Simulated Nuclear Waste Solutions Research Articles

Highly efficient and selective adsorption of palladium(II) from simulated nuclear waste solution using Amberlite XAD-7 resin impregnated with a phenanthroline-derived diamide

The increasing demand and limited natural resources for rare precious metal palladium render its separation and recovery from secondary sources such as acidic nuclear liquid waste greatly interesting. Herein, a novel Et-Tol-DAPhen extraction resin was prepared by impregnating Amberlite XAD-7 resin with a phenanthroline-derived diamide extractant N,N′-diethyl-N,N′-ditolyl-2,9-diamide-1,10-phenanthroline (Et-Tol-DAPhen), and the adsorption behavior of this new extraction resin towards Pd2+ in HNO3 solution was investigated through the batch and column experiments. The Et-Tol-DAPhen extraction resin exhibited strong adsorbability, high adsorption capacity, excellent selectivity and good reusability for Pd2+ in highly acidic HNO3 solution. In addition, the pseudo-second-order and Langmuir models were applied to understand the adsorption process of Pd2+ on Et-Tol-DAPhen extraction resin, respectively. It demonstrated the characteristics of a uniform single-layer chemisorption. The maximum uptake capacity could reach as high as 74.6 mg g−1 for Pd2+ in 3.0 mol/L HNO3 solution which is much higher than most adsorbent materials of palladium reported before. Furthermore, Et-Tol-DAPhen extraction resin could selectively uptake Pd2+ from a simulated acidic nuclear waste solution containing 20 different kinds of ions with the separation factor SFPd/M values as high as 100 to 10,000. Moreover, although the adsorption efficiency had a slight decrease after 5 times of adsorption-desorption process, it still remained at a high level. This new extraction resin displayed a potential application prospect in the field of separation and recovery of Pd2+ from acidic nuclear liquid waste.

Direct single-step synthesis of phase pure zeolite Na–P1, hydroxy sodalite and analcime from coal fly ash and assessment of their Cs+ and Sr2+ removal efficiencies

Combustion by-products of coal from thermal power plants cause environmental pollution; efforts are made to develop techniques and green processes that could reduce the waste volume and convert the waste to useful materials. Synthesis of phase pure zeolites from coal fly ash and its use as adsorbent and ion exchangers to remove Cs+ and Sr2+ from simulated nuclear waste solutions is the focus of the current study. Zeolite Na–P1, hydroxy sodalite and analcime were synthesized from fly ash obtained from an Indian thermal power plant by using hydrothermal method. Zeolite synthesis conditions were optimized by studying the effect of concentration of NaOH solution, homogenization time, hydrothermal temperature and time. The zeolite phase formation, surface morphology, thermal stability studies were carried out by XRD, SEM, TG-DTA techniques; specific surface area was determined by BET analysis. Crystalline pristine zeolite Na–P1 was prepared from fly ash on reaction with 1 M NaOH at 150 °C. Similarly, hydroxy sodalite and analcime were also synthesized from the fly ash by hydrothermal reaction with 3 M NaOH at 150 °C and 5 M NaOH at 200 °C respectively. The removal efficiencies of synthesized zeolites were investigated with aqueous simulated nuclear waste solutions containing Cs+ and Sr2+. The adsorption capacities and cation exchange capacity (CEC) of zeolite Na–P1, hydroxy sodalite and analcime for Cs+ and Sr2+ ions were demonstrated. The results show the adsorption capacity of Na–P1 for Cs+ is 39.3 mg/g and for Sr2+ is 92.48 mg/g respectively and is a better adsorbent than analcime and hydroxy sodalite.

Synthesis of nano-crystalline zeolite-A and zeolite-X from Indian coal fly ash, its characterization and performance evaluation for the removal of Cs+ and Sr2+ from simulated nuclear waste

