Radioactive Waste Solutions Research Articles

Radionuclide Solid:liquid partitioning in an aged, reducing-grout wasteform recovered from a disposal facility

The Saltstone Disposal Facility on the Savannah River Site in South Carolina disposes of Low-Level Waste in a reducing-grout waste form. Reducing grout is presently being evaluated as a subsurface disposal waste form at several other locations in the United States, as well as in Europe and Asia. The objective of this study was to collect core samples directly from the Saltstone Disposal Facility and measure desorption distribution coefficients (Kd; radionuclide concentration ratio of saltstone:liquid; (Bq/kg)/Bq/L)) and desorption apparent solubility values (ksp; radionuclide aqueous concentration (moles/L)). An important attribute of this study was that these tests were conducted with actual aged, grout waste form materials, not small-volume simulants prepared in a laboratory. The reducing grout is comprised of blast furnace slag, Class F fly ash, ordinary portland cement, and a radioactive salt waste solution generated during nuclear processing. The grout sample used in this study underwent hydrolyzation in the disposal facility for 30 months prior to measuring radionuclide leaching. Leaching experiments were conducted either in an inert (no oxygen) atmosphere to simulate conditions within the saltstone monolith prior to aging (becoming oxidized) or they were exposed to atmosphere conditions to simulate conditions of an aged saltstone. Importantly, these experiments were designed not to be diffusion limited, that is, the saltstone was ground finely and the suspensions were under constant agitation during the equilibration period. Under oxidized conditions, measured Tc Kd values were 10 mL/g, which was appreciably greater than the historical best-estimate value of 0.8 mL/g. This difference is likely the result of a fraction of the Tc remaining in the less soluble Tc(IV) form, even after extensive oxidation during the experiment. Under oxidized and reducing conditions, the measured Ba and Sr (both divalent alkaline earth metals) Kd value were more than an order of magnitude greater than historical best-estimate values of 100 mL/g. The unexpectedly high Ba and Sr Kd values were attributed to these radionuclides having sufficient time to age (form strong bonds) in the sulfur-rich saltstone sample. Apparent ksp values under reducing conditions were 10−9 mol/L Tc and 10−13 mol/L Pu, consistent with values measured with surrogate materials. Measured apparent Ba, Sr, and Th ksp values were significantly greater than historical best-estimates. The implications of the generally greater Kd values and lower ksp values in these measurements is that these cementitious waste forms have greater radionuclide retention than was previously estimated based on laboratory studies using surrogate materials. This work represents the first leaching study performed with an actual aged, reducing-grout sample and as such provides an important comparison to studies conducted with surrogate materials, and provides high pedigree data for other programs around the world evaluating reducing grouts as a wasteform for subsurface nuclear waste disposal.

Biosorption Capability of Chitosan for Removal of Cs-137 and/or Co-60 from Radioactive Waste Solution Simulates

Biosorption Capability of Chitosan for Removal of Cs-137 and/or Co-60 from Radioactive Waste Solution Simulates

Synthesis and Application of MnO-Fe2O3 Nanocomposites for the Removal of 137Cs and 60Co Radionuclides from Artificial Radioactive Aqueous Waste

For innovative application in wastewater treatment techniques, MnO-Fe2O3 nanocomposites were successfully synthesized using the sol–gel auto-combustion method at different temperatures for the adsorption of 137Cs and 60Co radionuclides from aqueous solution. The characterization of these nanocomposites was carried out through FT-IR, SEM-EDX, and X-ray diffraction. These nanocomposites were employed as adsorbent materials for the removal of 137Cs and 60Co radionuclides from simulated radioactive waste solutions. The study involved a series of experiments aiming to demonstrate the MnO-Fe2O3 nanoparticles’ exceptional adsorption potential concerning 137Cs and 60Co. Additionally, the investigation delved into how variations in temperature, dose amount, contact time, and pH value influence the adsorption dynamics. Due to their high specific surface area, the synthesized MnO-Fe2O3 nanoparticles had high adsorption capacity of more than 60% and 90% for 137Cs and 60Co, respectively. By investigation of kinetics and adsorption isotherms, pseudo-second-order reaction and the Langmuir model turned out to fit well for the adsorption of 137Cs and 60Co onto the MnO-Fe2O3 nanocomposites. Moreover, a thermodynamic analysis revealed that the adsorption process was spontaneous for both target metals and the adsorption of 60Co was endothermic, whereas the adsorption of 137Cs was exothermic.

