Adding back in description of reaction identifiers.

Author: jlconlin

Listing.md

@@ -6,6 +6,10 @@ If you are concerned about what this does you are free to examine the code yours
 ## Types of Nuclear Data
 In the table below are given the different kinds of data, generally they are referred to by the incident particle. Each type is a link to a notebook where the details can be explored. In the table below lists the different types of data and the associated `lib_type`, the one- or two-letter extension of the full ZAID. 
 
+
+### Ordering of Listing Tables
+The order of the elements in the listing tables is important; it is the same order as in the XSDIR file. When MCNP looks for data from the material card, it scans the XSDIR file---starting at the top---until it finds the *first* matching ZAID (or partial ZAID). Thus, the order of the XSDIR file is important and it is reflected in the order of the listing tables.
+
 | Type                                                           | `lib_type`  |
 |:-------------------------------------------------------------- | --------    |
 | [Continuous-Energy Neutron](ContinuousEnergyNeutron.ipynb)     | `c` or `nc` |
@@ -17,3 +21,10 @@ In the table below are given the different kinds of data, generally they are ref
 | [Dosimetry](Dosimetry.ipynb)                                   | `y`         |
 | [Electron](Electron.ipynb)                                     | `e`         |
 | [Charged-Particle](ChargedParticle.ipynb)                      | varied      |
+
+## Reaction Types and Tallying Reaction Rates
+In may of the notebooks linked above, you'll find a table of reaction types after the listing of the data. The reaction types are identified using the ENDF `MT` numbers. These numbers uniquely define a specific reaction. For example, `MT` 102, $(z,\gamma)$ is the "radiative capture" cross section induced by some incident particle $z$. For more information, see [Appendix B of the ENDF manual](https://www.nndc.bnl.gov/csewg/docs/endf-manual.pdf).
+
+The `MT` numbers can be used in MCNP with the tally multiplier card, `FM`. The `FM` card multiplies any tallied quantity (flux, current) by the specified cross section to produce reaction rates. The `MT` numbers are also used to identify reactions when plotting cross sections. In addition to the ENDF `MT` numbers, MCNP also defines `FM` numbers that are similar to the `MT` numbers. **Note:** `MT` numbers are always positive, while `FM` numbers are always negative. In each of the table of reation types, both the `MT` and `FM` numbers are given. 
+
+For more information about tallying reaction rates, see the MCNP manual.