Request for MaterialFromFilter for tally filters

[shimwell]

Description

Currently we have a nice way of tallying neutrons from particular cells that contribute to a tally with the openmc.CellFromFilter. This great for many use cases

However I have a use case where my detector cell is surrounded by void space (cell with no fill material) and I would like to know what cells or materials the neutrons interact with prior to arriving at the detector.

Hence it might be a solution to have a new filter called openmc.MaterialFromFilter which acts in a similar manner to the openmc.CellFromFilter. One difference is that neutrons won't scatter in the void material so the last interaction material will be able to show me where neutrons come from before interacting with my detector 🎉

Tagging @bam241 @rlbarker FYI

Alternatives

Add a flag to the CellFromFilter to ignore a subject of cells (like void cells)

Compatibility

Adding a new filter but not breaking API