Gregg J. Lumetta

Review: Solvent Systems Combining Neutral and Acidic Extractants for Separating Trivalent Lanthanides from the Transuranic Elements

Abstract This paper is a review of recent publications that have focused on combined extractant systems for separating trivalent actinides from the lanthanides. These mixed solvent systems combine an acidic extractant with a neutral extractant to achieve the actinide/lanthanide separation. Depending on the neutral extractant used, three categorizations of systems can be considered, including combinations of acidic extractants with 1) diamides, 2) carbamoylmethylphosphine oxides, and 3) polydentate nitrogen-donor ligands. This review of relevant publications indicates that, although there is significant potential for practical exploitation of mixed neutral/acidic extractant systems to achieve a single-step separation of trivalent actinides from acidic high-level waste solutions, the fundamental chemistry underlying these combined systems is not yet well understood. For example, although there is strong evidence suggesting that adducts form between the neutral and acidic extractants, the nature of these adducts generally is not known. Likewise, the structures of the mixed complexes formed between the metal ions and the two different extractants are not fully understood. Research into these basic phenomena likely will provide clues about how to design practical mixed-extractant systems that can be used to efficiently separate the transuranic elements from the lanthanides and other components of irradiated fuel.

Advanced separation techniques for nuclear fuel reprocessing and radioactive waste treatment

Part 1 Fundamentals of radioactive materials separations processes: chemistry, engineering and safeguards: Chemistry of radioactive materials in the nuclear fuel cycle Physical and chemical properties of actinides in nuclear fuel reprocessing Chemical engineering for advanced aqueous radioactive material separations Spectroscopic on-line monitoring for process control and safeguarding of radiochemical streams in nuclear fuel reprocessing Safeguards technology for radioactive materials processing and nuclear fuel reprocessing facilities. Part 2 Separation and extraction processes for nuclear fuel reprocessing and radioactive waste treatment: Standard and advanced separation: PUREX processes for nuclear fuel reprocessing Alternative separation and extraction: UREX+ processes for actinide and targeted fission product recovery Advanced reprocessing for fission product separation and extraction Combined processes for high level radioactive waste separations: UNEX and other extraction processes. Part 3 Emerging and innovative techniques in nuclear fuel reprocessing and radioactive waste treatment: Nuclear engineering for pyrochemical treatment of spent nuclear fuels Development of highly selective compounds and processes for solvent extraction of long-lived radionuclides from spent nuclear fuels Developments in the partitioning and transmutation of radioactive waste Solid-phase extraction technology for actinide and lanthanide separations in nuclear fuel reprocessing Supercritical fluid and ionic liquid extraction techniques for nuclear fuel reprocessing and radioactive waste treatment Development of biological treatment processes for the separation and recovery of radioactive wastes.

Development of Effective Solvent Modifiers for the Solvent Extraction of Cesium from Alkaline High‐Level Tank Waste

Abstract A series of novel alkylphenoxy fluorinated alcohols were prepared and investigated for their effectiveness as modifiers in solvents containing calix[4]arene‐bis‐(tert‐octylbenzo)‐crown‐6 for the extraction of cesium from alkaline nitrate media. The structure of the fluorinated portion of the modifier influences the chemical stability, and a modifier that contained a terminal 1,1,2,2‐tetrafluoroethoxy group was found to decompose following long‐term exposure to warm alkaline solutions. However, replacement of the tetrafluoroethoxy group with a 2,2,3,3‐tetrafluoropropoxy group led to a series of modifiers that possessed the alkaline stability required for a solvent extraction process. Within this series of modifiers, the structure of the alkyl substituent (tert‐octyl, tert‐butyl, tert‐amyl, and sec‐butyl) of the alkylphenoxy moiety was found to have a profound impact on the phase behavior of the solvent in liquid–liquid contacting experiments, and hence on the overall suitability of the modifier for a solvent extraction process. The sec‐butyl derivative [1‐(2,2,3,3‐tetrafluoropropoxy)‐3‐(4‐sec‐butylphenoxy)‐2‐propanol] (Cs‐7SB) was found to possess the best overall balance of properties with respect to third phase and coalescence behavior, cleanup following degradation, resistance to solids formation, and cesium distribution behavior. Accordingly, this modifier was selected for use as a component of the solvent employed in the Caustic‐Side Solvent Extraction (CSSX) process designed for the removal of cesium from high‐level nuclear waste (HLW) at the U.S. Department of Energys (DOE) Savannah River Site. In batch equilibrium experiments, this solvent has also been successfully shown to extract cesium from both simulated and actual solutions generated from caustic leaching of HLW tank sludge stored in tank B‐110 at the DOEs Hanford Site.

