Effect of gamma-irradiation on adsorption properties of Slovak bentonites

One of the basic prerequisites for the use of bentonite as engineering barrier in deep geological repositories for radioactive waste and spent nuclear fuel is their stability against ionizing radiation stemming from radionuclides present in radioactive waste and spent nuclear fuel. The aim of this study was to compare the changes in the adsorption properties of selected Slovak bentonites in relation to uranium fission products (¹³⁷Cs and ⁹⁰Sr), prior to and after irradiation of bentonites with a ⁶⁰Co γ-source and specifying the changes in the structure of Slovak bentonites induced by γ-radiation. The changes in irradiated natural forms of Slovak bentonites and the changes in their natrified analogues and fractions with different grain sizes were studied from five Slovak deposits: Jelšovy potok, Kopernica, Lastovce, Lieskovec and Dolná Ves. The EPR spectra of bentonites from deposits Jelšovy potok and Lieskovec with absorbed doses of 10⁴ and 10⁵ Gy γ-rays showed no changes in the structure of the studied Slovak bentonites. The changes, which in terms of structure destabilization can be considered insignificant, occurred only in bentonites with absorbed doses of γ-radiation as much as 1 MGy. The absorbed dose of 1 MGy γ-radiation did not have an effect on the adsorption of cesium on every studied bentonite. Changes that can also be regarded as insignificant occurred only during strontium adsorption, especially on Fe–bentonite from deposit Lieskovec and Ca–Mg–bentonite from deposit Jelšovy potok, when an increase in the adsorption capacity occurred. Attention should be paid in further research of this topic which would require carrying out experiments on bentonite samples with absorbed doses higher by several orders of magnitude.     

Comparative study of strontium adsorption on dioctahedral and trioctahedral smectites

Slovak bentonites characterized by good rheological, mineralogical and chemical stability are considered as suitable sealing barriers for construction of Slovak deep geological repository for high-level radioactive waste and spent nuclear fuel. There is several Slovak bentonite deposits, bentonites of which have appropriate adsorption properties meeting the geotechnical requirements for this type of barriers. Study of adsorption properties of bentonites (mainly smectites) is an essential step for developing the migration model long-lived corrosion and activation products, and fission products of uranium. Nuclear wastes contain the most important nuclear fission products, β-emitter ⁹⁰Sr with long half-life, biological half-life and high mobility. The present paper investigates and compares the strontium adsorption properties of bentonites of different mineral composition consisted mainly of dioctahedral and trioctahedral smectites.     

Comparative study of cesium adsorption on dioctahedral and trioctahedral smectites

Bentonites which are characterized by good rheological, mineralogical and chemical stability is considered used as sealing barriers in multibarrier Slovak system of deep geological repository for high-level radioactive waste and spent nuclear fuel. In Slovak Republic there are several significant deposits of bentonite, which are characterized by appropriate adsorption properties and meet the geotechnical requirements for this type of barriers. Study of adsorption properties of bentonites and other smectites is an essential step for developing the migration model long-lived corrosion and activation products, and fission products of uranium. Nuclear wastes contain the most important nuclear fission products, radioisotopes ¹³⁴Cs and ¹³⁷Cs. The present paper investigates and compares the cesium adsorption properties of Slovak and North America bentonites composed mainly of dioctahedral smectite montmorillonite (J, L, SAz-1 and STx-1) and trioctahedral smectites saponite (SapCa-2) and hectorite (SHCa-1).     

First shrinkage parameters of Slovak bentonites considered for engineered barriers in the deep geological repository of high-level radioactive waste and spent nuclear fuel

The high potential of bentonites to volume changes depending on the water content is considered as their advantage for the engineered barriers in the deep geological repository of high-level radioactive waste and spent nuclear fuel because of swelling and self-healing of cracks in contact with water. On the other hand, drying may lead to opening of cracks and spaces between the bentonite blocks. This would increase the permeability and contamination risk around the hot container with high-level radioactive waste and spent nuclear fuel, especially if the host rock mass is dry. First shrinkage tests on four Slovak bentonites studied for engineered barriers were carried out. The water content at the shrinkage limit and the relative linear shrinkage are the first available shrinkage parameters received for the bentonite paste. The shrinkage hazard is higher in the best bentonites with high swelling potential—from Kopernica and Jel?ový potok. The results indicated the necessity of further shrinkage tests to determine the relative linear and volume shrinkage of bentonite elements pressed of the loose bentonite powder of low water content.     

