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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Micro-Raman technology to interrogate two-phase extraction on a microfluidic device

Microfluidic devices provide ideal environments to study solvent extraction. When droplets form and generate plug flow down the microfluidic channel, the device acts as a microreactor in which the kinetics of chemical reactions and interfacial transfer can be examined. Among the many benefits of microreactors is the ability to maintain small sample volumes, which is especially important when studying solvent extraction in harsh environments, such as in separations related to the nuclear fuel cycle. In solvent extraction, the efficiency of the process depends on complex formation and rates of transfer in biphasic systems. Thus, it is important to understand the kinetic parameters in an extraction system to maintain a high efficiency and effectivity of the process. Here we present a methodology that combines chemometric analysis with on-line micro-Raman spectroscopy to monitor biphasic extractions within a microfluidic device. This monitoring provided concentration measurements in both organic and aquoeus plugs as they flowed through the microfluidic channel. The biphasic system studied was comprised of HNO3 as the aqueous phase and 30 % (v/v) tributyl phosphate in n-dodecane comprised the organic phase, which simulated the Plutonium Uranium Redox EXtraction (PUREX) process. Using pre-equilibrated solutions (post extraction) the validity of the technique and methodology is illustrated. Following this validation, solutions that were not equilibrated were examined and the kinetics of interfacial mass transfer within the biphasic system were established. Results were compared to kinetics already determined on a macro scale to prove the efficacy of the technique

extraction, interfacial transport, liquid:liquid e↗
DOE OSTI2021

Microfluidic Liquid-Liquid Extraction Chip with Integrated Raman Sensors (Phase I and Phase II Final Report)

As demand for electricity continues to increase worldwide, the world’s nuclear power generating capacity will continue to grow. Nuclear power is the most environmentally benign way of producing electricity on a large scale. The long-term successful use of nuclear power, however, is critically dependent upon adequate and safe processing and disposal of spent nuclear fuels. A very important feature of nuclear energy is that spent fuels can be reprocessed to recover fissile and fertile materials that can then be used as fresh fuel for nuclear power plants. The DOE-NE Fuel Cycle Research and Development (FCR&D) is currently developing nuclear material reprocessing technologies. In a typical nuclear fuel reprocessing system, centrifugal contactors are used as liquid-liquid extraction devices where two immiscible liquids are mixed at high speeds using a rotor, which creates a fine dispersion of droplets of organic phase in an aqueous phase that contains the analyte. Understanding the extraction efficiency at these contactors through modeling and simulation is important in the development of reprocessing technologies and the optimization of current technologies such as the PUREX process. The outcome of this program will be a liquid-liquid microfluidic flow cell chip with embedded spectroscopic sensors for the analysis of extracted analytes. This device will be useful in the optimization of new fuel reprocessing schemes as well as existing reprocessing processes by providing a microfluidic modeling platform to optimize extraction parameters. The Phase I and Phase II work developed a microfluidic chip design that allows integration of Raman and absorption fiber optic probes and successfully demonstrated the feasibility of using Raman and absorption probes to detect key analytes that partition into the aqueous and organic phases.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Response of soil bacteria to PUREX chemicals suggests biomarker utility and bioremediation potential

Chemicals involved in plutonium uranium reduction extraction (PUREX) can be released from nuclear reprocessing facilities and accumulate in the environment. We exposed chemically diverse soils to a range of concentrations of key chemicals used in the PUREX process. The responses of soil microbial communities are dependent on soil type, and tributyl phosphate exposure generates the most reproducible changes in microbial communities. We reconstructed the genomes of key bacteria and find several phosphotriesterase genes found only in Rhizobiaceae. The abundance of phosphotriesterase genes is significantly higher in samples exposed to tributyl phosphate. In conclusion, these phosphotriesterase genes may be involved in breakdown of tributyl phosphate, and a means of accessing phosphate for these bacteria.

