Decontamination of radioactive waste solutions using pottery

The sorption of microamounts of Eu spiked with ^152,154Eu on the artificial stone, pottery, was studied at varoius conditions. Three kinds of pottery (red, black, and white), in addition to the raw material (potter’s clay) of the red kind, were tested. The pottery shows high sorption power with respect to the radiotracer depending on the kind of pottery. The uptake percentages gradually increase with the contact time, weight of pottery, or pH of the aqueous solution, attaining saturation at certain values. The amounts of Eu³⁺ required for the saturation are 7.38, 7.00, 5.93, and 1.64 (mmol Eu) (g pottery)⁻¹ for raw, red, black, and white samples, respectively. This sequence is parallel to that of the uptake percentage, which is related to the surface area of each material. The sorption presumably occurs via adsorption and ion exchange. The results were applied to decontamination of low-level liquid radioactive waste by filtration through a pot of red pottery; the effluent was free from the radiotracer. Published in Russian in Radiokhimiya, 2006, Vol. 48, No. 4, pp. 352–356. The text was submitted by the authors in English.     

Low-temperature immobilization of liquid radioactive waste using silica from concentrated hydrothermal solutions

Experiments on vitrification of simulated liquid radioactive waste (LRW) were performed. To obtain solid glass-like phases, LRW components were treated with aqueous-alcoholic hydrochloric acid solutions containing hydrolyzates of alkyl silicates and aqueous silica sols obtained by membrane concentration of a natural hydrothermal solution. The pH of the mixture was 1.5–4. The characteristics of the solid samples obtained were studied by X-ray phase analysis, thermogravimetry, and scanning electron microscopy. A procedure was developed for low-temperature (5–60°C) immobilizaiton of low-level LRW using aqueous silica sols.     

Decontamination of liquid radioactive waste from transuranium elements and uranium by complexation with functionalized oligoethers, combined with ultrafiltration

Decontamination of liquid radioactive waste (LW) from transuranium elements (TUEs) and uranium by complexation with phosphorus-containing and guanidine phosphate oligoethers and by ultrafiltration was studied. The rejection coefficient of TUEs and U on these oligomers reaches 99.0–99.6% of the initial concentration.     

Kinetic and thermodynamic studies for cesium removal from low-level liquid radioactive waste using impregnated polymeric material

A novel polymeric impregnated material was prepared by loading 4,4′(5′)di-t-butylbenzo-18-crown-6 (DtBB18C6) onto poly(acrylamide-acrylic acid-acrylonitrile)-N,N′-methylenediacrylamide [P(AM-AA-AN)-DAM]. The sorption of ¹³⁷Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was studied using batch equilibrium technique with respect to the pH, contact time, and temperature. The applicability to treatment of low-level liquid radioactive waste was examined. The free energy (ΔG ⁰), enthalpy (ΔH ⁰), and entropy (ΔS ⁰) of the sorption were calculated. These parameters showed that the sorption of ¹³⁷Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was spontaneous and endothermic in nature. The Lagergren first-order, pseudo-second-order, and homogeneous particle diffusion models were tested kinetically to describe the reaction mechanism. The experimental data were fitted well by the pseudo-second-order kinetic model.     

Processes for treatment of liquid radioactive waste containing seawater

The main sources of formation of liquid radioactive waste (LRW) containing seawater are determined, and the main problems arising in management of such waste are analyzed. Sorption methods for removing long-lived Cs and Sr radionuclides from highly mineralized (>1 g L^–1) LRW are determined. The main physicochemical and sorption characteristics, advantages, and drawbacks of candidate sorbents for removing Cs and Sr radionuclides are described. Examples of using SRM and VS-5 chemical reaction sorption materials developed for removing Sr from LRW with the mineralization of up to 60 g L^–1 are given. The results of studying composite materials based on BaSiO₃ and resorcinol–formaldehyde resins, intended for removing Cs and Sr radionuclides from seawater, are analyzed. Composite sorbents of such type efficiently remove Cs and Sr radionuclides from seawater. Processes developed by the authors and brought into practice at various plants of the Far East for treatment of multicomponent LRW formed in the course of operation, repair, and decommissioning of nuclear-powered surface ships and submarines are described.     

