Treatment Of Radioactive Effluents Research Articles

Uranium capture from aqueous solution using palm-waste based activated carbon: sorption kinetics and equilibrium

Carbonaceous materials produced from agricultural waste (palm kernel shell) by pyrolysis can be a proper type of low-cost adsorbent for wide uses in radioactive effluent treatment. In this context, the as-produced bio-char (labeled as PBC) and its sub-driven sulfuric acid and zinc oxide activated carbons (labeled as PBC-SA, and PBC-Zn respectively) were employed as adsorbents for uranium sorption from aqueous solution. Various analytical techniques, including SEM (Scanning Electron Microscopy), EXD (X-ray Diffraction), BET (Brunauer–Emmett–Teller), FTIR (Fourier Transform Infrared Spectroscopy), and Zeta potential, provide insights into the material characteristics. Kinetic and isotherm investigations illuminated that the sorption process using the three sorbents is nicely fitted with Pseudo-second-order-kinetic and Langmuir isotherm models. The picked data display that the equilibrium time was 60 min, and the maximum sorption capacity was 9.89, 16.8, and 21.9 mg/g for PBC, PBC-SA, and PBC-Zn respectively, which reflects the highest affinity for zinc oxide, activated bio-char, among the three adsorbents, for uranium taking out from radioactive wastewater. Sorption thermodynamics declare that the sorption of U(VI) is an exothermic, spontaneous, and feasible process. About 92% of the uranium-loaded PBC-Zn sorbent was eluted using 1.0 M CH3COONa sodium ethanoate solution, and the sorbent demonstrated proper stability for 5 consecutive sorption/desorption cycles.

Photocatalytic Th(IV) removal: Unleashing the potential of amidoxime-modified graphitic carbon nitride photocatalyst

g-C3N4-AX photocatalyst was synthesized using a hydrothermal approach, with urea as a precursor, for the removal of Th(IV) in aqueous solution. The results of the experiment indicated that the addition of amidoxime to g-C3N4 altered its photocatalytic characteristics, resulting in a narrower band gap reduced from 2.75 eV to 2.7 eV, and faster formation of electron-hole pairs. The g-C3N4-AX photocatalyst exhibited remarkable uptake capacity with 71.09 mg.g−1 within 120 min. The excellent performance of g-C3N4-AX was mainly attribute to its effective absorption of visible light, facilitated by the narrow band gap. Therefore, this study proposes the potential of g-C3N4-AX can be efficiently adapted as a visible light-responsive catalyst for the application in radioactive effluent treatment.

Polysulfone-Gd2Zr2O7 mixed-matrix membranes with superior radiation resistant properties: Fabrication and application of a membrane device for radioactive effluent treatment

This is the first study undertaken towards development of mixed-matrix membranes (MMMs) with enhanced radiation resistant attributes by reinforcement of nanostructured Gd2Zr2O7 (GZO) within polysulfone (Psf) host-matrix. The study describes synthesis and characterization of GZO in disordered, defect-fluorite structure, having average crystallite size of 31(±3) nm. Membranes prepared with different loading of GZO (up to 2 w/w% of Psf) are exposed to γ-radiation up to a dose of 1000 kGy in aqueous environment. The effect of radiation on the structural, mechanical, and thermo-oxidative properties of MMMs has been compared with that of radiation-sensitive Psf membrane. The ultrafiltration performance of the (un)irradiated Psf and MMMs reveal that an optimum reinforcement of GZO at 1 (w/wPsf)% offers ~10 times radiation resistant MMM, compared to that of Psf membrane. The MMM with 1 (w/wPsf)% GZO was rolled into 2512 spiral configuration and a highly-compact device was developed for treatment of radioactive effluent. Based on the flux decline behaviour over 18 months duration, the fouling behaviour of the MMM module was modelled. The life-span of the MMM was predicted based on the absorbed radiation dose, GZO leaching study, and fouling behaviour. It is proposed that GZO mitigates a fraction of the impinged γ-energy in swapping of the Gd3+ and Zr4+ sites, which protects the polymeric host-matrix in-situ from radiation induced degradation. The abundance of Gd3+ and Zr4+ associated with polar O2− over the impregnated GZO further stabilizes MMMs through scavenging of the oxidizing and reducing radiolysed products of water. Thus, we report a smart approach to develop novel MMM with greater life-expectancy against high-energy radiation and further fabricate a membrane-based device for management of liquid radioactive waste.

Kinetics Studies on Free Radical Scavenging Property of Ceria in Polysulfone-Ceria Radiation Resistant Mixed-matrix Membrane

Cerium oxide (ceria) contains two stable states of cerium ions (Ce3+ and Ce4+). The presence of these two states, and the ability to swap from one state to another (Ce3+ ↔ Ce4+) by scavenging the highly reactive oxygen species (ROS) generated from radiolysis of water, ensures the enhanced stability of polysulfone (Psf) membranes in the γ-radiation environment. In this study, the ROS scavenging ability of ceria was studied. Ceria nanoparticles were found to scavenge hydroxyl radicals and hydrogen peroxide by mimicking peroxidase-like activity. This is responsible for offering protection to the Psf host matrix, and in turn the stability to the Psf-ceria mixed-matrix membranes (MMMs) in γ-radiation environment. Thus, presence of ceria nanoparticles provides an opportunity for the application of Psf-ceria MMMs in radioactive effluent treatment with increased life span, without compromise in the performance.

