Report unresolvable nuclide names instead of terminating

docs/source/usersguide/tallies.rst

@@ -90,8 +90,13 @@ U238 (separately), we'd set::
 
   tally.nuclides = ['U235', 'U238']
 
-You can also list 'all' as a nuclide which will give you a separate reaction
-rate for every nuclide in the model.
+To get a separate reaction rate for every nuclide in a material, ask the
+material for its nuclides::
+
+  tally.nuclides = material.get_nuclides()
+
+Note that these are nuclide names; an element symbol such as 'Al' is not a valid
+entry, even for photon tallies, whose data is tabulated per element.
 
 The following tables show all valid scores:
 

src/cross_sections.cpp

@@ -209,8 +209,10 @@ void read_ce_cross_sections(const vector<vector<double>>& nuc_temps,
 
       const auto& temps = nuc_temps[i_nuc];
       int err = openmc_load_nuclide(name.c_str(), temps.data(), temps.size());
-      if (err < 0)
-        throw std::runtime_error {get_errmsg()};
+      if (err < 0) {
+        // Nothing catches this either, so report it the same way
+        fatal_error(get_errmsg());
+      }
 
       already_read.insert(name);
     }

src/tallies/tally.cpp

@@ -743,8 +743,13 @@ void Tally::set_nuclides(const vector<std::string>& nuclides)
       auto search = data::nuclide_map.find(nuc);
       if (search == data::nuclide_map.end()) {
         int err = openmc_load_nuclide(nuc.c_str(), nullptr, 0);
-        if (err < 0)
-          throw std::runtime_error {get_errmsg()};
+        if (err < 0) {
+          // Nothing catches this, so report it rather than terminating on an
+          // uncaught exception. Note that a tally bin names a nuclide, so an
+          // element symbol does not resolve even when the element is present.
+          fatal_error(fmt::format(
+            "Could not add nuclide '{}' to a tally: {}", nuc, get_errmsg()));
+        }
       }
       nuclides_.push_back(data::nuclide_map.at(nuc));
     }

tests/unit_tests/test_tally_nuclide_names.py

@@ -0,0 +1,55 @@
+"""A tally nuclide bin names a nuclide.
+
+Anything else has to be reported rather than terminating on an uncaught
+exception, which is what a name the data library cannot resolve used to do.
+"""
+
+import openmc
+import pytest
+
+
+def model_with_tally_nuclides(nuclides):
+    openmc.reset_auto_ids()
+
+    mat = openmc.Material()
+    mat.add_nuclide('U235', 1.0)
+    mat.set_density('g/cm3', 10.0)
+
+    sphere = openmc.Sphere(r=5.0, boundary_type='vacuum')
+    model = openmc.Model()
+    model.geometry = openmc.Geometry([openmc.Cell(fill=mat, region=-sphere)])
+    model.settings.run_mode = 'fixed source'
+    model.settings.particles = 100
+    model.settings.batches = 1
+    model.settings.source = openmc.IndependentSource(space=openmc.stats.Point())
+
+    tally = openmc.Tally()
+    tally.nuclides = nuclides
+    tally.scores = ['total']
+    model.tallies = openmc.Tallies([tally])
+    return model
+
+
+@pytest.mark.parametrize('name', ['Al', 'U', 'all', 'not-a-nuclide'])
+def test_unresolvable_nuclide_is_reported(run_in_tmpdir, name):
+    """An element symbol is not a nuclide, and neither is a typo."""
+    model = model_with_tally_nuclides([name])
+    with pytest.raises(RuntimeError, match='Could not add nuclide'):
+        model.run()
+
+
+def test_element_symbol_of_a_present_element(run_in_tmpdir):
+    """Reported even when the element itself is in the material.
+
+    The bin is matched against nuclide names, so 'U' does not resolve in a
+    material made of U235.
+    """
+    model = model_with_tally_nuclides(['U'])
+    with pytest.raises(RuntimeError, match='Could not add nuclide'):
+        model.run()
+
+
+def test_total_and_nuclide_bins_together(run_in_tmpdir):
+    """The total bin stays valid alongside real nuclides."""
+    model = model_with_tally_nuclides(['total', 'U235'])
+    model.run()