Radioactive Liquid Waste Treatment Research Articles

In situ electrochemical synthesis of Ni–Mg–Al–LDHs for the treatment of simulated strontium-containing liquid radioactive waste from nuclear power plants

In situ electrochemical synthesis of Ni–Mg–Al–LDHs for the treatment of simulated strontium-containing liquid radioactive waste from nuclear power plants

Radioactive strontium ions sieving through reduced graphene oxide membrane

Partitioning of 90Sr from other radionuclides in radioactive liquid waste is crucial for treating high-level liquid waste (HLLW) generated during spent fuel reprocessing, as the unique high heat-generating and long half-life of 90Sr pose a significant challenge for the long-term storage of nuclear waste. However, it remains difficult to sieve 90Sr for membrane based on size-selective effect, as the radioactive liquid waste contains numerous typical divalent ions, such as 60Co, 63Ni, and 65Zn, which have the same valence and similar hydration diameters to 90Sr. Here, we experimentally demonstrate a rapid and efficient separation of radioactive 90Sr from other radioactive ions by a reduced graphene oxide membrane, which prepared by the amino-hydrothermal method from graphene oxide suspensions. For SrCl2/CoCl2, SrCl2/NiCl2, and SrCl2/ZnCl2 binary mixed solutions, SrCl2 can pass through efficiently, while the rejection rates for CoCl2, NiCl2, and ZnCl2 reached 99.9 %. The separation factors reached 3232, 4254, and 3314, which are about one or two orders of magnitude higher than the highest reported separation factors for membranes sieving divalent ions. Our findings offer a simple avenue for the precise treatment of radioactive liquid waste, which is vital for the advancement of nuclear energy technology.

Highly efficient and selective 99TcO4 – uptake from reprocessing low-level liquid waste using a zirconium-based metal−organic framework

ABSTRACT Efficient and selective sequestration of pertechnetate ions (99TcO4 –) from nuclear low-level waste (LLW) solution is very much desirable for the safe treatment of radioactive liquid waste. However, the separation of specific radionuclides from competitors is a huge challenge in separation science. In the nuclear fuel cycle, 99TcO4 – is one of troublesome fission products due to long half-life, large inventory in spent fuel, significant radiation hazards, and high environmental mobility of 99TcO4 – ions. Recently, the metal−organic framework (MOFs) has shown significant potentials toward 99TcO4 – removal. So far, most of the studies were reported using ReO4 – ions, an inactive surrogate of 99TcO4 – ions, and very few MOF materials are applied for real-time 99TcO4 – sequestration from nuclear waste solution in the presence of extreme conditions. Herein, we report 99TcO4 – ion removal from low-level liquid nuclear waste solution using zirconium-based metal−organic framework UIO-66-NH2. The material has exhibited fast kinetics and record selectivity toward 99TcO4 – ions, making the adsorbent highly promising.

Selective Separation of Zr(IV) from Simulated High-Level Liquid Wastes by Zeolites

ABSTRACT Effectively separating Zr(IV) from strong acidic and radioactive solutions is crucial for spent fuel reprocessing plants, but it remains a challenging task. This study investigated the adsorption of Zr(IV) in HNO3 solutions using zeolites as adsorbents. The zeolites were characterized by X-ray diffraction (XRD)，scanning electron microscopy (SEM), transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). The effects of adsorption time, HNO3 concentration, the initial concentration of Zr(IV), zeolites dosage, and temperature on the Zr(IV) adsorption behaviors were investigated. The results demonstrated that HY zeolite had a higher adsorption capacity for Zr(IV) than H-SSZ-13 zeolite. The adsorption capacity of HY zeolite was influenced by its Si/Al ratio, which determined the density of adsorption sites. The optimal HY zeolite (HY-25) exhibited a maximum adsorption capacity of 30.438 mg/g in 3 M HNO3 solution. Furthermore, the adsorption isotherms and kinetics of Zr(IV) adsorption were investigated. The adsorption of Zr(IV) on zeolites was endothermic and spontaneous, in accordance with the Freundlich’s isotherm model and pseudo-second-order kinetic model. In a simulated strong acidic solution of Zr(IV) and 10 co-existing cations (Ag+, Ba2+, Cs+, Ce3+, Eu3+, Fe3+, La3+, Nd3+, Sm3+, Sr2+), HY-25 exhibited good selective adsorption of Zr(IV), indicating its potential application in the treatment of high-level radioactive liquid waste.

