Neutronic Analysis of a Liquid Metal Fast Reactor Experiment in the Large-I Position of the Advanced Test Reactor

Graduate Thesis Or Dissertation

Public Deposited

Citeable URL: https://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/028713274

Descriptions

Attribute Name	Values
Creator	Lilley, Benjamin C.
Abstract	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.
License	All rights reserved
Resource Type	Masters Thesis
Date Issued	2020-05-29
Degree Level	Master's
Degree Name	Master of Science (M.S.)
Degree Field	Nuclear Engineering
Degree Grantor	Oregon State University
Commencement Year	2021
Advisor	Palmer, Todd S.
Committee Member	Marcum, Wade Reese, Steven R.
Academic Affiliation	Nuclear Science and Engineering
Déclaration de droits	In Copyright
Publisher	Oregon State University
Peer Reviewed	No
Language	English [eng]

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