The Actinide-Lanthanide Separation Concept

The Actinide-Lanthanide SEParation (ALSEP) concept is described. This concept is based on using an extractant phase consisting of either N,N,N’,N’-tetraoctyldiglycolamide (TODGA) or N,N,N’,N’-tetra(2-ethylhexyl)diglycolamide (T2EHDGA) combined with 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) to separate Am and Cm from lanthanide and other fission and activation products in a single solvent extraction cycle. The neutral TODGA or T2EHDGA serves to co-extract the trivalent actinide and lanthanide ions from nitric acid media. The distribution ratios for Am and the lanthanides increase with increasing nitric acid concentration. TODGA extracts these elements more strongly than T2EHDGA from nitric acid, but the weaker extracting ability of T2EHDGA could allow separation of Am from the light lanthanides during the extraction step. Switching the aqueous phase chemistry to a citrate-buffered diethylenetriaminepentaacetic acid (DTPA) solution at pH 2.5 to 4 results in selective transfer of the actinides to the aqueous phase, thus resulting in separation of these two groups of elements. Separation factors on the order of 20 to 40 can easily be achieved in the ALSEP systems.

Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

Abstract The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste generated during the production of nuclear weapons in the United States. The general strategy for treating the radioactive tank waste consists of pretreating the wastes by separating them into high-level and low-activity fractions. The high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for on-site. This paper reviews recent developments in the application of pretreatment technologies to the processing of the DOE radioactive tank wastes.

The TRUSPEAK Concept: Combining CMPO and HDEHP for Separating Trivalent Lanthanides from the Transuranic Elements

Combining octyl(phenyl)-N,N-diisobutyl-carbamoylmethyl-phosphine oxide (CMPO) and bis-(2-ethylhexyl) phosphoric acid (HDEHP) into a single process solvent for separating transuranic elements from liquid high-level waste is explored. Co-extraction of americium and the lanthanide elements from nitric acid solution is possible with a solvent mixture consisting of 0.1 M CMPO plus 1 M HDEHP in n-dodecane. Switching the aqueous-phase chemistry to a citrate-buffered solution of diethylene triamine pentaacetic acid (DTPA) allows for selective stripping of americium, separating it from the lanthanide elements. Potential strategies have been developed for managing molybdenum and zirconium (both of which co-extract with americium and the lanthanides). The work presented here demonstrates the feasibility of combining CMPO and HDEHP into a single extraction solvent for recovering americium from high-level waste and its separation from the lanthanides.

Synthesis and characterization of mono- and bis-(tetraalkylmalonamide)uranium(VI) complexes

The complex [UO 2 (NO 3 ) 2 (TMMA)] (TMMA= N , N , N ′, N ′-tetramethylmalonamide) was structurally characterized by single-crystal X-ray diffraction. The complex consists of two bidentate nitrate ions and one bidentate TMMA ligand coordinated to the UO 2 2+ ion. The complex [UO 2 (THMA) 2 ] 2+ (THMA= N , N , N ′, N ′-tetrahexylmalonamide) was prepared as the BF 4 − salt; this material tended to form an oil. However, [UO 2 (TMMA) 2 ](OTf) 2 (OTf=triflate) was isolated as a crystalline solid. Comparison of the Fourier transform infrared spectra of these complexes to the spectra of complexes formed in liquid–liquid extraction systems supports the hypothesis that complexes of the type [UO 2 (NO 3 ) 2 L] and [UO 2 L 2 ](NO 3 ) 2 (L=diamide extractant) form in the extraction systems.

Thermodynamic model for the solubility of Cr(OH)3(am) in concentrated NaOH and NaOH-NaNO3 solutions