Self-cleaning and regenerable nano zero-valent iron modified PCN-224 heterojunction for photo-enhanced radioactive waste reduction.

The expansion of large-scale nuclear power causes a substantial volume of radioactive wastewater containing uranium to be released into the environment. Because of uranium's toxicity and bioaccumulation, it is critical to develop the efficient and sustainable materials for selective removal of uranium (VI). Herein, a regenerable anti-biofouling nano zero-valent iron doped porphyrinic zirconium metal-organic framework (NZVI@PCN-224) heterojunction system was successfully fabricated. Due to the Schottky-junction effect at the NZVI/MOF interface, the NZVI nanomaterial immobilized on PCN-224 could improve interfacial electron transfer and separation efficiency, and enhance entire reduction of highly soluble U(VI) to less soluble U(IV), involving photocatalytic reduction and chemical reduction. Meanwhile, the photocatalytic effect also prompts the NZVI@PCN-224 to produce more biotoxic reactive oxygen species (ROS), resulting in high anti-microbial and anti-algae activities. Under dark conditions, NZVI@PCN-224 with a large specific surface area could provide sufficient oxo atoms as the uranium binding sites and show the highest uranium-adsorbing capability of 57.94mg/g at pH 4.0. After eight adsorption-desorption cycles, NZVI@PCN-224 still retained a high uranium adsorption capacity of 47.98mg/g and elimination efficiency (91.72%). This sorption/reduction/anti-biofouling synergistic strategy of combining chelation, chemical reduction and photocatalytic performance inspires new insights for highly efficient treatment of liquid radioactive waste.

TREATMENT OF RADIOACTIVE LIQUID WASTE USING SUN HEAT

This research aims to study the benefits of using the renewable energy in the field of treatment of liquid radioactive waste which is contaminated with a mixture of radionuclides such as Cesium Cs-137, Uranium U-238, Radium Ra-226,etc. which their evaporation degree is too high (more than 600C0 ) . The sun energy as example for renewable energy is used in the process of evaporation of liquid radioactive waste which formed from the decontamination by washing metallic pipes and washing a contaminated soil to convert liquid radioactive waste to more stable and less size radioactive waste in a form of sediment which treated later using cement and then to disposal. The effect of evaporation using sun energy on the human and environment is studied in the laboratory using chemical and radioactive measurement devices. A very good results had been gotten because no level of a radioactive contamination had appeared on the filter which had used in the evaporation process. An electric heater also had used for evaporation process and the results were the same which refers that radioactive nuclides weren't evaporate because of using low temperature. Liquid radioactive waste had distillated, the distillated liquid was measured, and it was clear from any contamination and this shows that there is no diffusion of radionuclides during evaporation process, and in this process the Decontamination Factor (DF) is very high.

Catalytic effect of cesium on the oxidation behavior of cation exchange resins in Li2CO3-Na2CO3-K2CO3 melt.

After the treatment of liquid radioactive waste, there is a certain amount of Cs in the waste resin, and these Cs-doped resins are prone to volatilize during the thermal treatment process and cause radionuclide leakage. The molten salt oxidation (MSO) can effectively prevent the volatilization of toxic metal, especially the volatilization of Cs. Under nitrogen and air conditions, it is found that the oxidation behavior between Cs-doped and clean cation exchange resins (CERs) is quite different. In the presence of oxygen and molten carbonate salt, Cs2CO3 is generated by the destruction of functional groups in Cs-doped CERs. The Cs2CO3 in Na2CO3-K2CO3-Li2CO3 reacts with oxygen to form Li2O2, which reduces the content of S in residue from 26.33 to 13.38% in air conditions at 400 °C and promotes the generation of sulfate in the molten carbonate salt. The elements Cs and S in the Cs doped CERs spontaneously form thermally stable Cs2SO4 in the molten carbonate salt.

Application of the radiotracer method to study the fouling of tubular microfiltration membranes

This paper presents a study on the fouling of ceramic microfiltration (MF) membrane. Such membranes are often used in the treatment of liquid radioactive waste. In the investigation of the phenomenon of fouling, a radiotracer method with a short-lived lanthanum isotope (140La) is applied. The effect of the molecular weight (MW) of the complexing agent, used to bind the radionuclides in the MF-enhanced process, was studied in radiotracer experiments. Poly(acrylic acid) with a MW of 30 kDa, 100 kDa and 250 kDa is used as the complexing agent. The distribution of particles with different MWs at various positions along the membrane module is examined. The test results show that the particles with the lowest MW (i.e. 30 kDa) blocks the membranes the quickest, causing the greatest decrease in the permeate flux. However, the amount of deposit formed on the membrane, which is reflected by the level of radioactivity measured, is the lowest in this case. This indicates the existence of different membrane blocking mechanisms, which depend on the size of the particles contained in the filtered medium.