Phase pure zeolite-A and zeolite-X were synthesized using coal fly ash (CFA) obtained from Indian thermal power plants by employing alkali fusion method followed by hydrothermal technique. The fusion of fly ash with Na2CO3 was accomplished by heating at 800 °C/2 h by maintaining fly ash to Na2CO3 ratio at 1.2. The fused mass was found to be nepheline (Na4Al4Si4O16); and on subsequent treatment of the fused mass with 3 M NaOH under hydrothermal condition transformed to zeolite-A (Na12Al12Si12O48.27H2O) and zeolite-X (Na88Al88Si104O384.194H2O). The effluent solution from zeolite-A synthesis was utilized to prepare cancrinite. The zeolites were characterized by XRD, FTIR, TG-DTA, SEM and surface area of the powders were measured by BET technique. The specific surface area of the zeolite-A and zeolite-X were found to be 58.29 ± 0.20 and 164.34 ± 5.4 m2g−1 respectively. The TG-DTA studies showed the conversion of nano-crystalline to micro-crystalline zeolites with loss of adsorbed water. The ion exchange capacities of these nano-crystalline zeolites were evaluated by using simulated nuclear waste solutions containing Cs+or Sr2+ ions. The adsorption capacity of zeolite-A was found to be 95.74 mg/g and 54.12 mg/g respectively for Sr2+ and Cs+ions. Similarly, zeolite-X shows the adsorption capacity of 93.14 mg/g and 53.14 mg/g respectively for Sr2+ and Cs+ ions.

Corrosion and mechanical properties of hot-rolled 0.5%Gd-0.8%B-stainless steels in a simulated nuclear waste treatment solution

Corrosion and mechanical behavior of the hot-rolled 0.5%Gd-0.8%B-stainless steel to develop a spent nuclear fuel storage material was studied in a simulated nuclear waste treatment condition with rolling condition. The austenite and ferrite phases of the 0.5%Gd-0.8%B-stainless steels are about 88:12. The average austenite and ferrite grain size of the plane normal to rolling, transverse and normal directions of the hot rolled specimens are about 5.08, 8.94, 19.35, 23.29, 26.00 and 18.11 [μm], respectively. The average micro-hardness of the as-cast specimen is 200.4 Hv, whereas, that of the hot-rolled specimen are 220.1, 204.7 and 203.5 [Hv] for the plane normal to RD, TD and ND, respectively. The UTS, YS and elongation of the as-cast and the hot-rolled specimen are 699, 484 [MPa], 34.0%, and 654, 432 [MPa] and 33.3%, respectively. The passivity was observed both for the as-cast and the hot rolled specimens in a simulated nuclear waste solution. The corrosion potential and corrosion rate of the as-casted specimens are −343 [mVSHE] and 3.26×10−7 [A/cm2], whereas, those of the hot rolled specimens with normal to ND, RD and TD are −630, −512 and −620 [mVSHE] and 6.12×10−7, 1.04×10−6 and 6.92×10−7 [A/cm2], respectively. Corrosion tends to occur preferentially Cr and B rich area.

A new approach for selective Cs+ separation from simulated nuclear waste solution using electrodriven cation transport through hollow fiber supported liquid membranes

The present work describes a novel and cleaner approach for large scale (litre) separation of cesium from simulated nuclear waste solutions by electrodriven transport through a hollow fiber supported liquid membrane (HFSLM) containing chlorinated cobalt dicarbollide as the carrier. To the best of our knowledge, this is the first report on electrodriven cation transport through HFSLM. The transport selectivity of Cs+ over Na+ and other fission products was studied from synthetic nuclear waste solutions. An attempt has also been made to pre-concentrate radio-cesium from an environmental sample containing very low level of 137Cs (≤ 10–12M) with excellent transport efficiency. A very high decontamination factor of cesium over Na+ other metal ions have been obtained for all types of feed solution. For neutral feed solution, the current efficiency for the Cs+ transport has been found to be 100%. The present work certainly adds a new dimension to the ongoing membrane based separations.

Surface studies on aluminized and thermally oxidized Alloy 690 substrates interacted with nitrate-based simulated nuclear waste and sodium borosilicate melt using Raman spectroscopy and X-ray photoelectron spectroscopy

Aluminized and thermally oxidized Alloy 690 specimen containing Al2O3 layer on the outermost surface and bare alloy substrates were exposed either in nitrate-based simulated nuclear waste solution at 373K for 216h or sodium borosilicate melt at 1248K for 192h. The exposed samples were characterized by micro-Raman spectroscopy and X-ray photoelectron spectroscopy (XPS). Aluminized and thermally oxidized sample exposed to simulated nuclear waste indicated the presence of Al2O3, mixed oxides of NiCr2O4 and NiFe2O4, NiO, a small amount of Cr2O3, mixed oxides of Cr+3 and Cr+6 and elemental aluminium. Similar surface compositions were recorded for coated samples exposed to sodium borosilicate melt. The formation of mixed oxides on coated samples could be attributable to oxidation treatment. Bare Alloy 690 substrates exposed to simulated nuclear waste, indicated the presence of γ-Fe2O3, NiO, a small amount of Cr2O3, mixed oxides of Cr+3 and Cr+6, elemental chromium and nickel. The alloy substrate exposed to sodium borosilicate melt showed similar compositions along with presence of CrO3, mixed oxides of NiCr2O4 and NiFe2O4. The presence of mixed oxides on bare alloy substrate exposed to molten glass resulted from high temperature exposure.