Bioremediation↗
DOE OSTI2022

Neptunium redox speciation and determination of its total concentration in dissolved fuel simulant solutions by spectrophotometry

Here, two new approaches to measure Np concentration in dissolved used nuclear fuel simulant (aqueous feed for PUREX process) by spectrophotometry are developed. The first approach is based on chemical reduction of Np in the feed to its tetravalent state using ascorbic acid with simultaneous conversion of Pu(IV) to Pu(III). Interfering effects from light absorbing fission and corrosion products are accounted for by measuring optical absorbance spectrum of aqueous raffinate after extraction of U, Np, and Pu by tributyl phosphate in dodecane. The second approach uses no chemical treatment at all and relies on spontaneous valency adjustment of Np to Np(V) by dilution of the feed with water to reduce its acidity to low decimolar range of nitric acid concentration. Results of Np determination in the feed by spectrophotometry are in good agreement with its concentration measured by ICP-MS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Selective Partitioning of Ruthenium from Nitric Acid Media

Ruthenium removal from PUREX raffinate solutions is imperative for the application of some advanced actinide partitioning technologies, particularly partitioning of hexavalent americium. Ruthenium interferes with the oxidation of americium and suppresses its extraction. Selective extraction of ruthenium from these solutions is complicated by the existence of multiple ruthenium species in nitric acid. In this work, ruthenium is selectively extracted from nitric acid media through oxidation to ruthenium tetroxide and absorption into polymer beads. Extraction of ruthenium tetroxide into a variety of organic solvents is also explored. The use of UV-Visible spectroscopy to quantify extraction of ruthenium tetroxide was demonstrated. This method for extraction of ruthenium from nitric acid was applied to PUREX raffinate simulant solution and was found to simultaneously remove ruthenium (through oxidation and absorption) and zirconium (through precipitation). This method not only removes ruthenium and zirconium from solution but also separates ruthenium and zirconium from each other. Thus, the two most problematic fission product elements for the PUREX process are separated in a single step. Further, americium was not extracted or sorbed, which should allow it to be partitioned following ruthenium/zirconium removal using existing oxidative treatments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Analysis of Venting of a Resin Slurry

A resin slurry venting analysis was conducted to address safety issues associated with over-pressurization of ion exchange columns used in the plutonium uranium redox extraction (PUREX) process at the U. S. Department of Energy’s (DOE’s) Savannah River Site (SRS). If flow to these columns were inadvertently interrupted, an exothermic runaway reaction could occur between the ion exchange resin and the nitric acid used in the feed stream. The nitric acid-resin reaction generates significant quantities of noncondensable gases, which would pressurize the column. To prevent the column from rupturing during such events, rupture disks are installed on the column vent lines. The venting analysis models accelerating rate calorimeter (ARC) tests and data from tests that were performed in a vented test vessel with a rupture disk. The tests showed that the pressure inside the test vessel continued to increase after the rupture disk opened, though at a slower rate than prior to the rupture. The increase in the vessel pressure is modeled as a transient phenomenon associated with expansion of the resin slurry/gas mixture upon rupture of the disk. It is postulated that the maximum pressure at the end of this expansion is limited by energy minimization to approximately 1.5 times the rupture disk burst pressure. The magnitude of this pressure increase is consistent with the measured pressure transients. The results of this analysis demonstrate the need to allow for a margin between the design pressure and the rupture disk burst pressure in similar applications.

42 ENGINEERING↗
DOE OSTI2017

Tc-99 Decontamination From Heat Treated Gaseous Diffusion Membrane -Phase I

Uranium gaseous diffusion cascades represent a significant environmental challenge to dismantle, containerize and dispose as low-level radioactive waste. Baseline technologies rely on manual manipulations involving direct access to technetium-contaminated piping and materials. There is a potential to utilize novel thermal decontamination technologies to remove the technetium and allow for on-site disposal of the very large uranium converters. Technetium entered these gaseous diffusion cascades as a hexafluoride complex in the same fashion as uranium. Technetium, as the isotope Tc-99, is an impurity that follows uranium in the first cycle of the Plutonium and Uranium Extraction (PUREX) process. The technetium speciation or exact form in the gas diffusion cascades is not well defined. Several forms of Tc-99 compounds, mostly the fluorinated technetium compounds with varying degrees of volatility have been speculated by the scientific community to be present in these cascades. Therefore, there may be a possibility of using thermal desorption, which is independent of the technetium oxidation states, to perform an in situ removal of the technetium as a volatile species and trap the radionuclide on sorbent traps which could be disposed as low-level waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Reducing Actinide Production Using Inert Matrix Fuels

