Transuranium Extraction Research Articles

Preparation and crystal structure of tetraoctylphosphonium tetrakis(pentafluorophenyl)borate ionic liquid for electrochemistry at its interface with water

With ongoing and expending global energy needs, nuclear power is likely to be a key energy source for many more generations. A key question is still the fate of spent nuclear fuel (SNF). SNF contains a great deal of valuable material. In order to improve safety and handling, herein is proposed relatively high melting point ionic liquid (IL) or organic ionic plastic crystals (OIPCs) for biphasic water|IL metal ion extraction. The proposed IL, tetraoctylphosphonium tetrakis(pentafluorophenyl)borate, P8888(C6F5)4, was discovered to have a melting point of ∼55°C and a high degree of order by differential scanning calorimetry and X-ray diffraction (CCDC 1550126), respectively. In this way, the IL/OIPC could be used in ligand assisted metal ion extraction from water solutions at elevated temperatures and then ‘frozen’ at ambient conditions for facile manipulation and transport of the SNF. To test the feasibility of this, electrochemistry such as cyclic voltammetry at a water|IL micro-interface (25 mm in diameter) was employed to assess the thermodynamics of K+ ligand assisted transfer. It was revealed that the TRUEX (Transuranium Extraction) ligand octyl(phenyl)-N,N’-diisobutylcarbamoylphosphine oxide (CMPO) has a high to moderate coordination to K+ at the w| P8888(C6F5)4, micro-interface. Two ligand to K+ stoichiometries were determined to be 3:1 and 2:1, respectively.

Electrochemical assessment of water|ionic liquid biphasic systems towards cesium extraction from nuclear waste

A room temperature ionic liquid (IL) composed of a quaternary alkylphosphonium (trihexyltetradecylphosphonium, P66614+) and tetrakis(pentafluorophenyl)borate anion (TB−) was employed within a water|P66614TB (w|P66614TB or w|IL) biphasic system to evaluate cesium ion extraction in comparison to that with a traditional water|organic solvent (w|o) combination. 137Cs is a major contributor to the radioactivity of spent nuclear fuel as it leaves the reactor, and its extraction efficiency is therefore of considerable importance. The extraction was facilitated by the ligand octyl(phenyl)-N,N′-diisobutylcarbamoylphosphine oxide (CMPO) used in TRans-Uranium EXtraction processes and investigated through well established liquid|liquid electrochemistry. This study gave access to the metal ion to ligand (1:n) stoichiometry and overall complexation constant, β, of the interfacial complexation reaction which were determined to be 1:3 and 1.6×1011 at the w|P66614TB interface while the study at w|o elicited an n equal to 1 with β equal to 86.5. Through a straightforward relationship, these complexation constant values were converted to distribution coefficients, δα, with the ligand concentrations studied for comparison to other studies present in the literature; the w|o and w|IL systems gave δα of 2 and 8.2×107, respectively, indicating a higher overall extraction efficiency for the latter. For the w|o system, the metal ion-ligand stoichiometries were confirmed through isotopic distribution analysis of mass spectra obtained by the direct injection of an emulsified water–organic solvent mixture into an electron spray ionization mass spectrometer.

Development of the Diamex Process for Treating PHWR High-Level Liquid Waste

The extraction behavior of actinides like U(VI), Pu(IV), and Am(III) as well as of fission products like Tc(VII), Zr(IV), Eu(III), and structural material Fe(III) using nitric acid/simulated pressurized heavy water reactor-high-level waste solution as aqueous phase and N,N’-dimethyl-N,N’ dibutyl tetra decyl malonamide (DMDBTDMA) in n-dodecane as solvent was investigated. The present work contributes significantly toward the development of nonphosphorous (environmentally friendly) extractant capable of partitioning long-lived actinides and fission products like 99Tc from relatively short lived fission products like 90Sr and 137Cs as well as inactive components present in high-level waste. The DAm values determined in the temperature range between 15 and 45°C showed a gradual decrease with increase in temperature throughout the acidity range. The effect of N,N di-2-ethylhexyl acetamide (D2EHAA) as phase modifier on the physicochemical properties of DMDBTDMA/n-dodecane solvent was investigated and could be correlated with parameters like the limiting organic concentration value of HNO3, DAm, or phase disengagement time. Overall comparison of diamide extraction (Diamex) and transuranium extraction (Truex) solvents has been made.

