1. Work a problem involving charged particle penetration and energy
deposition using Table 3.6 data.

2. List and describe any or all of the 5 neutron source mechanisms
discussed in class..

3. Calculate the neutron yield due to spontaneous fission and/or emission.

4. List and describe any or all of the 7 gamma ray source mechanisms
discussed in class.

5. Calculate the gamma ray yield from radioactive sources.

6. Define and discuss any of the terms defined in Lesson 14 and 15.

7. Define the response function.

8. Calculate deposited energy from: (1) Isotropic elastic scatter of
neutrons or (2) Isotropic inelastic scatter of neutrons

9. Discuss (and distinguish) free field flux response functions versus
local flux response functions.

10. Be able to work problems like any of the homework problems.

For the MCNP portion of the course, be able to:

1. Construct an SDEF card from a described source's SPATIAL (point,
surface, box, sphere, or cylinder shaped) and ENERGY characteristics.
I will NOT give you any cheat sheet about the SDEF parameters (ERG,
POS, RAD, ...)

2. Explain and set up a tally of the F1, F2, or F4 varieties.

3. Create a response function from provided data (continuous). You may
need your calculator for (energy, probability) values.

4. Create an energy-independent, unmultiplied, rectangular mesh tally.
I will give you enough information to figure out the tally grid: (x,y,z) min values, max values, and number of divisions.