Ultrahigh Water Permeance of Reduced Graphene Oxide Membrane for Radioactive Liquid Waste Treatment.

Membrane methods exhibit great potential for application in radioactive liquid waste treatment. In this work, we prepared a reduced graphene oxide using the amino-hydrothermal method (AH-rGO) that exhibited effective rejection rates of 99.9% for CoCl2, ZnCl2, NiCl2, and radionuclide 60Co solutions with an ultrahigh water permeance of >71.9 L m−2 h−1 bar−1. The thickness of the AH-rGO membranes affects the water permeance, as the membrane with a thickness of ≈250 nm has the highest water permeance of up to 125.1 L m−2 h−1 bar−1 with the corresponding rejection rate of 86.8%. Importantly, this is the most permeable membrane with a satisfactory level of the rejection rate for typical radioactive ions of Co2+, Zn2+, and Ni2+. Moreover, the AH-rGO membranes presented excellent stability. These findings demonstrate the potential of reduced graphene oxide (rGO) membranes for radioactive liquid waste treatment.

A low-cost photo-evaporation inorganic membrane preparation and treatment of the simulated high salinity radioactive waste water

Solar-driven desalination is an energy-saving and environmentally benign wastewater treatment technology. A method of in situ self-reduction of graphene oxide (rGO) by cheap geopolymer was introduced, and a photo evaporation membrane device (rGOPGC) for treatment of the simulated high salt liquid radioactive waste (HSLRW) was prepared in the present study. Compared with other rGO based photo evaporation membrane materials, geopolymer matrix has the advantages of low cost, reductant free, simple preparation process and mild conditions. After desalination of simulated seawater, the concentrations of Na+, K+, Ca2+ and Mg2+ ions reached the WHO standard, and the removal rates of radioactive I-, Cs+ and Sr2+ in the simulated high salinity wastewater reached 99.62%, 99.71% and 99.99% respectively; The evaporation rate of rGOPGC remained stable at 1.5 kg/m2/h after 16 cycles in high salinity environment. There was no obvious salt accumulation on the upper surface of the device, indicating its high stability. Furthermore, the evaporation performance at high temperature near the nuclear power plant (NPP) waste water was simulated and tested. Under one solar intensity and 35 °C ambient temperature, the evaporation rate of 1.75 kg/m2/h and the evaporation efficiency of 98.51% were achieved. The results indicated that the rGOPGC device is potential in the concentration evaluation of HSLRW.

Removal of cesium from liquid radioactive waste by in situ electrochemical synthesis of cesium zinc ferrocyanide

Cesium ion in the simulated liquid radioactive waste was treated with cesium zinc ferrocyanide (CsZnFC) precipitate synthesized by in situ electrochemical synthesis method (ISESM) in this study. The best removal rate of Cs+ was 99.42%, and the residual Cs+ in the effluent was 29.13 μg L‐1 under the optimal experimental conditions of initial pH (pHint) of 7.0, reaction temperature (T) of 35 °C, current density (J) of 0.7 mA cm−2, electrolytic time (t) of 30 min, and molar ratio of [Fe(CN)6]4- to Cs (RFCN/Cs) of 1.0. The synthesized precipitate was characterized by SEM-EDS, XRD and FT-IR. The precipitated particles are cubic with a particle size of around 190 nm. The XRD pattern shows that the phase composition of the synthesized compound is Cs2Zn[Fe(CN)6], and the absorption peaks of O–H bonds, C–N bonds, and C–Fe bonds appear in IR spectrum. The method of in situ electrochemical synthesis of CsZnFC is promising for the treatment of liquid radioactive wastes.

IAEA-Assisted Treatment of Liquid Radioactive Waste at the Saakadze Site in Georgia

50 m3 of legacy liquid radioactive waste at the Saakadze site in Georgia was treated using a modular type facility with apparatuses encased in three metallic 200 L drums using as purification method the sorption/ion exchange technology. The main contaminant of water in the underground tank was the long-lived radionuclide 226Ra. The casing of processing equipment enabled an effective conditioning of all secondary waste at the end of treatment campaign which resulted in the fully purified water stored on site for further reuse or discharge, and three 200 L metallic drums with cemented radioactive waste which are currently safely stored.

