[en] Providing clean water and energy for about nine billion people on the earth by midcentury is a daunting challenge. Major investments in efficiency of energy and water use and deployment of all economical energy sources will be needed. Separations technology has an important role to play in producing both clean energy and water. Some examples are carbon dioxide capture and sequestration from fossil energy power plants and advanced nuclear fuel cycle scemes. Membrane separations systems are under development to improve the economics of carbon capture that would be required at a huge scale. For nuclear fuel cycles, only the PUREX liquid-liquid extraction process has been deployed on a large scale to recover uranium and plutonium from used fuel. Most current R and D on separations technology for used nuclear fuel focuses on ehhancements to a PUREX-type plant to recover the minor actinides (neptunium, americiu, and curium) and more efficiently disposition the fission products. Are there more efficient routes to recycle the actinides on the horizon? Some new approaches and barriers to development will be briefly reviewed.

[en] The scope of this project is to determine the feasibility of washing plutonium-containing combustible residues using ultrasonic disruption as a method for dislodging particulate. Removal of plutonium particulate and, to a lesser extent, solubilized plutonium from the organic substrate should substantially reduce potential fire, explosion or radioactive release hazards due to radiolytic hydrogen generation or high flammability. Tests were conducted on polypropylene filters which were used as pre-filters in the rich-residue ion-exchange process at the Los Alamos Plutonium Facility. These filters are similar to the Ful-Floreg_sign cartridges used at Rocky Flats that make up a substantial fraction of the combustible residues with the highest hazard rating. Batch experiments were run on crushed filter material in order to determine the amount of Pu removed by stirring, stirring and sonication, and stirring and sonication with the introduction of Pu-chelating water-soluble polymers or surfactants. Significantly more Pu is removed using sonication and sonication with chelators than is removed with mechanical stirring alone

[en] Several new types of bifunctional anion-exchange resin have been designed, synthesized, and tested for their ability to remove Pu(IV) from nitric acid. The functional group present in all of these resins is a pyridinium unit derived from the host poly(4-vinylpyridine) resin. Bifunctionality was achieved by adding a second anion-exchange site to the pyridine nitrogen via an alkylene spacer that encompassed ethylene through hexylene. The attached anion-exchange sites include trimethylammonium, pyridinium, and trimethylphosphonium. The sorption kinetics and distribution coefficients of Pu(IV) from 5 to 9 M nitric acid generally were best when the two anion-exchange sites were separated by a five-carbon spacer. The order of affinity for Pu(IV) from 7 M nitric acid was pyridinium>phosphonium>trimethylammonium. Replacing the central methylene unit in a five-carbon spacer with an ether-oxygen further enhanced the sorption of Pu(IV) onto the trimethylammonium-containing resin. Because these bifunctional resins are modifications of existing commercial ReillexTM resins, they are much easier to prepare than resins that require complete synthesis. We plan to evaluate these and related bifunctional resin structures for their ability to selectively remove other targeted anions from groundwater and industrial waste streams

[en] Soluble anion-exchange polymers have been designed, synthesized, and evaluated for their ability to take up Pu(IV) from nitric acid solutions. These polymers, based on linear poly(4-vinylpyridine) (PVP) and polyethyleneimine (PEI) are soluble in aqueous and strong acid solutions. Weak-base sites on the polymers are protonated under experimental conditions, and, in most cases, have been converted by alkylation to form mono- and bifunctional strong-base anion-exchange sites. Distribution of Pu(IV) onto these polymers was determined by comparing visible spectroscopic data in the presence and absence of the soluble polymer. Overall plutonium affinity for the anion-exchange sites in the soluble materials is found to be much lower than for comparable solid resins, but the distribution behavior follows similar trends in that bifunctionalized materials are superior to monofunctionalized and a five-atom 'spacer' between the two cationic sites is superior to other spacer lengths