Concept of autonomous processing of liquid radioactive waste from ship nuclear power installations

The concept of autonomous processing of liquid radioactive waste from ship nuclear power installations is substantiated, and its implementation in the form of a module membrane sorption installation consisting of micro- and ultrafiltration, reverse-osmosis, and ion-exchange modules is suggested. Data on treatment of liquid radioactive waste of complex physicochemical composition using this installation are reported.     

Geochemical aspects of environmentally safe conservation of liquid radioactive waste

Physicochemical features of the sorption of Sr, Cs, and U radionuclides on natural mineral sorbents (montmorillonites of Na and Ca type, kaolinites, illites) were studied. The main processes responsible for binding and retention of radionuclides are ion exchange and formation of complexes on the mineral surface. The influence of pH, salt composition of the solution, specific surface area of the sorbent, and its pore size on the radionuclide immobilization efficiency was examined.     

Decontamination of irradiated Be from radioactive impurities by precipitation of hydroxides

Decontamination of TV-56 grade Be irradiated in an SM reactor to a neutron fluence of 6 × 10²² cm^?2 (E > 0.1 MeV) from radioactive impurities by precipitation of its hydroxide with ammonia in the presence of a complexone, diethylenetriaminepentaacetic acid (DTPA), was studied. The decontamination procedure involved dissolution in hydrochloric acid, addition of the complexone, precipitation of beryllium hydroxide with ammonia, separation of the hydroxide from solution, and washing of the hydroxide precipitate. The radioactivity of the solutions obtained after each decontamination step was determined.     

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 °C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined.     

Chromatographic separation of 65Zn from radioactive tellurium waste solution using alumina column

High-purity ⁶⁵Zn was separated from a mixture containing ^{121,121m,123m,127}Te, ⁶⁵Zn, ⁵⁴Mn, ⁶⁰Co, ^110m Ag, ¹²⁵Sb, and ¹³⁴Cs using a small chromatographic alumina column. Samples of aged radioactive tellurium waste were dissolved in alkali solutions (1 and 5 M NaOH) and fed into preconditioned 1.0 g alumina columns at ～50°C. The columns were washed with 1 M NaOH or successively with 5 and 1 M NaOH. ⁶⁵Zn was quantitatively retained in the alumina column in the course of feeding and washing the column with 1 M NaOH. Solutions of NH₄Cl-NH₃, NH₄Cl-HCl, and HCl were studied as eluents for ⁶⁵Zn from the alumina column. 2 M HCl ensured the highest elution yield (88.7 ± 1.7%) with the ⁶⁵Zn radionuclidic purity of 99.4 ± 0.02%.     

Development of a process for treatment and solidification of liquid radioactive concentrates from the Leipunskii Institute of Energy Physics

Treatment of liquid radioactive concentrates (LRCs) from the Leipunskii Institute of Energy Physics using Termoksid-35 ferrocyanide sorbent was studied. After passing LRC with a volume activity of 3.9 × 10⁸ Bq l⁻¹ through a column packed with the sorbent, the volume activity of the filtrate does not exceed 3.7 × 10⁴ Bq l⁻¹. The LRC decontaminated from the major amount of radionuclides is fed to cementation. A formulation of a cement compound with a polyfunctional additive consisting of finely dispersed cement, bentonite clay, biocidal additive, plasticizer, and defoamer was developed. For the storage of a container filter with the spent ferrocyanide sorbent, it is suggested to place it in a 1 m³ metallic container in which the cement compound with the LRC decontaminated from the major fraction of cesium acts as biological protection. The γ-radiation dose rate from the 1 m³ container filled with the cement compound, with the filter with the spent sorbent placed inside, was calculated. A technology for processing of liquid radioactive concentrates from the Leipunskii Institute of Energy Physics was suggested and substantiated.     