The main objective of this study was to develop a thermodynamic model for predicting Cr(III) behavior in concentrated NaOH and in mixed NaOH–NaNO3 solutions for application to developing effective caustic leaching strategies for high-level nuclear waste sludges. To meet this objective, the solubility of Cr(OH)3(am) was measured in 0.003 to 10.5 m NaOH, 3.0 m NaOH with NaNO3 varying from 0.1 to 7.5 m, and 4.6 m NaNO3 with NaOH varying from 0.1 to 3.5 m at room temperature (22 ± 2°C). A combination of techniques, X-ray absorption spectroscopy (XAS) and absorptive stripping voltammetry analyses, were used to determine the oxidation state and nature of aqueous Cr. A thermodynamic model, based on the Pitzer equations, was developed from the solubility measurements to account for dramatic increases in aqueous Cr with increases in NaOH concentration. The model includes only two aqueous Cr species, Cr(OH)4− and Cr2O2(OH)4− (although the possible presence of a small percentage of higher oligomers at >5.0 m NaOH cannot be discounted) and their ion–interaction parameters with Na+. The logarithms of the equilibrium constants for the reactions involving Cr(OH)4− [Cr(OH)3(am) + OH− ⇌ Cr(OH)4−] and Cr2O2(OH)42− [2Cr(OH)3(am) + 2OH− ⇌ Cr2O2(OH)42− + 2H2O] were determined to be −4.36 ± 0.24 and −5.24 ± 0.24, respectively. This model was further tested and provided close agreement between the observed Cr concentrations in equilibrium with Cr(OH)3(am) in mixed NaOH–NaNO3 solutions and with high-level tank sludges leached with and primarily containing NaOH as the major electrolyte.

PRELIMINARY EVALUATION OF CHROMATOGRAPHIC TECHNIQUES FOR THE SEPARATION OF RADIONUCLIDES FROM HIGH-LEVEL RADIOACTIVE WASTE

ABSTRACT Three commercially available chromatographic materials were tested in a column mode for use in separating Am, Pu, and ™Sr from actual neutralized cladding removal waste (NCRW) taken from a Hanford waste tank. For these experiments, the NCRW was dissolved in HNO3/HF solutions. TRU-Spec™ was evaluated for the separation of transuranic elements. TRU-Spec™ removed Am and Pu from a dissolved NCRW solution, but not as effectively as the TRUEX solvent extraction process. Sr. Spec™ and SuperLig™ 601 were evaluated for the separation of Sr, Sr- Spec™ performed very well in the separation of SK from dissolved NCRW solution; Sr was not detectable in the column effluent after six bed volumes of solution had been passed through the column. SuperLig™ 601 was less effective than Sr- Spec™ at separating Sr from dissolved NCRW solution under the conditions used in this experiment. The poor performance of SuperLig Spec™ 601 has been attributed to slow kinetics of absorption of Sr.

COMPLEXATION OF URANYL ION BY TETRAHEXYLMALONAMIDES: AN EQUILIBRIUM MODELING AND INFRARED SPECTROSCOPIC STUDY

Abstract We investigated the extraction of uranyl nitrate from aqueous sodium nitrate with a series of tetrahexylmalonamides. The tetrahexylmalonamides considered were N , N , N ′, N ′-tetrahexylmalonamide (THMA), N , N , N ′, N ′-tetrahexyl-2-methylmalonamide (MeTHMA), and N , N , N ′, N ′-tetrahexyl-2,2-dimethylmalonamide (DiMeTHMA). This series allowed for a systematic determination of the effects of alkyl substitution of the methylene carbon. Equilibrium modeling of the extraction data indicates that at 1 M NaNO 3 , two extracted species are formed: UO 2 (NO 3 ) 2 L 2 and UO 2 (NO 3 ) 2 L 3 . The relative abundance of these two species depends on the nature of the tetrahexylmalonamide ligand. The UO 2 (NO 3 ) 2 L 2 species is dominant in the DiMeTHMA system, with very little formation of the UO 2 (NO 3 ) 2 L 3 species. In contrast, the UO 2 (NO 3 ) 2 L 3 species is more predominant in the MeTHMA case. The case of THMA lies in between. The greater propensity of MeTHMA versus THMA to bind in a 3:1 fashion to uranyl ion might reflect the greater basicity of the carbonyl oxygens in MeTHMA. The fact that DiMeTHMA binds primarily in 2:1 fashion suggests that steric constraints are more important in that ligand. As the nitrate concentration is increased, the ligand-to-metal ratios tend to decrease, i.e. the UO 2 (NO 3 ) 2 L 2 species tends to predominate, while the UO 2 (NO 3 ) 2 L 3 species becomes less important. In the case of THMA and MeTHMA, equilibrium modeling suggests the existence of a UO 2 (NO 3 ) 2 L species at higher nitrate concentrations. FTIR spectral studies confirm that at least two uranyl–THMA complexes formed, one of which has been identified as UO 2 (NO 3 ) 2 (THMA) by thermogravimetric analysis (TGA). The identity of the second species has not been definitively determined, but is most likely UO 2 (NO 3 ) 2 (THMA) 2 .