Spent Ion Exchange Resins Research Articles

Uptake of heavy metal ions from aqueous solutions by sorbents obtained from the spent ion exchange resins

Obtaining useful sorbents from spent materials is of great importance. The paper presents the results of tests determining the kinetic and adsorption parameters on the removal of heavy metal Cu(II), Cd(II), Co(II) and Pb(II) ions on the carbonaceous adsorbents W1 KPS, W2 KPS, D1 CC3 and D2 CC3 obtained from spent ion exchangers. Among others the sorption effects of contact time, initial concentration, solution pH, influence of interfering ions were studied. W1 KPS obtained by the microwave heating exhibits a higher affinity for heavy metal ions than W2 KPS. However, for the sorbent obtained in the resistance furnace higher values of the sorption capacity were obtained for D2 CC3 rather than D1 CC3. It was found that the sorption capacity increases with the increasing initial concentration of the solution and phase contact time. The maximum pH adsorption was found at pH 5.0. A better fit of the kinetic data from the experimental one was presented using the pseudo second order model. In this paper, desorption using acidic desorbing agents was also examined. The results show that 2 M HNO3 is the most effective desorbing agent. Carbonaceous sorbents were characterized using elemental analysis, Fourier transform infrared spectroscopy and point of zero charge measurements. Differential scanning calorimetry analysis was also conducted.

Pyrolysis of undoped and multi-element doped ion exchange resins with regard to storage properties

Pyrolysis can be used as pretreatment for permanent storage of radioactively contaminated ion exchange resins. Therefore, the thermal decomposition of different ion exchange resins was investigated in fixed and fluidized bed reactors. Here, the anion and cation exchange resin mixtures for pressurized water reactors, in this case Lewatit, and boiling water reactors, in this case Powdex, were pyrolyzed up to 650°C with varied heating rates at atmospheric pressure in different test facilities. A maximum weight loss of 63 or 75wt.% (dry) and a reduction of the volume of the bed of 40 or 20 vol.% were founded. After specifying optimized pyrolysis conditions for the treatment, the resins were doped with cesium and various other elements to imitate the spent ion exchange resin from the nuclear power plants. With regard to cost reduction for exhaust gas cleaning, the retention of the doped elements in the pyrolysis residue was examined.

Accelerated Leach Test for Low-level Radioactive Waste Forms in the Hungarian NPP Paks

An accelerated leach test method was used for low-level radioactive waste forms in the Hungarian NPP Paks. These experiments were performed using cylinders prepared form Hungarian cement type CEM I 32,5 LH and CEM III/B 32, N-LH/SR. Each cylinder was made using cement or cement plus additive using radioactive waste water. The cemented radioactive material was evaporator bottom residue or sludge as well as evaporator cleaning acid solution, spent ion exchange resin, decontamination solution from NPP Paks, containing 134Cs, 137Cs and 60Co as main radioactive components. Leach tests were performed according to ASTM C 1308-08 standard. A computer program (ILT15) associated with the accelerated leach test was developed based on the ASTM C 1308-08 standard. Literature test and measured leaching data were analyzed to assess whether the model for diffusion from a finite cylinder describes leaching from cement based waste forms. In this paper some of the experimental and modeling work used to validate the test method are presented.

Kinetic study and optimization of electro-Fenton process for dissolution and mineralization of ion exchange resins

Spent ion exchange resins have become a crucial radioactive solid waste from the nuclear industry. Developing effective and safe disposal methods as alternatives to the present cementation method remains challenging. This investigation demonstrates the treatment of a mixed resin (sulfonated and quaternary ammonium polystyrene beads with a weight ratio of 40%:60%) by the electro-Fenton process. Mesh-type titanium metal that was coated with IrO2/RuO2 (Ti-DSA) was used as the anode and a stainless steel net was used as the cathode. The conversion of resins to soluble fragments and the removal of total organic carbon reached 92% and 99.4%, respectively, under conditions of solid loading=40gL−1, pH 2, applied current=2A, H2O2 flow rate=1.2mLmin−1, FeSO4=20mMat 85°C. A pseudo first-order kinetic model of consecutive reactions specified that the efficacy of the electro-Fenton depended strongly on the slowly generated styrene in the aqueous phase by H2O2 and strong acid, which was rapidly mineralized by the hydroxyl radicals. The electro-Fenton process with reused iron catalyst was effective for treating ion exchange resin for at least for three runs, greatly reducing the volume of waste resin liquid.

