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Solid Sorbent Cost Sensitivity Analysis: A Framework for UNF Reprocessing Sorbent Cost Comparison

Solid sorbents have been the subject of research and development across the U.S. Department of Energy national laboratory complex for many years. They are generally accepted as a safer alternative to cryogenic distillation for noble gas capture, and they present an easier pathway in development of long-term waste forms after iodine capture. As more types of sorbents have been proposed for capture of volatile radionuclides, it has become necessary to compare them based on performance and cost criteria. This report details cost and performance information for three promising sorbents and provides a cost sensitivity analysis. The goal of this analysis is to establish a framework which can be utilized to directly compare future sorbents, with differing properties, to the sorbents discussed in this report. Direct comparison is instrumental to making informed decisions to efficiently guide research and minimize laborious detours. In all cases, sorbent capacity is a major cost driver as it influences the mass of sorbent, operational footprint, and disposal cost requirements., However, sorbent price can greatly increase the cost of a capture technology. For sorbents used in krypton capture, the purity of krypton released to storage is the highest cost driver.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Impact of f-element complexation on the radiolytic robustness of separations ligands

Impact of f-element complexation on the radiolytic robustness of separations ligands Gregory P. Horne, Makayla R. Baxter, Corey D. Pilgrim, Travis S. Grimes, Center for Radiation Chemistry Research, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415, USA Cristian Celis Barros, Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL 32306, USA. E-mail: ccelisbarros@fsu.edu Andrew R. Cook, Department of Chemistry, Brookhaven National Laboratory, Upton, New York, 11973, USA Stephen P. Mezyk, Department of Chemistry and Biochemistry, California State University Long Beach, 1250 Bellflower Boulevard, Long Beach, California, 90840-9507, USA The effects of ionizing radiation are ubiquitous throughout all aspects of a nuclear fuel cycle. However, the complexity and intensity of these effects are greatest during the management of used nuclear fuel, owing to the presence of a wide spectrum of radionuclides from neutron capture and fission processes. With regards to used nuclear fuel (UNF) reprocessing, radiation-induced processes typically promote the destruction of active compounds (e.g., complexants and additives) with the concomitant formation of potentially detrimental degradation products and corresponding changes in physical and chemical properties, which ultimately impact the effectiveness and longevity of a given reprocessing system. Concerning UNF complexants, radiation chemistry studies have historically focused on their radiation robustness in the absence of the metal ions they were designed to selectively complex. This knowledge gap is worrisome as previous studies on aqueous phase complexants have demonstrated significant changes in radiolytic behavior upon metal ion complexation.1-4 More recently, the rate of reaction of the n-dodecane radical cation—believed to be the major organic phase radiation-induced transient species responsible for complexant radiolysis in n-dodecane based solvent systems—with hexa-n-octylnitrilo-triacetamide (HONTA) was shown to increase by an order-of-magnitude upon complexation of europium or americium.5 These findings have significant implications on the projected longevity of complexants in UNF reprocessing solvent systems. Consequently, a thorough understanding of metal ion complexation effects on the radiolytic integrity of UNF complexants is essential to evaluate their potential for process application. Presented here are two recent studies from the Idaho National Laboratory Center for Radiation Chemistry Research group that demonstrate the various impacts of f-element complexation (uranium, americium, and lanthanides) on the radiolytic integrity (gamma and electron pulse) of tributyl phosphate (TBP), N,N-di-(2-ethylhexyl)butyramide (DEHBA), N,N-di-2-ethylhexylisobutryamide (DEHiBA), and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) under UNF reprocessing conditions. References 1) Bhattacharyya and Kundu, Int. J. Radiat. Phys. Chem., 1971, 3, 1. 2) Kundu and Matuura, Int. J. Radiat. Phys. Chem., 1975, 7, 565. 3) Ilan and Czapski, Biochimica et Biophysica Acta, 1977, 498, 386. 4) Buettner, Doherty, and Patterson, Fed. Euro. Biochem. Soc., 1983, 158 (1), 143. 5) Toigawa, Peterman, Meeker, Grimes, Zalupski, Mezyk, Cook, Yamashita, Kumagai, Matsumura, Horne, PCCP, 2021, 23, 1343.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Electrochemical Manipulation and Radiolytic Evaluation of Organic Phase Neptunium