Radionuclide 106Ru behavior in aqueous solutions by ion exchange, ultrafiltration, and centrifugation methods

The paper presents the results of 106Ru radionuclide behavior regularities study in aqueous solutions in a wide pH range by ultrafiltration, ion exchange and centrifugation methods. The regions of 106Ru various species existence in solution have been established: cationic 106Ru species at pH < 3.5; the transition region of non-ionic species formation in the range of pH 3.5–4.2 and the region of non-ionic species predominant formation at pH > 4.2. A characteristic feature of the studied solutions is the formation of non-ionic particles by microconcentrations of 106Ru via pseudocolloids at lower pH values as compared to ruthenium solutions with a concentration of 10-6–10-4 mol/dm3. The established regularities of the behavior of ruthenium radionuclides can be utilized to increase the efficiency of ion exchange and membrane separation methods at nuclear and radiation facilities for technological solutions and liquid radioactive waste treatment.

Efficient removal and immobilization of radioactive iodide and iodate from aqueous solutions by bismuth-based composite beads

Effectively remove radioactive iodine from aqueous solutions is highly desirable for the purpose of radioactive liquid waste treatment. However, anionic radioiodine species do not prefer to sorb on minerals or inorganic sorbents. Herein, we report a stable millimeter-level bismuth-based polymer composite beads (δ-Bi2O3@PES) with fast removal kinetics and high adsorption capacities for both iodide (I−) and iodate (IO3−). Abundant micro-flower-like δ-Bi2O3 was embedded in the porous interconnection network of polymer composite beads, making it an effective adsorbent for removing radioactive iodine. δ-Bi2O3@PES composite beads showed high adsorption capacities of 95.4 mg/g and 170.6 mg/g for I− and IO3− by forming an insoluble Bi4I2O5 and BiOIO3 phase, respectively, which also can selectively remove iodine anions in the presence of large excesses of NO3– and SO42−. Furthermore, δ-Bi2O3@PES composite beads can highly reduce the Bi3+ leaching of δ-Bi2O3 powder after immersed in the acid solution and can be easily separated from the liquid after sorption. The performance of δ-Bi2O3@PES in dynamic column experiments indicates that economical Bi-based composite bead is a promising candidate sorbent for environmental remediation of actual iodine-contaminated water.

Oxidative decomposition of ammonium ion with ozone in the presence of cobalt and chloride ions for the treatment of radioactive liquid waste

To prevent unexpected accidents at nuclear facilities caused by accumulated ammonium nitrate in an aqueous liquid waste containing ammonium salts and nitric acid, NH4+ in the liquid waste must be decomposed under mild reaction conditions. In this study, we investigated the oxidative decomposition of NH4+ with O3 at 333 K in the presence of a homogeneous Co2+ catalyst and Cl− in the wide pH range of the test solution. The reaction behavior was greatly affected by pH of the test solution. In a basic solution at pH 12, high conversion of NH4+ was obtained even in the absence of Co2+ and Cl−, and the main product was NO3−. However, Co2+ and Cl− in the solution greatly enhanced the decomposition rate of NH4+ in acidic to mild basic solutions (pH 1–8), while only low conversion of NH4+ was observed unless both Co2+ and Cl− were present. For the reaction with Co2+ and Cl− in the solutions, NH4+ was transformed mainly into chloramines (NHxCl3−x, x = 1–3) by the reaction with HClO, which was formed by the reaction of Cl− with O3 catalyzed by the homogeneous Co2+ catalyst, and led to the high decomposition rate of NH4+. Cl− suppressed the formation of the precipitate CoO(OH) during the reaction and consequently the Co2+ catalyst stably existed in the reaction solution, which was another reason for the high decomposition rate of NH4+ in the presence of Cl−. Owing to the swift decomposition of NH4+ under mild reaction conditions and small formation of secondary waste, the oxidative decomposition of NH4+ in the presence of the homogeneous Co2+ catalyst and Cl− is suitable and applicable for the treatment of the aqueous liquid waste containing ammonium salts and nitric acid.

Liquid Low-Level Radioactive Waste Treatment Using an Electrodialysis Process.