Reduction of radioactive waste volume using selective crystallization processes

Satisfactory combination of small volume of compositions incorporating ¹³⁷Cs, ⁹⁰Sr, and ²⁴⁴Cm with the admissible heat release rate can be attained by crystallization of definite compounds from high-level solutions. The concentrate of cesium isotopes is obtained by their precipitation in the form of phosphoromolybdate and, after additional storage, in the form of CsMgPO₄·6H₂O; the concentrate of Am, Cm, Sr, Ba, and Ln can be obtained using the known but somewhat modified (increased pH of solution) OXAL process. It is appropriate to remove Am and long-lived Cm isotopes from this concentrate also after storage ensuring the decay of ⁹⁰Sr and ²⁴⁴Cm. The remaining rare-earth elements with traces of Am and Cm and stable isotopes of Sr and Ba can be disposed of, e.g., in the form of a mineral-like phosphate.     

Treatment of radioactive liquid waste by sorption on natural zeolite in Turkey

Liquid radioactive waste has been generated from the use of radioactive materials in industrial applications, research and medicine in Turkey. Natural zeolites (clinoptilolite) have been studied for the removal of several key radionuclides ((137)Cs, (60)Co, (90)Sr and (110m)Ag) from liquid radioactive waste. The aim of the present study is to investigate effectiveness of zeolite treatment on decontamination factor (DF) in a combined process (chemical precipitation and adsorption) at the laboratory tests and scale up to the waste treatment plant. In this study, sorption and precipitation techniques were adapted to decontamination of liquid low level waste (LLW). Effective decontamination was achieved when sorbents are used during the chemical precipitation. Natural zeolite samples were taken from different zeolite formations in Turkey. Comparison of the ion-exchange properties of zeolite minerals from different formations shows that Gordes clinoptilolite was the most suitable natural sorbent for radionuclides under dynamic treatment conditions and as an additive for chemical precipitation process. Clinoptilolite were shown to have a high selectivity for (137)Cs and (110m)Ag as sorbent. In the absence of potassium ions, native clinoptilolite removed (60)Co and (90)Sr very effectively from the liquid waste. In the end of this liquid waste treatment, decontamination factor was provided as 430 by using 0.5 mm clinoptilolite at 30 degrees C.     

Role of simple anionic ligands in deep decontamination of liquid radioactive waste

The influence of the NO₉^-, SO₄^2–, HCO₃^–, and Ac^– anions on the efficiency of the sorption of bivalent metals by sulfonic cation exchangers was studied. The main method was elution of the Mg²⁺, Ca²⁺, and Co²⁺ cations with individual solutions of the corresponding sodium salts and their mixed solutions with sodium nitrate from the Dowex-50 ion-exchange resin at fixed temperatures. The exchange efficiency is influenced both by neutral complexes of the metal cations mainly with inorganic acid ligands and by the single-charged species. The latter species affect the efficiency of the ion-exchange waste reprocessing considerably more strongly. An increase in the temperature of the solution being treated leads to a significant shift of the peak position toward larger retention volumes, which indicates that it is appropriate to perform the water treatment on sulfonic cation exchangers at elevated temperatures. The first formation constant of the single-charged cobalt dihydrogen phosphate complex species at an ionic strength of 0 M was obtained: logK₁(CoH₂PO₄⁺) = 1.36.     

Inorganic sorbents based on modified microspheres for treatment of liquid radioactive waste

Sorbents based on hollow microspheres of entrained ash, with the surface modified by various chemical compounds (ferrocyanides, phosphates, oxides, etc.), were synthesized. Their physical and ion-exchange characteristics were examined. These sorbents show promise for treatment of low-and intermediate-level liquid waste of various origins in the dynamic and static sorption modes. The use of microspheres as matrices is not only technically but also environmentally efficient, because waste from thermal power engineering is utilized for disposal of nuclear waste. Original Russian Text L.D. Danilin, V.S. Drozhzhin, 2007, published in Radiokhimiya, 2007, Vol. 49, No. 3, pp. 283–286.     

