Distribution of technetium in PUREX process streams

Extraction of technetium has been carried out from the aqueous medium containing nitric acid under different experimental conditions to investigate its extraction behaviour in 30% tri-n-butyl phosphate in n-dodecane. In order to study the distribution behaviour of technetium in different streams of PUREX process, experiments were carried out under process conditions. The distribution of technetium was also studied using anion-exchange resin. Based on these results, the path of technetium in the PUREX process streams has been established which will be useful in the development of an advanced PUREX flow-sheet for containment and isolation of technetium in an environmental friendly fuel cycle.     

EXTRACTION BEHAVIOUR OF TECHNETIUM AND ACTINIDES IN THE ALIQUAT-336/NITRIC ACID SYSTEM

ABSTRACT

The extraction behaviour of technetium, thorium, uranium, neptunium, plutonium, americium and curium in the Aliquat-336 (diluted with 1,3-diisopropyl benzene) - nitric acid system have been studied. Aliquat-336 (tricapryl-methyl ammonium nitrate) is a quaternary ammonium salt extracting different species with an anion exchange mechanism. Distribution data obtained are modeled by anion exchange (technetium) and ion-pair formation mechanisms (actinides) with the extraction of nitric acid included to account for the lowering of the free extractant concentration. Reasonably high distribution ratios were obtained for technetium and the tetravalent elements (Th, Np and Pu) and Aliquat-336 can therefore be useful for partitioning of these elements.     

An Empirical Correlation for Pu(III) Distribution Coefficients in 30% TBP/n‐Dodecane PUREX System in the Presence of U(VI), U(IV), Pu(IV), Pu(III), and Hydrazine Nitrate

Abstract

For modeling and simulation of the partitioning step (either chemical or electrochemical) of the PUREX process, distribution coefficients of the involved species are needed. In this contribution, reported experimental data on Pu(III) extraction under PUREX conditions have been analyzed, and an empirical model is reported. The model reported here is considered more reliable than the model currently available in the literature.     

Batch Tests for Optimisation of Solvent Composition and Process Flexibility of the CHALMEX FS-13 Process

ABSTRACT

Studies have been performed with the purpose of determining the optimal solvent composition of a Chalmers grouped actinide extraction (CHALMEX) solvent for the selective co-extraction of transuranic elements in a novel Grouped ActiNide EXtraction (GANEX) process. The solvent is composed of 6,6’-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-[1,2,4]-triazin-3-yl)-[2,2’]-bipyridine (CyMe₄-BTBP) and tri-n-butyl phosphate (TBP) in phenyl trifluoromethyl sulfone (FS-13). The performance of the system has been shown to significantly depend on the ratios of the two extracting agents and the diluent to one another. Furthermore, the performance of the determined optimal solvent (10 mM CyMe₄-BTBP in 30% v/v TBP and 70% v/v FS-13) on various simulated PUREX raffinate solutions was tested. It was found that the solvent extracts all transuranic elements with high efficiency and good selectivity with regard to most other elements (fission products/activation products) present in the simulated PUREX raffinate solutions. Moreover, the solvent was found to extract a significant amount of acid. Palladium, silver, and cadmium were co-extracted along with the TRU-radionuclides, which has also been observed in other similar CHALMEX systems. The extraction of plutonium and uranium was preserved for all tested simulated PUREX raffinate solutions compared to experiments using trace amounts.     

Behaviors of plutonium and neptunium in nitric acid solutions containing hydrazine and technetium ions

