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The extraction of Pu(IV) using 1.1 M solution of tri-iso-amyl phosphate (TiAP)/n-dodecane (DD) from plutonium nitrate solutions in nitric acid media was examined as a function of equilibrium aqueous phase metal ion concentration and equilibrium aqueous phase acidity at 303 K. The nitric acid concentration in the organic phase was measured as a function of equilibrium organic phase plutonium concentration. The co-extraction of U(VI) and Pu(IV) was studied using 1.1 M TiAP/DD system as a function of their equilibrium aqueous phase metal ion concentration and compared with 1.1 M tri-n-butyl phosphate (TBP)/n-DD system under identical conditions. Co-extraction and co-stripping of U(VI) and Pu(IV) were studied using 1.1 M TiAP/DD and 1.1 M TBP/DD systems in cross current mode to evaluate the number of stages required for the extraction and stripping of heavy metal ions (uranium and plutonium). The extraction and stripping efficiencies were calculated for both the systems. The saturation limit of the organic phase was also established in these studies.

The alkyl derivatives of N,N,-dialkyl-2-hydroxyacetamide (DAHyA), namely, N,N,-dihexyl-2-hydroxyacetamide (DHHyA), N,N,-dioctyl-2-hydroxyacetamide (DOHyA), N,N,-didecyl-2-hydroxyacetamide (DDHyA), were synthesized and characterized by FT-IR, and NMR spectroscopy. The extraction behaviour of Am(III) in a solution of DAHyA in n-dodecane was studied as a function of various parameters to evaluate the feasibility of using DAHyA for partitioning of trivalent actinides from high-level liquid waste (HLLW). The distribution ratio of Am(III) (DAm(III)) in 0.1 M DAHyA/n-DD increased from 10−4to > 200 with increase in nitric acid concentration from 1 M to 6 M. A DAm(III)value 10–70 was obtained in a nitric acid concentration range from 3–4 M. The extraction of Am(III) in 0.1 M DAHyA/n-dodecane decreased in the order DHHyA > DOHyA > DDHyA. The third-phase formation behaviour of nitric acid and neodymium (III) in 0.2 M DAHyA/n-dodecane was studied. Near stoichiometric amount of neodymium (III) can be loaded in octyl and decyl derivatives without any third-phase formation from 3–4 M nitric acid medium. Unlike other extractants, the loaded trivalent metal ion in DAHyA can be recovered even with 1 M nitric acid. Therefore, the study revealed that N,N,-dialkyl-2-hydroxyacetamides are promising candidates for partitioning the trivalent actinides from HLLW.

Piperidinium nitrate (Pip − NO3) and pyrrolidinium nitrate (Pyr − NO3) functional groups were anchored on a poly(styrene-divinylbenzene) matrix and evaluated the resultant anion exchange resin for the extraction of plutonium from nitric acid medium. The distribution coefficient (Kd, mL/g) of Pu(IV) in these resins increased with the concentration of nitric acid, reaching a maximum Kdat 7 M nitric acid, followed by decrease. The extraction of Pu(IV) increased with the duration of equilibration followed by the establishment of equilibrium, occurred within four hours of equilibration. The kinetic data were fitted with pseudo-first order and pseudo-second order rate equations. The apparent plutonium exchange capacity was determined to be ∼256 mg/g for Pip − NO3resin and 285 mg/g for Pyr − NO3resin at 7 M nitric acid. The radiolytic degradation of Pip − NO3and Pyr − NO3in presence of nitric acid (7 M) was studied upto a dose of 200 KGy and the results are reported in this paper.

The radioisotope 89Sr as 89SrCl2is medically useful for bone pain palliation and is produced in fast reactors using the 89Y(n, p)89Sr reaction. A procedure for isolation of the radionuclide 89Sr by chemical processing of the irradiated Y2O3target has been standardised and trial runs have been carried out at the Fast Breeder Test Reactor (FBTR), Kalpakkam. The chemical processing of the irradiated Y2O3target involves i) the removal of target Y(III) by TBP extraction and ii) further purification of the separated 89Sr fraction by cationic exchange chromatography. However a selective isolation of 89Sr by the Sr-specific crown ether makes the above chemical processing faster and relatively simple. This work presents a study on the selective removal of Sr from the irradiated target dissolver solution using the Sr-specific crown ether 4,4′(5′) di-tert-butyl-cyclohexano-18-crown-6 (DtBuCH18C6) in octanol medium. The separation behaviour of the other impurities such as Ce(IV), Y(III), Tb(III), Eu(III), Zn(II), Mn(II) and Rb(I) present along with Sr(II) in the irradiated sample was also investigated. The method of separation by using the crown ether DtBuCH18C6 is proved to be a potential tool for the purification of 89Sr(II) source produced from yttria target in fast reactors.