Decontamination of spent ion-exchange resins contaminated with cesium radionuclides

Possible causes of the occurrence of radioactive contamination irremovable by acid–base regeneration of ion-exchange resins used in treatment of process media and liquid radioactive waste have been studied. It has been found that most of the irremovable cesium radionuclides are bound to inorganic deposits on the surface and in the bulk of ion-exchange resin granules. The nature of inorganic inclusions has been investigated on real and model spent ion-exchange resins (SIER). A method has been proposed for decontaminating SIER using resorcinol–formaldehyde resins selective to cesium radionuclides. Such an approach has been shown to be promising for deep decontamination of SIER.

Heat transfer of spent ion exchange resin in iron ore sintering process

Spent ion exchange resin (SIER) is a kind of solid waste which derived from water treatment process in iron and steel industry. Utilization of SIER as a replacement of fuel in sintering process is a promising method of SIER treatment. In order to investigate the feasibility of this method, heat transfer effects of SIER in iron ore sintering process were studied in this paper via numerical simulation. A 3D unsteady numerical reference model was developed on the basis of the porous media model and local non-equilibrium thermodynamics model and verified by the data form measurement. Heat transfer effects of different SIER mass fractions (0–8%) in sintering material during the sintering process were studied on the numerical model established in this paper. The results showed that with the increasing mass fraction of SIER, the maximum combustion zone thickness and flame front speed are both increased, the heating-up point, the moment when solid temperature reached the top and the maximum temperatures were all became earlier in different location of sintering bed. When SIER content is 8%, the maximum sintering temperature exceeds the maximum limit of best sintering temperature.

Effect of blastfurnace slag addition to Portland cement for cationic exchange resins encapsulation

In the nuclear industry, cement-based materials are extensively used to encapsulate spent ion exchange resins (IERs) before their final disposal in a repository. It is well known that the cement has to be carefully selected to prevent any deleterious expansion of the solidified waste form, but the reasons for this possible expansion are not clearly established. This work aims at filling the gap. The swelling pressure of IERs is first investigated as a function of ions exchange and ionic strength. It is shown that pressures of a few tenths of MPa can be produced by decreases in the ionic strength of the bulk solution, or by ion exchanges (2Na + instead of Ca 2+ , Na + instead of K + ). Then, the chemical evolution of cationic resins initially in the Na + form is characterized in CEM I (Portland cement) and CEM III (Portland cement + blastfurnace slag) cements at early age and an explanation is proposed for the better stability of CEM III material.