Copper ferrocyanide loaded track etched membrane: an effective cesium adsorbent

For the first time, the use of a membrane based copper ferrocyanide loaded sorbent for removal of Cs+ from neutral aqueous solution is reported in this work. The results indicated high Cs+ exchange capacity (1.75 mmol/g), large distribution coefficient, as well as fast absorption of the cation in the membrane. The membrane, inspite of presence of other metal ion, exclusively absorbed 137Cs from a simulated nuclear waste solution, though the absorption rate is affected due to their presence. This composite membrane showed promising properties for Cs+ removal from nuclear waste solution.

Electrodriven selective transport of Cs+ using chlorinated cobalt dicarbollide in polymer inclusion membrane: a novel approach for cesium removal from simulated nuclear waste solution.

The work describes a novel and cleaner approach of electrodriven selective transport of Cs from simulated nuclear waste solutions through cellulose tri acetate (CTA)/poly vinyl chloride (PVC) based polymer inclusion membrane. The electrodriven cation transport together with the use of highly Cs+ selective hexachlorinated derivative of cobalt bis dicarbollide, allows to achieve selective separation of Cs+ from high concentration of Na+ and other fission products in nuclear waste solutions. The transport selectivity has been studied using radiotracer technique as well as atomic emission spectroscopic technique. Transport studies using CTA based membrane have been carried out from neutral solution as well as 0.4 M HNO3, while that with PVC based membrane has been carried out from 3 M HNO3. High decontamination factor for Cs+ over Na+ has been obtained in all the cases. Experiment with simulated high level waste solution shows selective transport of Cs+ from most of other fission products also. Significantly fast Cs+ transport rate along with high selectivity is an interesting feature observed in this membrane. The current efficiency for Cs+ transport has been found to be ∼100%. The promising results show the possibility of using this kind of electrodriven membrane transport methods for nuclear waste treatment.

The Effect of Alkyl Substitution on the Extraction Properties of BTPhen Ligands for the Partitioning of Trivalent Minor Actinides and Lanthanides in Spent Nuclear Fuels

Bis-triazinylphenanthroline ligands (BTPhens), which contain additional alkyl (n-butyl and sec-butyl) groups attached to the triazine rings, have been synthesized, and the effects of this alkyl substitution on their extraction properties with Ln(III) and An(III) cations in simulated nuclear waste solutions have been studied. The speciation of n-butyl-substituted ligand (C4- BTPhen) with some trivalent lanthanide nitrates was elucidated by 1 H-NMR spectroscopic titrations. These experiments have shown that the dominant species in solution were the 1:2 complexes [Ln(III)(BTPhen)2], even at higher Ln(III) concentrations, and the relative stability of 2:1 to 1:1 BTPhen-Ln(III) complexes varied with different lanthanides. As expected, sec-butylsubstituted ligand (sec-C4 BTPhen) showed higher solubility than C4-BTPhen in certain diluents. A greater separation factor (SFAm/Eu = ca. 210) was observed for sec-C4-BTPhen compared to C4-BTPhen (SFAm/Eu = ca. 125) in 1-octanol at 4 M HNO3 solutions. The greater separation factor may be due to the higher solubility of the 2:1 complex for sec-C4-BTPhen at the interface than the 1:1 complex of C4-BTPhen.

Application of nanofiltration membrane and ethyleneimine oligomer mixture for selective separation of Strontium from a simulated nuclear waste solution

Efficient radioactive waste management calls for maximum volume reduction and selective separation of elements from a radioactive waste solution. Ethylenimine (EI) oligomer mixture and GE Osmonics nanofiltration membrane has been used for selective strontium separation from an aqueous synthetic nuclear waste solution. Effect of concentration of EI, presence of cesium and sodium ions in strontium rejection, and effect of feed pH on strontium rejection, volume reduction, and permeate flux has been studied.

