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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Advanced Aqueous Separation Systems for Actinide Partitioning

One of the most challenging aspects of advanced processing of used nuclear fuel is the separation of transplutonium actinides from fission product lanthanides. This separation is essential if actinide transmutation options are to be pursued in advanced fuel cycles, as lanthanides compete with actinides for neutrons in both thermal and fast reactors, thus limiting efficiency. The separation is difficult because the chemistry of Am 3+ and Cm 3+ is nearly identical to that of the trivalent lanthanides (Ln 3+ ). The prior literature teaches that two approaches offer the greatest probability of devising a successful group separation process based on aqueous processes: 1) the application of complexing agents containing ligand donor atoms that are softer than oxygen (N, S, Cl - ) or 2) changing the oxidation state of Am to the IV, V, or VI state to increase the essential differences between Am and lanthanide chemistry (an approach utilized in the PUREX process to selectively remove Pu 4+ and UO 2 2+ from fission products). The latter approach offers the additional benefit of enabling a separation of Am from Cm, as Cm(III) is resistant to oxidation and so can easily be made to follow the lanthanides. The fundamental limitations of these approaches are that 1) the soft(er) donor atoms that interact more strongly with actinide cations than lanthanides form substantially weaker bonds than oxygen atoms, thus necessitating modification of extraction conditions for adequate phase transfer efficiency, 2) soft donor reagents have been seen to suffer slow phase transfer kinetics and hydro-/radiolytic stability limitations and 3) the upper oxidation states of Am are all moderately strong oxidants, hence of only transient stability in media representative of conventional aqueous separations systems. There are examples in the literature of both approaches having been described. However, it is not clear at present that any extant process is sufficiently robust for application at the scale necessary for commercial fuel processing supporting transmutation of transplutonium elements. This research project continued basic themes investigated by this research group during the past decade. In the Fuel Cycle Research and Development program at DOE, the current favorite process for accomplishing the separation of trivalent actinides from fission product lanthanides is the TALSPEAK process. TALSPEAK is a solvent extraction method (developed at Oak Ridge National Lab in the 1960s) based on the combination of a cation exchanging extractant (e.g., HDEHP), an actinide-selective aminopolycarboxylate complexing agent (e.g., DTPA), and a carboxylic acid buffer to control pH in the range of 3-4. Considerable effort has been expended in this research group during the past 8 years to elaborate the details of TALSPEAK in the interest of developing improved approaches to the operation of TALSPEAK-like systems. In this project we focused on defining aggregation phenomena in conventional TALSPEAK separations, on supporting the development of Advanced TALSPEAK processes, on profiling the aqueous complexation kinetics of lanthanides in TALSPEAK relevant aqueous media, on the design of new diglycolamide and N-donor extractants, and on characterizing cation-cation complexes of pentavalent actinides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Criticality concerns of a group actinide co-crystallization separations approach to used nuclear fuel recycling

We report the co-crystallization of hexavalent actinides with uranyl nitrate hexahydrate has been established as a viable approach to achieving group actinide separation in used nuclear fuel recycle while avoiding the difficulties associated with the solvent-extraction process altogether. A preliminary neutronics safety analysis has been performed to alleviate criticality concerns associated with this group hexavalent actinide co-crystallization approach. The basic uranyl nitrate hexahydrate material compositions used are representative of used fuel found at commercial pressurized water reactors all around the U.S. This material is subsequently doped with various percentages of either typical reactor grade plutonium or with the total transuranic vector, replacing the uranium atoms in the crystal lattice structure. Infinite and effective neutron multiplication factors were calculated using the state-of-the-art Monte Carlo neutron transport code MCNP in a variety of plausible material configurations, assuming an acceptable safety limit for criticality to be an effective neutron multiplication factor less than 0.95. The results indicate that the crystal structures form a very subcritical intrinsic geometry even with six water molecules being incorporated into the crystal structure. In conclusion, the results of this research suggest that significant opportunity exists for the safe transfer of the group actinide separation approach from a laboratory to an operational scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

OECD/NEA Ongoing activities related to the nuclear fuel cycle

As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclear systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Selectivity in Ligand Binding to Uranyl Compounds: A Synthetic, Structural, Thermodynamic and Computational Study