Diglycolamic acid was anchored on a polyamine matrix and the product (PA-DGAH) was characterized by ion-exchange capacity measurement, TG-DTA and FT-IR spectroscopy. The extraction behavior of Am(III) and Eu(III) in PA-DGAH was studied from dilute nitric acid medium to examine the feasibility using the anchored adsorbent for their mutual separation. The effect of various parameters such as the duration of equilibration, concentration of europium, nitric acid and diethylenetriaminepentaacetic acid (DTPA) in aqueous phase on the distribution coefficient (Kd) of Am(III) and Eu(III) was studied. The distribution coefficient decreased with increase in the concentration of nitric acid. Rapid extraction of metal ions in the initial stages of equilibration followed by the establishment of equilibrium occurred within 4 h. The data on the rate of uptake of Am(III) and Eu(III) were fitted into pseudo-first order and pseudo-second order rate equation. The extraction isotherm was fitted to Langmuir and Freundlich adsorption models and the apparent europium extraction capacity was determined. The mechanism of extraction was elucidated and the conditions needed for efficient separation of Am(III) from Eu(III) was optimized using DTPA. The study indicated the possibility of using PA-DGAH for the separation of Eu(III) from Am(III) with high separation factors.

This paper reports the extraction behavior of uranium and thorium by diamylbutyl phosphonate (DABP). This extractant was synthesized, characterized by various techniques like carbon and hydrogen analysis, IR, NMR (1H, 13C and 31P) and mass spectra. This paper reports the physicochemical properties like density and viscosity of this ligand. The tracer and macro extraction of uranium and thorium and stoichiometry of the extracted species of actinides by DABP is presented in this paper. This compound is an unsymmetrical phosphonate and is compared with symmetrical member of the same family, diamylamyl phosphonate (DAAP) and a member of trialkyl phosphate family, viz.tributyl phosphate (TBP).

In recent years, Supercritical Fluid Extraction (SFE) technique has been widely used for the extraction of metal ions. In the present study, extraction of uranium from nitric acid medium was investigated using supercritical carbon dioxide (Sc-CO2) containing various organophosphorous compounds such as trialkyl phosphates e.g.tri-iso-amyl phosphate (TiAP), tri-sec-butyl phosphate (TsBP) and tri-n-butyl phosphate (TBP), dialkylalkyl phosphonates, e.g.diamylamyl phosphonate (DAAP) and dibutyl butyl phosphonate (DBBP), dialkyl hydrogen phosphonates, e.g.dioctyl hydrogen phosphonate (DOHP), dioctylphosphineoxide (DOPO), trioctyl phosphine oxide (TOPO), n-octylphenyl N,N-diisobutyl carbamoylmethylphosphine oxide (CMPO) and di-2-ethyl-hexyl phosphoric acid (HDEHP). Some of these ligands have been investigated for the first time in the supercritical phase for the extraction of uranium. The extraction efficiency of uranium was studied with TiAP, DAAP and DBBP as a function of nitric acid concentration; the kinetics of the equilibration period (static extraction) and transportation of the metal complex (dynamic extraction) was investigated. The influence of pressure and temperature on the extraction behaviour of uranium with DAAP was studied from 4 N HNO3. The extraction efficiency of uranium from 4 N nitric acid medium was found to increase in the order of phosphates < phosphonates < HDEHP < TOPO < CMPO. In the case of phosphates and phosphonates, the maximum extraction of uranium was found to be from 4 N HNO3medium. The acidic extractants, HDEHP and DOHP showed relatively higher extraction at lower acidities. The relative extraction of uranium and thorium from their mixture was also examined using Sc-CO2containing phosphates, phosphonates and TOPO. The ligand, TsBP provided better fractionation between uranium and thorium compared to trialkyl phosphates, dialkyl alkyl phosphonates and TOPO.

The extraction and stripping behavior of various metal ions present in the fast reactor simulated high-level liquid waste (FR-SHLLW) was studied using a solvent phase composed of a neutral extractant, N,N,-didodecyl-N′,N′-dioctyl-3-oxapentane-1,5-diamide (D3DODGA) and an acidic extractant, di-2-ethylhexyl diglycolamic acid (HDEHDGA) in n-dodecane (n-DD). The third phase formation behavior of the solvent formulation D3DODGA + HDEHDGA/n-DD, was studied when it was contacted with FR-SHLLW, and the concentration of neutral and acidic extractant needed to avoid the third phase formation was optimized. The distribution ratio of various metal ions present in FR-SHLLW was measured in a solution of 0.1 M D3DODGA + 0.2 M HDEHDGA/n-DD. The extraction of Am(III) was accompanied by the co-extraction of lanthanides and unwanted metal ions such as Zr(IV), Y(III), and Pd(II). A procedure was developed to minimize the extraction of unwanted metal ions by using aqueous soluble complexing agents in FR-SHLLW. Based on those results, the counter-current mixer-settler run was performed in a 20-stage mixer-settler. Quantitative extraction of Am(III), Ln(III), Y(III), and Sr(II) in 0.1 M D3DODGA + 0.2 M HDEHDGA/n-DD was observed. The recovery of Am(III) from the loaded organic phase was carried out by the optimized aqueous formulation composed of 0.01 M diethylenetriaminepentaacetic acid (DTPA) + 0.5 M citric acid (CA) at pH 1.5. The stripping of Am(III) was accompanied by co-stripping of some early lanthanides. However the later lanthanides (Eu(III) and beyond) were not back extracted to Am(III) product. Therefore, the studies foresee the possibility of intra-lanthanides as well as lanthanide-actinide separation in a single-processing cycle.

