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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Hydroxypyridinone-based stabilization of Np(IV) enabling efficient U/Np/Pu separations in the Adapted PUREX process

The classical process to recover uranium (U) and plutonium (Pu) from used nuclear fuel using tributyl phosphate (TBP), namely the Plutonium Uranium Redox EXtraction (PUREX) process, is complicated by the persistent presence of neptunium (Np) and thus requires extra purification steps. The concept of Adapted PUREX seeks to achieve Np recovery by adjusting the valence of the metal more effectively, thereby controlling its behavior more precisely. This study introduces the use of an aqueous hydroxypyridinone chelator, 3,4,3-LI(1,2-HOPO) (abbreviated as HOPO), to dictate the behavior of Np for recovery and meanwhile simplify cumbersome reprocessing steps. The interactions between Np and HOPO were probed mechanistically by way of absorption spectrophotometry, in conjunction with cyclic voltammetry. UV–Vis-NIR spectra illustrated the reduction of NpO 2 2+ to Np 4+ , with a fast reaction rate. Cyclic voltammetry revealed quasi-reversible processes between the oxidized and reduced forms of the ligand and its Np complexes. The corresponding heterogeneous rate constants (k 0 ) were estimated from the peak-to-peak separation potentials (ΔE p ), at ~ 4 – 35 μm/s for both HOPO and NpHOPO, with scan rates of 0.01 – 0.4 V/s. Meanwhile, the electromotive force (E MF ) as well as the change of Gibbs free energy (ΔG) were assessed from the half-wave potential (E 1/2 ), demonstrating the completeness of NpO 2 2+ reduction to Np 4+ by HOPO. The cumulative formation constant of the resulting NpHOPO complex (logβ 101 ) was determined by metal competition titration to be 42.0 ± 0.6, corroborating the extraordinarily high affinity of HOPO to tetravalent metal ions. Here, the prowess of valence control by HOPO and the high stability of the formed complex resulted in enhanced separations of Np from U and of Pu from U, with a maximum separation factor of ~7000 for both, nearly 90- and 10300-fold higher, respectively, than the values obtained using conventional PUREX formulae.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Aqueous Reprocessing of Used Nuclear Fuel

Aqueous reprocessing of used nuclear fuel has a long history of successful operation worldwide. The history of aqueous reprocessing and a description of current and past operating aqueous reprocessing facilities are provided and the main types of solvent extraction equipment utilized in the nuclear industry are described. While the PUREX process is well established for separating U and Pu from used nuclear fuel, numerous variants to the PUREX process are being developed and are described. Finally, advanced separation processes in which minor actinides are separated from the used nuclear fuel are a key research focus worldwide. These processes and the current state of development are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Organic Separation Test Results

Separable organics have been defined as “those organic compounds of very limited solubility in the bulk waste and that can form a separate liquid phase or layer” (Smalley and Nguyen 2013), and result from three main solvent extraction processes: U Plant Uranium Recovery Process, B Plant Waste Fractionation Process, and Plutonium Uranium Extraction (PUREX) Process. The primary organic solvents associated with tank solids are TBP, D2EHPA, and NPH. There is concern that, while this organic material is bound to the sludge particles as it is stored in the tanks, waste feed delivery activities, specifically transfer pump and mixer pump operations, could cause the organics to form a separated layer in the tank farms feed tank. Therefore, Washington River Protection Solutions (WRPS) is experimentally evaluating the potential of organic solvents separating from the tank solids (sludge) during waste feed delivery activities, specifically the waste mixing and transfer processes. Given the Hanford Tank Waste Treatment and Immobilization Plant (WTP) waste acceptance criteria per the Waste Feed Acceptance Criteria document (24590-WTP-RPT-MGT-11-014) that there is to be “no visible layer” of separable organics in the waste feed, this would result in the batch being unacceptable to transfer to WTP. This study is of particular importance to WRPS because of these WTP requirements.