Presentation for Seed LDRD proposal presentation: This Seed LDRD research proposes to test an innovative approach to precisely control the oxidation state distribution of neptunium (Np) under used nuclear fuel (UNF) reprocessing conditions using novel electrochemical methods. Current UNF reprocessing technologies are not financially viable in the US, requiring significant scientific and technological innovation to improve cost efficiency. Reducing the number of process cycles is one avenue for increasing cost efficiency, of which resolving the challenge associated with isolating Np is critical. Under envisioned process conditions, Np is present in a mixture of extractable (Np(IV) and Np(VI)) and inextractable (Np(V)) oxidation states, the distribution of which is dependent on several factors, including ionizing radiation dose, that change throughout a reprocessing scheme. Consequently, Np unintentionally partitions into various UNF reprocessing phases and product streams, ultimately increasing the number of process cycles to isolate UNF components. Here, we propose to employ novel, high surface area, optically transparent, Ligand Modified tin-doped indium oxide Electrodes (LMES). These LMEs bind Np, and therefore enable concurrent generation and spectroscopic characterization of Np oxidation states in organic solutions. These proof-of-concept experiments will: (i) facilitate optimization of the proposed electrochemical system conditions for Np oxidation state manipulation; (ii) identify the accessible electrochemical window for Np complexes in the organic phase; (iii) report characteristic optical spectra for each accessible complexed Np oxidation state; and (iv) determine the lifetime and partitioning of a given atypical Np oxidation state in the organic phase in the presence and absence of ionizing radiation fields. The data gathered by this Seed LDRD will provide support for the design of an electrochemical process concept for the precise manipulation of Np oxidation states in UNF reprocessing solvent systems, with the intention of providing advanced control over Np mass transfer, and thus greater process efficiency and economy.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Generation and Study of Am(IV) by Temperature-Controlled Electron Pulse Radiolysis

Used nuclear fuel (UNF) separation techniques that strive to separate radiotoxic americium (Am) from trivalent lanthanide fission products through oxidation state control have increased research efforts surrounding Am(V) and Am(VI). However, equivalent knowledge of the tetravalent state, Am(IV), has remained elusive, particularly in conditions more representative of UNF reprocessing, i.e., in concentrated nitric acid (HNO3). With this in mind, we have used electron pulse radiolysis to study the radiation-induced redox reaction of Am(III) with the oxidizing nitrate radical (NO3?) in 6 M HNO3: Am(III) + NO3? ? Am(IV) + NO3? . These experiments enabled us to observe the growth and decay of Am(IV) in a concentrated acidic solution for the first time. The transient Am(IV) species was found to have a lifetime of ~16 µs?sufficiently long-lived to play a critical mechanistic role in UNF reprocessing systems. Additionally, we performed the first-ever temperature-dependent kinetics study of an actinide element, elucidating unprecedented Arrhenius and Eyring activation parameters for the reaction of Am(III) with NO3?. This new knowledge provides much-needed molecular-level insights into the radiation-induced behavior of Am.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Radiation Effects in Next Generation Used Nuclear Fuel Reprocessing Strategies

