mit-crpg/openmc — pull request #280 by paulromano (2014-06-03 01:56:05+00:00)
Title: Fix tallying <nuclides> with void materials

If you have a tally that includes a void material and you try to use `` to specify that a specific nuclide, or all, should be scored, a segfault occurs. This happens when the tallying routine tries to reference `materials(p % material)`; if the material is void, `p%material = MATERIAL_VOID = -1`.

This pull request checks to make sure that the material isn't a void material when tallying with ``. A test was also added to the test suite to make sure this is covered in the future.