complexants↗
DOE OSTI2016

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cast storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI), a highly oxidizing system is required, which has only been demonstrated on a limited basis. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cask storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather, the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI) which has only been demonstrated on a limited basis, a highly oxidizing system is required. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Modulation of Tetravalent Actinide Oxalate Morphology through Organic Compounds

Spent nuclear fuel of pressurized water reactors mainly consists of uranium, plutonium, minor actinides and fission products. The long term radiotoxicity of wastes is mainly due to plutonium and minor actinides such as neptunium, americium and curium. So, in order both to reduce the radiotoxicity of the ultimate wastes and to save uranium resources, researches for current and new generations of nuclear plants focus on the recycling of plutonium and minor actinides, to fabricate new mixed oxides nuclear fuel pellets (MOX) or transmutation targets. In this aim, production of actinide oxides is necessary. Currently, according to the Purex process, the spent fuel is dissolved in nitric acid and then valuable actinides (uranium and plutonium) are separated and purified from fission products and minor actinides in the Areva plant 'La Hague'. The resulting plutonium nitric solution is converted into a solid phase by oxalic conversion: the oxalic precipitation is followed by the calcination of the obtained oxalate into plutonium oxide. Considering either the MOX fuel or the transmutation targets fabrication from actinide oxides, the control of the particle structure and microstructure is a key factor related to the ability of oxide powder to sinter and shape as pellets. In the oxalic conversion process, the morphology of the precipitated oxalate is strongly influenced by the precipitation conditions and is kept during the calcination from oxalate into oxide. In this context, the aim of this work is to control the morphology of the final actinide oxide by changing the oxalic precipitation conditions. A lot of possibilities can be considered to modulate the morphology of actinide oxalates. Numerous parameters such as temperature, pressure, the presence of a magnetic field, the operating process (seeding, emulsions)... can modify the precipitated particles shape. Additives are also able to play a major role at the precipitation stage by means of different mechanisms involving either complexation, surface covering or structure insertion. This study focuses on this way to modify the morphology of actinides oxalates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

TC-99 Decontaminant from heat treated gaseous diffusion membrane -Phase I, Part B

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 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 gaseous 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 or leaching desorption, which is independent of the technetium oxidation states, to perform an insitu removal of the technetium as a volatile species and trap the radionuclide on sorbent traps which could be disposed as low-level waste. Based on the positive results of the first part of this work1 the use of steam as a thermal decontamination agent was further explored with a second piece of used barrier material from a different location. This new series of tests included exposing more of the material surface to the flow of high temperature steam through the change in the reactor design, subjecting it to alternating periods of stream and vacuum, as well as determining if a lower temperature steam, i.e., 121°C (250°F) would be effective, too. Along with these methods, one other simpler method involving the leaching of the Tc-99 contaminated barrier material with a 1.0 M aqueous solution of ammonium carbonate, with and without sonication, was evaluated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Probing air-water interfaces of dibutyl phosphoric acid (HDBP) aqueous solutions using vibrational sum frequency generation (vSFG) spectroscopy

Liquid-liquid extraction is a separation technique implemented in a wide variety of areas, achieving particular success in both the nuclear and biomedical fields. In this work, vibrational sum frequency generation spectroscopy (VSFG) and surface tension measurements were used to investigate the adsorption of dibutyl phosphate (DBP) at air-aqueous interfaces to simulate liquid-liquid systems relevant to the Plutonium Uranium Redox Extraction (PUREX) Process. The objective of this work is to establish qualitative relationships between changes in the bulk aqueous phase concentrations of DBP and its concentration and structure at air-liquid interface as probed with VSFG. Nitric acid concentration and solution ionic strength were varied to examine their effect on the interfacial DBP.. Introduction of DBP into neat water resulted in reduction of the VSFG spectral intensity in the dangling O-H region (3680 – 3800 cm -1 ) but large increase in the H-bonded O-H stretch frequency region (3000 – 3500 cm-1) and the appearance of the CH 3 symmetric stretch and CH 3 Fermi resonance peaks at ~ 2880 and 2945 cm -1 , respectively, indicating DBP at the air-water interface. The intensity of the C-H strecth peaks increased as DBP concentration increased from 0.24 to 32 mM, accompanied by a decreasing surface tension values. At fixed DBP concentration, the addition of either or both of HNO 3 and NaNO 3 to an ionic strength of 1 M or 3 M led to significant reduction of the O-H VSFG peaks and enhancement of the C-H peaks. The origins of these experimental observations are attributed to both the increased HDBP molecules partitioning and adsorption to the interface and the protonation of the interfacial DBP- molecules.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Management of HL deposits for the D and D of UP1 reprocessing plant - retrieval and treatment scenarios - 15295

