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Medical Isotope Production without Highly Enriched Uranium
Direct replacement of the HEU in the target with LEU (with an increase in the number of targets that are irradiated).
Increase the mass of U-235 in the LEU target by increasing target size.
Increase the mass of U-235 in the target by changing the composition of the target meat.
These approaches are described in the following sections.
Direct Replacement of the HEU in the Target with LEU
HEU and LEU have essentially the same physical and chemical properties, so the direct replacement of HEU by LEU in the target meat would pose no particular target design, fabrication, or testing challenges. The LEU target would have the same geometry, heat transfer, and chemical processing properties as the equivalent HEU target and could be irradiated and processed in essentially the same manner. Assuming the same target design and uranium density, the yield of Mo-99 from the LEU target would be only about 20 percent of the HEU target it replaces owing to its reduced uranium-235 (U-235) mass and increased neutron capture.1 Consequently, approximately five times as many LEU targets would have to be irradiated and processed to produce the same amount of Mo-99 as a single HEU target, and up to five times as much volume of waste from target processing might be produced as a result. Some producers have suggested that their facilities might not be able to accommodate these higher throughput requirements without substantial modification.
Facility modifications might not be necessary, however, if certain process changes are made. Current target dissolution processes (see Chapter 2) operate well below solubility limits using containers that are small relative to the hot cells in which they sit. Higher throughputs could be accommodated by increasing container sizes and/or increasing material concentrations in the solvent.2 Liquid waste can also be converted to solids by precipitation, evaporation, or calcination to substantially reduce its volume. Moreover, given its lower U-235 enrichment, this solid waste can be more closely packed together in storage containers and facilities without increasing criticality risks. Because the LEU targets produce the same amount of fission heat and heat-producing fission products as HEU targets
Most HEU targets have 93 percent U-235 enrichments; LEU targets would have 19.75 percent enrichments. Neutron capture in LEU targets (primarily by uranium-238 [U-238]) would be about 15 percent higher than in equivalent-sized HEU targets.
Increasing material concentrations in the solvent could lead to criticality problems if HEU targets are used but would not be a problem for LEU targets.