A novel 2-oxy-N,N-dioctylacetamide anchored on Merrifield resin was synthesized and characterized by elemental analysis, TG-DTA and IR spectroscopy, for the first time. The extraction behavior of Am(III) and Eu(III) from nitric acid medium in the dioctylacetamide anchored resin (R-DOA) was studied as a function of various parameters such as duration of equilibration, concentration of nitric acid and sodium nitrate etc. The distribution coefficient (Kd) of Am(III) and Eu(III) in R-DOA increased with increase in concentration of nitric acid. The Kdvalues reached a maximum at 3–4?M nitric acid concentration in both cases, followed by decrease in Kdvalues. Rapid extraction of Am(III) and Eu(III) from nitric acid medium was observed in the initial stages of equilibration followed by the establishment of equilibrium occurred within 60?min. The extraction data was fitted in to the first order rate equation. The recovery of Am(III) and Eu(III) from the loaded organic phase was studied using dilute nitric acid. The results revealed that 2-oxyacetamide functionalized resin is the promising candidate for the separation of trivalent actinides and lanthanides from nitric acid medium.

As part of the studies on the development of alternate extractants for fast reactor fuel reprocessing, a mixer-settler facility for performing counter-current solvent extraction runs with radioactive materials has been set up and commissioned in our laboratory for flow sheet development experiments. Mixer-settler runs were carried out with 1.1?M TiAP/Heavy Normal Paraffin (HNP) as the solvent using the above facility to test the equipment as well as the behaviour of the solvent for continuous solvent extraction processes. Stage profile data generated for the extraction and stripping of nitric acid with TiAP based solvent have been reported. Studies carried out on the continuous counter-current extraction and stripping of U(VI) for 1.1?M TiAP/HNP-U(VI)-HNO3system under high solvent loading conditions are also reported in this paper. Overall and stage-wise mass balance data for the above runs are also discussed.

AbstractThe extraction behavior of U(VI) in the PUREX solvent namely 1.1 M tri-n-butyl phosphate (TBP) in n-dodecane (n-DD) was studied in the presence and absence of the ionic liquid, tri-n-octylmethylammonium bis-(2-ethylhexyl)phosphate ([A336]+[DEHP]–). The discrepancy in the extraction behavior observed in these systems was probed by measuring the distribution ratio of U(VI) as a function of various parameters. The distribution ratio of U(VI) in 0.03 M [A336]+[DEHP]–/n-DD decreased with increase in the concentration of nitric acid. The extraction of U(VI) was accompanied by the extraction of nitric acid leading to the formation of ionic liquid constituents namely tri-n-octylmethylammonium nitrate ([A336]+[NO3]–and bis(2-ethylhexyl)phosphoric acid (HDEHP). Therefore, the distribution ratio of U(VI) in the PUREX solvent was measured in the presence of these constituents also. The study revealed that extraction of U(VI) was controlled essentially by the synergistic contribution of TBP and the constituents of ionic liquid.

AbstractBatch extraction of U(VI) by 1.1 M solution of tri-iso-amyl phosphate (TiAP) in n-dodecane (n-DD) from uranyl nitrate solutions in nitric acid media has been investigated as a function of metal loading and equilibrium aqueous phase acidity at 303 K. A method adopted for free acidity determination by using a mixture of potassium oxalate and sodium fluoride for complexing U(VI) has been validated by comparing the data generated in the present study with literature data available for tri-n-butyl phosphate (TBP) under identical conditions. Extraction and stripping behaviour of 1.1 M TiAP/n-DD-U(VI)/HNO3and 1.1 M TBP/n-DD-U(VI)/HNO3systems have been compared by carrying out extraction, scrubbing and a series of stripping in cross-current mode. Stripping behaviour of both the solvents loaded with U(VI) and HNO3has also been investigated as a function of time to understand the effects of nitric acid induced degradation on stripping.