EXTRACTION OF MOLYBDENUM FROM NITRIC ACID BY OCTYL(PHENYL)-N,N-DIISOBUTYLCARBAMOYLMETHYL-PHOSPHINE OXIDE

The extraction behavior of molybdenum(VI) under the conditions of transuranium extraction (TRUEX) process was studied by using octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide. The extraction behavior of molybdenum(VI) under the condition of PUREX process was also investigated for comparison. The extraction of molybdenum with the copresence of uranium(VI), dysprosium(III), and the effect of oxalic acid were examined. Under the PUREX conditions, molybdenum was less extractable, while under the TRUEX process conditions, a higher distribution ratio of molybdenum was observed. Under the TRUEX conditions, the addition of oxalic acid was effective in depressing molybdenum extraction. It was concluded that, at the concentration of nitric acid around 1 mol dm−3 or higher, Mo(VI) is extracted predominantly as cationic species, while at lower concentration of nitric acid, the coextraction of anionic molybdenum as a mixed complex with other metal, that is, UO2 2+, occurs.

Numerical Simulation of Extraction Behavior of Americium and Nitric Acid in the TRUEX Process

A numerical simulation code for the transuranium extraction (TRUEX) process was developed to determine the optimum operational conditions for the separation and recovery of transuranium (TRU) elements, such as americium. Equilibrium constants for the americium extraction by octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were determined by the least-squares method using experimental data. The activity coefficients for nitric acid and water were calculated with equations derived from experimental data. The calculated concentration profiles of americium, by using the best available equations, did not agree well with the experimental data in the extraction section. The stage efficiency was added in this code as a parameter to correct the differences between the calculated and the experimental profiles. The calculated concentration profiles using the modified code were in good agreement with the experimental data when 0.9 was used for the stage efficiency.

DISSOLUTION OF ORNL HLW SLUDGE AND PARTITIONING OF THE ACTINIDES USING THE TRUEX PROCESS

ABSTRACT Experiments were conducted to evaluate acid dissolution of actual high-level radioactive waste sludge and to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from the dissolved sludge solution. All tests were performed at a temperature of 24°C. Samples of sludge from Melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign, the rinsed sludge was leached in nitric acid and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of ∼1.8 M with a nitric acid concentration of 2.9 M. The solution began gelling immediately, and a visible gel layer was observed after 8 days. In the other campaign, the sludge was neutralized with nitric acid to destroy the carbonates and was subsequently leached with 2.6 M NaOH for ∼6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained a total cation concentration (excluding H+) of ∼0.6 M and a nitric acid concentration of 1.7 M. This solution became hazy in ∼8 days, indicating gel formation, but did not display separated gel layers after aging for 20 days. Batch liquid-liquid equilibrium tests of the extraction and the stripping operations were conducted. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge, and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th, and U removed from aqueous solution in only one extraction stage were >95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of >92%. In one test, vanadium appeared to be moderately extracted.

The Truex Process and the Management of Liquid Tru Uwaste

Abstract The TRUEX (TRansUranium Extraction) process is a new generic actinide extraction/recovery process for the removal of all actinides from acidic nitrate and chloride nuclear waste solutions. A brief review of the relevant chemistry of the TRUEX process and a summary of the current status of development and deployment of TRUEX process flowsheets to treat specific acidic waste solutions at several U.S. DOE sites is presented.

BEHAVIOR OF AMERICIUM IN THE STRIP STAGES OF THE TRUEX PROCESS

ABSTRACT The TRUEX (TRansUranium Extraction) process, recently developed at Argonne National Laboratory, separates and concentrates transuranium elements from acidic nuclear wastes. It is based on the use of the neutral extractant octyl(phenyl)–N,N–diisobutylcarbamoylmethylphosphine oxide (CMPO). CMPO Is combined with tributylphosphate (TBP) and a diluent to formulate the TRUEX process solvent. An attractive feature of the TRUEX process is that it allows the separation of americium from plutonium by using two different strip solutions. A detailed knowledge of the chemistry of americium in the stripping stages is essential for a successful Am/Pu separation. In this work the distribution ratio of Am(III) between TRUEX process solvent and diluted nitric acid has been measured in up to six successive strip experiments. The behavior of americlum during stripping has been correlated to the purity of the CMPO and the composition of the feed solution. With simulated waste solutions and with SX-grade CMPO, it is p...