Koistinen,

**Quote:**

Neutron flux is typically not evenly distributed in practical reactors.

I now have a simple model of my reactor, including some neutron flux assumptions.

The core is a 1 meter radius sphere, within a 2 meter radius blanket. The neutron flux at the very center is 2.4 “units” - using whatever multiplier is needed to match a notional average of 2.4 freed neutrons per fission. Most of the neutrons are fast, and escape the inner core. 1.0 of them cause fission in the inner core, which I assume reaches 50% of the way to the core/blanket barrier. In the outer core, keff is less than 1.0, dropping to an assumed 0.5 at the barrier. Note that theses two assumptions use the 50% “first cut estimate” in operations research - if you don't the answer, set upper and lower boundaries, and take the midpoint.

I assume the flux is 1.2 at the barrier, another 50% assumption. But this assumption is based on geometry, which is a bit better justified. Consider a nucleus at the very top of the core. It gets hit by neutrons only from below, so it “sees” only half as many neutrons as a nucleus at the center where the flux is 2.4.

I also assume the flux at the outer blanket boundary is not zero and some neutrons are truly wasted on the containment walls. It's tempting to say 0.2 are wasted, since I'm assuming a 1.00 breeding ratio. However, I fear that would be mixing neutrons and neutron flux.

The outer core is 8x the volume of the inner core. The blanket is 8x the volume of the whole core. This is just a coincidence. It comes from assumption 1 in my first post: that we will need 4 times more TH than (5%) U. This is modified by the model I just described, where the flux in the blanket is about half that in the core. Given Titanium-48’s suggestion, 8x may be closer to 9x, and the blanket radius will be just over 2 meters. 2 meters might still be good if Thorium is slightly more soluble in molten salt than Uranium. 2 meters will not be enough if we go with a tube rather than a sphere.

Actually I'm not sure I need this level of design, though it's fun. I'm trying to simulate the waste streams of various MSR designs. The breeding stream (hardly waste!) is a sideshow, mostly included because it is one of the simplest actinide evolutions. For that, I can just assume that each year in 1 GW iso-breeder, 1000 kg of 232-Th evolves to 1000 kg 233-U. I might have some "waste" (232-U) in this stream if I ever get around to incorporating the (n, 2n) reaction in my sim.