In this work, the possibility of using electrodialysis for the treatment of liquid low-level radioactive waste was investigated. The first aim of the research was to evaluate the influence of the process parameters on the treatment of model solutions with different compositions. Subsequent experimental tests were conducted using solutions containing selected radionuclides (60Co and 137Cs), which are potential contaminants of effluents from nuclear power plants, as well as components often found in waste generated from industrial and medical radioisotope applications. The results of the experiments performed on real radioactive waste confirmed that electrodialysis was a suitable method for the treatment of such effluents because it ensured high levels of desalination and rates of decontamination. The most important parameters impacting the process were the applied voltage and electrical current. Moreover, this research shows that the application of the ED process enables the separation of non-ionic organic contaminants of LLW, which are unfavorable in further stages of waste predisposal.

Development of New Processes for the Decommissioning Decontamination and for Treatment and Disposal of the Secondary Low- and Intermediate-Level Radioactive Waste

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

Synthesis of titanium phosphates from unconventional solid precursor and their ion-exchange and electrochemical properties

A new low-cost and simple synthesis of titanium phosphate functional materials of desired composition has been developed. The route of material transformation is unconventional solid precursor-sorbent-anode material. The synthesis is based on the heterogeneous interaction between phosphoric acid and the solid (NH4)2TiO(SO4)2·H2O precursor. The influence of the synthesis conditions on the titanium phosphate composition has been thoroughly studied using NMR, FTIR spectroscopy, XRD, and DTA techniques, and mechanism of TiP formation has been established. Optimal synthesis conditions to provide obtaining of a pure TiO(OH)(H2PO4)·H2O (TiHP) phase have been found. In contrast to existing methods that require rigid synthesis conditions and high reagent consumption, the new synthesis lasts 4 h and no special equipment is needed. The sorption and electrochemical capabilities of TiHP have been investigated. The obtained material exhibits the highest sorption capacity toward Cs+ and Sr2+ ions among other TiP-based ion exchangers. Furthermore, TiHP has been tested as a sorbent for treatment of multicomponent liquid radioactive waste and distribution coefficients of the radionuclides have been found to be 105 mL·g–1. New approach to the synthesis of precursor for Li-ion batteries has been proposed. The substitution of protons of the dihydrogen phosphate group in TiHP by lanthanum (III) cations and subsequent calcination at 900 °C result in formation of nanoparticles of the final powder and provide the good electrochemical characteristics of obtained electrode material.

HYDROPHOBIZATION OF PET TRACK-ETCHED MEMBRANES FOR DIRECT CONTACT MEMBRANE DISTILLATION OF LIQUID RADIOACTIVE WASTES

This article provides the results of liquid low-level radioactive wastes treatment by direct contact membrane distillation using polyethylene terephthalate hydrophobic track-etched membranes. The hydrophobization of track-etched membranes was carried out by UV-induced graft polymerization of triethoxyvinylsilane with styrene and coating with fluorine-containing silanes. Hydrophobic membranes were investigated by scanning electron microscope, Fourier-transform infrared spectroscopy, contact anglemeasurements, and liquid entry pressure analysis. Prepared membranes were tested in treatment of liquid low-level radioactive wastes by membrane distillation. The influence of pore sizes on water flux and rejection degree was studied. Rejection degree was evaluated by conductometry and atomic emission method. Decontamination factors evaluated by gamma-ray spectroscopy for 60Co, 137Cs, and 241Am are 85.4, 1900 and 5.4 for membranes modified with polystyrene and triethoxyvinylsilanewith pore diameters of 142 nm; 85.0, 1462 and 4 for membranes modified with perfluorododecyltrichlorosilanewith pore diameters of 150 nm respectively.

The Ways of Implementation of the Optimization Principle in the Personnel Radiological Protection

An overview of the joint works undertaken by the specialists of Ural Federal University, the Institute of Mathematics and Mechanics of the Russian Academy of Sciences, and the Institute of Reactor Materials (IRM) in certain areas of implementing the optimization principle in radiological protection of nuclear industry personnel is provided. The effectiveness and the capacity of using routing optimization when carrying out works in unsteady radiation fields (when relocating and dismantling equipment of radioactive systems) are evaluated. The results of works on designing the composition of homogeneous radiation-protective materials (RPM) and their capacity to implement the optimization principle in the planning of radiation-hazardous works, including by the example of AO IRM and in ion-selective treatment of liquid radioactive waste, are analyzed. The application of the optimization principle for individual personnel radiation risk management at AO IRM is discussed.