With the global community committed to significantly expanding nuclear energy capacity, the development of efficient used nuclear fuel (UNF) management strategies has become more critical than ever. These strategies are vital to fostering the widespread adoption of closed fuel cycles, which are essential for sustainable nuclear energy production and security. Achieving this ambitious goal necessitates a comprehensive understanding of radiation effects on next-generation technologies, as radiolysis can often limit the longevity and performance of these systems. This seminar will provide an overview of next-generation UNF reprocessing strategies, highlighting the latest advancements and innovative approaches in the field. Particular attention will be given to two key areas of recent research: 1. Radiation robustness and performance of advanced sulfur chloride-based chlorination technologies. We will explore the efficacy of sulfur chloride-based chlorination processes in the presence of surrogate cladding materials, specifically aluminum. These processes have shown promise in the dissolution, decontamination, and recovery of cladding materials for reuse. Detailed findings on how the composition and performance of these sulfur chloride solvents respond to radiation exposure will be discussed. 2. Impacts of metal ion complexation and direct dissolution conditions on monoamide-based reprocessing strategies. We will delve into the time-resolved and dose accumulation effects of irradiation on the direct dissolution of voloxidized uranium and rhenium using N,N-di-(2-ethylhexyl) butyramide (DEHBA) or N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) in pre-equilibrated n-dodecane solvent. The implications of these interactions on dissolution efficiency, radiolytic stability, and overall process performance will be examined. These studies aim to underscore the importance of understanding radiation effects in the development of next-generation UNF reprocessing technologies and the global transition towards more sustainable and efficient nuclear energy systems.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗

Investigating Radiation-Induced Actinide Species in Solution

Used nuclear fuel (UNF) separation techniques that strive to separate radiotoxic americium (Am) from trivalent lanthanide fission products through oxidation state control have increased research efforts surrounding Am(V) and Am(VI). However, equivalent knowledge of the tetravalent state, Am(IV), has remained elusive, particularly in conditions more representative of UNF reprocessing, i.e., in concentrated nitric acid (HNO3). With this in mind, we have used electron pulse radiolysis to study the radiation-induced redox reaction of Am(III) with the oxidizing nitrate radical (NO3?) in 6 M HNO3: Am(III) + NO3? ? Am(IV) + NO3? . These experiments enabled us to observe the growth and decay of Am(IV) in a concentrated acidic solution for the first time. The transient Am(IV) species was found to have a lifetime of ~16 µs?sufficiently long-lived to play a critical mechanistic role in UNF reprocessing systems. Additionally, we performed the first-ever temperature-dependent kinetics study of an actinide element, elucidating unprecedented Arrhenius and Eyring activation parameters for the reaction of Am(III) with NO3?. This new knowledge provides much-needed molecular-level insights into the radiation-induced behavior of Am.

actinide↗

Capture of volatile RuO 4 from oxidized simulated used nuclear fuel solutions

Ruthenium is a challenging fission product in used nuclear fuel (UNF) reprocessing due to its complex redox chemistry, variable speciation in nitric acid, partial extractability, and volatility. This work presents a strategy for Ru removal based on the volatilization of RuO 4 using the strong oxidant sodium bismuthate, followed by RuO 2 deposition onto various substrates. Among the materials tested, polymer-based substrates such as polyolefin wax film (Parafilm®), exhibited superior performance, achieving up to near-quantitative Ru removal from solution. After dissolution of the substrate, 99.6% of the Ru was recovered as RuO 2 . The deposition mechanism onto the polyolefin wax film involves both physisorption and chemical reaction through oxidative cleavage of olefinic bonds within the polymer. In contrast, on inorganic substrates such as Al foil, RuO 4 undergoes incomplete surface adsorption and reduction, as confirmed by FTIR and XPS analyses. This approach remains effective under conditions of simulated oxidation of UNF solutions, selectively removing Ru while leaving other fission products in solution. As a result, capture of RuO 4 offers a practical and efficient strategy for ruthenium decontamination and recovery, offering a route for integration into existing UNF reprocessing flowsheets to enhance overall process safety and performance.