Utilizing deep eutectic solvent for facile, selective and sustainable sequestration of plutonium

In the recent surge of green and cheap energy for future generation; nuclear industry is going to play a pivotal role and reprocessing of nuclear fuel through safe and systematic approach is the key. In that direction recent literature has shown that there is lack of better and improved methods towards efficient extraction of highly radiotoxic and hazardous plutonium from nuclear waste. Herein, we have utilized a greener and cost effective indigenously synthesized deep eutectic solvent (DES) for the selective extraction of plutonium from (Am, Pu, U) waste. The maximum D value achieved for Pu was 67.42 at 5 mol L−1 HNO3 media and the equilibrium uptake capacity was found to be 107.28 ± 5 μg of Pu/g of C12DES. This was followed by solidified floating organic drop micro-extraction (SFODME) technique for its quantification. High radiation stability of this DES (upto 750 kGy) highlighted its great potential in nuclear fuel reprocessing. Several other reprocess parameters were studied and optimized such as DES volume, alkyl chain length attached to hydrogen bond acceptor (HBA), actinide concentration and speciation and its acidity to achieve best extraction of plutonium ion form the waste stream. Regeneration capability and reusability of the DES employing stripping agents gave encouraging results as 95% of Pu could be stripped in a single cycle using 1 M hydroxylamine hydrochloride as the stripping agent. The developed method was applied for recovery and quantification of Pu from real samples, viz. (U,Pu) Carbide fuel solution, radioactive research waste and synthetic dissolver solution. We believe this work is a way forward in resolving a variety of issues that exist in health physics and reprocessing laboratories of nuclear installation pertaining to selective and efficient extraction of Pu (IV).

An adsorption agent based on chitosan–zeolite composite: environmental and radioactive liquid waste remediation

Chitosan–zeolite composite as a low-cost and efficient material in the removal of caesium in synthetic and liquid organic radioactive waste solutions.

Adsorption of 60Co(II) and 152+154Eu(III) radionuclides by a sustainable nanobentonite@sodium alginate@oleylamine nanocomposite

A sustainable and efficient nanobentonite@sodium alginate@oleylamine (Nbent@Alg@OA) nanocomposite has been successfully synthesized via coating reaction of nanobentonite (Nbent) with alginate (Alg) and oleylamine (OA). The nanocomposite has been characterized and examined for the adsorption of 60Co(II) and 152+154Eu(III) radionuclides from simulated radioactive waste solution. FT-IR, XRD, SEM, and HR-TEM techniques have been applied to confirm the structural and morphological characteristics of the Nbent@Alg@OA nanocomposite. The effects of various parameters, such as pH of the medium, initial concentration of the radionuclides, contact time, and temperature on the adsorption of 60Co(II) and 152+154Eu(III) radionuclides were investigated by the batch adsorption technique. The results revealed that the optimum pH values for the adsorption of 152+154Eu (III) and 60Co (II) radionuclides were 4 and 5, respectively. The adsorption capacity of 152+154Eu(III) (65.6219 mg/g) was found greater than that of 60Co(II) (47.3469 mg/g). The adsorption process was found to be well described by the pseudo-second-order kinetic model. Furthermore, the equilibrium isotherm evaluation revealed that the Langmuir model was adequately matched with the adsorption data. According to the thermodynamic characteristics, the adsorption process was endothermic and spontaneous. Regeneration and reuse of Nbent@Alg@OA nanocomposite confirmed that the recycled nanocomposite was sufficiently efficient in several successive practical applications.