The valence behaviors of plutonium and neptunium in the interaction of Pu(IV) and Np(V) with hydrazine and tetravalent uranium in technetium(VII)-containing aqueous nitric acid are reported. At [HNO₃] = 1 mol/l and Pu(IV) and Tc(VII) concentrations of ～0.1 and 0.01–0.2 mol/l, respectively, Pu(IV) is reduced to Pu(III) and is then entirely reoxidized to Pu(IV). Neptunium(V) in 1–3 M HNO₃ undergoes reduction to Np(IV) and then turns back into Np(V). The resulting solution usually contains a mixture of Np(IV) and Np(V). The reduction of Pu(IV) to Pu(III) and the reduction of Np(V) to Np(IV) are accompanied by hydrazine decomposition and by the reduction of most of the Tc(VII) to its lower valence forms. The conversions of Pu(III) into Pu(IV) and of Np(IV) into Np(V) are accompanied by the oxidation of these forms of technetium to Tc(VII). The introduction of diethylenetriaminepentaacetic acid into the reaction system makes Pu(III) more stable against reoxidation into Pu(IV) by reducing the hydrazine decomposition rate, enhances the conversion of Np(V) into Np(IV), and hampers Np(IV) oxidation to Np(V).     

EXTRACTION STUDY FOR TRUEX PROCESS USING SHORT-LIVED RADIONUCLIDES PRODUCED BY NEUTRON IRRADIATION OF URANIUM

The extraction behavior of short-lived fission products and neptunium was studied by using octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide under the conditions of the transuranic elements extraction (TRUEX) process. The short-lived fission products and neptunium were produced by neutron irradiation of UO₂ of natural uranium, and the extraction behavior of ⁹³Y, ⁹⁹Mo, ⁹⁷Zr, ¹²²Sb, ¹³²Te, ¹³³I, ¹⁴³Ce, and ²³⁹Np was simultaneously studied, where ¹²²Sb was produced by neutron irradiation of antimony metal. The extraction of fission products and Np under the conditions of the PUREX process was also studied for comparison. The extraction of nuclides in the presence of large amounts of uranium(VI), and the presence of oxalic acid was examined. The conditions and performance of the TRUEX extraction were discussed by considering the obtained results.     

Role of Acetohydroxamic Acid in Selective Extraction of Technetium and Uranium Employing N,N-Dihexyloctanamide as Extractant

Straight chain N,N-dihexyloctanamide (DHOA) has been identified as a promising alternate extractant to tributyl phosphate (TBP) for the reprocessing of uranium based spent fuels. The present work compares extraction behavior of technetium using DHOA and TBP solutions in n-dodecane, under varying experimental conditions such as acidity (0.5–6 M HNO₃); extractant concentration (1.1 and 1.5 M), and uranium loading (50 g/L, relevant for Pu rich spent fuel feed solutions). The effect of acetohydroxamic acid concentration on U, Pu, Np, and Tc extraction behavior has also been investigated. Pu(IV)-AHA interaction and its influence on extraction using TBP and DHOA extractants has been studied spectrophotometrically. The experimental data suggest that 1.1 M DHOA is better than 1.1 M TBP with respect to co-extraction of Tc and U, and U decontamination with respect to Np/Pu.     

SELECTIVE SEPARATION OF URANYL ION FROM TRU'S IN A COMBINED SOLVENT EXTRACTION PROCESS USING TETRAHYDROFURAN-2,3,4,5-TETRACARBOXYLICACID

ABSTRACT

Selective partitioning of uranyl from transuranic elements in a solvent extraction system which employs a neutral organophosphorus extractant and an aqueous complexant has been demonstrated in a previous report. The extractant solution combines octyl(phenyl)-N,N-diisobutylcarbamoylrnethylphosphine oxide (CMPO), diamyt(amyl)phosphonate (or tributylphosphate), and di(t-butylcyclohexano)-18-crown-6 in Isopar L, and is designed for simultaneous removal of strontium, technetium, lanthanides and actinides from radioactive wastes. The aqueous complexant is tetrahydrofuran-2,3,4,5-tetracarboxylic acid (THFTCA). In this report, the separation of UO₂ ²⁺ from Np(IV), Eu(III), Am(III), and Pu(IV) using the Combined Process Solvent has been optimized. Potentiometric titration and NMR spectroscopic results describe the distribution of THFTCA into the organic phase as a function of acidity and [THFTCA]. Further potentiometric titration experiments have determined the stoichiometry and stability of uranyl complexes in the aqueous phase. The thermodynamic data indicate that the uranyl complexes are anomalously weak which partially accounts for the selectivity. Ternary complexes involving, UO₂ ²⁺ CMPO, and THFTCA in the extractant phase also appear to play a role.     

