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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

An Advanced TALSPEAK Concept for Separating Minor Actinides. Part 2. Flowsheet Test with Actinide-spiked Simulant

A solvent extraction system has been developed for separating trivalent actinides from lanthanides. This “Advanced TALSPEAK” system uses 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester to extract the lanthanides into a n-dodecane-based solvent phase, while the actinides are retained in a citrate-buffered aqueous phase by complexation to N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid. Batch distribution measurements indicate that the separation of americium from the light lanthanides decreases as the pH decreases. For example, the separation factor between La and Am increases from 2.5 at pH 2.0 to 19.3 at pH 3.0. However, previous investigations indicated that the extraction rates for the heavier lanthanides decrease with increasing pH. So, a balance between these two competing effects is required. An aqueous phase in which the pH was set at 2.6 was chosen for further process development because this offered optimal separation, with a minimum separation factor of ~8.4, based on the separation between La and Am. Centrifugal contactor single-stage efficiencies were measured to characterize the performance of the system under flow conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 OSTI2017

An Advanced TALSPEAK Concept for Separating Minor Actinides. Part 1. Process Optimization and Flowsheet Development

A system is being developed to separate trivalent actinides from lanthanide fission product elements that uses 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester to extract the lanthanide ions into an organic phase, while the actinide ions are held in the citrate-buffered aqueous phase by complexation to N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid (HEDTA). Earlier investigations of this system using a 2-cm centrifugal contactor revealed that the relatively slow extraction of Sm3+, Eu3+, and Gd3+ resulted in low separation factors from Am3+. In the work reported here, adjustments to the aqueous phase chemistry were made to improve the extraction rates. The results suggest that increasing the concentration of the citric acid buffer from 0.2 to 0.6 mol/L, and lowering the pH from 3.1 to 2.6, significantly improved lanthanide extraction rates resulting in an actinide/lanthanide separation system suitable for deployment in centrifugal contactors. Experiments performed to evaluate whether the lanthanide extraction rates can be improved by replacing aqueous HEDTA with nitrilotriacetic acid (NTA) exhibited promising results. However, NTA exhibited an unsatisfactorily high distribution value for Am3+ under the extraction conditions examined.

actinide extraction↗
DOE OSTI2019

Real time monitoring of the chemistry of hydroxylamine nitrate and iron as surrogates for nuclear materials processing

The separation of actinides in aqueous solution often involves changes in redox chemistry achieved by reagents such as hydroxylamine nitrate (HAN). As a result, in situ monitoring of the chemistry of HAN in contact with plutonium and neptunium in concentrated nitric acid at 35°C is important for the efficacy of Pu-238 extraction. A spectroscopic method of following the chemistry of HAN in contact with iron (as a simulant for plutonium) was developed, using a quantum cascade laser-attenuated total reflectance (QCL-ATR) system. The mid-infrared (3.67–12.5 μm) was chosen for its ability to distinguish the molecular vibrations of analytes and exclude the background absorption of the concentrated HNO 3 matrix. The concentrations of chemical species found in the reaction were drawn from the collected QCL power data using Beer’s law after a baseline correction. These data were qualitatively compared to a kinetic rate model based on a simplified reaction mechanism. The QCL-ATR method can help assess the areas of stability of HAN, nitrate/nitrite interactions, and assist process control of actinide separations.

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

Theoretical Prediction of Am(III)/Eu(III) Selectivity to Aid the Design of Actinide-Lanthanide Separation Agents

Selective extraction of minor actinides from lanthanides is a critical step in the reduction of radiotoxicity of spent nuclear fuels. However, the design of suitable ligands for separating chemically similar 4f- and 5f-block trivalent metal ions poses a significant challenge. Furthermore, first-principles calculations should play an important role in the design of new separation agents, but their ability to predict metal ion selectivity has not been systematically evaluated. We examine the ability of several density functional theory methods to predict selectivity of Am(III) and Eu(III) with oxygen, mixed oxygen–nitrogen, and sulfur donor ligands. The results establish a computational method capable of predicting the correct order of selectivities obtained from liquid–liquid extraction and aqueous phase complexation studies. To allow reasonably accurate predictions, it was critical to employ sufficiently flexible basis sets and provide proper account of solvation effects. The approach is utilized to estimate the selectivity of novel amide-functionalized diazine and 1,2,3-triazole ligands.