Cation immobilization in pyrolyzed simulated spent ion exchange resins

Significant quantities of spent ion exchange resins that are contaminated by an assortment of radioactive elements are produced by the nuclear industry each year. The baseline technology for the conditioning of these spent resins is encapsulation in ordinary Portland cement which has various shortcomings none the least of which is the relatively low loading of resin in the cement and the poor immobilization of highly mobile elements such as cesium. The present study was conducted with cationic resin samples (Lewatit S100) loaded with Cs+, Sr2+, Co2+, Ni2+ in roughly equimolar proportions at levels at or below 30% of the total cation exchange capacity. Low temperature thermal treatment of the resins was conducted in inert (Ar), or reducing (CH4) gas atmospheres, or supercritical ethanol to convert the hydrated polymeric resin beads into carbonaceous materials that contained no water. This pyrolytic treatment resulted in at least a 50% volume reduction to give mechanically robust spherical materials. Scanning electron microscope investigations of cross-sections of the beads combined with energy dispersive analysis showed that initially all elements were uniformly distributed through the resin matrix but that at higher temperatures the distribution of Cs became inhomogeneous. Although Cs was found in the entire cross-section, a significant proportion of the Cs occurred within internal rings while a proportion migrated toward the outer surfaces to form a crustal deposit. Leaching experiments conducted in water at 25°C showed that the divalent contaminant elements were very difficult to leach from the beads heated in inert atmospheres in the range 200–600°C. Cumulative fractional loses of the order of 0.001 were observed for these divalent elements for temperatures below 500°C. Regardless of the processing temperature, the cumulative fractional loss of Cs in water at 25°C reached a plateau or steady-state within the first 24h increasing only marginally up 120h. For unheated beads, a steady-state was reached at a cumulative fractional Cs loss of about 0.0045 while for beads heated to 400°C the steady-state was reached at a value of 0.1. Under comparable conditions leaching of Sr, Co and Ni were close to the detection limits. It is concluded that the fraction of the Cs inventory that was more readily leached derived mainly from Cs phases in the crustal deposit.

Development of pyrolytic monolithic carbon composites for the conditioning of spent ion exchange resins

ABSTRACTThe pyrolysis of ion exchange resin beads that are used for the purification of water in reactor primary- and secondary-cooling circuits can result in stable and leach resistant carbonaceous products. However, free flowing beads are less desirable waste forms for disposal in sub-surface or surface repositories than monolithic masses of low porosity. We have investigated the pyrolysis of polymeric resin – cation exchange resin composites to give mechanically robust and chemically durable monolithic carbonaceous waste forms that are suitable for repository disposition. Also investigated was the dependence of product properties on various processing parameters (temperature ramp and final temperature). As a first approach, epoxy resins were used for the preparation of monoliths since such resins cure at room temperature and result in a relatively high carbon yield. Carbonaceous monolithic products were prepared at 400, 500, 600, 700 and 800 °C using a temperature ramp of 2°C/min. The products were maintained at the chosen temperatures for a period of one hour. Mass losses, volume reduction, hardness and compressive strength were measured and mathematical functions are proposed to describe the measured values of these properties. The carbon monoliths were observed to be mechanically robust.

ICONE19-43160 Development of Spent Ion Exchange Resin Processing in Nuclear Power Stations

Commercial operation of the Tsuruga nuclear power station Unit 1, owned by the Japan Atomic Power Company (JAPC), will be terminated in 2016. For safe decommissioning of the station, technologies for processing stored radioactive wastes such as spent Ion EXchange resin (IEX) and Filter Sludge (FS) have been jointly developed by JAPC and JGC Corporation. The Wet-Oxidation process (WetOx) was applied for decomposition of the spent IEX and FS, and Super Cement (SC) solidification was chosen for immobilization of the WetOx residue. Pilot scale tests for the WetOx treatment have been successfully conducted with simulated wastes. The WetOx residue is a concentrate with Suspended Solid (SS) and sodium sulfate, and was solidified by SC in lab scale cementation apparatus. The compressive strength of the solidified waste was confirmed to meet the desired level for radioactive waste disposal.

Modeling of Incineration of Spent Ion Exchange Resins of Boiling Water and Pressurized Water Nuclear Reactors

Incineration of spent ion exchange resin was simulated using the ChemSheet chemical calculation program. The simulation of the incineration was modeled for typical spent resin produced by pressurized water reactors (PWRs) and boiling water reactors (BWRs) in Finland. The objective of the study was to find the volume and mass reduction and the chemical compounds formed during incineration. The simulation showed that active elements did not play any role in incineration owing to small amount of Cs, Co, etc. The ash contained metal oxides—mainly hematite, iron oxide Fe2O3. Other products of the incineration were water, carbon dioxide, sulfuric acid, and nitrogen oxides. The volume reductions 1/100 and 1/14 of the spent resin were obtained for PWRs and BWRs, respectively. The annual ash production from incineration was calculated to be 408 kg and 746 kg for the currently operating Finnish PWR and BWR plants in Loviisa and Olkiluoto, respectively.