Graphene oxide for effective radionuclide removal

Here we show the efficacy of graphene oxide (GO) for rapid removal of some of the most toxic and radioactive long-lived human-made radionuclides from contaminated water, even from acidic solutions (pH < 2). The interaction of GO with actinides including Am(III), Th(IV), Pu(IV), Np(V), U(VI) and typical fission products Sr(II), Eu(III) and Tc(VII) were studied, along with their sorption kinetics. Cation/GO coagulation occurs with the formation of nanoparticle aggregates of GO sheets, facilitating their removal. GO is far more effective in removal of transuranium elements from simulated nuclear waste solutions than other routinely used sorbents such as bentonite clays and activated carbon. These results point toward a simple methodology to mollify the severity of nuclear waste contamination, thereby leading to effective measures for environmental remediation.

Synthesis and Evaluation of Lipophilic BTBP Ligands for An/Ln Separation in Nuclear Waste Treatment: The Effect of Alkyl Substitution on Extraction Properties and Implications for Ligand Design

AbstractFour new 6,6′‐bis(1,2,4‐triazin‐3‐yl)‐2,2′‐bipyridine (BTBP) ligands, which contain either additional alkyl groups on the pyridine rings or seven‐membered aliphatic rings attached to the triazine rings, have been synthesized, and the effects of the additional alkyl substitution in the 4‐ and 4′‐positions of the pyridine rings on their extraction properties with LnIII and AnIII cations in simulated nuclear waste solutions have been studied. The speciation of ligand 13 with some trivalent lanthanide nitrates was elucidated by 1H NMR spectroscopic titrations and ESI‐MS. Although 13 formed both 1:1 and 1:2 complexes with LaIII and YIII, only 1:2 complexes were observed with EuIII and CeIII. Quite unexpectedly, both alkyl‐substituted ligands 12 and 13 showed lower solubilities in certain diluents than the unsubstituted ligand CyMe4‐BTBP. Compared to CyMe4‐BTBP, alkyl‐substitution was found to decrease the rates of metal‐ion extraction of the ligands in both 1‐octanol and cyclohexanone. A highly efficient (DAm &gt; 10) and selective (SFAm/Eu &gt; 90) extraction was observed for 12 and 13 in cyclohexanone and for 13 in 1‐octanol in the presence of a phase‐transfer agent. The implications of these results for the design of improved extractants for radioactive waste treatment are discussed.

Technetium‐99 MAS NMR Spectroscopy of a Cationic Framework Material that Traps TcO4− Ions

No chance for technetium: 99Tc magic-angle spinning (MAS) NMR spectra show that TcO4− ions, which are generated by nuclear fission and can contaminate the environment, can be trapped within the channels and cavities of a cationic framework material (see picture). These spectra are among the first 99Tc MAS NMR spectra reported to date, and show that the TcO4− ions can be efficiently removed from simulated nuclear waste solutions.

Assessing the mechanism of the synergistic action of calixarenes and Co-dicarbollides in lanthanide extractions

The extraction pathway of lanthanide ions from simulated nuclear waste solutions and the synergistic effects of a calixarene and a cobalt dicarbollide anion (CD) have been investigated on a model system by the combined use of electrospray ionization MS, FTMS ICR, together with 1H and 11B high field NMR spectroscopy techniques. Moreover, the extracting capabilities of the calixarene/BrCD mixture were determined by liquid–liquid extraction tests. An enhanced extraction efficiency of the calixarene/BrCD mixture with respect to the calixarene alone was observed. MS analysis allowed the assessment of a supramolecular ternary association, while the NMR data gave information on the calixarene binding sites for lanthanide ions. Herein, we report the experimental evidence of the formation of a ternary lanthanide/calixarene/Co-dicarbollide aggregate, which allowed us for the first time to assess the mechanism of the synergistic action of calixarenes and CD anions in lanthanide ion extraction.

Adsorptive Separation of Palladium from a Simulated Nuclear Waste Solution with Activated Carbon Fibers

Adsorptive separation of palladium from a nitric acid solution was performed by using a pitch‐based activated carbon fiber modified with NaOH (NaOH‐ACF). The NaOH‐ACF showed a markedly improved adsorption performance for the palladium ions as compared with the untreated ACF. It also showed a much higher preferential adsorption for the palladium ion over the rhodium and rhenium ions. The equilibrium capacity of the NaOH‐ACF for palladium was 1.25 meq · g−1. The breakthrough behavior obtained by using the bed packed with the NaOH‐ACF showed a complete separation of palladium from the coexisting ions of rhodium and rhenium. The spent bed was eluted effectively with nitric acid and a palladium nitrate with a purity of higher than 99% could be recovered.