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Oxidation and Extraction of Americium(VI) from Lanthanides

Partitioning and transmutation of americium and curium has been the subject of significant research in the nuclear fuel cycle over the past two decades. Numerous approaches have been studied to separate these elements from the lanthanide elements to enable their recycle and transmutation. Americium is one of the primary long-term heat sources in used nuclear fuel, which may impact repository loading. The lanthanides are both a) neutron poisons, reducing reactor performance, if recycled, and b) can chemically interact with fuel cladding to cause early failure, if present in sufficient concentration. Chemical separation approaches for heterogeneous fuel recycle, typically utilize proven technology (PUREX or modifications to the PUREX process) to remove uranium and plutonium from the dissolved fuel (and if desired, can remove neptunium as well). Follow-on processing is then required to separate the americium and curium, or the americium alone (for target fabrication). In a homogeneous recycle approach, the bulk of the uranium is typically separated first, followed by additional processing to remove the transuranic elements as a group. Because of the chemical similarity of the americium with the lanthanides, and particularly with curium, these separate processes are difficult. Utilizing the unique valence states of americium (primarily the (VI) oxidation state), opens up new opportunities for the separation of americium for either heterogeneous recycle (enabling americium separation alone) or homogeneous recycle (separating americium together with plutonium and neptunium). The US Department of Energy Fuel Cycle Technologies program has been researching methods to oxidize americium and to extract it from lanthanides as americium(VI), preferably from molar nitric acid solutions. Early efforts identified sodium bismuthate as an effective oxidant, which enabled testing of the extraction of americium(VI) by tributyl phosphate and diamylamylphosphonate. Successful extraction of americium(VI) was demonstrated in this system, in both batch contact laboratory tests and in centrifugal contactors; however, the sodium bismuthate is a solid, and must be in excess of saturation with the solution. This creates some engineering challenges in the process. A significant effort to identify alternative oxidants has led to a number of promising approaches. Silver-catalyzed ozone, copper(III) periodate, and electrochemical oxidation of americium from the trivalent to hexavalent state have all been successfully demonstrated in laboratory experiments. These approaches would enable oxidation and extraction of americium without having to filter solids. Once the americium is oxidized, a key issue is how long it will remain in the hexavalent state. Investigations have shown that some materials of construction (e.g., tantalum) do not reduce americium as quickly as stainless steel. Also, alternative extractants, such as amides, do not appear to reduce the americium as rapidly as the phosphate or phosphonate extractants. While still early in their development, these advances in the technology offer some positive indications that a workable extraction process for the separation of americium(VI) can be realized in the near future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Mock Nuclear Processing Facility-Safeguards Training Requirements

This document outlines specific training requirements in the topical areas of Material Control and Accounting (MC&A) and Physical Protection(PP) which are to be used as technical input for designing a mock Integrated Security Facility (ISF) at Sandia National Laboratories (SNL). The overall project objective for these requirements is to enhance the ability to deliver training on Material Protection Control and Accounting (MC&A) concepts regarding hazardous material such as irradiated materials with respect to bulk processing facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2016

Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Automation Toward Tether-less Robotic Operations in Nuclear Facilities - 17474

Automation capabilities that enable communication-denied nuclear robotic operations have come of age. These layer over tethered operation or enable un-tethered systems for some operations. Tele-operated, tethered devices are traditionally preferred for nuclear remote systems. Tethers provide comm, power, and a means for mechanical extraction. They also restrict mobility and create additional failure points, and in some cases, circumstances preclude mechanical recovery by tether. Tethering has superb advantages where it is viable, but a subset of nuclear EM challenges compel mobile remote systems that forgo tethering. Tethering was essential at a time when autonomy or wireless comm could not accomplish viable operations, but now an emerging class of un-tethered exploration and service robots are compelling for situations where tethers compound waste and exposure, or are not viable. Technical advances in localization, modeling, planning, autonomy, integration and reliability have made it possible to augment tethered operations or operate altogether without tether in selective operations. Tether preclusion technologies include localization, perception, navigation, safeguarding and task prescription. These effectively accomplish tasks like way-point following, wall-following, coverage patterning, navigate-to-goal, next-best view and many others. Although these are not yet general for the broad agenda of the nuclear complex, these are already useful for many operations. These provide a foundational base that can be built upon over time for more complex operations yet to come. As a context and analog study, the technologies are exhibited by robotically (without tether) modeling a coal mine with tracks and train cars. This is analogous to technology and operation that might apply for inspection of waste storage at WIPP or the PUREX tunnels. Autonomous exploration of this nature is within state-of-art and possible with or without tether. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Vitrification of High-Level Waste: how to Select the Best Suited Technology - 18117

