P. R. Vasudeva Rao

A REVIEW OF THIRD PHASE FORMATION IN EXTRACTION OF ACTINIDES BY NEUTRAL ORGANOPHOSPHORUS EXTRACTANTS

ABSTRACT Data are reviewed on the formation of third phase in the extraction of acti-nide(IV.VI) nitrates by neutral organophosphorus extractants, mainly tributyl phosphate. The data are critically evaluated and the effect of variables on the third phase formation is discussed. The variables are the concentrations of nitric acid and the extractant, temperature, the nature of diluent, addition of modifiers and the ionic strength of the aqueous phase. Also discussed are systems involving two extracted actinide ions.

Potential Applications of Room Temperature Ionic Liquids for Fission Products and Actinide Separation

Room temperature ionic liquids (RTILs) are receiving increased attention for possible applications in nuclear fuel cycle. RTILs are being investigated as possible substitutes to molecular diluents in solvent extraction, and as an alternative to high temperature molten salts in non-aqueous reprocessing applications. It is well recognized that ionic liquids often display unique extraction behavior. Unusual extraction of various metal ions has been reported in literature when RTILs are used as solvent medium. These aspects make the subject of RTILs fascinating and challenging. This review provides the current status of room temperature ionic liquid research related to the extraction of actinides and fission products (cesium and strontium) from nitric acid medium, with special emphasis on the studies carried out in our laboratory.

Liquid–liquid extraction of Pu(IV), U(VI) and Am(III) using malonamide in room temperature ionic liquid as diluent

The extraction behavior of U(VI), Pu(IV) and Am(III) from nitric acid medium by a solution of N,N-dimethyl-N,N-dioctyl-2-(2-hexyloxyethyl)malonamide (DMDOHEMA) in the room temperature ionic liquid, 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (C(4)mimNTf(2)), was studied. The distribution ratio of these actinides in DMDOHEMA/C(4)mimNTf(2) was measured as a function of various parameters such as the concentration of nitric acid, DMDOHEMA, NTf(2)(-), alkyl chain length of ionic liquid. The extraction of actinides in the absence of DMDOHEMA was insignificant and the distribution ratio achieved in conjunction with C(4)mimNTf(2), was remarkable. The separation factor of U(VI) and Pu(IV) achieved with the use of DMDOHEMA, ionic liquid was compared with Am(III) and other fission products. The stoichiometry of the metal-solvate was determined to be 1:2 for U(VI) and Pu(IV) and 1:3 for Am(III).

Extraction of uranium by amine, amide and benzamide grafted covalently on silica gel

Organic complexing moieties such as amine, 2°-amide and benzamide functional groups were covalently linked to silica gel by performing organomodification on commercially available silica gel. The extraction of uranium as a function of pH, time, concentration of uranium and the effect of ions such as Fe2+, Co2+, Ni2+ and CO32- on the distribution coefficient of uranium by sorbent was studied. The data was fitted into Langmuir adsorption isotherm. The selectivity of the sorbents towards uranium were found to decrease in the order gel-amine >> gel-benzamide ≅ gel-2°-amide at all pHs. The presence of metal ions such as Fe2+, Co2+, Ni2+ and CO32- were found to decrease the distribution coefficients (Kd, ml/g) of uranium and the decrease was very high when Fe2+ was present. The performance of the sorbent, gel-amine, under a given column condition was tested by following the breakthrough curve of uranium up to C/C0 = 0.025. The sorbent was found to be good for the quantitative pre-concentration of uranium from a large amount of sodium.

Surface complexation modeling of uranyl ion sorption on mesoporous silica

Sorption of uranium on MCM-41 was studied as a function of pH, time, [U(VI)] and [CO32−]. In the absence of carbonate the sorption edge occurs at pH 2 and the percentage sorption increased to 95% when the pH of the solution is 6. Further increase in pH results in complete sorption (99%) of uranium. In the presence of carbonate the percentage sorption decreased from 94 to 75% when the pH was varied from 6 to 11. The acid–base property of MCM-41 was investigated by titration of the sorbent with HNO3/NaOH at constant ionic strength. Three surface complexation models, namely, the constant capacitance model (CCM), diffuse layer model and non-electrostatic chemical equilibrium model (CEM) were employed to simulate the amphoteric behaviour of MCM-41 and uranium sorption on it. Various model parameters used for describing the sorption property of the sorbent were obtained from the non-linear regression of the experimental data. The sensitivity analysis and goodness-of-fit indicated that only CEM and CCM could be used to the description of the experimental data.

