+        nucs = POINTER(c_int)()
+        n = c_int()
+        _dll.openmc_tally_nuclides(self._index, nucs, n)
+        return [NuclideView(nucs[i]).name if nucs[i] > 0 else 'total'
+                for i in range(n.value)]
+
     @property
     def results(self):
         """Get tally results array
diff --git a/src/api.F90 b/src/api.F90
index 6d50ca51e..cd8a8a064 100644
--- a/src/api.F90
+++ b/src/api.F90
@@ -44,6 +44,7 @@ module openmc_api
   public :: openmc_reset
   public :: openmc_run
   public :: openmc_tally_id
+  public :: openmc_tally_nuclides
   public :: openmc_tally_results
 
   ! Error codes
@@ -335,7 +336,7 @@ contains
   end function openmc_get_tally
 
 !===============================================================================
-! OPENMC_LOAD_LOAD loads a nuclide from the cross section library
+! OPENMC_LOAD_NUCLIDE loads a nuclide from the cross section library
 !===============================================================================
 
   function openmc_load_nuclide(name) result(err) bind(C)
@@ -657,6 +658,30 @@ contains
     end if
   end function openmc_tally_id
 
+!===============================================================================
+! OPENMC_TALLY_NUCLIDES returns the list of nuclides assigned to a tally
+!===============================================================================
+
+  function openmc_tally_nuclides(index, ptr, n) result(err) bind(C)
+    integer(C_INT32_T), value :: index
+    type(C_PTR), intent(out) :: ptr
+    integer(C_INT), intent(out) :: n
+    integer(C_INT) :: err
+
+    err = E_UNASSIGNED
+    if (index >= 1 .and. index <= size(tallies)) then
+      associate (t => tallies(index))
+        if (allocated(t % nuclide_bins)) then
+          ptr = C_LOC(t % nuclide_bins(1))
+          n = size(t % nuclide_bins)
+          err = 0
+        end if
+      end associate
+    else
+      err = E_OUT_OF_BOUNDS
+    end if
+  end function openmc_tally_nuclides
+
 !===============================================================================
 ! OPENMC_TALLY_RESULTS returns a pointer to a tally results array along with its
 ! shape. This allows a user to obtain in-memory tally results from Python