Effect of Solution Concentration on the Properties of a Cementitious Grout Wasteform for Low-Level Nuclear Waste

A simulated low-level nuclear waste solution was studied for possible solidification in a cement-based matrix. The waste composition was based on an alkaline mixture of Na3(PO4)·12H2O, NaNO2, Na2CO3 and Al(NO3)3·9H2O, and the binder composition was cement (21%), fly ash (68%), and attapulgite clay (11%). The materials were mixed at a high solution-to-binder ratio of 1.0 l/kg, and curing temperatures varied from 45 to 90°C. The effect of changes in solution concentration was studied. Solution concentration ranged from a dilution to 5.5% (designed to simulate a possible off-gas condensate obtained during vitrification of the waste) to the full concentration of the simulated waste. Compressive strength and early age heat development increased as the concentration was increased up to 67%, but at higher concentrations both compressive strength and heat development decreased. X-ray diffraction and 29Si and 27Al magic angle spinning nuclear magnetic resonance spectroscopy pointed to a high degree of reaction of the fly ash in the mixes and formation of zeolites at the higher concentrations. Na-P1 zeolite formed in increasing quantities as the concentration was raised to 67%, but at the highest concentrations the zeolite formed was sodalite.

The Extraction of 137Cs and 89Sr from Waste Simulants Using Pillared Montmorillonite

ABSTRACT Two samples of a silica-pillared montmorillonite produced using 3-aminopropyltrimethoxy silane and an alumina-pillared montmorillonite were evaluated for the removal of 137Cs and 89Sr from a simulated nuclear waste solution and a simulated groundwater, and the results were compared to the parent montmorillonite and two zeolites, AW500 (chabazite) and clinoptilolite. The parent and pillared clays were characterized using x-ray powder diffraction and surface area analysis by nitrogen adsorption/desorption studies. The pillared clays exhibited d-spacings of between 173 and 182 A after calcination, and surface areas ranging from 71 to 264 m2g-1. Both of the silica-pillared clays and the alumina-pillared clay exhibited excellent Kdsfor 137Cs from simulated groundwater with values of 23,650, 23,260 and 144,570 mL/g, respectively. These were far better than the Kds obtained by clinoptilolite and AW500 which had Kds of only 14,560 and 9650 mL/g, respectively. None of the pillared clays showed a high selectivity for 89Sr from groundwater or l37Cs from simulated alkaline tank waste. They did, however, show a slight selectivity for 89Sr in the simulated Hanford tank wastes, but this is thought to be due to a precipitation mechanism rather than to ion exchange.

Physico-chemical and ion exchange characteristics of ammonium molybdophosphate irradiated with electrons

Physico-chemical changes accompanying the irradiation of ammonium molybdophosphate (AMP) with electrons up to a dose of 200 MGy have been studied by X-ray diffraction, SEM, EDX, thermal analyses, EPR and NMR measurements. The ion exchange characteristics of the control and irradiated AMP with respect to Cs+ have been studied using simulated acidic nuclear waste solutions in a column. Apparent equilibrium constants for the Cs+–H+ exchange have been calculated from breakthrough curves. Structural breakdown occurred on irradiating AMP to a dose of 100–200 MGy and was partially converted to MoO3 and ammonium molybdate. Although the useful exchange capacities of the irradiated AMP were reduced, the apparent ion exchange selectivity for Cs+ (with respect to H+) improved as compared to the control AMP.

Estimation studies of strontium in simulated nuclear waste solutions by WDXRF

Estimation of strontium has been studied by wavelength dispersive X-ray fluorescence spectrometery (WDXRF) as an analytical technique using Sr Kα as an analytical line. Lower limits of detection, precision, and intensities were plotted against counting time and concentrations to assess optimum time for measurement, expected scatter of the results, and calibration curves for intensity to concentration relationship. For its assessment the best equations determined by regression and least squares fitting along with a standard multiple addition technique were applied to a complex fuel admixture and simulated nuclear fuel solutions at a bum up of 650 GJ/Kg after a 0.5-year cooling time.