AbstractNew unsymmetrical diglycolamides (UDGAs), N, N-di-butyl-N′, N′-di-dodecyl-3-oxapentane-1,5-diamide (C12-C4), N, N-di-dodecyl-N′, N′-di-hexyl-3-oxapentane-1,5-diamide (C12-C6), N, N-di-decyl-N′, N′-di-dodecyl-3-oxapentane-1,5-diamide (C12-C10) have been synthesized, and evaluated for the separation of americium(III) and europium(III) from nitric acid medium. The extraction behavior of Am(III), Eu(III), and Sr(II) in a solution of these UDGAs in n-dodecane was studied as a function of concentration of nitric acid in the aqueous phase. The distribution ratio of Am(III) and Eu(III) increased with increase in the concentration of nitric acid. The third phase formation behavior of nitric acid and neodymium(III) in 0.1 M UDGA/n-dodecane was studied. The third phase formation was not observed in all these UDGAs in n-dodecane (0.1 M), when the concentration of Nd(III) was ∼ 500 mM in 3–4 M nitric acid. The stoichiometry of Am(III)-UDGA was determined from the slope analysis of the extraction data, which indicated the formation of 1:3 complex in all cases. Our studies revealed that the UDGA ligands with dodecyl group attached to one amidic nitrogen atom is inevitable for preventing third phase formation and the alkyl group at the other amidic nitrogen can be varied from butyl to decyl group for obtaining efficient extraction of trivalent actinides from high-level nuclear waste.

AbstractA new approach, namely “Single-cycle method for partitioning of Minor Actinides using completely incinerable ReagenTs” (SMART), has been explored for the separation of Am(III) from Eu(III) present in nitric acid medium. The extraction behavior of Am(III) and Eu(III) in a solution of an unsymmetrical diglycolamide, N,N,-didodecyl-N′,N′-dioctyl-3-oxapentane-1,5-diamide (D3DODGA), and an acidic extractant, N, N-di-2-ethylhexyl diglycolamic acid (HDEHDGA), in n-dodecane was studied. The distribution ratio of both these metal ions in D3DODGA-HDEHDGA/n-dodecane initially decreased with increase in the concentration of nitric acid reached a minimum at 0.1 M nitric acid followed by increase. Synergic extraction of Am(III) and Eu(III) was observed at nitric acid concentrations above 0.1 M and antagonism at lower acidities. Contrasting behavior observed at different acidities was probed by the slope analysis of the extraction data. The study revealed the involvement of both D3DODGA and HDEHDGA during synergism and increased participation of HDEHDGA during antagonism. The stripping behavior of Am(III) and Eu(III) from the loaded organic phase was studied as a function of nitric acid, DTPA, and citric acid concentrations. The conditions needed for the mutual separation of Am(III) and Eu(III) from the loaded organic phase were optimized. Our studies revealed the possibility of separating trivalent actinides from HLLW using these completely incinerable reagents.

AbstractA solution of 0.1 M N,N-di-2-ethylhexyl-N´,N´-di-octyl-3-oxa-1,5-diamide (DEHDODGA)-0.5 M N,N-dihexyloctanamide (DHOA) in n-dodecane has been evaluated for the separation of trivalent actinides from high-level liquid waste (HLLW). The extraction and stripping behaviour of Am(III) and other metal ions present in HLLW was studied. The distribution ratio of various metal ions was measured as a function of concentration of nitric acid and interfering ion. The extraction of metal ions decreased in the order M4+= M3+» M2+~ MO22+» M+. The extraction of Am(III), Eu(III), Y(III), and Zr(IV) was quantitative. However, the distribution ratio of unwanted metal ions such as Fe(III), Co(II), Sb(III), Mn(II), and Cs(I) was negligible. The distribution ratios of Cd(II), Ru(III), Pd(II), Mo(VI), Cr(VI), Ba(II), Ni(II), and Sr(II) were not insignificant, but were quite low. The conditions needed for the quantitative recovery of trivalents from the loaded organic phase were optimized with the use of diethylenetriaminepentaacetic acid (DTPA)-citric acid at pH 3 to facilitate the mutual separation of lanthanides-actinides in the subsequent step.

AbstractThe present study describes a correlation that is developed from retention of lanthanide and actinide complexes with the stability constant. In these studies, an ion-pairing reagent, camphor-10-sulphonic acid (CSA) was used as the modifier and organic acids such as a-hydroxy isobutyric acid (a-HIBA), mandelic acid, lactic acid and tartaric acid were used as complexing reagent for elution. From these studies, a correlation has been established between capacity factor of a metal ion, concentration of ion-pairing reagent and complexing agent with the stability constant of metal complex. Based on these studies, it has been shown that the stability constant of lanthanide and actinide complexes can be estimated using a single lanthanide calibrant. Validation of the method was carried out with the complexing agents such as a-HIBA and lactic acid.It was also demonstrated that data from a single chromatogram can be used for estimation of stability constant at various ionic strengths. These studies also demonstrated that the method can be applied for estimation of stability constant of actinides with a ligand whose value is not reported yet, e.g., ligands of importance in the lanthanide–actinide separations, chelation therapy etc.The chromatographic separation method is fast and the estimation of stability constant can be done in a very short time, which is a significant advantage especially in dealing with radioactive elements. The stability constant data was used to derive speciation data of plutonium in different oxidation states as well as that of americium with a-HIBA. The elution behavior of actinides such as Pu and Am from reversed phase chromatographic technique could be explained based on these studies.