The uranyl cation (UO 2 2+ ) is the most abundant form of uranium on the planet. It is estimated that 4.5 billion tons of uranium in this form exist in sea water. The ability to bind and extract the uranyl cation from aqueous solution while separating it from other elements would provide a limitless source of nuclear fuel. A large body of research concerns the selective recognition and extraction of uranyl. A stable molecule, the cation has a linear O=U=O geometry. The short U-O bonds (1.78 Å) arise from the combination of uranium 5f/6d and oxygen 2p orbitals. Due to the oxygen moieties being multiply bonded, these sites were not thought to be basic enough for Lewis acidic coordination to be a viable approach to sequestration. We believe that the goal of developing a practical system for uranium separation from seawater will not be attained without new insights into our existing fundamental knowledge of actinide chemistry. We posit that detailed studies of the kinetic and thermodynamic factors that influence interactions between f-elements and ligands with a range of donor atoms is essential to any major advance in this important area. The goal of this research is thus to broaden the coordination chemistry of the uranyl ion by studying new ligand systems via synthetic, structural, thermodynamic and computational methods. We anticipate that this fundamental science will find use beyond actinide separation technologies in areas such as nuclear waste remediation and nuclear materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Pentavalent Curium, Berkelium, and Californium in Nitrate Complexes: Extending Actinide Chemistry and Oxidation States

Pentavalent actinyl nitrate complexes An V O 2 (NO 3 ) 2 - were produced by elimination of two NO 2 from An III (NO 3 ) 4 - for An = Pu, Am, Cm, Bk, and Cf. Density functional theory (B3LYP) and relativistic multireference (CASPT2) calculations confirmed the AnO 2 (NO 3 ) 2 - as AnVO 2 + actinyl moieties coordinated by nitrates. Computations of alternative An III O 2 (NO 3 ) 2 - and AnI V O 2 (NO 3 ) 2 - revealed significantly higher energies. Previous computations for bare AnO 2 + indicated AnVO 2 + for An = Pu, Am, Cf, and Bk, but Cm III O 2 + : electron donation from nitrate ligands has here stabilized the first Cm V complex, Cm V O 2 (NO 3 ) 2 - . Structural parameters and bonding analyses indicate increasing An-NO 3 bond covalency from Pu to Cf, in accordance with principles for actinide separations. Atomic ionization energies effectively predict relative stabilities of oxidation states; more reliable energies are needed for the actinides.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Fluoride-conversion synthesis of homogeneous actinide oxide solid solutions

Here, a novel route to synthesize (U, Th)O2 solid solutions at a relatively low temperature of 1100 C is demonstrated. First, the separate actinide oxides reacted with ammonium bifluoride to form ammonium actinide fluorides at room temperature. Subsequently, this mixture was converted to the actinide oxide solid solution using a two-phased heat treatment, first at 610 C in static air, then at 1100 C in flowing argon. Solid solutions obeying Vegard s Law were synthesized for ThO2 content from 10 to 90 wt%. Microscopy showed that the (U, Th)O2 solid solutions synthesized with this method to have considerably high crystallinity and homogeneity, suggesting the suitability of material thus synthesized for sintering into nuclear fuel pellets at low temperatures.

(U↗
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↗
DOE OSTI2016

Pursuing Cost Effective Tank Waste Characterization at the Savannah River Site - 16172

The U. S. Department of Energy Office of Environmental Management has tasked the Savannah River National Laboratory (SRNL) with developing strategies and technologies to understand, optimize, scale, and speed up tank waste characterization. This scope is part of a multi-year project with the end goal of implementing programmatic changes that accelerate tank waste processing and tank closure schedules, and at the same time reduce characterization costs, while maintaining data integrity. The project is currently in its second year. During the first year of the project, SRNL completed a series of activities focused on understanding the range of current characterization practices, needs, gaps, risks, and potential alternative approaches. Based on the results, conclusions were drawn regarding aspects of the programs that are most costly and time consuming; particular waste constituents that drive costs, schedule, and program risks; alternative characterization approaches (other than sampling and analysis) that are less expensive and more rapid; new laboratory methods holding the greatest promise for reducing characterization costs and schedule; and in whole, the primary areas where current characterization practices can be improved. SRNL's current scope is focusing on three primary activities: 1) development of a technical basis/strategy for improving the cost effectiveness and schedule of SRNL's tank closure characterization program; 2) design/assembly of hardware, plumbing, and software for automating select radiochemical separation and waste removal processes; and 3) development and testing of alternative radiochemical separation protocols most likely to improve high resource demand/time consuming analysis methods (such as Ra-226, Pa-231, Tc-99/I-129, and/or Y/trivalent actinide separations). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Prioritized List of Research Needs to support MRWFD Case Study Flowsheet Advancement