Since more than 30 years now, a number of countries have selected and operated vitrification as the best industrial practice for safely and durably immobilizing the most active liquid waste. This now significant worldwide experience covers a wide range of waste compositions, from concentrated alkaline defense waste sludges with a variety of salts and chemicals (Na, Fe, Al, Cr, sulphate,... ) to relatively clear acidic HLLW from commercial, low salt, PUREX-type processes, waste that include dissolved cladding material (Al, Mg, stainless steel), waste that include high levels of noble metals and fines... The activity levels vary from 40- year old defense waste to relatively young waste from high burn-up commercial fuel. The current throughputs vary from about 5 kg/h of glass to about 90 kg/h of glass per melter. Even higher throughputs are currently being considered for Hanford facility for instance. In order to cope with the range of waste compositions, glass compositions have also evolved from low melting temperature borosilicate or phosphate glass to glass or glass ceramics with higher melting temperatures. More complex matrices are currently under study in a number of countries, to immobilize more complex waste and / or increase waste loadings. Each challenge is unique and, for each challenge, it is necessary to select the best combination of waste form and technology. For instance for very active solutions with high levels of noble metal particles that could settle, increase apparent glass viscosity, and impact electrical conductivity, one would select borosilicate glass in the mechanically-stirred hot wall induction melter with bottom pouring that has proven its efficiency in industrial conditions at La Hague. For high glass throughputs with low solids in a borosilicate melt (typically Defense waste), ceramic melters have proven their reliability and efficiency. For matrices that need higher temperatures, or with very high corrosiveness, Cold-Crucible Induction Melters are now considered, with one first industrial implementation at La Hague to produce borosilicate glass and a glass-ceramic. All these technologies have been improved over the years to increase throughput, waste loading, melting temperature, reliability... Other types of waste are now being studied worldwide, and for each, it is necessary to carefully select the adapted waste form and technology. The technologies studied will range from traditional technologies (such as hot-wall induction melter technology and liquid-fed ceramic melter technologies) to the more recent advanced technologies (such as Cold Crucible Induction Melter or Advanced Ceramic melters). In addition, the field of vitrification is extending to other types of waste; the current advances in this field will also be described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Countercurrent Actinide Lanthanide Separation Process (ALSEP) Demonstration Test with a Simulated PUREX Raffinate in Centrifugal Contactors on the Laboratory Scale

An Actinide Lanthanide Separation Process (ALSEP) for the separation of trivalent actinides (An(III)) from simulated raffinate solution was successfully demonstrated using a 32-stage 1 cm annular centrifugal contactor setup. The ALSEP solvent was composed of a mixture of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and N,N,N',N'-tetra-(2-ethylhexyl)-diglycolamide (T2EHDGA) in n-dodecane. Flowsheet calculations and evaluation of the results were done using the Argonne’s Model for Universal Solvent Extraction (AMUSE) code using single-stage distribution data. The co-extraction of Zr(IV) and Pd(II) was prevented using CDTA (trans-1,2-diaminocyclohexane-N,N,N',N'-tetraacetic acid) as a masking agent in the feed. For the scrubbing of co-extracted Mo; citrate-buffered acetohydroxamic acid was used. The separation of An(III) from the trivalent lanthanides (Ln(III)) was achieved using citrate-buffered diethylene-triamine-N,N,N',N",N"-pentaacetic acid (DTPA), and Ln(III) were efficiently back extracted using N,N,N',N'-tetraethyl-diglycolamide (TEDGA). A clean An(III) product was obtained with a recovery of 95% americium and curium. The Ln(III) were efficiently stripped; but the Ln(III) product contained 5% of the co-stripped An(III). The carryover of Am and Cm into the Ln(III) product is attributed to too few actinide stripping stages, which was constrained by the number of centrifugal contactors available. Improved separation would be achieved by increasing the number of An strip stages. The heavier lanthanides (Pr, Nd, Sm, Eu, and Gd) and yttrium were mainly routed to the Ln product, whereas the lighter lanthanides (La and Ce) were mostly routed to the raffinate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI1961