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

Synthesis and radiometric evaluation of diglycolamide functionalized mesoporous silica for the chromatographic separation of actinides Th, Pa and U

The separation of Th, Pa, and U is of critical importance in many applications including nuclear power, nuclear waste, environmental and geochemistry, nuclear forensics and nuclear medicine. Diglycolamide (DGA)-based resins have shown the ability to separate many elements, yet, these resins consist of non-covalent impregnation of the DGA molecules on the resin backbone resulting in co-elution of the extraction molecule during separation cycles, therefore limiting their long-term and repeated use. Covalently binding the DGA molecules onto silica is one way to overcome this issue. Herein, measured equilibrium distribution coefficients of normal extraction chromatographic DGA resin and a covalently bound form (KIT-6-N-DGA sorbent) are reported. Several differences are observed between the two systems, the most significant being observed for uranium, which demonstrated significantly lower sorption behavior on KIT-6-N-DGA. These results indicate that U can effectively be separated from Th and Pa using KIT-6-N-DGA, a task that could not be completed with the use of normal DGA alone.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Advancing the scientific basis of trivalent actinide-lanthanide separations

For advanced fuel cycles designed to support transmutation of transplutonium actinides, several options have been demonstrated for process-scale aqueous separations for U, Np, Pu management and for partitioning of trivalent actinides and fission product lanthanides away from other fission products. The more difficult mutual separation of Am/Cm from La-Tb remains the subject of considerable fundamental and applied research. The chemical separations literature teaches that the most productive alternatives to pursue are those based on ligand donor atoms less electronegative than O, specifically N- and S-containing complexants and chloride ion (Cl{sup -}). These 'soft-donor' atoms have exhibited usable selectivity in their bonding interactions with trivalent actinides relative to lanthanides. In this report, selected features of soft donor reagent design, characterization and application development will be discussed. The roles of thiocyanate, aminopoly-carboxylic acids and lactate in separation processes are detailed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Closing the Nuclear Fuel Cycle with a Simplified Minor Actinide Lanthanide Separation Process (ALSEP) and Additive Manufacturing

Expanded low-carbon baseload power production through the use of nuclear fission can be enabled by recycling long-lived actinide isotopes within the nuclear fuel cycle. This approach provides the benefits of (a) more completely utilizing the energy potential of mined uranium, (b) reducing the footprint of nuclear geological repositories, and (c) reducing the time required for the radiotoxicity of the disposed waste to decrease to the level of uranium ore from one hundred thousand years to a few hundred years. A key step in achieving this goal is the separation of long-lived isotopes of americium (Am) and curium (Cm) for recycle into fast reactors. To achieve this goal, a novel process was successfully demonstrated on a laboratory scale using a bank of 1.25-cm centrifugal contactors, fabricated by additive manufacturing, and a simulant containing the major fission product elements. Americium and Cm were separated from the lanthanides with over 99.9% completion. The sum of the impurities of the Am/Cm product stream using the simulated raffinate was found to be 3.2 × 10 –3 g/L. The process performance was validated using a genuine high burnup used nuclear fuel raffinate in a batch regime. Separation factors of nearly 100 for 154 Eu over 241 Am were achieved. All these results indicate the process scalability to an engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) as a general strategy and associated Waste Treatment Baseline Study (WTBS). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to both a nominal case study and near-term target reprocessing flowsheets. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include: • Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers (HICs). • Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. • Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. • Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. • An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased crystalline synthetic rock (Synroc). • An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or Synroc. • Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative with high promise the purification and reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations. • Electrochemical process salt HLW which is incorporated into a glass bonded Sodalite waste form known as the ceramic waste form (CWF). • Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from the reference mass balances are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Separation of Actinides from Lanthanides in Molten Fluorides by the Modulated-Current Electrolysis