In FY-13, a case study evaluation was performed of full recycle technologies for both the processing of light-water reactor (LWR) used nuclear fuels as well as fast reactor (FR) fuel in the full recycle option. This effort focused on the identification of the case study processes and the initial preparation of material balance flowsheets for the identified technologies. In identifying the case study flowsheets, it was decided that two cases would be developed: one which identifies the flowsheet as currently developed and another near-term target flowsheet which identifies the flowsheet as envisioned within two years, pending the results of ongoing research. The case study focus is on homogeneous aqueous recycle of the U/TRU resulting from the processing of LWR fuel as feed for metal fuel fabrication. The metal fuel is utilized in a sodium-cooled fast reactor, and the used fast reactor fuel is processed using electrochemical separations. The recovered U/TRU from electrochemical separations is recycled to fuel fabrication and the fast reactor. Waste streams from the aqueous and electrochemical processing are treated and prepared for disposition. Off-gas from the separations and waste processing are also treated. As part of the FY-13 effort, preliminary process unknowns and research needs to advance the near-term target flowsheets were identified. In FY-14, these research needs were updated, expanded and prioritized. This report again updates the prioritized list of research needs based upon results to date in FY-15. The research needs are listed for each of the main portions of the flowsheet: 1) Aqueous headend, 2) Headend tritium pretreatment off-gas, 3) Aqueous U/Pu/Np recovery, 4) Aqueous TRU product solidification, 5) Aqueous actinide/lanthanide separation, 6) Aqueous off-gas treatment, 7) Aqueous HLW management, 8) Treatment of aqueous process wastes, 9) E-chem actinide separations, 10) E-chem off-gas, 11) E-chem HLW management. The identified research needs were prioritized within each of these areas. No effort was made to perform an overall prioritization. This information will be used by the MRWFD Campaign leadership in research planning for FY-16. Additionally, this information will be incorporated into the next version of the Case Study Report scheduled to be issued September 2015.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rescue of Unique Liquid IP-2 Portable Tanks at Savannah River Site – 25078

Westinghouse Savannah River Company, the SRS Management and Operations Contractor at the time, procured six large IP-2 portable tanks capable of transporting liquids in 2003 to support the layup of the F Canyon at SRS. F Canyon was one of two full-scale hardened radiochemical separations facilities built at SRS in the 1950s as part of the initial construction of the Site. F Canyon and H Canyon provided actinide separations capabilities for SRS through the end of the Cold War. Following the end of the Cold War era, the decision was eventually made that only one of these radiochemical separations facilities at SRS would still be needed moving forward. As H Canyon was considered better suited for ongoing SRS missions, F Canyon was placed into a layup state in the first decade of the 21st Century, with F Area deactivation completing in 2006. This process resulted in the creation of excess, radioactively contaminated solutions, which required disposition. These solutions included contaminated organic solutions as well as DU-bearing aqueous solutions. The six portable tanks were procured to support the transportation of these solutions.

Fee, Nathaniel C. [Savannah River Nuclear Solution↗
DOE OSTI2013

Reprocessing system with nuclide separation based on chromatography in hydrochloric acid solution

We have proposed the reprocessing system with nuclide separation processes based on the chromatographic technique in the hydrochloric acid solution system. Our proposed system consists of the dissolution process, the reprocessing process, the minor actinide separation process, and nuclide separation processes. In the reprocessing and separation processes, the pyridine resin is used as a main separation media. It was confirmed that the dissolution in the hydrochloric acid solution is easily achieved by the plasma voloxidation and by the addition of oxygen peroxide into the hydrochloric acid solution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Physics-based model for electro-chemical process

Considering the kinetics of electrochemical reactions and mass transfer at the surface and near-surface of the electrode, a physics-based separation model for separating actinides from fission products in an electro-refiner is developed. The model, taking into account the physical, chemical and electrochemical processes at the electrode surface, can be applied to study electrorefining kinetics. One of the methods used for validation has been to apply the developed model to the computation of the cyclic voltammetry process of PuCl{sub 3} and UCl{sub 3} at a solid electrode in molten KCl-LiCl. The computed results appear to be similar to experimental measures. The separation model can be applied to predict materials flows under normal and abnormal operation conditions. Parametric studies can be conducted based on the model to identify the most important factors that affect the electrorefining processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Automated Separation of Uranium and Plutonium from Environmental Swipe Samples for Multiple Collector Inductively Coupled Plasma Mass Spectrometry

A fully automated method for the separation of low-concentration uranium from plutonium in environmental swipe samples has been developed. The offline chromatography system features renewable 1 mL Eichrom TEVA and UTEVA column generation from bulk resin slurry. Discrete fractions of the separated actinides are delivered into user defined vials for future analysis. Clean room background levels were achieved outside of a cleanroom environment with this method. Purification of uranium and plutonium from various sample matrixes and at various concentrations was successful. Major and minor isotope ratios for both elements were measured via multiple collector inductively coupled plasma mass spectrometry and were in good agreement with certified reference values. Lastly, validation of the separation method was conducted on archived environmental samples and agreed with values previously reported using standard column chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Rapid separation and purification of uranium and plutonium from dilute-matrix samples