Treatment of radioactive ionic exchange resins by super- and sub-critical water oxidation (SCWO)

As the usage of ion exchange resins increases the inventory of spent ion exchange resins increases in nuclear power plants. This study is to find an environmental-friendly process to treat theses spent resins. The test samples were prepared by diluting the slurry made by wet ball milling the spent cationic exchange resins for 24 h. The spent cationic exchange resins were separated from mixed ion exchange resins by a fluidized bed gravimetric separator. The decomposition of the samples was investigated with super-critical water oxidation (SCWO) equipment. A statistical test method – the central composite design as a statistical design of experiments – was adopted to find the optimum condition to decompose the spent exchange resins. The optimum condition was 60% of excess oxygen, 22.5 min of residence time, 0.615 wt% of NaOH, 358 of reaction temperature, and 3600 psi of reaction pressure, which is a sub-critical condition. The liquid product of the decomposition has the characteristics of 80–185 ppm of COD (Chemical Oxygen Demand), 4.0–6.0 of pH, and <1.0 ppm of corrosive components (Ni, Fe, Cr, and Mo). The exhaust gas from the SCWO equipment contained NOx of 0 ppm, SOx of 3 ppm (environment exhaust standard in Korea: NOx 200 ppm, SOx 300 ppm). Co-substituted mock samples were prepared to simulate spent cationic exchange resins from nuclear power plants which can contain radioactive Co isotopes. The conditions to obtain organic compound destruction ratio which conforms the effluent stand for the mock samples were found. The treated water filtered with 0.2-filter contained less than 1 ppm of Co. Thus Co recovery rate of more 99% was achieved.

Supercritical water oxidation of ion exchange resins: Degradation mechanisms

Spent ion exchange resins are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation could offer a viable treatment alternative to destroy the organic structure of resins and contain radioactivity. IER degradation experiments were carried out in a continuous supercritical water reactor. Total organic carbon degradation rates in the range of 95–98% were obtained depending on operating conditions. GC–MS chromatography analyses were carried out to determine intermediate products formed during the reaction. Around 50 species were identified for cationic and anionic resins. Degradation of polystyrenic structure leads to the formation of low molecular weight compounds. Benzoic acid, phenol and acetic acid are the main compounds. However, other products are detected in appreciable yields such as phenolic species or heterocycles, for anionic IERs degradation. Intermediates produced by intramolecular rearrangements are also obtained. A radical degradation mechanism is proposed for each resin. In this overall mechanism, several hypotheses are foreseen, according to HOO radical attack sites.

Ion exchange resins destruction in a stirred supercritical water oxidation reactor

Spent ion exchange resins (IERs) are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation offers a viable treatment alternative to destroy the organic structure of resins, used to remove radioactivity. Up to now, studies carried out in supercritical water for IER destruction showed that degradation rates higher than 99% are difficult to obtain even using a catalyst or a large oxidant excess. In this study, a co-fuel, isopropanol, has been used in order to improve degradation rates by initiating the oxidation reaction and increasing temperature of the reaction medium. Concentrations up to 20 wt% were tested for anionic and cationic resins. Total organic carbon reduction rates higher than 99% were obtained from this process, without the use of a catalyst. The influence of operating parameters such as IERs feed concentration, nature and counterions of exchanged IERs were also studied.