DEVELOPING AND TESTING AN ALKALINE-SIDE SOLVENT EXTRACTION PROCESS FOR TECHNETIUM SEPARATION FROM TANK WASTE

ABSTRACT

Engineering development and testing of the SRTALK solvent extraction process are discussed in this paper. This process provides a way to carry out alkaline-side removal and recovery of technetium in the form of pertechnetate anion from nuclear waste tanks within the DOE complex. The SRTALK extractant consists of a crown ether, bis-4,4′(5′)[(tert-butyl)cyclohexano]-18-crown-6, in a modifier, tributyl phosphate, and a diluent, Isopar®L. The SRTALK flowsheet given here separates technetium from the waste and concentrates it by a factor of ten to minimize the load on the downstream evaporator for the technetium effluent. In this work; we initially generated and correlated the technetium extraction data, measured the dispersion number for various processing conditions, and determined hydraulic performance in a single-stage 2-cm centrifugal contactor. Then we used extraction-factor analysis, single-stage contactor tests, and stage-to-stage process calculations to develop a SRTALK flowsheet. Key features of the flowsheet are (1) a low organic-to-aqueous (O/A) flow ratio in the extraction section and a high O/A flow ratio in the strip section to concentrate the technetium and (2) the use of a scrub section to reduce the salt load in the concentrated technetium effluent. Finally, the SRTALK process was evaluated in a multistage test using a synthetic tank waste. This test was very successful. Initial batch tests with actual waste from the Hanford nuclear waste tanks show the same technetium extractability as determined with the synthetic waste feed. Therefore, technetium removal from actual tank wastes should also work well using the SRTALK process.

     

Simulation of Neptunium Extraction in an Advanced PUREX Process—Model Improvement

Routing neptunium to a single product in spent nuclear fuel reprocessing is a significant challenge. In this work, we have further improved the simulation of neptunium extraction in an advanced PUREX flowsheet by applying a revised model of the Np(V)–Np(VI) redox reaction kinetics, a new nitric acid radiolysis model, and by evaluating various models for the nitrous acid distribution coefficient. The Np disproportionation reaction is shown to have a negligible effect. The models are validated against published ‘cold test’ experimental results; the ‘hot test’ simulation suggests that high neptunium radiolysis could help to achieve high recoveries using this flowsheet.     

Neptunium Extraction and Stability in the GANEX Solvent: 0.2 M TODGA/0.5 M DMDOHEMA/Kerosene

Batch distribution studies show that above ~1 M HNO₃ Np(IV) and Np(VI) are well extracted into a solvent of 0.2 M TODGA/0.5 M DMDOHEMA in kerosene that has been formulated for the extraction of transuranic actinides in the GANEX (grouped actinide extraction) process. Np(IV) and Np(VI) are quite stable in the solvent phase, although Np(VI) is slowly reduced to Np(IV) on standing. Stripping of Np(IV,VI) ions from the GANEX solvent has been shown to be quite efficient using acetohydroxamic acid providing aqueous HNO₃ concentrations are below ~0.3 M. In contrast, Np(V) shows much lower distribution ratios and is unstable in the GANEX solvent phase with quite rapid formation of Np(IV) observed. Closer analysis shows this to be due to Np(V) disproportionation, which is enhanced at higher organic phase acidities. Disproportionation of Np(V) is also shown to occur in separate TODGA and DMDOHEMA kerosene solutions. These observations thus enable conditions for neptunium extraction to be optimized during the design of a GANEX flowsheet.     

Demonstration of a TODGA based Extraction Process for the Partitioning of Minor Actinides from a PUREX Raffinate

Abstract: Efficient recovery of minor actinides (MA) from genuine PUREX raffinate has been successfully demonstrated by the TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. The process was carried out in centrifugal contactors using an optimized flow‐sheet involving a total of 32 stages, divided into 4 stages for extraction, 12 stages for scrubbing and 16 stages for back‐extraction. Very high feed decontamination factors were obtained (Am, Cm ～ 40 000) and the recovery of these elements was higher than 99.99%. Of the non‐lanthanide fission products only Y and a small part of Ru were co‐separated into the product fraction together with the lanthanides and the MA.     