AbstractThe alkyl derivatives of diglycolamide (DGA) such as N,N,N´,N´-tetraoctyldiglycolamide (TODGA), N,N,N´,N´-tetraethylhexyldiglycolamide (TEHDGA) and N,N-diethylhexyl- N´,N´-dioctyldiglycolamide (DEHDODGA) are chain isomers differing only in the arrangement of carbon chain of the alkyl group. The extraction behavior of Am(III) in the solution of these diglycolamides in ionic liquid medium was studied as a function of concentration of nitric acid and diglycolamide. The results were compared with those obtained in a molecular diluent, n-dodecane. The extraction of Am(III) in ionic liquid medium increased in the order TEHDGA

AbstractA typical high active waste (HAW) arising from reprocessing of the nuclear fuel –(U0.3Pu0.7)C irradiated to the burn-up of 155 GWd/Te in a Fast Breeder Test Reactor (FBTR) was characterized by various analytical techniques. The target element was separated from other interferences or pre-concentrated by a suitable technique such as solvent extraction, chromatographic methods etc.The results obtained from various techniques are compared. The implications of the estimated composition in handling, treatment and disposal of HAW are discussed.

The extractants, bis(2-ethylhexyl)diglycolamicacid (HDEHDGA) and bis(2-ethylhexy)thiodiglycolamic acid (HDEHSDGA) were synthesized and characterized by 1H and 13C NMR, mass and IR spectroscopy. The extraction behaviour of (152+154)Eu(III) and 241Am(III) from nitric acid medium by a solution of HDEHDGA (or HDEHSDGA) in n-dodecane (n-DD) was studied for the mutual separation of actinides and lanthanides. The effect of various parameters such as the pH, concentrations of HDEHDGA, HDEHSDGA, sodium nitrate, N,N,N´,N´-tetrakis(2-pyridylmethyl)ethylenediamine (TPEN) and diethylenetriaminepentaacetic acid (DTPA) on the separation factor (SF) of americium(III) over europium(III) and vice versawas studied, and the conditions needed for the preferential separation were optimised. The results show that HDEHDGA exhibits higher extraction for (152+154)Eu(III) and HDEHSDGA shows the superior selectivity for 241Am(III).

AbstractA procedure has been developed for the decontamination of radioruthenium from the lean organic phase composed of a solution 0.2 M n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylyphosphineoxide (CMPO) and 1.2 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), which was used for the partitioning of minor actinides from actual high active waste solution (155 GWd/Te). For this purpose, the stripping behavior of radioruthenium from 0.2 M CMPO–1.2 M TBP in n-DD was studied at 298 K by using various aqueous reagents and adsorbents. Among the different reagents investigated, the aqueous solution of sodium hydroxide and sodium carbonate and adsorbents such as neutral alumina and anion exchange resin (OH-form) were identified as the promising candidates. Nearly 90–95% of radioruthenium was removed from the lean organic phase in seven contacts using sodium carbonate or sodium hydroxide solution. The residual radioactivity in the organic phase was removed by treatment with neutral alumina or anion exchange resin. The quality of the organic phase was ascertained by 241Am(III) retention test.

AbstractThe retention behaviour of uranium, thorium, americium and lanthanides has been investigated, using a reverse phase C18column modified with 2,6-bis(5,6-dipropyl-1,2,4-triazin-3-yl)pyridine (n-Pr-BTP) with the aim of investigating rapid lanthanide-actinide separations by HPLC technique. Nitric acid and α-hydroxy isobutyric acid (α-HIBA) have been employed as the mobile phase. The influence of mobile phase pH and concentration on the retention of metal ions has been studied. The aim of this study is to investigate the use of BTP coated reverse phase columns as analytical columns for the fast separation of the minor actinides from the lanthanides in the high-level waste (HLW), and to explore whether such columns can be used on a preparative scale for the purification and the recovery of minor actinides from the high-level waste (HLW). It was observed that reverse phase columns modified with BTP showed greater affinity for actinides but exhibited no retention for lanthanides and the lanthanide-actinide separations could be achieved in nitric acid mobile phase, without the need for complexing agents. Americium was separated from the lanthanides with good peak profile and baseline resolution in both nitric acid as well as α-HIBA as the mobile phase. Based on this, the separation of americium from the lanthanides in the HLW was also demonstrated. The retention of a significant amount of thorium by the column in breakthrough studies also indicated that these columns can be used for preparative scale purification of the actinides.