Unusual extraction of plutonium(IV) from uranium(VI) and americium(III) using phosphonate based task specific ionic liquid

Abstract A 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 ImPNTf2 in 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, ImPNTf2 and temperature. The distribution ratio of Pu(IV) in ImPNTf2/bmimNTf2 decreased 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)-ImPNTf2 was 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/amimNTf2 and amimNTf2 were 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.

The Effect of the Structure of Trialkyl Phosphates on their Physicochemical Properties and Extraction Behavior

The density of various trialkyl phosphates (TalP) such as tri‐n‐butyl phosphate (TBP), tri‐iso‐butyl phosphate (TiBP), tri‐sec‐butyl phosphate (TsBP), tri‐n‐amyl phosphate (TAP), tri‐2‐methylbutyl phosphate (T2MBP), tri‐iso‐amyl phosphate (tri‐3‐methylbutyl phosphate, TiAP), tri‐sec‐amyl phosphate (tri‐2‐amyl phosphate, TsAP), tri‐cyclo‐amyl phosphate (TcyAP), tri‐n‐hexyl phosphate (THP), and 1.1 M solutions of some of these phosphates in various diluents, solubility of water in trialkyl phosphates, and the aqueous solubility of trialkyl phosphates have been measured. Extraction of nitric acid, Th(IV), and U(VI) by trialkyl phosphates has also been studied by the batch extraction method. Metal‐solvate stoichiometry in the extraction of Th(IV) and U(VI) by some of the phosphates has been evaluated. Data on the extraction of U(VI) by various trialkyl phosphates as a function of equilibrium aqueous‐phase nitric acid concentration at 303 K are presented in this paper. Data on the extraction of Th(IV) and U(VI) from 1 M and 5 M HNO3 by trialkyl phosphates as a function of equilibrium aqueous‐phase metal concentration at 303 K are also presented in this paper. The effects of the structure of trialkyl phosphates on their physicochemical properties and extraction behavior are described in this paper.

Studies on the Use of N,N,N´,N´-Tetra(2-ethylhexyl) Diglycolamide (TEHDGA) for Actinide Partitioning. I: Investigation on Third-Phase Formation and Extraction Behavior

Abstract Diglycolamides have emerged as an interesting class of extractants for actinide partitioning from high-level waste (HLW). N,N,N´,N´-tetraoctyl diglycolamide (TODGA) has been extensively studied for lanthanide-actinide co-extraction behavior. The present work deals with a branched isomer of TODGA, that is, N,N,N´,N´-tetra(2-ethylhexyl) diglycolamide (TEHDGA). TEHDGA was studied for the extraction of 241Am and third-phase formation. The effect of using different phase modifiers on the prevention of the formation of a third phase during nitric acid extraction by TEHDGA along with the acid uptake behavior by TEHDGA in the presence of the modifiers was studied. The modifiers used for this purpose were di(n-hexyl)octanamide (DHOA), isodecanol, and n-decanol. The effect of the modifiers on the uptake of 241Am as a function of acid concentration and as a function of modifier concentration was also examined. DHOA was found to be a suitable modifier, in spite of its high acid uptake. The uptake of lanthanides Ce, La, Eu, Gd, and Nd and elements such as Fe, Ni, Mn, Mo, Ru, Sr, and Cs with DHOA-modified TEHDGA–n-dodecane solvent systems were investigated. The results obtained indicated that, while DHOA-modified TEHDGA/n-dodecane extracted lanthanides and actinides, it did not show any significant uptake of other elements. Thus, the TEHDGA-DHOA/n-dodecane solvent system can be used effectively for the partitioning of lanthanides and actinides from HLW.

A new procedure for the spectrophotometric determination of uranium(VI) in the presence of a large excess of thorium(IV).

Spectrophotometric determination of microgram amounts of U(VI) with 2-(5-Bromo-2-Pyridylazo)-5-diethylaminophenol (Br-PADAP), originally developed by Johnson and Florence has been modified to enable the determination of U(VI) in the presence of a large excess of Th(IV). The effects of thorium, tri-n-butyl phosphate (TBP) and ethanol on the estimation of uranium have been studied in detail and are presented in this paper. This modified method can be applied for the analysis of U(VI) both in aqueous and organic samples containing a large excess of Th(IV) (Th:U = 10000:1).

Development of fuels and structural materials for fast breeder reactors

Fast breeder reactors (FBRs) are destined to play a crucial role in the Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion.The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.