Introduction: Current fission-based methods of 99Mo production use targets that are based on high output without taking into consideration the amount of radioactive waste produced. We examine the idea of using a low enriched target (<20%) to reduce the amount of nuclear and chemical wastes produced by the manufacture of 99Mo via fission of uranium-based targets.Methods: MCNP6.2 was used to model targets of 20%, 10%, 7%, 3%, and 1% enrichment for 99Mo output and sustainability.Results and Discussion: The 1% enriched target at the lowest density of 0.2 g/cm3 of UO2 was found to have the highest sustainability score at 6 days irradiation but it had a low 99Mo output. On the other hand, the highest output target was found to be 20% enriched with a density of 8.0 g UO2/cm3 with an irradiation time of 20 days. Target security and safeguards concerns were modelled and found to be of minimal concern.

Introduction: The most common current uranium target type for 99Mo production uses plate geometry. This paper investigates the effects of geometry on 99Mo output and target sustainability.Methods: MCNP6.2 was used to model rectangular, spherical, and cylindrical targets ranging from 12.03 cm3 to 120.34 cm3 to determine the target geometry and volume effects on 99Mo output, sustainability, proliferation concerns, and heating. 4-7 days irradiation was used with a consistent target density of 2 g/cm3 UO2 for all target types.Results and Discussion: The cylindrical target was found to have the best performance due to having the second highest sustainability, the highest 99Mo output and produce the lowest amounts of 239Pu compared to the other target geometries. The heating comparison showed that there were negligible heating concerns for all target volumes and geometries.

A new target material combination was modelled to replace the existing uranium-aluminium design used for 99Mo manufacture to increase the sustainability of the production process. Previous efforts to develop a more sustainable uranium target for 99Mo production, resulted in the levels of 239Pu in the target after irradiation being elevated due to the increase in 238U present. MCNP6.2 was used to model 4 different cylindrical targets based on 4–7 days irradiation to further understand this effect. To reduce the resultant 239Pu levels, ratios of 0–99% of Ce were used as a replacement for 238U. The results show that the addition of 140Ce and the removal of 238U reduced the 239Pu levels in the target significantly thus increasing the sustainability of the target and giving a slight increase to the 99Mo output of the targets.