Determine natural element expansion based on cross section library

This pull request adds the capability for a user to specify what cross section library they are using so that when an `<element>` is present in a material, OpenMC knows how to properly split it into different isotopes. This will make it easier to have material input files that will work with different cross section libraries. In an ideal world, all cross section libraries would have all naturally-occurring isotopes; however, this is not the case. When an isotope is missing, its abundance is lumped into other nuclides. Some libraries also have cross sections for natural elements themselves. The most common example is Carbon, which is still a natural element in all modern libraries.

The extra input is just `<natural_elements>` in settings.xml which can take values such as "ENDF/B-VII.1" or "JEFF-3.2".