Many next-generation light water reactor (LWR) concepts, such as mobile small modular reactors, are seeking to use smaller core dimensions than conventional reactor types. Smaller reactor cores require an increase in fissile material to maintain reactivity. For nonproliferation purposes, enrichment increases are limited to less than 20% high-assay low-enriched uranium (HALEU), and so, fuels with higher uranium density than UO2 must be considered. To this end, uranium-molybdenum (U-Mo) alloys were tested using the Fission Accelerated Steady-state Test (FAST) approach. The experiment test matrix is focused on identifying the temperature transition from low swelling and high fission gas retention to breakaway swelling and low fission gas retention. This paper documents the results of irradiation tests and postirradiation examinations including neutron radiography, rodlet profilometry, fission gas collection analysis, and optical metallography. The results of these tests showed that unconstrained U-Mo fuels (solid, Na-bonded rodlets) have a swelling threshold temperature between 400°C to 450°C with minimal fission gas release (FGR) below this point. Higher-temperature solid fuel showed microstructural zoning with small pore networks while lower-temperature solid fuels have a uniform microstructure with large pore networks. Annular U-Mo fuels, where swelling had some self-constraint imposed upon it, were shown to have much reduced swelling compared to their solid counterparts due to the compressive strains imposed during swelling, which correlated with the very low FGR for irradiation temperatures up to 500°C. These initial results show that the use of U-Mo in constrained fuel geometries could be used as a high uranium density HALEU fuel for LWRs.