Utilization of Slovak bentonites in deposition of high-level radioactive waste and spent nuclear fuel

The basic strategic aims in the field of managing high-level radioactive waste and liquidation of nuclear power plants are all contained in the Energy policy of the Slovak Republic. Its aim is to resolve the concept of the backside of the nuclear energetics fuel cycle??long-term deposition of high-level radioactive waste and spent nuclear fuel (SNF). The most important form of high-level radioactive waste and SNF long-term deposition is their deposition in deep geological formations created by natural as well as engineering barriers used to isolate the long-lived radionuclides from the biosphere. The basic components of these barriers are clays, of which bentonite is generally referred to as the most suitable clay material. There are a few significant bentonite deposits in the Slovak Republic: Jel?ový potok, Kopernica, Lastovce, Lieskovec, Dolná Ves. The review article summarizes the information on geotechnical properties of Slovak bentonites published up-to-date, which is inevitable to know for the intention of their use. It highlights the advantages and shows drawbacks of five Slovak deposits. It suggests further research direction, to draw a thorough hydraulical, microbial and radiation profile of Slovak bentonites.     

Adsorption of cesium on domestic bentonites

Bentonite is a natural clay and one of the most promising candidates for use as a buffer material in the geological disposal systems for spent nuclear fuel and high-level nuclear waste. It is intended to isolate metal canisters with highly radioactive waste products from the surrounding rocks because of its ability to retard the movement of radionuclides by adsorption. Slovak Republic avails of many significant deposits of bentonites. Adsorption of Cs on five Slovak bentonites of deposits (Jelšovy potok, Kopernica, Lieskovec, Lastovce and Dolná Ves) has been studied with the use of batch technique. In the case of Dolná Ves deposit, the mixed-layer illite–smectite has been identified as the main clay component. Natural and irradiated samples, in two different kinds of grain size: 45 and 250 μm have been used in the experiments. The adsorptions of Cs on bentonite under various experimental conditions, such as contact time, adsorbent and adsorbate concentrations have been studied. The Cation Exchange Capacity values for particular deposits drop in the following order: Jelšovy potok > Kopernica > Lieskovec > Lastovce > Dolná Ves. Bentonites irradiated samples with 390 kGy have shown higher specific surface and higher values of the adsorption capacity. Distribution coefficients have been determined for bentonite-cesium solution system as a function of contact time and adsorbate and adsorbent concentration. The data have been interpreted in terms of Langmuir isotherm. The uptake of Cs has been rapid and the adsorption of cesium has increased with increasing metal concentrations. The adsorption percentage has decreased with increasing of metal concentrations. Adsorption of Cs has been suppressed by presence of Ca²⁺ more than Na⁺ cation. Sorption experiments carried out show that the most suitable materials intended for use as barriers surrounding a canister of spent nuclear fuel are bentonites of the Jelšovy potok and Kopernica deposits.     

Cesium sorption on bentonites and montmorillonite K10

The physical and chemical properties of illitic clay minerals from Slovak deposit suitable for application in engineering barriers for high level radioactive waste repositories and spent nuclear fuels were studied. The isolation of spent nuclear fuels and high level radioactive wastes from the outer environment in a deep repository is gained by means of a system of multiple engineering and natural sealing barriers. Vital segments in a multiple barrier system are clay rocks, of which bentonites represent the most suitable clay material. Cs-adsorption on fine fractions of adsorbents (bentonites from three Slovak deposits: Jelšovy potok J15, Kopernica K15, Lieskovec L15 and montmorillonite K10) has been studied with using batch of radiometric techniques. Adsorption parameters have been determined for adsorbent-cesium solution system as a function of contact time and adsorbate concentration. The influences of pH change, the effect of competitive cations, complex-forming organic chelating agents on the adsorption of Cs have also been studied.     