Studies on decontamination of spent ion exchange resin used for plutonium purification in PUREX stream

This paper deals with the studies on decontaminations of spent ion exchange resin used for purification of plutonium in PUREX process stream. Studies were carried out to optimize the chemical procedure for removal of plutonium and fission products activities form spent Ion Exchange resin. Different metal complexing reagents were tested for leaching out of radionuclides entrapped in irradiated spent ion exchange resin. The experimental results indicate that 0.01 M NaF solution was found the most suitable for removal of plutonium. The mixture of Na2CO3 and sodium salt of EDTA solution was found to be better for decontamination of spent ion exchange resin from beta and gamma activities. Optimized mixture of 0.5 M Na2CO3 and 0.1 M sodium salt of EDTA solution was found to be the most effective for fission product activities removal. After successive multiple contacts using these suitable reagents, the Pu and fission product activities in spent ion exchange resin were brought down to a minimum possible level, making it quite suitable for its long term storage.

EVALUATION OF FERROCYANIDE ANION EXCHANGE RESINS REGARDING THE UPTAKE OF Cs+IONS AND THEIR REGENERATION

Ferrocyanide-anion exchange resin was prepared and the prepared ion exchange resins were tested on the ability to uptake <TEX>$Cs^+$</TEX> ion. The prepared ion exchange resins were resin-KCoFC, resin-KNiFC, and resin-KCuFC. The three tested ion exchange resins showed ion exchange selectivity on the <TEX>$Cs^+$</TEX> ion of the surrogate soil decontamination solution, and resin-KCoFC showed the best <TEX>$Cs^+$</TEX> ion uptake capability among the tested ion exchange resins. The ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation. A regeneration feasibility study of the spent ion exchange resins was also performed by the successive application of hydrogen peroxide and hydrazine. The desorption of the <TEX>$Cs^+$</TEX> ion from the ion exchange resin satisfied the electroneutrality condition in the oxidation step; the desorption of the <TEX>$Fe^{2+}$</TEX> ion in the reduction step could also be reduced by adding the <TEX>$K^+$</TEX> ion.

Studies in liquid phase mineralization of cation exchange resin by a hybrid process of Fenton dissolution followed by wet oxidation

Abstract The spent ion exchange resin poses a disposal problem. A novel process, dissolution followed by wet oxidation (WO), for mineralization of the cation exchange resin (gel type) was studied by using homogeneous CuSO 4 as the catalyst. The waste obtained after copper catalyzed dissolution of ion exchange resin by H 2 O 2 was then further treated by wet oxidation for mineralization. The process parameters for dissolution were studied and optimum parameters were established. The wet oxidation was studied in the temperature range of 453–493 K and the oxygen partial pressure range of 0.34–1.38 MPa at pH 6.5. The kinetic data were modeled using the power law rate expression in terms of chemical oxygen demand (COD).

Characterization of 14C in Swedish Light Water Reactors

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

ICONE15-10785 Feasibility Study of Solidification for LowLevel Liquid Waste Generated by Sulfuric Acid Elution Treatment of Spent Ion Exchange Resin

ICONE15-10785 Feasibility Study of Solidification for LowLevel Liquid Waste Generated by Sulfuric Acid Elution Treatment of Spent Ion Exchange Resin

Radiochemical analysis of chlorine-36

The radioactive chlorine isotope, 36Cl, decays with a half-life of 3×105 years by emitting a beta particle (98 %) and by electron capture. The aim of this paper is to propose a radiochemical separation method of 36Cl from the other beta-gamma emitters present in low and medium radioactive wastes such as spent ion exchange resins and evaporator concentrates, that arise from Nuclear Power Plants and particularly in the wastes that come from decommissioning activities of graphite reactors, in order to provide data for 36Cl inventory calculations. The separation method proposed is based on an oxidation technique where chlorine is trapped by NaOH. 36Cl beta emissions are measured by liquid scintillation counting by the dual label technique in order to avoid the contamination produced by 14C which is also trapped by NaOH and which is the main contaminant present in graphite samples. The sensitivity of this method is sufficient to achieve the needed thresholds for the radiological characterization of the radioactive materials to which this method can be applied.