Fast Room‐Temperature Synthesis of an Extremely Alkaline‐Resistant Cationic Metal–Organic Framework for Sequestering TcO4− with Exceptional Selectivity

AbstractLeveraging metal–organic framework (MOF) to eliminate radioactive contaminants has invariably received much attention, but low preparation efficiency and poor selectivity have still limited its actual application. Herein, it is found that a fourfold interpenetrated cationic MOF (Ag‐TPPE) can be rapidly synthesized by only mixing, stirring, or sonication at room temperature. More impressively, the preparation process can be completed in <1 min. Up to now, this is the first report that cationic MOFs can be obtained by directly mixing the corresponding raw materials at room temperature. In addition to holding the rare merits of structural stability in extremely strong bases (8 m NaOH), Ag‐TPPE can selectively remove TcO4− in the presence of large excess SO42− or NO3−. Based on its ultra‐high selectivity, Ag‐TPPE exhibits exceptional removal rate for TcO4− from simulated Hanford and Savannah River Site waste streams, which refreshes the record of selective sorption of TcO4− at low solid/liquid ratio, overcoming the drawback of overused sorbent in treatment of radioactive waste solution. Such superior sorption capabilities are thoroughly elucidated by the density functional theory calculations on a molecular level, clearly disclosing that TcO4− can enter into the framework through breathing effect and is trapped in the large cavity through dense hydrogen bonds.

Efficient enrichment of U(VI) by two-dimensional layered transition metal carbide composite

Abstract With the rapid development of nuclear energy, how to safely and efficiently dispose of radioactive waste solution has become an urgent environmental problem of public concern. It is of great significance to construct a new type of high-efficiency adsorbent material to recover uranium from nuclear waste solution. In this work, the Ti3C2Tx material (an emerging two-dimensional inorganic layered material) with a stable layered structure was used as the matrix, and the amidoxime functionalized MXene composite material (PAO/Ti3C2Tx) was synthesized by in-situ polymerization. The amidoxime-functionalized Ti3C2Tx showed excellent capacity to capture U(VI), with a maximum adsorption capacity of 98.04 mg/g at 25 °C, which was significantly better than that of Ti3C2Tx, and the adsorption selectivity for U(VI) was greatly improved. The adsorption was conformed to Langmuir isotherm model and pseudo-second-order kinetic model. In addition, the adsorbed UO22+ could be effectively desorbed by 0.1 M HNO3, and the adsorption performance of PAO/Ti3C2Tx did not decrease significantly after 5 adsorption/desorption cycles. The results of ionic strength experiment, FT-IR, SEM, and XPS jointly indicated that adsorption mechanism of U(VI) on PAO/Ti3C2Tx was the combined effect of the amidoxime group and -O and -OH active groups on the surface of Ti3C2Tx, mainly inner complexation. These advantages make PAO/Ti3C2Tx composite a highly potential U(VI) adsorbent with great application prospects.

Nitric acid concentration strongly influences low level uranium determination on PEDOT-PSS coated glassy carbon electrodes.

Understanding the parametric optimization and addition of modifications in existing techniques are the keys to successful research in the analytical sciences. The present study reports an electroanalytical technique by modifying an electrode with PEDOT-PSS and adjusting the analyte solution to lower acidity (0.05 M) levels. The proposed methodology achieved a detection limit of 1.24 ppb of uranium, which is far below the available concentrations of uranium in seawater; the WHO permissible limit for uranium in drinking water; and the allowed disposal limit for uranium in radioactive waste solutions. Redox characteristics and speciation analysis supported by theoretical prediction helped to understand the basic mechanism which lay behind the achievement of such low detection limits. The technique was successfully employed for uranium determination in real water samples and the results were validated by another independent technique (ICP-MS).