THE TRUEX PROCESS - A PROCESS FOR THE EXTRACTION OF THE TKANSURANIC ELEMENTS EROM NITRIC AC In WASTES UTILIZING MODIFIED PUREX SOLVENT*

ABSTRACT

A generic transurantc (TRU) element extraction/recovery process was developed based on the use of octyl(phenyl)-N,N-diiso-butylcarbamoylmetliylphosphine oxide, 0φD(iB)CMPO, dissolved in PUREX process solvent (tribntyl phosphate, TBP, in normal paraffluic hydrocarbon, NPH). The process (called TRUEX) is capable of reducing the TRU concentration by many orders of magnitude In waste solutions containing a wide range of nitric acid, salt, and fission product concentrations. A major feature of the process is that it is readily adaptable for waste processing in existing fuel reprocessing facilities.     

Note: Effect of Temperature in the Extraction of Uranium and Plutonium by Triisoamyl Phosphate

Abstract

The extraction of uranium(VI) by triisoamyl phosphate (TiAP) has been studied to derive the thermodynamic parameters such as entropy change and the free-energy change. The extraction of U(VI) and Pu(IV) has also been studied with 1.1 M solutions of tri-n-butyl phosphate (TBP), tri-n-amyl phosphate (TAP), and TiAP as a function of temperature. While the enthalpy of U(VI) extraction was found to be exothermic, the enthalpy for the extraction of Pu(IV) was always found to be endothermic. The temperature at which the distribution ratios of U(VI) and Pu(IV) cross each other (the temperature of inversion) has been derived for TBP, TAP, and TiAP, and the results reveal the lowest temperature of inversion occurs for TiAP. The results indicate the advantage of TiAP as an extractant in avoiding plutonium reflux during the PUREX process involving high plutonium feed solutions, in addition to lower aqueous solubility, freedom from the third-phase formation problem, lower degradation, and better economics.     

Partitioning of Actinides from High Level Waste of PUREX Origin Using Octylphenyl-N,N'-diisobutylcarbamoylmethyl Phosphine Oxide (CMPO)-Based Supported Liquid Membrane

Abstract

The present studies deal with the application of the supported liquid membrane (SLM) technique for partitioning of actinides from high level waste of PUREX origin. The process uses a solution of octylphenyl-N,N'-diisobutylcarbamoylmethyl phosphine oxide (CMPO) in n-dodecane as a carrier with a polytetrafluoroethylene support and a mixture of citric acid, formic acid, and hydrazine hydrate as the receiving phase. The studies involve the investigation of such parameters as carrier concentration in SLM, acidity of the feed, and the feed composition. The studies indicated good transport of actinides like neptunium, americium, and plutonium across the membrane from nitric acid medium. A high concentration of uranium in the feed retards the transport of americium, suggesting the need for prior removal of uranium from the waste. The separation of actinides from uranium-lean simulated samples as well as actual high level waste has been found to be feasible using the above technique.     

Direct Selective Extraction of Actinides (III) from PUREX Raffinate using a Mixture of CyMe4BTBP and TODGA as 1-cycle SANEX Solvent

Abstract

Within the framework of our research activities related to the partitioning of spent nuclear-fuel solutions, the direct selective extraction of trivalent actinides from a simulated PUREX raffinate was studied using a mixture of CyMe₄BTBP and TODGA (1-cycle SANEX). The solvent showed a high selectivity for trivalent actinides with a high lanthanide separation factor. However, the coextraction of some fission product elements (Cu, Ni, Zr, Mo, Pd, Ag, and Cd) from a simulated PUREX raffinate was observed, with distribution ratios up to 30 (Cu). The extraction of Zr and Mo could be suppressed using oxalic acid but the use of the well-known Pd complexant N-(2-Hydroxyethyl)-ethylendiamin-N,N′,N′-triacetic acid (HEDTA) was unsuccessful. During screening experiments with different amino acids and derivatives, the sulfur-bearing amino acid L-Cysteine showed good complexation of Pd and prevented its extraction into the organic phase without influencing the extraction of the trivalent actinides Am (III) and Cm (III). The optimization studies included the influence of the L-Cysteine and HNO₃ concentration and the kinetics of the extraction. The development of a process-like extraction series showed very promising results in view of further optimizing the process. A strategy for a single-cycle process is proposed within this article.     