Built in 1958 on the Marcoule CEA site, the UP1 facility, was the 1. French industrial (500 t U / year) reprocessing plant based on the PUREX process. The plant, which has stopped operation in 1997, constitutes with its auxiliary workshop (vitrification workshop, glass canister storage), one of the largest and most challenging decommissioning project in France. In about 40 years of operation, various type of fuels elements such as metallic uranium, U-Al fuel rods, Phenix breeder elements or PWR fuel have been reprocessed, thereby requiring several plant and process adaptations. Resulting from these complex process chemistry and operation history, significant fouling of the chemical plant circuits and components has occurred. The fission product (FP) concentration process is particularly concerned. Preliminary investigation performed in 2010 in the evaporator's feeding tanks revealed the presence of significant amounts of sludge and dose rates as high as 250 Gy/h. Therefore a thorough investigation campaign, including visual inspection, volume quantification, deposits sampling, and radiological characterization of the components of this workshop has been initiated in 2013. Chemical analysis performed so far show that the solids are mainly composed of zirconium phosphates, 'MoZr' compounds and nitrate salt precipitates. But the presence of insoluble fines arising from the dissolution workshop and organic residues resulting from the TBP decomposition has also been evidenced in some of the tanks. The numerous constraints associated with the management of such highly active sludge, with Cs-137 specific activity ranging from 1 to 10 GBq/g, and a complex physico-chemical composition, lead CEA to investigate the implementation of an in situ conditioning treatment able to produce a waste package complying with existing evacuation routes and/or on-site storage facilities. The full paper presents an overview of the UP1 D and D project and then focuses on the possible dismantling scenarios for the FP concentration workshop, including the evaluation of different strategies for the highly active deposits management. The investigations and deposits characterization program will be developed so as the sampling methodologies that has been specifically tuned. In the light of the best actual knowledge of the deposits properties, the recovery and conditioning treatment scenarios will be exposed, considering, as a case study, the development of a specific vitrification tool to be implemented In Situ and specifically designed to match the requirements and constraints of the D and D project: adequately dimensioned, compact, flexible with respect to the feed composition, for a limited duration of use and designed to be dismantled. (authors)

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

Solvent Exchange

This report summarizes the work of evaluating the chemical stability of phosphoramidic acid, N-N-diethyl-, bis(2-ethylhexyl) ester under the Solvent Exchange LDRD project, as a potential solvent exchange chemistry and analog to solvent exchange chemistries used in the TALSPEAK and ALSEP processes as alternatives to the use of tri-butyl phosphate in the PUREX process.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Fault Detection in the Solvent Extraction Process with Non-traditional Sensors

Solvent extraction is used to separate metals or other complexes into two different immiscible liquids and is an essential component of the PUREX Process. Improvements to the solvent extraction process can directly contribute to an organization’s ability to ensure the purity and recovery of special nuclear materials from spent nuclear fuel. Such improvements can benefit nuclear reprocessing efforts, increase fuel reutilization, and limit the concentration of actinides in nuclear waste repositories.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Zurückgezogen: Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗
DOE OSTI2016

Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗
DOE OSTI2016

Spent Fuel Reprocessing Development and Usage of Fission Products as Starting Precursor for Ceramic Waste-Forms - 16498

