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A ceramic-metal composite (cermet) of uranium nitride (UN) particles embedded in a tungsten-molybdenum (W/Mo) alloy matrix is being considered as reactor fuel for nuclear thermal propulsion (NTP). One possible issue is the loss of fissile uranium atoms during reactor operation. We begin by reviewing historical data that suggest that a likely mechanism for fuel loss is transport of free uranium along metal grain boundaries and through cracks to free surfaces. We then employ simple two-dimensional (2D) mesoscale simulations to provide insights into crack formation and free uranium transport in a W/ Mo-UN cermet. Phase-field fracture simulations show that cracks form during cooling at the fuel-matrix interface and then within the fuel particles. Transport simulations show that cracks at the fuel-matrix interface and within the fuel accelerate fuel loss. Mechanical and diffusion data are needed for UN and W/Mo alloys to make these preliminary predictions more accurate.

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