Uranium sorption from waste solutions by Talc Phosphogypsum ferri-silicate synthetic new sorbent

Abstract In this investigation, a synthetic Talc Phosphogypsum ferri-silicate TPFS sorbent was prepared by thermal activation then evaluated the uranium ions removal from sulfate waste solution containing uranium. Generally, the synthetic adsorbents from raw and waste materials have a significant attention from scientists because the environmental concern and economic development, particularly, the uranium elimination from radioactive waste solutions. The uranium removal percentage and loading capacity were determined by optimization the conditions of adsorption such as the pH range, adsorbent/adsorbate ratio, uranium concentration of radioactive waste solutions, equilibrium time and temperature. The resultant adsorption efficiency and loading capacity were 87.2% and 375 mg g−1, respectively. The adsorption isothermally was in accordance with Langmuir isotherm model, in addition pseudo-second-order kinetic model, with theoretical capacity of 384.6 and 333 mg g−1, respectively. Uranium (VI) adsorption on TPFS was inhibited at elevated temperatures. The removal of uranium from sulfate waste solution by TPES sorbent according to the thermodynamic functions values was exothermic (∆H of −16.095) and non-spontaneous in nature (∆G of −17.27 at 303 K). In addition, there was a decrease in the randomness at the TPFS/uranium waste solution interface with ∆S value of 3.88.

Potential application of graphene oxide and Aspergillus niger spores with high adsorption capacity for recovery of europium from red phosphor, compact fluorescent lamp and simulated radioactive waste

The recovery of rare earth elements (REEs) is a global challenge and the mining of rare earths has serious environmental implications due to the toxic waste released post mining. Hence, the rising demand for rare earths and their far reaching electronic applications necessitates an effective strategy to recover the REEs from more viable sources. In this work, the graphene oxide–Aspergillus niger spores (GO–A. niger spores) blend was utilized for adsorptive recovery of a precious rare earth Eu(III) and the adsorption variables like pH of the medium, adsorbent dosage, sorption kinetics, thermodynamics, and isotherm were optimized for the developed biosorbent. The adsorption process suits the Langmuir isotherm model with a maximum adsorption capacity of 147.3 mg/g. The pseudo–second–order kinetics is a perfect fit to describe the adsorption process. The results obtained through the Van't Hoff plot show negative free energy change (ΔG0) which implies the spontaneity of the adsorption process. The negative standard enthalpy change (ΔH0) values show that the nature of the adsorption process is exothermic. The analytical characterizations including Fourier transform infrared spectroscopy (FTIR), Raman, X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Brunauer-Emmett-Teller (BET), thermal gravimetric analysis (TGA), and scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDX) were employed to study the biosorbent. The features of GO–A. niger spores biosorbent were applied to recover Eu(III) from real samples such as fluorescent lamp phosphor, red phosphor powder, and a simulated radioactive waste solution.

Pure and Sb-doped ZrO2 for removal of IO3− from radioactive waste solutions

Radioactive 129I with a long half-life (1.57 × 107 y) and high mobility is a serious radiohazard and one of the top risk radionuclides associated with its accidental and planned releases to nature. The complex speciation chemistry of iodine makes its removal a complicated task, and usually a single method is not able to remove all iodine species. Especially its oxidized form iodate (IO3−) lacks a selective and effective removal method. Here, the granular aggregates of hydrous zirconium oxides with and without antimony doping were tested for IO3− removal and the effects of contact time, competing anions in different concentrations and pH were examined. The materials showed high selectivity for IO3− (Kd over up to 50,000 ml/g) in the presence of competing ions and relatively fast uptake kinetics (eq. < 1 h). However, B(OH)4− and SO42−, as competing ions, lowered the iodate uptake significantly in basic and acidic solution, respectively. The suitability of the materials for practical applications was tested in a series of column experiments where the materials showed remarkably high apparent capacity for the IO3− uptake (3.2–3.5 mmol/g).Graphic abstract

Synthesis and characterization of Chitosan-sodium titanate nanocomposite beads for separation of radionuclides from aqueous radioactive waste

This paper reports development of a new nanocomposite “Chitosan-sodium titanate’ containing 50% of sodium titanate. SEM and TEM/EDS analysis confirmed formation of homogeneously dispersed un-agglomerated product. Ion exchange property of the composite is evaluated towards separation of strontium from simulated radioactive waste solution. The synthesized composite has been used in a fixed bed column for separation of strontium and 5000 bed volumes of simulated waste were decontaminated. It is inferred that the synthesized composite has potential for the separation of Sr from alkaline waste streams.