AbstractBurn-up measurement for the fast reactor fuels in general and measurements on short cooled dissolver solutions in particular pose a great challenge due to the high level of radioactivity associated with the solution. An HPLC technique using monolith support has been developed for the first time for the determination of burn-up of fast reactor fuels and applied to the dissolver solution of the FBTR (Fast Breeder Test Reactor, Kalpakkam, India) fuel discharged at nominal burn-up of 150 GWd/t. A dynamic ion-exchange technique developed for the individual separation of lanthanides as well as actinides in connection with the burn-up measurements, has led to separation of all 14 lanthanides in about 2.77 min, the fastest LC technique as of now in literature and has resulted in a significant reduction in analysis time, radiation dose to operator and liquid waste generation. The dissolver solution was injected directly into the HPLC and assayed for lanthanide fission products, uranium, and plutonium using appropriate dilutions. The atom percent fission was calculated based on these measurements and the results are discussed. A reversed-phase chromatographic technique was also employed for the determination of uranium and plutonium in the dissolver solution. The sorption behaviour of uranium and thorium was examined on a monolith support modified with bis-2-ethylhexyl succinamic acid (BEHSA). Rapid separation of uranium from thorium could be achieved in about 20 s on the modified monolith support.

AbstractA method has been developed for partitioning of minor actinides from fast reactor (FR) fuel solution by a TRUEX solvent composed of 0.2 M n-octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (CMPO)-1.2 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), and subsequently demonstrated with genuine fast reactor dissolver solution (155 GWd/Te) using a novel 16-stage ejector mixer settler in hot cells. Cesium, plutonium and uranium present in the dissolver solution were removed, prior to minor actinide partitioning, by using ammonium molybdophosphate impregnated XAD-7 (AMP-XAD), methylated poly(4-vinylpyridine) (PVP-Me), and macroporous bifunctional phosphinic acid (MPBPA) resins respectively. Extraction of europium(III) and cerium(III) from simulated and real dissolver solution, and their stripping behavior from loaded organic phase was studied in batch method using various citric acid–nitric acid formulations. Based on these results, partitioning of minor actinides from fast reactor dissolver solution was demonstrated in hot cells. The extraction and stripping profiles of 154Eu, 144Ce, 106Ru and 137Cs, and mass balance of 241Am(III) achieved in the demonstration run have been reported in this paper.

AbstractA task specific ionic liquid (TSIL), diethyl-2-(3-methylimidazolium)ethylphosphonate bis(trifluoromethanesulfonyl)imide, ImPNTf2, was synthesized and characterized by 1H, 13C NMR, mass and IR spectroscopy. Extraction of Pu(IV), U(VI) and Am(III) from nitric acid medium by a solution of ImPNTf2in room temperature ionic liquids (RTILs), 1-alkyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (alkyl = butyl or octyl), was studied as a function of concentrations of nitric acid, ImPNTf2and temperature. The distribution ratio of Pu(IV) in ImPNTf2/bmimNTf2decreased with increase in the concentration of nitric acid reached a minimum around 4−5 M in nitric acid followed by increase. The stoichiometry of Pu(IV)-ImPNTf2was determined by the slope analysis of extraction data. In contrast to Pu(IV) extraction, the distribution ratios of Am(III) and U(VI) in ImPNTf2/amimNTf2and amimNTf2were significantly low at all nitric acid concentrations that led to unusually high separation factors for plutonium(IV) from other actinides. Viscosity of ionic liquid phase at various temperatures, and enthalpy change accompanied by the extraction of plutonium(IV) were determined and the results are reported in this paper.

This paper presents the results on the use of short columns (3−5 cm long) with small particle size (1.8 μm) for high performance liquid chromatographic separation of individual lanthanides and uranium from plutonium as well as uranium from thorium to achieve rapid separations i.e.separation time as short as 3.6 min for individual lanthanides, 1 min for thorium-uranium and 4.2 min for uranium from plutonium. These advantages can be exploited to significantly reduce analysis time, liquid waste generation as well as dose to operator when radioactive samples are analysed e.g.burn-up determination.

Extraction of americium(III) from nitric acid medium by a solution of tri-n-butylphosphate (TBP) and n-octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in room temperature ionic liquid, 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (bmimNTf2), was studied and the results were compared with that obtained with CMPO-TBP in n-doddecane (n-DD). The distribution ratio of 241Am(III) in TBP-CMPO/bmimNTf2was measured as a function of various parameters such as concentrations of nitric acid, CMPO, bmimNO3, NaNO3and TBP and temperature. Remarkably large distribution ratios were observed for the extraction of americium(III) when bmimNTf2acted as diluent and the extraction was insignificant in the absence of CMPO. The stoichiometry of metal–solvate in organic phase was determined by the slope analysis of extraction data and it indicated the formation of 1:3 (Am:CMPO) complex in organic phase. Viscosity of TBP-CMPO/bmimNTf2at various temperatures and enthalpy change accompanied by the extraction of americium(III) were determined and reported in this paper.