Radioactive Tracer Tests of Two Improved TRPO Processes for High-Level Liquid Waste Using Annular Centrifugal Contactors

High-level liquid waste (HLLW) produced from the reprocessing of the spent nuclear fuel still contains moderate amounts of uranium, transuranium (TRU) actinides, and fission products, and thus constitutes a permanent hazard to the environment. The partitioning and transmutation (P&T) strategy has increasingly attracted interest for the safe treatment and disposal of HLLW, in which the partitioning of HLLW is one of the critical technical issues. Two improved trialkylphosphine oxide (TRPO) processes for the removal of actinides have been developed to treat Chinese HLLW, based on the original TRPO process. In one improved process N,N-diethylhydroxylamine as a reducing agent was used for reducing Np(V) and Np(VI) to Np(IV) in order to improve the extraction efficiency of Np. In the other improved process, ammonium vanadate as an oxidizing agent was used for oxidizing Np(V) and Np(IV) to Np(VI) in order to improve the extraction efficiency of Np. Radioactive tracer tests of two improved TRPO processes were carried out using 30-stage 10-mm-diam annular centrifugal contactors and simulated HLLW containing U, Np, Pu, and Am. The test results showed that the decontamination factor of total α activity was >1 × 10⁵. During the test, 30-stage 10-mm-diam annular centrifugal contactors worked in a stable manner continuously, with no stage failing or any interruption of the operation.     

Actinide Third Phase Formation in 1.1 M TBP/Nitric Acid/Alkane Diluent Systems

Abstract

Additional information on the organic phase speciation of Np and Pu was obtained in order to further understand the impact on third phase formation. In the Np(VI) extraction system, indications of the presence of a species associated with the absorbance at 1210 nm appears to be consistent with an increased tendency for third phase formation. Attempts to couple this absorption peak to a higher order nitrate species were inconclusive, and further study is required. For Pu(VI), continued evidence has emerged suggesting a role of higher order nitrate species in third phase formation.     

AKUFVE STUDIES ON EXTRACTION BEHAVIOUR OF NEPTUNIUM FROM SIMULATED HLW SOLUTIONS BY 30% TBP

ABSTRACT

The extraction behaviour of neptunium from 3 M nitric acid as well as simulated pressurised heavy water reactor high level radioactive waste (PHWR-HLW) solution by 30% TBP/dodecane was studied using AKUFVE. Np(IV)/Np(V) was oxidised to Np(VI) using oxidising agents, such as K₂Cr₂O₇, VO₂ ⁺ and NaNO₂. Stripping of neptunium from the loaded TBP phase was studied using reducing agents like hydrogen peroxide, ascorbic acid, hydroxyl-amine hydrochloride and hydrazine sulphate. Results of these extraction and stripping studies have been discussed in this paper.     

The Solvent Extraction of Heptavalent Technetium by Tributyl Phosphate

Abstract

The solvent extraction of heptavalent technetium from aqueous nitric or hydrochloric acid by tributyl phosphate in n-dodecane (TBP-NDD) has been studied over a wide range of TBP and acid concentrations at 25, 50, and 60°C. The extraction was found to proceed according to the reaction 3TBP + H⁺ + TcO₄ ^? → (HTcO₄ · 3TBP). A discussion of possible reaction mechanisms is presented, along with values for ΔG, ΔH, ΔS, and the equilibrium constant for the extraction reaction. Finally, evidence for the coextraction of technetium by uranyl ions is discussed.