Binders for radioactive waste forms made from pretreated calcined sodium bearing waste

Radioactive waste generated during the reprocessing of fuel rods by the U.S. Department of Energy (DOE) is stored in underground tanks at Hanford, Savannah River and INEEL. The liquid fraction commonly referred to as sodium bearing waste (SBW), is a highly alkaline solution containing large amounts of sodium hydroxide, sodium nitrate and sodium nitrite. It has been shown that SBW can be mixed with a reducing agent and metakaolin and then calcined at 500°–700°C to form a calcine containing sodium aluminosilicate phases such as zeolite A, hydroxysodalite and/or cancrinite. Although calcination of the pretreated SBW produces a reasonable waste form in its own right, existing regulations require that granular calcines must be solidified before they can be shipped off site. It is possible to solidify the calcine in a number of ways. The calcine can be mixed with additional metakaolin and NaOH solution followed by mild curing (90°–200°C). The solid that forms (aka hydroceramic) has both strength and suitably low leachability. The current study examines the feasibility of using a more conventional Portland cement binder to solidify the calcine. Although strength was adequate, the leachabilities of the Portland cement solidified samples were higher than those of companion samples made with metakaolin. The zeolitic phases in the calcine acted like pozzolans and reacted with the Ca(OH)₂ in the Portland cement binder forming additional calcium silicate hydrate (C—S—H). Typically C—S—H is unable to host large amounts of sodium ions in its structure, thus a majority of the sodium present in the zeolites became concentrated in the pore solution present in the Portland cement binder and readily entered the leachant during PCT testing. In this instance metakaolin mixed with NaOH proved to be a superior binder for solidification purposes.     

Exploring innovative solutions for aged concrete utilization: treatment of liquid radioactive waste

Clean Technologies and Environmental Policy - In this study, the applicability of waste concrete as a sorbent material for the liquid radioactive waste management was considered. The sample was...     

Retrospective dosimetry: dose evaluation using unheated and heated quartz from a radioactive waste storage building

In the assessment of dose received from a nuclear accident, considerable attention has been paid to retrospective dosimetry using heated materials such as household ceramics and bricks. However, unheated materials such as mortar and concrete are more commonly found in industrial sites and particularly in nuclear installations. These materials contain natural dosemeters such as quartz, which usually is less sensitive than its heated counterpart. The potential of quartz extracted from mortar in a wall of a low-level radioactive-waste storage facility containing distributed sources of 60Co and 137Cs has been investigated. Dose-depth proliles based on small aliquots and single grains from the quartz extracted from the mortar samples are reported here. These are compared with results from heated quartz and polymineral fine grains extracted from an adjacent brick, and the integrated dose recorded by environmental TLDs.     

Recovery of 60Co from EDTA-containing aqueous solutions

The possibility of recovering radioactive cobalt from EDTA-containing aqueous solutions using oxidation (O₃) and photooxidation (UV/H₂O₂) methods was examined. The optimal amounts of reagents ensuring complete decomposition of EDTA and recovery of radioactive cobalt were determined. The procedures developed for oxidative decomposition of EDTA and for recovery of radioactive isotopes from aqueous media were tested with real liquid radioactive solutions.     

Extraction of Co(II) from aqueous solution using emulsion liquid membrane

The extraction equilibrium of Co(II) from thiocyanate medium by CYANEX 923 (mixture of straight chain alkylated phosphine oxides) in cyclohexane was studied. The stoichiometry of the extraction reaction was postulated based on slope analysis method and the extraction constant Kex was calculated. The stripping percentage of Co(II) with sulphuric acid from the loaded CYANEX 923 was found to increase with the increase in acid concentration. The extraction of Co(II) from aqueous thiocyanate medium into emulsion liquid membrane using CYANEX 923 extractant was also studied. The influence of different parameters such as stirring speed, surfactant concentration, pH of the extractant phase, carrier concentration, internal phase stripping acid concentration, initial Co(II) concentration as well as temperature on the emulsion stability were investigated. The applicability of the emulsion liquid membrane (ELM) process using CYANEX 923 as extractant and SPAN 80 as surfactant for the removal and the concentration of Co(II) from thiocyanate solution was investigated. The results show that it is possible to recover 95% of cobalt in the inner phase after 10 min of contacting time with a concentration factor of 5.