AbstractA novel method has been developed for the controlled delivery of various ligands including solid extractants using supercritical carbon dioxide (Sc-CO2). A ligand delivery vessel with a restricted geometry where its inlet and outlet are located beside each other, accomplished the task of the controlled delivery of ligands. The proposed method eliminates the usage of modifier solvent such as methanol for the delivery of various ligands during the supercritical fluid extraction of metal ions. Using neat Sc-CO2, the delivery profiles for the various ligands such as octyl(phenyl)-N,N-diisobutylcarbamoylmethyl-phosphineoxide (CMPO), tri-n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO), theonyltrifluroacetone (TTA) and di(2-ethylhexyl)isobutyramide (D2EHIBA) were established. The delivery profiles were optimised by investigating the influence of various extraction parameters such as ligand delivery vessel geometry, temperature, pressure, flow rate of Sc-CO2and ligand content. Applications of these ligand delivery modes were demonstrated for the supercritical fluid extraction of uranium from tissue paper matrix and preferential extraction of uranium over thorium. These results are compared with the extraction studies involving methanol or hexane modifier under similar experimental conditions. The extraction of neodymium nitrate was also demonstrated using modifier free delivery mode.

The extraction of plutonium from tissue paper matrix was carried out using supercritical carbon dioxide (Sc-CO2). The removal of plutonium was also studied from other matrices such as teflon, glass and stainless steel. Initial experiments were carried out under subcritical conditions. Supercritical extraction conditions were also employed in some experiments for the recovery of plutonium from tissue paper matrix. Complete extraction of Pu(III) as well as Pu(IV) in their nitrate form from tissue paper was achieved for the first time using supercritical carbon dioxide modified with n-octyl(phenyl)-N,N-diisobutyl carbamoylmethyl phosphine oxide (φCMPO) in methanol. Pu(IV) was also completely removed from other matrices such as teflon, stainless steel and glass. Studies on the extraction of Am(III) using supercritical carbon dioxide modified with CMPO in methanol resulted in its complete recovery from tissue paper.

Removal of radioactive nuclides from effluents is important for the safe disposal of the radioactive waste. In the present work ion exchange studies were carried out using simulated alpha active organic liquid scintillator and ethyl alcohol waste solutions containing Pu and Am using different resins. The time of equilibration and batch data were obtained under different conditions for individual resins as well as their mixtures. The indigenously developed Macroporous bifunctional phosphinic acid (MPBPA) resin was found to be suitable for the removal of alpha activity mainly due to Pu and Amberlite IR-120 and Biorad 50W × 8 for removal of Am. The results indicated that the presence of alcohol in the scintillator solution helps in increasing the distribution ratio. Experiments were carried out with column containing either MPBPA resin alone or with additional resin bed of Amberlite IR-120 using active organic liquid scintillator waste (with and without dilution with alcohol). Studies were also carried out on alcoholic waste generated in the laboratory during the washing of the used liquid scintillator vials. The alcohol after removal of activity either could be reused or could be disposed off directly as a ‘potentially’ active waste.

Alkali metal uranyl chloride (M2UO2Cl4(M = Na or Cs)) was dissolved in 1-butyl-3-methylimidazolium chloride (bmimCl) and the redox behavior of uranyl ion present in the resultant solution was investigated by cyclic voltammetry at 343 K. The cyclic voltammogram consisted of a reduction wave occurring at the peak potential of −0.85 V, due to the reduction of U(VI) and two oxidation waves occurring at the peak potentials of −0.6 V and +0.2 V. Controlled potential electrolysis of uranium(VI) loaded bmimCl gave a black deposit, which was characterized as uranium oxide by EDXRF and XRD analysis. Extraction of uranium(VI) from nitric acid medium by 0.05 M tri-n-octylmethylammonium chloride (TOMAC) in chloroform was studied and the cyclic voltammogram of uranyl ion present in the extracted phase exhibited a single reduction (−0.8 V) and an oxidation wave (0.06 V) at 298 K. Controlled potential electrolysis of uranium(VI) loaded TOMAC at −1.0 V also gave a black uranium oxide deposit similar to that observed in the previous case. The results indicated that uranyl ion in organic phase undergoes a single step two electron quasi-reversible reduction at the working electrode, which can be conveniently exploited for the direct recovery of uranium from the spent fuel or from high level liquid wastes.

Third phase formation in the extraction of Th(IV) by tri-2-methyl butyl phosphate (T2MBP) in various diluents such as n-dodecane, n-tetradecane and n-hexadecane from nitric acid media has been investigated for the first time. The limiting organic concentration (LOC) and maximum organic concentration (MOC) as well as critical aqueous concentration (CAC) and saturated aqueous concentration (SAC) were measured as a function of equilibrium aqueous acidity. Data on LOC and MOC values have been generated with various concentrations of T2MBP (1.1 M, 1.38 M and 1.65 M) in n-dodecane at 303 K. LOC and CAC values have been measured for the extraction of Th(IV) by 1.1 M T2MBP/n-hexadecane at various temperatures (298 K, 303 K and 313 K) as a function of equilibrium aqueous acidity. Effect of temperature on third phase formation has been studied in the extraction of Th(IV) from its solution with near zero free acidity by 1.1 M T2MBP in n-tetradecane and n-hexadecane as diluents. Data on extraction of nitric acid in the absence and presence of Th(IV) by T2MBP/n-dodecane is also reported in this paper.