[en] The aging of the US nuclear stockpile presents a number of challenges, including the increasing radioactivity of plutonium residues due to the ingrowth of ²⁴¹Am from the β-decay of ²⁴¹Pu. We investigated parameters that affect the sorption of Am onto anion-exchange resins from concentrated effluents derived from nitric acid processing of plutonium residues. These postevaporator wastes are nearly saturated solutions of acidic nitrate salts, and americium removal is complicated by physical factors, such as solution viscosity and particulates, as well as by the presence of large quantities of competing metals and acid. Single- and double-contact batch distribution coefficients for americium and neodymium from simple and complex surrogate solutions are presented. Varied parameters include the nitrate salt concentration and composition and the nitric acid concentration. We find that under these extremely concentrated conditions, Am(III) removal efficiencies can surpass 50% per contact. Distribution coefficients for both neodymium and americium are insensitive to solution acidity and appear to be driven primarily by low water activities of the solutions

[en] Americium (III) exhibits an unexpectedly high affinity for anion-exchange material from the high-salt evaporator bottoms solutions--an effect which has not been duplicated using simple salt solutions. Similar behavior is observed for its lanthanide homologue, Nd(III), in complex evaporator bottoms surrogate solutions. There appears to be no single controlling factor--acid concentration, total nitrate concentration or solution ionic strength--which accounts for the approximately 2-fold increase in retention of the trivalent ions from complex solutions relative to simple solutions. Calculation of species activities (i.e., water, proton and nitrate) in such concentrated mixed salt solutions is difficult and of questionable accuracy, but it is likely that the answer to forcing formation of anionic nitrate complexes of americium lies in the relative activities of water and nitrate. From a practical viewpoint, the modest americium removal needs (ca. 50--75%) from nitric acid evaporator bottoms allow sufficient latitude for the use of non-optimized conditions such as running existing columns filled with older, well-used Reillex HPQ. Newer materials, such as HPQ-100 and the experimental bifunctional resins, which exhibit higher distribution coefficients, would allow for either increased Am removal or the use of smaller columns. It is also of interest that one of the experimental neutral-donor solid-support extractants, DHDECMP, exhibits a similarly high level of americium (total alpha) removal from EV bottoms and is much less sensitive to total acid content than commercially-available material

[en] The processing of used nuclear fuel in alkaline solution provides potentially useful new selectivity for separating the actinides from each other and f rom the fission products. Over the ast decade, several research teams around the world have considered dissolution of used fuel in alkaline solution and further partitioning in this medium as an alternative to acid dissolution. The chemistry of the actinides and fission products in alkaline soilltion requires extensive investigation to more carefully evaluate its potential for developing useful separation methods for used nuclear fueI.

[en] Three compositionally identical Pu-bearing reference glasses were prepared in July 1982, each containing 1 wt.% PuO2; however, the 238Pu/239Pu isotopic ratio was different in each glass. As a result, the alpha-activities in the as-prepared glasses varied by nearly a factor of 200. The actual activities measured are within 15% of the intended values. In the 20 some years since their preparation, several studies have been performed on these glasses. The final results of the most recent studies are summarized in this paper.

[en] Project goals: Can used nuclear fuel be partitioned by dissolution in alkaline aqueous solution to give a solution of uranium, neptunium, plutonium, americium and curium and a filterable solid containing nearly all of the lanthanide fission products and certain other fission products? What is the chemistry of Am/Cm/Ln in oxidative carbonate solutions? Can higher oxidation states of Am be stabilized and exploited? Conclusions: Am(VI) is kinetically stable in 0.5-2.0 M carbonate solutions for hours. Aliquat 336 in toluene has been successfully shown to extract U(VI) and Pu(VI) from carbonate solutions. (Stepanov et al 2011). Higher carbonate concentration gives lower D, SF_U/Eu for = 4 in 1 M K₂CO₃. Experiments with Am(VI) were unsuccessful due to reduction by the organics. Multiple sources of reducing organics...more optimization. Reduction experiments of Am(VI) in dodecane/octanol/Aliquat 336 show that after 5 minutes of contact, only 30-40% of the Am(VI) has been reduced. Long enough to perform an extraction. Shorter contact times, lower T, and lower Aliquat 336 concentration still did not result in any significant extraction of Am. Anion exchange experiments using a strong base anion exchanger show uptake of U(VI) with minimal uptake of Nd(III). Experiments with Am(VI) indicate Am sorption with a Kd of 9 (10 minute contact) but sorption mechanism is not yet understood. SF_U/Nd for = 7 and SF_U/Eu for = 19 after 24 hours in 1 M K₂CO₃.

[en] The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.