Enhancing [formula omitted]-radiation resistant property of polysulfone membranes with carboxylated nanodiamond: Impact and effect of surface tunability

Polysulfone (Psf) is the most commonly used polymer for the ultrafiltration membrane applications. Though it is regarded as high-performance engineering thermoplastic, it has γ-radiation stability only up to 100 kGy. When exposed to γ-radiation, the polymeric chains undergo scission and cross-linking, leading to a change in the cross-linking density and degradation in membrane performance. For the first time, carboxylated nanodiamonds (CND) are incorporated into Psf matrix in order to enhance the radiation resistant property of the membranes. Psf-CND composite membranes were fabricated with different loadings of CND (0.1–1 wt%) and exposed to different doses of γ-radiation up to 1000 kGy. The nanomaterials and (un) irradiated membranes were characterised using different instrumental techniques like FESEM, TEM, XPS, DSA, UTM, and DSC. It is claimed that CND helps in homogeneous dispersion of nanoparticles in the membrane matrix while protecting it from the radiation environment by scavenging of radiolysed products of water. The membrane with an optimum loading of 0.5% CND was found to be stable up to 1000 kGy of radiation dose which is 10 times as compared to that of control Psf membrane. The present study has significantly opened up opportunities for enhancing the deployment of polymeric membranes in radioactive effluent treatment.

U(VI) sorption during ferrihydrite formation: Underpinning radioactive effluent treatment.

Iron (oxyhydr)oxide nanoparticles are known to sorb metals, including radionuclides, from solution in various environmental and industrial systems. Effluent treatment processes including the Enhanced Actinide Removal Plant (EARP) (Sellafield, UK) use a neutralisation process to induce the precipitation of iron (oxyhydr)oxides to remove radionuclides from solution. There is a paucity of information on mechanism(s) of U(VI) removal under conditions relevant to such industrial processes. Here, we investigated removal of U(VI) from simulated effluents containing 7.16 mM Fe(III) with 4.2 × 10-4-1.05 mM U(VI), during the base induced hydrolysis of Fe(III). The solid product was ferrihydrite under all conditions. Acid dissolutions, Fourier Transform infrared spectroscopy and thermodynamic modelling indicated that U(VI) was removed from solution by adsorption to the ferrihydrite. The sorption mechanism was supported by X-ray Absorption Spectroscopy which showed U(VI) was adsorbed to ferrihydrite via a bidentate edge-sharing inner-sphere species with carbonate forming a ternary surface complex. At concentrations ≤0.42 mM U(VI) was removed entirely via adsorption, however at 1.05 mM U(VI) there was also evidence for precipitation of a discrete U(VI) phase. Overall these results confirm that U(VI) sequestered via adsorption to ferrihydrite over a concentration range from 4.2 × 10-4-0.42 mM confirming a remarkably consistent removal mechanism in this industrially relevant system.

Mechanisms of strontium removal by a Ba-titanate material for the wastewater treatment

Growing interest in radioactive effluent treatment, especially after the Fukushima nuclear accident, has led to the development of new, increasingly efficient mineral sorbents. The sorbents currently attracting the most attention are materials dedicated to strontium extraction. The Ba-titanate material investigated here, combining several sorption mechanisms, has a high capacity and selectivity for strontium with a distribution coefficient, Kd, of 863mLg−1, obtained from a batch sorption test with seawater ([Sr]=9×10−5molL−1; titanate concentration, 1gL−1). This Kd is six times higher than that of Na-titanate, one of the strontium sorbents used for multi-nuclide removal at the Fukushima-Daiichi nuclear power station. Upon contact with the sulfate-containing effluent, Sr extraction proceeds by ion exchange with the Ba ions inserted in the material and by coprecipitation with insoluble barium sulfate formed at the surface of the titanate grains.

The conceptual flowsheet of effluent treatment during preparing spherical fuel elements of HTR

High temperature gas-cooled reactor (HTR) is one of the advanced nuclear reactors owing to its inherent safety and broad applications. For HTR, one of the key components is the ceramic fuel element. During the preparation of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was established. According to the above treatment process, the uranium concentration was decreased from 200mg/l to the level of discharged standard.

Biosorption of radionuclides by Rhizopus arrhizus

The sorption of various radionuclides viz. 233U, 239Pu, 241Am, 144Ce, 147Pm, 152+154Eu and 95Zr from aqueous nitrate medium has been studied using biomass Rhizopus arrhizus. The biosorption of 233U and 239Pu was found to be maximum at pH 6-7 whereas for other trivalent actinides and fission products viz. 241Am, 144Ce, 147Pm, 152+154Eu and tetravalent 95Zr, it was more effective at pH 2. This biomass is a promising sorbant for the treatment of radioactive effluents from nuclear industry. © Rapid Science Ltd. 1998

Radioactive waste treatment using natural Syrian bentonite

The uptake of137Cs and90Sr/90Y onto Syrian bentonite has been studed, using batch and column tracer techniques in order to investigate its utilization for aqueous radioactive effluent treatment. Parameters influencing the percent uptake considered and studied in this work are: nuclide concentration, pH of the aqueous phase, heat treatment and particle size. Leaching experiments using natural sea and ground waters were carried out on bentonite/cement composites. They demonstrated the effectiveness of calcination and cement containment.