Recently, the Molten Salt Reactor (MSR) is taking a renewed attraction as one of the reactor systems highlighted by the Generation IV International Forum. It is a non-classical (liquid fuel) system which can be operated as a breeder within the {sup 232}Th-{sup 233}U fuel cycle with the very interesting breeding factor. The necessary precondition for its proper functionality is the online reprocessing within the MSR's fuel cycle. A lot of effort is given to the development of the on-line pyrochemical reprocessing technology. Among the processes considered to be suitable for this process, the electrochemical separations are highlighted. Within this context, it is necessary to study electrochemical behavior of both fissile material and fission products in the molten salt systems and to develop the processes of their extraction from the system at various stages of the whole cycle. In this contribution, the study of the electrochemical separation of U from Gd in the LiF-CaF{sub 2} molten system by the modulated current electrolysis is presented. The study of MSR-related electrochemistry will continue as a part of the new Czech MSR R and D program. The knowledge of the electrochemical behavior of actinides and lanthanides in molten fluorides is necessary to the development of any electrochemical separation process. In our previous work, electrochemical characteristics of uranium, thorium, as well as that of several lanthanides, were measured in LiF-BeF{sub 2} (the basic melt considered for the MSR concept) and Be-free melts (LiF-NaFKF, LiF-NaF, LiF-CaF{sub 2}), and added to available literature data. If the fluoride carrier melt is electrochemically stable enough, it is possible to deposit the element of interest in its basic, metallic state (e.g. uranium in all types of fluoride melts tested) by the electrolysis. In the case of insufficient electrochemical stability of the fluoride melt in which the actinides and lanthanides are dissolved for the experiments, it is possible to use a reactive working electrode made of a material which creates alloys with the element to be deposited. An alloying reaction can shift the deposition potential in a favorable direction and therefore achieve the separation of an element from the system. The use of reactive electrode can also improve the adhesion of the deposit. When two or more electroactive species are present in the system at the same time, the efficiency of their separation by an electrolysis depends on how different the deposition potentials of these two species are. If we input certain modulation to the electrolytic current, we can improve overall result of an electrolytic experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Organic and Aqueous Redox Speciation of Cu(III) Periodate Oxidized Transuranium Actinides

A hexavalent group actinide separation process could streamline used nuclear fuel recycle and waste management. The limiting factor to such a process compatible with current fuel dissolution practices is obtaining and maintaining hexavalent Am, in molar nitric acid due to the high reduction potential of the Am(VI)/Am(III) couple (1.68 V vs SCE). Two strong oxidants, sodium bismuthate and Cu(III) periodate, have demonstrated quantitative oxidation of Am under molar acid conditions and better than 50% recovery by diamyl amylphosphonate (DAAP) is possible under these same conditions. This work considers the use of Cu(III) periodate to oxidize Np(V) to Np(VI) and Pu(IV) to Pu(VI) and recover these elements by extraction with DAAP. A metal:oxidant ratio of 1:1.2 and 1:3 was necessary to quantitatively oxidize Np(V) and Pu(IV), respectively, to the hexavalent state. Extraction of hexavalent Np, Pu, and Am by 1 M DAAP in n-dodecane was measured using UV-Vis [Pu(VI), Am (VI)] and NIR [Np(VI)]. Distribution values of Am(VI) were found to match previous tracer level studies. The organic phase spectra of Np, Pu, and Am are presented and molar absorptivities are calculated for characteristic peaks. Hexavalent Pu was found to be stable in the organic phase while Np(VI) showed some reduction to Np(V) and Am was present as Am(III), Am(V), and Am(VI) species in aqueous and organic phases during the extraction experiments. These results demonstrate, for the first time, the ability to recover macroscopic amounts of americium that would be present during fuel reprocessing and are the first characterization of Am organic phase oxidation state speciation relevant to a hexavalent group actinide separation process under acidic conditions.

americium↗
DOE OSTI2020

Sulfur donating extractants for the separation of trivalent actinides and lanthanides

The effective separation of trivalent actinides and lanthanides is capable of reducing the long term radiation hazard associated with used nuclear fuel. Here, this class of separation exploits the tendency of ligands containing large and polarizable soft donor atoms to preferentially bind to the trivalent actinides instead of the lanthanides. Among the soft donors, nitrogen and sulfur based ligands have received the most attention with sulfur donors generally having greater selectivity for the actinides. Herein, the speciation, mechanism, and selectivity for the extraction of lanthanides and actinides from aqueous media into an organic phase by various sulfur containing extractants is reviewed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Ultra-selective ligand-driven separation of strategic actinides

