In this study the coordination chemistry of three ligands, C5H4NOCONRR' (where, R, R' = (C3H7)-C-i (L1); R, R' = (C4H9)-C-i (L2); and R = H, R' = (C4H9)-C-t (L3) composed of N-oxide and carboxamide groups have been explored with uranyl nitrate and some selected lanthanide (La, Sm, and Eu) nitrates. All the synthesized ligands as well as their complexes (1-12) of type UO2(NO3)(2)L (where, L = L1, L2, and L3 for 1, 2, and 3 respectively) and Ln(NO3)(3)(H2O)L-2 (where, Ln = La, L = L1 for 4, L = L2 for 5, and L = L3 for 6; Ln = Sm, L = L1 for 7, L = L2 for 8, and L = L3 for 9; Ln = Eu, L = L1 for 10, L = L2 for 11, and L = L3 for 12) have been characterized by elemental analysis, spectroscopic analyses such as FTIR, H-1 NMR, and electrospray ionization mass spectrometry (ESI-MS). Solid-state structural analysis of L1, 3, and 10 is carried out by X-ray crystallographic technique. The CO and NO groups of L1 are placed almost mutually perpendicular to each other in the crystal structure of L1. The X-ray data show that in [UO2(NO3)(2){C5H4NOCONH ((C4H9)-C-t)}] (3), the ligand acts as a bidentate chelating ligand and is bonded through both the N-oxo and amide oxygen atoms, whereas, in [Eu(NO3)(3)(H2O){C5H4NOCON((C3H7)-C-i)(2)}(2)] (10), the ligands show monodentate behavior and are bonded only through N-oxo oxygen atoms. Quantum mechanical calculation at DFT level corroborates the possibility of various bonding modes of these ligands towards uranium and europium nitrate with the preference of bonding as observed in the synthesized complexes. Solvent extraction studies using N,N-dioctyl N-oxo pyridine 2-carboxamide ligand (L4) in n-dodecane with UO22+, Pu4+, Am3+ and Eu3+ indicate the trend Pu4+ > UO22+ > Am3+ > Eu3+ at acidity range from 0.01 M to 6 M HNO3. The ligands show good radiation stability at gamma dose up to 500 kGy and chemical stability at 3 M HNO3 for up to 200 h without much affecting the metal ion extraction. Theoretical calculations show the possibility of presence of different metal species in the organic phase, other than the products obtained from dichloromethane during the solvent extraction of UO22+ and Eu3+ in water/dodecane biphasic media. Energy decomposition analysis supports the higher extraction coefficient of UO22+ than Eu3+ with an evidence of higher orbital interaction of the ligands with UO22+. (C) 2021 Elsevier Ltd. All rights reserved.

Zinc oxide is an active, inorganic material for a number of important applications. Zinc Oxide have wide properties for industrial applications which has been further increased by making the use of it in nanoscale size that fit in the corrosion resistant coatings. Nanoparticles have unique properties such as large surface area and nanoscale size due to which it possess various application in coatings and paints. Coating is a covering which when applied to the surface forms a layer and act as a protective barrier to the surface based on the property of the coating. Nanocomposite coatings are comparatively low cost and have more applications like dealing with the corrosion and radiation resistivity. Incorporation of nanoparticles in coating enhance its property by reducing the void and expected to induce barrier property for corrosion. For this purpose, zinc oxide nanoparticles were synthesized using wet chemical method. The effect of reactants (base) used in ZnO synthesis was studied from pH 9 to 11 and was characterized using XRD, FT-IR, SEM-EDX, etc. To increase the hydrophobicity and to make them suitable to disperse in polymeric binder, synthesized ZnO nanoparticles, have been surface modified with different silanes such as aminopropyl triethoxy silane (APTES), hexadecyl trimethoxy silane (HDTMS) and vinyltriethoxy silane (VTES). Surface modification was examined by FTIR, TGA, SEM-EDX and Contact angle analysis and VTES was found suitable for surface modification. Surface modified ZnO nanoparticles have been dispersed in polymeric binder (liquid EPDM) with other additives to form coating formulations. These coating formulations were applied on MS panels and it was observed that developed coating formulations have good mechanical and physical properties and high corrosion resistance.