Microsphere Sorbents Based on Cenosphere Supported Zirconium Molybdates and Zirconium Silicates for Cesium-137 and Strontium-90 Removal from Radioactive Waste Solutions

Microsphere Sorbents Based on Cenosphere Supported Zirconium Molybdates and Zirconium Silicates for Cesium-137 and Strontium-90 Removal from Radioactive Waste Solutions

Microsphere Sorbents Based on Cenosphere Supported Zirconium Molybdates and Zirconium Silicates for Cesium-137 and Strontium-90 Removal from Radioactive Waste Solutions

Microsphere Sorbents Based on Cenosphere Supported Zirconium Molybdates and Zirconium Silicates for Cesium-137 and Strontium-90 Removal from Radioactive Waste Solutions

Synthesis of chitosan hydrogel polymer for removal of radioactive organic research waste prior to treatment

ABSTRACT Chitosan hydrogel polymer (CHGP) has been exploited as the second most abundant natural material in the environment for preparing a new adsorbent with a low cost, higher removal capacity of dyes, good stability overworking pH range and environmentally safe for treating aqueous solutions of organic pollutants Arsenazo-III (Ar-III) and Alizarin Red S (ARS). The characteristics of the prepared polymer were revealed through infrared analysis (FTIR) as well as a scanning electron microscope (SEM). Various factors affecting the treatment process were studied through batch experiments. These factors are the agitation time, pH, different polymer weights, and the initial concentrations of Ar-III and ARS dyes in the solution plus temperature. The Freundlich and Langmuir isothermal models were applied and the equilibrium data showed that they fit well with the Langmuir model for both the reagents removal processes. Based on this model, the maximum capacity values were estimated at 99.9 and 62.5 mg/g for Ar-III and ARS at ambient temperature, respectively. In terms of kinetics, we find that the adsorption process of both dyes onto the prepared CHGP polymer corresponds to the pseudo-first-order mechanism. Also, a suggested reaction mechanism was established based on the electrostatic, hydrogen bonding, and mainly pH. Through practical experience the optimum conditions for efficient removal of both Ar-III and ARS reagents (dyes) from contaminated radioactive solutions by using CHGP are the initial dye concentration 100 mg L-1, pH = 6, shaking time 120 min. and the polymer dose 0.01 g at ambient temperature. Finally, the results showed that the prepared polymer (CHGP) could be promisingly applied in removing both Ar-III and ARS dyes from radioactive waste solutions.

Capability of Aspergillus niger to bioconcentrate cesium-137 and cobalt- 60 from medium and low level radioactive waste solution simulates

Introduction: The bioremediation, as a treatment process for low- and medium radioactive liquid wastes generated from various applications of nuclear technology in our daily life, represents a biotechnological innovation as well as an excellent tool for bioconcentration of radionuclides. Aim: study the capability of Aspergillus niger (A. niger) to removal of Cesium-137 (Cs-137) and/or Cobalt - 60 (Co-60) from simulated spiked solutions. Materials and Methods: This part of experimental work was carried out to evaluate the factors influencing the performance of the fungus and optimizing the bioconcentration of those two radionuclides. The impacts of incubation periods, irradiation of fungus spores prior seeding, age of A. niger and the initial concentrations of the radioactivity added on the bioconcentration factor (CF) for the two radioions were studied on laboratory scale experiments. Results and Discussion: Based on the data reached it was found that A. niger can bioconcentrate, from the spiked solution, more than 80% of Co-60 and about 25% of Cs-137 within 48 hours. Conclusion: Therefore, Aspergillus niger can be easily grown in substantial manner using unsophisticated techniques, inexpensive growth media. Besides, the acceptable bioconcentration factor achieved by the fungi constituted an economical pattern for treating hazardous effluents spiked with radioactive ions, and it might have a synergetic role during applying other treatment methods. Keywords: Aspergillus niger, Bioconcentration factor, Co-60, Cs-137.