Abstract Metal ion separations are critical to numerous fields, including nuclear medicine, waste recycling, space exploration, and fundamental research. Nonetheless, operational conditions and performance are limited, imposing compromises between recovery, purity, and cost. Siderophore-inspired ligands show unprecedented charge-based selectivity and compatibility with harsh industry conditions, affording excellent separation efficiency, robustness and process control. Here, we successfully demonstrate a general separation strategy on three distinct systems, for Ac, Pu, and Bk purification. Separation factors (SF) obtained with model compound 3,4,3-LI(1,2-HOPO) are orders of magnitude higher than with any other ligand currently employed: 10 6 between Ac and relevant metal impurities, and over 10 8 for redox-free Pu purification against uranyl ions and trivalent actinides or fission products. Finally, a one-step separation method (SF > 3 × 10 6 and radiopurity > 99.999%) enables the isolation of Bk from adjacent actinides and fission products. The proposed approach offers a paradigm change for the production of strategic elements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Experimental and modeling studies of PR and ND oxalate solubility to high ionic strengths: Insight into actinide(III) oxalates

Actinide oxalates are chemical compounds important to nuclear industry, ranging from actinide separation in waste reprocessing, to production of specialty actinides, and to disposal of high level nuclear waste (HLW) and spent nuclear fuel (SNF). In this study, the solubility constants for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O by performing solubility experiments in HNO 3 and mixtures of HNO 3 and H 2 C 2 O 4 at 23.0 ± 0.2 °C have been determined. The targeted starting materials, Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O, were successfully synthesized at room temperature using PrCl 3 , NdCl 3 and oxalic acid as the source metrials. Then, we utilized the targeted solubility-controlling phases to conduct solubility measurements. There was no phase change over the entire periods of experiments, demonstrating that Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O were the solubility-controlling phases in our respective experiments. Based on our experimental data, we have developed a thermodynamic model for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O in the mixtures of HNO 3 and H 2 C 2 O 4 to high ionic strengths. The model for Pr 2 (C 2 O 4 ) 3 •10H 2 O reproduces well the reported experimental data for Pu 2 (C 2 O 4 ) 3 •10H 2 O, which are not utilized for the model development, demonstrating that Pr(III) is an excellent analog for Pu(III). Similarly, the model for Nd 2 (C 2 O 4 ) 3 •10H 2 O reproduces the solubility of Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The Pitzer model was used for the calculation of activity coefficients. Based on the published, well established model for dissociation constants for oxalic acid and stability constants for actinide-oxalate complexes [i.e., AmC 2 O 4 + , and Am(C 2 O 4 ) 2 - ] to high ionic strengths, we have obtained the solubility constants (log 10 K 0 ) for the following reactions at 25 °C, Pr 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Pr 3+ + 3C 2 O 4 2- + 10H 2 O(l). Nd 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Nd 3+ + 3C 2 O 4 2- + 10H 2 O(l). to be -30.82 ± 0.30 (2σ), and - 31.14 ± 0.35 (2σ), respectively. These values for can be directly applied to Pu 2 (C 2 O 4 ) 3 •10H 2 O, Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The model established for actinide oxalates by this study provides the needed knowledge with regard to solubilities of actinide/REE oxalates at various ionic strengths, and is expected to find applications in many fields, including the geological disposal of nuclear waste and the mobility of REE under the surface conditions, as Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O can be regarded as the pure Pr and Nd end-members of deveroite, a recently discovered natural REE oxalate with the following stoichiometry, (Ce 1.01 Nd 0.33 La 0.32 Pr 0.11 Y 0.11 Sm 0.01 Pb 0.04 U 0.03 Th 0.01 Ca 0.04 ) 2.01 (C 2 O 4 ) 2.99 •9.99H 2 O. Regarding its importance in the geological disposal of nuclear waste, Am 2 (C 2 O 4 ) 3 •10H 2 O/Pu 2 (C 2 O 4 ) 3 •10H 2 O/Cm 2 (C 2 O 4 ) 3 •10H 2 O can be the source-term phase for actinides, as demonstrated by the instance in the disposal in clay/shale formations. This is exemplified by the stability of Am 2 (C 2 O 4 ) 3 •10H 2 O in comparison with Am(OH) 3 (am), Am(OH) 3 (s) and AmCO 3 (OH)(s) under the relevant geological repository conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