The di-butyl phosphate (DBP) containing alkaline low level radioactive waste generated during PUREX solvent cleanup with Na2CO3 was decontaminated" using a new process called "in-situ causticization" where NaOH was produced in-situ by addition of Ca(OH)(2) resulting from precipitation of CaCO3. Alpha activity from the carbonate free solution was quantitatively precipitated by NaOH leading to a high alpha decontamination factor (DF) of similar to 240 in the supernatant solution. DF for beta-activity was low (similar to 3) due to partial precipitation of cesium and ruthenium. DBP remained in the solution as Na-DBP. XAD-4 resin was used to remove Na-DBP and any other dissolved organic. The remaining cesium was removed by resorcinol -formaldehyde resin and the solution containing only ruthenium was suitably diluted and discharged to sea. The calcium carbonate sludge was solidified with ordinary Portland cement, the activity load, compressive strength and leach rate of cement waste product (CWP) was within the acceptable limits of our near surface disposal facilities. The final volume of CWP obtained by this process is at least 2.5 times less than the volume obtained by direct cementation of the waste. (c) 2016 Elsevier B.V. All rights reserved.

Because of their significant retention capability, clay minerals have been proposed as a potential engineered barrier in high level nuclear radioactive waste disposal repositories. Smectite-rich natural clay is being considered as a backfill and buffer material for the Indian repository program. In the present study the sorption of Am by the clay, from granitic ground water, has been investigated. To identify the minerals in the clay controlling the sorption process, the adsorption isotherm of Eu(III), a chemical analogue of Am(III), was determined on montmorillonite-kaolinite clay mixtures having 0-20 wt% kaolinite. The effect of experimental parameters, such as, pH, ionic strength, and the presence of cation and anions on Am(III) sorption was further investigated to develop a sorption model for the natural clay. Overlapping adsorption isotherms of Eu(III) obtained for different montmorillonite-kaolinite clay suspensions established montmorillonite as the main sorbent for Eu/ Am(III) in the natural clay. Americium(III) sorption increases with pH in three distinct stages: at lower pH values (<4) the sorption is virtually insensitive to pH, then rises sharply (4-7) and subsequently attains a constant value at higher pH values (>8). Decreasing ionic strength increases the sorption at pH < 6 indicating the dominant role of ion exchange reactions at lower pH. A surface complexation model, developed for natural clay by including ion exchange site and amphoteric sites present at edges, simulates the sorption profiles at varying pH and ionic strength well and confirms the montmorillonite fraction as the sorbent controlling Am(III) sorption. The presence of Ca(II) as well as anions (Cl NO3-) does not affect Am(III) sorption on clay under granitic ground water pH and ionic strength conditions. However, the profile of Am(III) sorption to Ca(II)equilibrated clay differs from that for Na-equilibrated clay corroborating weaker exchange of Ca(II)Am( III) in comparison to Na(I)-Am(III). The presence of SO42- in the sorption system lowers Am(III) sorption at lower pH values. Modeling the sorption data indicated the participation of SO42- containing Am surface species. The thermodynamic model developed for sorption onto natural clay was checked for Eu(III) sorption from granitic ground water at pH 6.1. The model simulates the sorption at lower metal ion concentration while there is deviation at higher metal ion concentration. Inclusion of more types of surface sites and the effect of organic material need to be tested to correct the model for this deviation. (C) 2013 Elsevier Ltd. All rights reserved.

Presence of abundant pores and highly entangled nanotubes make bucky paper (BP) a natural candidate for filtration related applications. Both single- and multiwall carbon nanotube (SWNT and MWNT) BPs were fabricated via self-assembly. Average diameter of the pores on the surface of MWNT BP appeared to be 33 +/- 15 nm. However, due to the high tortuousity of BP, the cutoff size, estimated by filtration of colloidal dispersions of Au and CdS nanoparticles of different diameters, turned out to be 4-5 nm. The particle sizes were verified using microscopic and spectroscopic methods. The flux of Au nanoparticle solutions through BP was about 1000 L h(-1) m(-2) bar(-1). Highly flexible and robust SWNT BPs were prepared very easily, even without proper dispersion. It was found that the crystallinity and purity of the SWNTs had a bigger role than the quality of dispersion in determining the final mechanical strength of BP. The density of SWNT BP was estimated to be 1.3 +/- 0.3 g/cm(3) after correcting for the weights of the impurities. This is among the highest ever reported. The average wall to wall separation of adjacent tubes was estimated to be 0.35 +/- 0.2 nm, which is very close to the ideal value. The SWNT BP was also found to be impervious to liquids such as water, n-hexane, acetone, and isopropyl alcohol, indirectly verifying its close knit structure and small pore size. This is indicative of its possible use as a molecular sieve membrane.

