Material testing experiments are needed to enable the next generation of nuclear fission reactors. Concluding in 2009, the Boosted Fast Flux Loop (BFFL) project was devised as a way to test fast reactor materials and fuels in the Advanced Test Reactor (ATR), however the experiment location it used is now not feasible. This work discusses the neutronic analysis of a modified BFFL design using sodium coolant in the underutilized large-I position in the ATR. The computer codes Attila, MC**2, and MCNP were used to model the geometry and neutron interactions of the design.
The fast neutron flux magnitude is 2 x 10^13 n / (cm^2 s), well below the minimum BFFL experiment criteria of 3 x 10^15 n / (cm^2 s). The fast-to-thermal flux ratio of 8 is also below the minimum design criteria of 15. With further modification, this design may meet minimum materials testing performance in the large-I position.