Delbert L. Bowers

Extraction behaviour of actinides and lanthanides in TALSPEAK, TRUEX and NPEX processes of UREX+

Abstract Bench-scale studies to determine the extraction behavior of Pu, Np, Am and lanthanides with the organophosphorus extractants TBP, CMPO and HDEHP have been carried out. Based on the results obtained using actual spent nuclear fuel solutions, enhancements to the NPEX, TRUEX and TALSPEAK processes have been successful. In NPEX, >99.94% of both Np and Pu were separated from the fission products. In TRUEX, essentially complete recovery of the actinides (An) and the lanthanides (Ln) was achieved. In TALSPEAK, the complete separation of Pu, Np and Am from the lanthanides was demonstrated several times under various process conditions. The recovery of transuranics (TRU), including Am and Cm, is nearly 100% (below detection limit in the Ln stream), while the total recovery of Ln in the product stream exceeded 99.97%.

The production, separation, and use of 67Cu for radioimmunotherapy: a review.

A review of the literature pertaining to the production and separation of (67)Cu. This isotope is useful from both therapeutic and diagnostic standpoints due to its medium energy beta particle, gamma emissions, and 2.6-day half-life. It has been produced via proton, neutron, and gamma irradiations on zinc followed by solvent extraction, ion exchange, electrodeposition, and/or sublimation. Widespread use of this isotope for clinical studies and preliminary treatments has been limited by unreliable supplies, cost, and difficulty in obtaining therapeutic quantities.

The Determination of Impurities in Plutonium Metal by Anion Exchange and ICP/AES

The determination of trace-metal impurities in plutonium metal was investigated with the use of anion exchange separation and analysis by inductively coupled plasma/atomic emission spectrometry. Data on recoveries and independent measurements for some 30 elements show precisions and accuracies in the 1–10% range, values that are a significant improvement over those of traditional carrier distillation techniques. Americium can be determined on the same sample aliquot, with figures of merit rivaling those of nuclear counting procedures.

ICP/AES Actinide Detection Limits

Inductively coupled plasma-atomic emission spectrometry (ICP-AES) has been applied to the determination of some actinides and the isotopic composition of plutonium. The analysis of plutonium for metal impurities, including americium, has been reported. Prevalent figures of merit stated in the ICP-AES literature are experimentally determined detection limits. As a consequence, these were studied for the most sensitive lines of americium, curium, neptunium, and plutonium on a spectrometer system of moderate resolution. Our results are compared to published detection limits for some of the actinides. Data for curium, an element not previously investigated, are also included.

A Solution-Based Approach for Mo-99 Production: Considerations for Nitrate versus Sulfate Media

Molybdenum-99 is the parent of Technetium-99m, which is used in nearly 80% of all nuclear medicine procedures. The medical community has been plagued by Mo-99 shortages due to aging reactors, such as the NRU (National Research Universal) reactor in Canada. There are currently no US producers of Mo-99, and NRU is scheduled for shutdown in 2016, which means that another Mo-99 shortage is imminent unless a potential domestic Mo-99 producer fills the void. Argonne National Laboratory is assisting two potential domestic suppliers of Mo-99 by examining the effects of a uranyl nitrate versus a uranyl sulfate target solution configuration on Mo-99 production. Uranyl nitrate solutions are easier to prepare and do not generate detectable amounts of peroxide upon irradiation, but a high radiation field can lead to a large increase in pH, which can lead to the precipitation of fission products and uranyl hydroxides. Uranyl sulfate solutions are more difficult to prepare, and enough peroxide is generated during irradiation to cause precipitation of uranyl peroxide, but this can be prevented by adding a catalyst to the solution. A titania sorbent can be used to recover Mo-99 from a highly concentrated uranyl nitrate or uranyl sulfate solution; however, different approaches must be taken to prevent precipitation during Mo-99 production.

FY-15 Progress Report on Cleanup of irradiated SHINE Target Solutions Containing 140g-U/L Uranyl Sulfate

During FY 2012 and 2013, a process was developed to convert the SHINE Target Solution (STS) of irradiated uranyl sulfate (140 g U/L) to uranyl nitrate. This process is necessary so that the uranium solution can be processed by the UREX (Uranium Extraction) separation process, which will remove impurities from the uranium so that it can be recycled. The uranyl sulfate solution must contain <0.02 M SO42- so that the uranium will be extractable into the UREXsolvent. In addition, it is desired that the barium content be below 0.0007 M, as this is the limit in the Resource Conservation and Recovery Act (RCRA).