Laboratory Development of a Pot Calcination Process for Converting Liquid Wastes to Solids

Simulated Darex, Purex, Thorex, and TBP-25 wastes were batch-evaporated and calcined to solids. The weight of residue was decreased to ~ 8% of the original waste for Darex, to ~ 4% for Purex, to ~ 2% for Thorex, and ~ 7% for TBP-25 by calcination to 1200°C. Semicontinuous evaporation and calcination to solids at 900°C of Darex and Purex wastes gave volume reduction factors of ~ 8 and 7 to 10, respectively. The nitrate contents of the residues from batch calcination to 800°C were from ~ 0.07 to 0.5 wt % while the nitrate contents of the residues from semicontinuous calcination to 900°C were from ~ 0.1 to 0.7 wt %. Sodium, calcium, and magnesium additives to Purex waste decreased the percent of sulfate volatilized into the condensate from ~ 30% to <0.5%. Replacement of the atmosphere above the waste with nitric oxide decreased ruthenium volatility from 50 to 60% of that originally present in Purex waste to 0.5 to 3.5%. The thermal conductivities for the calcined wastes, measured in situ, were all >0.1 Btu/hr ft °F at ~ 400°F and increased almost linearly with increasing temperature to >0.3 Btu/hr ft °F at ~ 1600°F in all cases studied.

052001* -- Nuclear Fuels-- Waste Processing↗
DOE OSTI2017

Managing Zirconium Chemistry and Phase Compatibility in Combined Process Separations for Minor Actinide Partitioning

In response to the NEUP Program Supporting Fuel Cycle R&D Separations and Waste Forms call DEFOA- 0000799, this report describes the results of an R&D project focusing on streamlining separation processes for advanced fuel cycles. An example of such a process relevant to the U.S. DOE FCR&D program would be one combining the functions of the TRUEX process for partitioning of lanthanides and minor actinides from PUREX(UREX) raffinates with that of the TALSPEAK process for separating transplutonium actinides from fission product lanthanides. A fully-developed PUREX(UREX)/TRUEX/TALSPEAK suite would generate actinides as product(s) for reuse (or transmutation) and fission products as waste. As standalone, consecutive unit-operations, TRUEX and TALSPEAK employ different extractant solutions (solvating (CMPO, octyl(phenyl)-N,Ndiisobutylcarbamoylmethylphosphine oxide) vs. cation exchanging (HDEHP, di-2(ethyl)hexylphosphoric acid) extractants), and distinct aqueous phases (2-4 M HNO 3 vs. concentrated pH 3.5 carboxylic acid buffers containing actinide selective chelating agents). The separate processes may also operate with different phase transfer kinetic constraints. Experience teaches (and it has been demonstrated at the lab scale) that, with proper control, multiple process separation systems can operate successfully. However, it is also recognized that considerable economies of scale could be achieved if multiple operations could be merged into a single process based on a combined extractant solvent. The task of accountability of nuclear materials through the process(es) also becomes more robust with fewer steps, providing that the processes can be accurately modeled. Work is underway in the U.S. and Europe on developing several new options for combined processes (TRUSPEAK, ALSEP, SANEX, GANEX, ExAm are examples). There are unique challenges associated with the operation of such processes, some relating to organic phase chemistry, others arising from the variable composition of the aqueous medium. This project targets in particular two problematic issues in designing combined process systems: managing the chemistry of challenging aqueous species (like Zr 4+ ) and optimizing the composition and properties of combined extractant organic phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