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Add a neutron flux spectrum tally top the OpenMC model defined in #2 and record a 175 group neutron flux spectrum.

Compare the number densities of each nuclide between the ALARA and OpenMC activation problems.

• One approach is to plot the ratio as a function of the nuclide ID.
• Another approach is to plot the ratio as a function of the number density (chose one result as a reference). This will allow us to be less concerned about differences where the number density is very low.

Full define all parts of a simple spherical shell problem:

• source
• materials
• geometry
• boundary conditions
• tallies

Define an ALARA input file that is consistent with the definitions of the OpenMC problem in #2. Perform an activation problem with a reasonable total flux magnitude and irradiation schedule to achieve some substantial activation.

• total source strength to normalize neutron flux magnitude
• irradiation history

Learn how to perform an activation simulation with OpenMC and extend the problem defined in #2 with the irradiation schedule defined in #4

Using the problem defined in #1, setup and run an OpenMC simulation that successfully transports neutrons through the geometry.