This work presents a streamlined separation and purification approach for trace uranium and plutonium from dilute (carrier-free) matrices. The method, effective for nanogram quantities of U and femtogram to picogram quantities of Pu, is ideally suited for environmental swipe samples that contain a small amount of collected bulk material. As such, it may be applicable for processing swipe samples such as those collected in IAEA inspection activities as well as swipes that are loaded with unknown analytes, such as those implemented in interlaboratory round-robin or proficiency tests. Additionally, the simplified actinide separation could find use in internal laboratory monitoring of clean room conditions prior to or following more extensive chemical processing. We describe key modifications to conventional techniques that result in a relatively rapid, cost-effective, and efficient U and Pu separation process. We demonstrate the efficacy of implementing anion exchange chromatography in a single column approach. We also show that hydrobromic acid is an effective substitute in lieu of hydroiodoic acid for eluting Pu. Lastly, we show that nitric acid is an effective digestion agent in lieu of perchloric acid and/or hydrofluoric acid. A step by step procedure of this process is detailed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Selectivity in ligand binding to uranyl compounds: A synthetic, structural, thermodynamic and computational study

The uranyl cation (UO₂²⁺) is the most abundant form of uranium on the planet. It is estimated that 4.5 billion tons of uranium in this form exist in sea water. The ability to bind and extract the uranyl cation from aqueous solution while separating it from other elements would provide a limitless source of nuclear fuel. A large body of research concerns the selective recognition and extraction of uranyl. A stable molecule, the cation has a linear O=U=O geometry. The short U-O bonds (1.78 Å) arise from the combination of uranium 5f/6d and oxygen 2p orbitals. Due to the oxygen moieties being multiply bonded, these sites were not thought to be basic enough for Lewis acidic coordination to be a viable approach to sequestration. The goal of this research is thus to broaden the coordination chemistry of the uranyl ion by studying new ligand systems via synthetic, structural, thermodynamic and computational methods. It is anticipated that this fundamental science will find use beyond actinide separation technologies in areas such as nuclear waste remediation and nuclear materials. The focus of this study is to synthesize uranyl complexes incorporating amidinate and guanidinate ligands. Both synthetic and computational methods are used to investigate novel equatorial ligand coordination and how this affects the basicity of the oxo ligands. Such an understanding will later apply to designing ligands incorporating functionalities that can bind uranyl both equatorially and axially for highly selective sequestration. Efficient and durable chromatography supports for lanthanide separation will be generated by (1) identifying robust peptoid-based ligands capable of binding different lanthanides with variable affinities, and (2) developing practical synthetic methods for the attachment of these ligands to Dowex ion exchange resins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Annual Report on the Activities and Publications of the DHS-DNDO-NTNFC-Sponsored Post-doctoral Fellow at Los Alamos National Laboratory

This report is a summary update on the projects Magen Coleman is working on as a postdoctoral fellow funded by DHS. These research projects are designed to explore different radioanalytical facets of nuclear forensics, including destructive and non-destructive analysis, chronometry measurements and modeling, spectral analysis, and actinide separations. Method development and instrumental analysis are integral parts of these projects, along with strong radiochemical knowledge. The current projects discussed here include method development for 232U analysis, the assembly of a gamma mapping system, the testing of neutron bubble detectors for neutron spectrometry, and continued work on the chronometry of plutonium and uranium materials. This report documents the work that has been performed since April 2011.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Testing of Oxidizing Chemicals and Sorbents in support of the WIPP Basis of Knowledge - 19617

Oxidizing chemicals are commonly found in waste generated from actinide separations and other chemical processes. They can generate and/or increase the combustion of combustible materials, as experienced in the 2014 radiological event at the WIPP. Consequently, a broad combination of oxidizer/sorbent mixtures have been experimentally evaluated using a modified approach to the EPA SW-846 Method 1040, in order to provide a highly reliable assessment of their potential hazards. Some results of this work are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The separation of lanthanides and actinides in supercritical fluid carbon dioxide

Supercritical fluid carbon dioxide presents an attractive alternative to conventional solvents for recovery of the actinides and lanthanides. Carbon dioxide is a good solvent for fluorine and phosphate-containing ligands, including the traditional tributylphosphate ligand used in process-scale uranium separations. Actinide and lanthanide oxides may even be directly dissolved in carbon dioxide containing the complexes formed between these ligands and mineral acids, obviating the need for large volumes of acids for leaching and dissolution, and the corresponding organic liquid–liquid solvent extraction solutions. As a result, examples of the application of this novel technology for actinide and lanthanide separations are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