Ruthenium↗

Radiolytic degradation of n -dodecane: Insights from time resolved pulsed electron radiolysis using aromatic scavengers

This study uses time-resolved picosecond pulsed electron radiolysis and transient absorbance spectroscopy to investigate early radiolytic processes in n-dodecane, a prototypical organic solvent for reprocessing used nuclear fuel (UNF). The primary radical cation, solvated electron, and excited states in neat n-dodecane are formed and consumed on timescales that are difficult to observe directly. Aromatic solutes, biphenyl and benzophenone, were utilized as scavengers to convert these short-lived species into longer-lived, optically detectable transient products to serve as indirect probes of the fraction of the initial n-dodecane radiolytic products that can be intercepted before undergoing ultrafast decay or competing reactions. The transient absorbance spectra of these scavengers and their radiolytic products were analyzed. Results indicate that while biphenyl and benzophenone effectively capture several radiolytic species, they do not fully scavenge all n-dodecane excited states, suggesting potential inefficiencies or additional side reactions. This study provides insight into the formation and accessibility of radiation-induced excited states and charge carriers in n-dodecane, which play a key role in the degradation of solvent extraction systems used in UNF reprocessing.

58 GEOSCIENCES↗

Concepts for Actinide Recovery from TRISO Used Nuclear Fuel

The work described in this report has developed a preliminary conceptual flowsheet based on limited published literature for the head-end of a TRISO oxycarbide UNF reprocessing plant. The essential function of the head-end is to prepare the UNF for subsequent separation of actinides. The main steps of the conceptual TRISO UNF head-end reprocessing flowsheet are: 1. Fragmentation of the graphite moderator to expose the fuel particles. 2. Separation of fuel particles from fragmented graphite moderator by sieving and fluidization. 3. Fragmentation of the fuel particles to expose the fuel kernel. 4. Dissolution of the fuel kernel in nitric acid. Noteworthy at this step is the formation of organic acids (mainly oxalic and mellitic acids) from the carbon associated with the fuel and TRISO coatings. These acids can interfere with actinide separations and solvent extraction hydraulic performance and carbon dioxide gas generation. 5. Clarification of the dissolved fuel solution to separate the coating fragments. 6. Dissolver off-gas treatment with specific emphasis on managing generation of carbon dioxide gas containing carbon-14. The quantity of carbon dioxide is projected to be 100 to 1000 times greater than that arising from dissolving LWR UNF. Pre or post treatment of the exposed fuel kernels could be undertaken to mitigate the formation of organic acids but there still exists a significant off-gas treatment challenge. The literature on reprocessing uranium nitride UNF was reviewed but a flowsheet was not specifically developed since its scope and uncertainties are expected to be encompassed by the oxycarbide flowsheet. Overall, the predominant feature of processing TRISO UNF is carbon management to reduce waste volume and mitigate the formation of carbon dioxide and organic acids. This report recommends broad areas of research that will help to further define the head-end flowsheet.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Investigating the Impacts of Direct Dissolution Conditions on the Radiolytic Longevity of Butyramide Extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Investigating the impacts of used nuclear fuel direct dissolution on the radiolytic longevity of solvent and butyramide extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Influence of Metal Ion Complexation on the Radiolytic Longevity of Butyramide Extractants under Direct Dissolution Process Conditions

The direct dissolution of voloxidized used nuclear fuel (UNF) into an organic solution–comprised of diluent and specialized extractants–poses a promising alternative to the traditional liquid–liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in conventional extraction flowsheets. Given the limited knowledge of radiation effects under direct dissolution conditions, here we present a time-resolved and dose-accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants, N,N-di(2-ethylhexyl) butyramide (DEHBA) and N,N-di(2-ethylhexyl)isobutyramide (DEHiBA), in pre-equilibrated n-dodecane solvent in the presence and absence of process-relevant metal ions, specifically, uranium and rhenium. Loss G(DEHBA) and G(DEHiBA) values were found to be comparable to each other, with an average of 0.37 ± 0.02 μmol J –1 , and to previous data from the γ irradiation of DEHBA and DEHiBA under conventional solvent extraction conditions. Rhenium, and by extension technetium, extraction had a modest decrease (~10%) in the overall radiolytic stability of DEHiBA only, despite >2× observed increases in chemical kinetic reactivity of the corresponding complexes with the n-dodecane radical cation. Uranium loading, on the other hand, significantly improved the lifetime of both ligands (>30%) under γ irradiation, with a greater stabilization observed for DEHBA over DEHiBA. The observed radioprotective effect afforded by uranium loading is fortuitous for the longevity of direct dissolution solvent.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Influence of metal ion complexation on the radiolytic longevity of butyramide extractants under direct dissolution conditions