Removal of Cs+ and Sr2+ ions from simulated radioactive waste solutions using Zeolite-A synthesized from kaolin and their structural stability at high pressures

Zeolite-A was synthesized from kaolin activated at 500 °C at relatively lower temperature than the studies reported in the literature.The metakaolin formed at 500 °C was used to synthesise zeolite-A by treating with 1–5 M NaOH solutions at 80–100 °C for 3–24 h under hydrothermal reaction condition. The zeolite-A was characterized by XRD, FT-IR and SEM. The percentage of crystallinity and crystallite sizes of the zeolite-A were determined from XRD analysis; the crystal system of zeolite-A was found to be cubic with space group (Fm3‾c ). The ion-exchange property of zeolite-A for Cs+ and Sr+2 was explored at ambient temperature (25 °C). The adsorption efficiency of zeolite-A is higher for Sr+2 than that of Cs+; and adsorption capacities for Cs+ and Sr2+ ions were found to be 91.51% and 99.16% respectively. Strontium substituted zeolite has higher crystallinity and lattice strain than that of caesium substituted zeolite-A. Structural stability investigations were carried out on pristine zeolite-A and its Cs+ and Sr2+ substituted analogues by using high-pressure (0.1 MPa-11.4 GPa) X-ray diffractometry technique for the first time. The pristine zeolite-A shows comparatively lower bulk modulus than that of caesium and strontium loaded zeolite. It is also seen that the compressibility is least for the strontium loaded zeolite-A (SrZtA) followed by caesium loaded zeolite-A (CsZtA) and pristine zeolite-A (ZtA) (i.e compressibility of SrZtA < compressibility of CsZtA < compressibility of ZtA).

Adsorptive removal of PAR and Arsenazo-III from radioactive waste solutions by modified sugarcane bagasse as eco-friendly sorbent

Abstract In this paper, sugarcane bagasse (SCB) was modified using phosphoric acid. The modified sugarcane bagasse (MSCB) has been used to remove 4-(2-pyridylazo)resorcinol (PAR) and Arsenazo-III (Ar-III) from liquid radioactive waste. The surface morphology and functional groups of the MSCB were studied using scanning electron microscopy (SEM), X-ray diffraction (XRD), fourier transform infrared spectroscopy (FTIR), and thermogravimetric analysis (TGA). Adsorption ability of MSCB has been tested by batch mode through some relevant factors like initial pH, reaction time, initial coloring reagents (PAR and Ar-III) concentrations, and adsorbent weight. At adsorption equilibrium time 180 min and pH values of 3 and 1 for PAR and Ar-III; the maximum removal (%) for both PAR and Ar-III were 93 and 57%, respectively. The adsorption isotherm data are representative well to Freundlich isotherm model. The mean free energy of adsorption, E (kJ/mol), has been estimated as 5.75 and 2.28 kJ/mol for PAR and Ar-III, respectively, which suggests that the adsorption occurred physically. The maximum adsorption capacity of MSCB for PAR and Ar-III is 96.62 and 15.18 mg/g, respectively. The adsorption kinetics are better fitted by the pseudo-second-order model. The partial film along with intra-particle diffusion controlled the diffusion of coloring reagents from the solution bulk to the particle interior pores. Application of MSCB for removing PAR and Ar-III from simulated liquid radioactive waste containing U(VI) and Th(VI) ions has been achieved successfully.

All-chromatographic method for the recovery of Americium-241 from solutions of complex composition

This paper reported creating the all-chromatographic technique for the recovery of americium radionuclides from solutions of the complex composition generated in the radiochemical industry. The technique was based entirely on the displacement complexing chromatography; neither additional precipitation nor extraction stages were necessary. The technology used the elution with DTPA and NTA-based solutions containing the citric acid additives and the sequential elution with these eluents, emphasizing the separation of americium from the heavy rare earth elements, lead, and cadmium. The complete separation of americium from lead and cadmium was a peculiar exercise; the elution order of lead and cadmium changed in the presence of the citric acid additives. The recovery and separation of americium from actual radioactive waste solutions of different origins demonstrated 96 and 99% chemical purity of the final product.