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile isotopes and waste preparing for disposal. Reprocessing is a series of chemical operations that separates plutonium and uranium from other constituents in nuclear waste. Indeed, the plutonium separated can be used to fuel reactors, but also to make nuclear weapons. Since reprocessing operations are far more expensive than using uranium fuel and disposing of the spent fuel directly, this process is not readily considered. Without doubts the reprocessing price of $1000 per kilogram of heavy metal (kgHM), like uranium must be less by at least 2-3 times compared to directly deposing the spent fuel. And some more simple technology (in comparison with PUREX process) must be developed, for example, technology based on precipitation processes. One of the options to change reprocessing technology is the idea to use certain fission products occurring in spent fuel in considerable amount and capable to form insoluble substances as suitable materials for long-term storage or final disposal of radioactive wastes. The development of new materials for HLW immobilization are intended to reduce disposal cost. The present work was aimed at the search for the simple engineering decisions that would provide opportunity of fission products usage as starting precursor for ceramic waste-forms. The results of experiments on synthesis of Pd-based metallo-ceramic compositions are presented in this paper. (authors)

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

Microfluidic In-Situ Spectrophotometric Approaches to Tackle Actinides Analysis in Multiple Oxidation States

The study and development of present and future processes for the treatment/recycling of spent nuclear fuels require many steps, from design in the laboratory to setting up on an industrial scale. In all of these steps, analysis and instrumentation are key points. For scientific reasons (small-scale studies, control of phenomena, etc.) but also with regard to minimizing costs, risks, and waste, such developments are increasingly carried out on milli- or microfluidic devices. The logic is the same for the chemical analyses associated with their follow-up and interpretation. Due to this, over the last few years, opto–microfluidic analysis devices adapted to the monitoring of different processes (dissolution, liquid–liquid extraction, precipitation, etc.) have been increasingly designed and developed. In this work, we prove that photonic lab-on-a-chip (PhLoC) technology is fully suitable for all actinides concentration monitoring along the plutonium uranium refining extraction (plutonium, uranium, reduction, extraction, or Purex) process. Several PhLoC microfluidic platforms were specifically designed and used in different nuclear research and development (R&D) laboratories, to tackle actinides analysis in multiple oxidation states even in mixtures. The detection limits reached (tens of µmol·L −1 ) are fully compliant with on-line process monitoring, whereas a range of analyzable concentrations of three orders of magnitude can be covered with less than 150 µL of analyte. Finally, this work confirms the possibility and the potential of coupling Raman and ultraviolet–visible (UV–Vis) spectroscopies at the microfluidic scale, opening the perspective of measuring very complex mixtures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2024

Batch Extraction Studies to Evaluate Trace Element Behavior in PUREX Conditions

The multilab Intentional Forensics Venture is working to identify which stable elements (i.e., taggants) at trace concentrations relative to U would persist throughout the nuclear fuel cycle in a voluntary fuel tagging scheme. A taggant would provide the nuclear forensics community with a “barcode” to help identify nuclear materials found outside of regulatory control. A portion of this project was focused on reprocessing effects and determining which, if any, elements would coextract with U(VI) in standard Pu–U reduction extraction (PUREX) conditions. Elements with a propensity to coextract could, in theory, be used as taggants from a PUREX perspective. Although retention is not a performance requirement, the taggant signature would need to partition predictably from the U stream after the PUREX process to maintain forensic utility. This report documents results from several batch extraction studies with numerous trace elements from HNO 3 (1.5–5 M), with and without U(VI), into 30% tri-n-butyl phosphate (TBP) in kerosene. Extraction and back-extraction tests were used to evaluate nearly 60 elements in surrogate conditions for PUREX, and distribution coefficients (i.e., D-values) for most species were <0.1, indicating few species are likely to co-extract with U through PUREX. Additional studies are needed to optimize sample volumes and dilutions to dial in these low D-values. The D-values (D) were determined for several of the more promising elements, including Re and Se. Ultimately, we conclude that only a limited number of the ~ 60 elements investigated are extractable in the U stream of PUREX, based on measured D values, meaning most candidate elemental taggants would likely be lost at this stage of the nuclear fuel cycle, even when considering a range of acid concentrations.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