The direct dissolution of volox-treated used nuclear fuel (UNF) into an organic solution—comprised of diluent and specialized extractants—poses a promising alternative to the traditional liquid-liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in liquid-liquid solvent extraction flowsheets. With this in mind, and given the limited knowledge of radiation effects under direct dissolution conditions, we present a time-resolved and dose accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants—N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA)—in pre-equilibrated n-dodecane solvent in the presence and absence of process relevant metal ions, uranium and rhenium. Rhenium, and by extension technetium, extraction had little impact (=10%) on the overall radiolytic stability of these ligands, despite observed increases in chemical kinetic reactivity (>2×) of the corresponding complexes with the n-dodecane radical cation. Uranium-loading on the other hand, significantly improved the lifetime of both ligands (>30%) under gamma irradiation, with a greater stabilization observed for DEHBA over DEHiBA. This draft manuscript has been prepared in fulfillment of NTRD-MRWFD-2024 M3FT-24IN030101115.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Krypton Concentration using HZ-PAN

Idaho National Laboratory (INL) has developed and tested engineered sorbents to separate and capture volatile fission products such iodine, xenon (Xe), and krypton (Kr) from off-gas streams. As noble gases, Xe and Kr can be difficult to capture and separate. Historically, cryogenic distillation has been used to execute the separation. However, performing this separation from bulk air streams is expensive and can pose significant hazards. INL has successfully developed two different sorbents for Xe and Kr capture, silver mordenite polyacrylonitrile and hydrogen mordenite polyacrylonitrile (HZ-PAN). Adsorption studies to date successfully separated and captured Kr from carrier gas streams, but those studies focused primarily on initial separation and capture. Successful utilization in a used nuclear fuel (UNF) reprocessing facility, however, requires further concentration of the Kr to minimize long term storage volumes. This study focuses on the concentration of Kr utilizing HZ PAN as the concentrating media. Concentrating Kr during desorption is ideal for two reasons: one, to produce Kr that can be used for commercial and research applications, or two, to minimize the volume of the radioactive gaseous waste stream for disposal. This report investigates the Kr concentrating potential of HZ-PAN by running multiple Kr adsorption-desorption cycles, collecting the desorbed effluent from a saturated HZ-PAN column, and loading it onto a fresh column. This study demonstrates that four adsorption-desorption cycles can transform a 150 ppmv Kr stream into a 37.5% Kr stream, an overall concentration factor of 2497. Each concentration step results in successively smaller volumes desorbed, reducing volume by a factor of approximately 40,000. Depending on the operational goals, during desorption one could collect smaller volume fractions of high concentration Kr (~ 67%). This demonstration should be considered proof-of-concept. Further refinement is necessary to develop optimum operating schemes to integrate into UNF off-gas treatment.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

Alternative Materials Literature Review: A Review of Alternative Materials for Xenon Capture in Comparison to AgZ-PAN

A collaborative literature review between Idaho National Laboratory (INL) and Pacific National Laboratory (PNNL) was conducted to research alternative metals/materials for xenon adsorption in gaseous effluents generated from used nuclear fuel (UNF) reprocessing. The review was conducted to research sorbents that could outperform silver mordenite (AgZ) at ambient temperature. The use of silver requires complex and expensive waste disposal processes due to it being an RCRA metal, therefore alternative materials need to be researched and evaluated. The report details two different types of sorbent material: alternative zeolite materials and non-zeolite-based materials (i.e., MOFs, COFs, etc.). This report documents a brief history of the materials, what testing has been conducted on them, and what results suggest the possibility of outperforming the silver-based materials. Sorbent performance characteristics such as durability, capacity, and operating temperatures were evaluated. The simplicity of the sorbents was also evaluated, meaning how they are acquired (synthesized vs. naturally abundant). Recommendations were made for the most promising sorbents based on available research in the subject area and these recommended sorbents will be tested in the upcoming fiscal year. The recommended sorbents that are zeolite based are CHA zeolites, ZIF-69, and Na-ETS-10. The recommended sorbents that are non-zeolite based are SBMOF-1 and CC3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Group hexavalent actinide separation from lanthanides using sodium bismuthate chromatography