After gaining considerable expertise in establishing and exercising waste management strategy by immobilization of high-level radioactive liquid waste (HLW) in borosilicate glass matrix followed by its interim storage, it is envisaged to focus on geologic disposal of radioactive waste packages. The multi barrier disposal system is considered one of the options for isolation of radioactive waste from human environment for an extended period of time. The engineered barrier comprises of waste form, canisters, over packs, buffer and backfill material whereas host rock serves as the natural barrier. The buffer and backfill materials play an important role because it can retard the migration of radionuclides by delaying the contact with ground water and associated sorption. Characterization and demonstration of efficacy of swelling clays in repository condition forms an integral part of studies ensuring the safe disposal of high level radioactive waste. Natural clays may be used solely or as a phase in a buffer/backfilling mixture because of its low hydraulic conductivity, high specific surface area and in some cases swelling ability. Long term performance of these materials has to be assessed with respect to their suitability as candidate backfill material. Twenty six nos. of swelling clay samples were collected from the natural deposits of different locations from western zone of the country. Samples collected have been taken up for detailed characterization of clay with respect to mineralogical, physico chemical, chemical, microscopic and geotechnical properties. From the X-ray diffraction patterns presented and the d-values associated with the patterns, montmorrilonite and quartz have been identified as major phases present in all the samples of swelling clay. Energy Dispersive X-Ray (EDX) analysis of the swelling clay samples mainly indicated presence of Si, Al, Na, K, Mg supporting thereby the X-Ray diffraction data. SEM micrographs of two typical clay samples at different magnifications indicate flake like surface morphology. The clay samples were found to be having swelling ability indicating smectite rich composition and optimum CEC. While chemical analysis indicates the presence of SiO2, Al2O3, Fe2O3, MgO, Na2O, CaO as major elements, higher amount of chlorides is INTERNATIONAL JOURNAL OF ENVIRONMENTAL ENGINEERING SCIENCE [IJEES] 226 attributed to the marine origin of the clay. Present paper given an account the experimental studies carried out to evaluate the potentiality of using swelling clays as buffer/backfill material for geological repositories of immobilized highlevel radioactive liquid waste.

The underlying objective governing the management of radioactive waste is protection of human being and the environment, now as well as in the future. As a waste management philosophy, utmost emphasis is given to waste volume minimization at all stages of design, operation and maintenance [1]. The development of innovative treatment processes for low and intermediate level wastes (LLW and ILW) in recent times has focused on volume reduction as one of the main objectives [2]. During reprocessing of spent fuel, an optimized mixture of tri-n-butyl phosphate (TBP) and dodecane is used as an extracting agent for actinides. During their repeated use, solvent undergoes chemical/radiolytic degradation, loosing its efficiency and hence to be discarded as spent organic solvent waste. This waste is presently being treated by alkaline hydrolysis process where spent solvent is refluxed with concentrated sodium hydroxide solution at 110°C [3]. TBP component of the spent solvent waste gets hydrolyzed and is converted into sodium salt of di-butyl phosphate (DBP). The other products obtained during hydrolysis include sodium salt of mono-butyl phosphate (MBP), butanol and phosphoric acid. Dodecane component of the spent solvent waste does not take part in the hydrolysis reaction and appears as a clear phase. The dodecane waste thus separated is subjected for thermal destruction by incineration. The emulsified aqueous layer thus obtained retains most of the activity present in the spent solvent waste. The present paper describes experimental studies performed to look for various approaches for management of this type of waste.