A new class of polymeric resin called MCM-BEHSA, for extraction of U(VI), Th(IV) and La(III) was synthesized by grafting Merrifield chloromethylated polymer with N,N-bis-(2-ethyl-hexyl)-succinamic acid. The chemical modifications at each step of the grafting was confirmed by FT-IR spectroscopy, CPMAS solid state NMR spectroscopy, CHN elemental analysis and also by Thermo Gravimetric analysis. All the parameters influencing the quantitative metal ion extraction were optimized by both static and dynamic methods. The resin showed good extractability over a wide range of acidity (0.01-10 M) with fast exchange rate (saturation possible within 20 minutes) and high sorption capacities for U(VI), Th(IV) and La(III). Quantitative metal desorption was achieved by using 0.5 M (NH4)2CO3for all the analytes. A significant feature of this grafted polymer was its ability to extract Th(IV) from very high nitric acid concentrations. Interference studies with commonly encountered metal ions, rare earth ions and electrolytes showed very high tolerance limit for those ions. Enrichment factor values of 400, 550 and 350 with limit of quantification (LOQ) as 10 ng ml-1, 10 ng ml-1and 20 ng ml-1were observed for U(VI), Th(IV) and La(III), respectively. All the analytical data were within 3.2% rsd, reflecting the reproducibility and reliability of the developed method.

Room temperature molten salts for possible application of recovery of fission palladium from irradiated nuclear fuel/wastes have been investigated. The redox behavior of a solution of palladium(II) chloride and 1-butyl-3-methylimidazolium chloride (bmimCl) at glassy carbon working electrode has been studied at various temperatures using cyclic voltammetry. The voltammogram of bmimCl-PdCl2consists of a single reduction and two oxidation waves indicating that Pd2+undergoes a single step two-electron quasi-reversible reduction process at the working electrode. Controlled potential deposition of palladium on platinum electrode gave a black deposit, which was characterized to be metallic palladium. Extraction of palladium by Aliquat-336 has been studied as a function of nitric acid concentration and the redox behavior of palladium in Aliquat-336 phase has been investigated by cyclic voltammetry. The results suggest that some aqueous insoluble room temperature ionic liquids can act both as extractant and electrolytic medium and also have the potential of recovering of palladium from nuclear wastes.

The removal of uranyl nitrate from tissue matrix has been studied with supercritical carbon dioxide modified with methanol alone as well as complexing reagents dissolved in methanol. A systematic study of various complexing agents led to the development of an extraction procedure for the quantitative recovery of uranium from tissue matrix with supercritical carbon dioxide modified with methanol containing small quantities of acetylacetone. The drying time and temperature employed in loading of uranyl nitrate onto tissue paper were found to influence the extraction efficiency significantly.

N-octanoyl-N-phenylhydroxamic acid (OHA) has been prepared by partial reduction of nitrobenzene with Zn dust/NH4Cl, followed by acylation with n-octanoyl chloride. Structure of OHA was confirmed by 1H NMR, IR spectra and CHNS microanalysis. The extraction of uranium by OHA has been studied as a function of pH, temperature and concentration of OHA. The extraction of uranium was found to increase with increase in pH. Uranium forms a 1:2 complex with OHA. IR Spectra & NMR Spectra of the uranium-OHA complex and slope analysis obtained from extraction studies indicated that the extraction of uranium was due to the formation of a chelate. The enthalpy change accompanied by the extraction of uranium by OHA has been obtained by temperature variation method. The extraction process was found to be endothermic at pH 2 and exothermic at pH 6.

Extraction of uranium from tissue matrix was studied using supercritical carbon dioxide containing modifier solvent. The extraction efficiency was investigated with carbon dioxide containing methanol, tri-n-octyl phosphine oxide (TOPO) in methanol and tri-n-butyl phosphate (TBP) in methanol as modifiers. The extraction behaviour of the UO2(NO3)2· 2TOPO complex loaded on tissue was also investigated. Supercritical carbon dioxide modified with methanol alone was found to extract a maximum of about 76% uranium from tissue matrix.

In the above mentioned paper we had reported a non-linear dependence of limiting organic concentration (LOC) with temperature in the case of tri-2-methyl butyl phosphate in n-tetradecane as well as n-hexadecane diluents. Subsequently we extended the studies to other trialkyl phosphates such as TBP, TiBP, TsBP, TAP, TiAP and TsAP in various diluents (n-decane, n-dodecane, n-tetradecane, n-hexadecane and n-octadecane) and found recently that the dependence was always linear, irrespective of the extractant-diluent combination. This behaviour urged us to have a thorough review of the data reported earlier and when we repeated the studies again in the same systems, we found that the dependence was only linear and not as reported earlier.