Advanced used nuclear fuel (UNF) reprocessing strategies are limited by the complex radiochemical separations and engineering required to achieve the separation of actinides (An) from neutron scavenging lanthanides (Ln). The accessibility of the hexavalent oxidation state for the actinides (U – Am) provides a pathway to achieving a group hexavalent actinide separation from the trivalent lanthanides and Cm. The solid oxidant and ion exchanger, sodium bismuthate (NaBiO 3 ), has been demonstrated to quantitatively oxidize and separate Am from trivalent Cm in a column chromatographic system. This work expands on the use of NaBiO 3 chromatography to characterize the adsorption, kinetic, and elution behavior of U, Pu, and Eu. Separation factors over 200 with rapid kinetics were observed at dilute nitric acid concentrations with a complete An/Ln separation achieved in under an hour. In conclusion, the adsorption and chromatographic behavior of key fission products present in various reprocessing raffinates was characterized which demonstrated potential application of a NaBiO 3 -based separation following a TRUEX process.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Extraction, Characterization, and Stability Studies of Bistriazinyl-Derived Carboxylic Acids

Separation of An(III) and Ln(III) ions will benefit the recycling of used nuclear fuel (UNF). For this purpose, many ligands have been tested over the years, and several separation processes have been successfully demonstrated on the laboratory scale. Current research aims at the development of new ligands that are built only with carbon, hydrogen, oxygen, and nitrogen (CHON), as they can be incinerated completely without secondary waste production. Here, we tested a new class of water-soluble ligands, the bistriazinyl-octa-carboxylic acids. One member, in particular, 2,6-bis-[5,6-di(3,4-dicarboxyphenyl)-1,2,4- triazin-3-yl]-pyridine (BTPOA), was found to be suitable for the selective separation of Am(III) and Cm(III) ions from Ln(III) ions and may act as a CHON alternative to its sulfonated analogue (SO 3 -Ph-BTP). BTPOA exhibited good extraction results and a high selectivity for Am(III) over Eu(III) ions. This ligand’s complexation of metal ions was further studied using potentiometric spectroscopy, as well as time-resolved laser-induced spectroscopy with Cm(III) and Eu(III) in aqueous HClO 4 and HNO 3 media. Conditional stability constants of each formed species were determined. In the HClO 4 system, Cm(III) formed three species (1:1, 1:2, and 1:3) through the stepwise addition of a single BTPOA molecule. On the other hand, in HNO 3 , Cm(III) formed two 1:2 complexes and one 1:3 complex, while the stepwise formation of three species was observed for Eu(III). The stability constants are comparable to the values for SO 3 -Ph-BTP. The radiolytic behavior of BTPOA was also investigated using electron pulse irradiation measurements to determine absolute rate coefficients (k) under ambient temperature conditions for the reaction of BTPOA with typical UNF reprocessing radical radiolysis products the hydrated electron (e aq − , k = (1.60 ± 0.02) × 10 10 M −1 s −1 ), the hydrogen atom (H • , k = (2.17 ± 0.03) × 10 9 M −1 s −1 ), and hydroxyl ( • OH, k = (6.95 ± 0.06) × 10 9 M −1 s −1 ) and nitrate (NO 3 • , k = (0.37 ± 0.02) × 10 7 M −1 s−1) radicals. These rate coefficients indicate that the radiolytic longevity of BTPOA should increase with HNO 3 concentration, owing to the consumption of e aq − /H • , by nitrate anions, and the replacement of • OH by the less reactive NO 3 • .

Diaz Gomez, Laura [Forschungszentrum Juelich (Germ↗