Adsorption equilibrium and kinetic studies of strontium on Mg-bentonite, Fe-bentonite and illite/smectite

Radionuclide adsorption on clay rocks has in recent years been studied mainly in connection with their use as sealing barriers in nuclear waste and spent nuclear fuel repositories. In Slovakia we find deposits of bentonites which should be used for the above mentioned purpose. The usability of adsorbents in practical applications depends on the speed of the adsorption process of the adsorbate on the adsorbent surface and distribution ratio. The work objective was the study of the kinetics of Sr adsorption on clay adsorbents with different geological origin. The geological origin of bentonite significantly influences its mineralogical and chemical composition and therein its adsorption properties. The adsorption process of strontium was fast. Adsorption equilibrium was reached for all three samples studied within 1 min from the beginning of the contact between solid and liquid phases. After the adsorption equilibrium was reached there were no more changes in the values of distribution coefficients and the adsorption percentage, and comparable values were reached in the contact-phase time span studied within 10 days. The values of adsorbed strontium were decreasing in the following order: J250 > L250 > DV45. The pseudo second-order kinetic models was used to describe model the kinetic data and provided excellent kinetic data fitting (R ² > 0.999).     

On the separation of cesium and strontium from oxalate filtrate by neutralization

At the treatment of HLLW (high-level liquid radioactive waste) by oxalate precipitation, the filtrate is occured as a waste. But various elements including Cs and Sr are contained in the oxalate filtrate. It is very important, therefore, to separate from each other in a viewpoint of ultimate isolation of heat generating nuclides such as Cs and Sr from HLLW. In this study, the simulated oxalate filtrate, consisted of five elements (Cs, Sr, Ru, Fe and Mo) and organics (oxalic acid and ascorbic acid), was prepared, from which the separation of Cs and Sr was carried out by neutralization with NaOH. As a result, the simulated solution could be divided into Cs, Sr and Ru containing groups with pH. Sr was readily precipitated as forms of Sr(C₂O₄)·nH₂O at pH 3. Iron and ruthenium were also co-precipitated around pH 8, and Cs and Mo remained in the resulting solution.     

Sorption of the long-lived radionuclides cesium-134, strontium-85 and cobalt-60 on bentonite

The sorption of long-lived radionuclides of cesium, strontium and cobalt (¹³⁴Cs, ⁸⁵Sr and ⁶⁰Co) on bentonite under various experimental conditions, such as contact time, pH, sorbent and sorbate concentrations have been studied. The uptake of Cs and Sr was rapid and equilibrium was reached almost instantaneously in both the cases, while Co sorption was time dependent. The sorption of these nuclides increased by increasing pH. The uptake of Cs, Sr and Co increased with increasing the amount of the bentonite clay. The percentage sorption for Cs, Sr and Co decreased with increasing metal concentrations. The desorption studies with 0.01M CaCl₂ and ground water at low-metal loadings on bentonite showed that about 95% of Cs, 85-90% of Sr and 97% of Co were irreversibly sorbed. These results could be helpful for nuclear waste management, for waste water effluents containing low concentrations of cesium, strontium and cobalt.     

Fixation of radioactive isotopes on bentonite as radioactive waste treatment step

Fixation of¹³⁷Cs,¹⁴⁴Ce,⁶⁰Co,⁹⁰Sr,²⁴⁰Th and²³³U from aqueous and phosphate media on bentonite clay was studied. The fixation of the radioactive ions on bentonite surfaces was dependent on the pH behavior of the metal ions. A method was proposed to use bentonite as an absorbent of ions from simulated radioactive waste as a treatment step.     

Quality assessment of hydrosphere in the vicinity of Czech nuclear power plants by radioanalytical methods

Summary The paper deals with the impact of nuclear plants and radioactive waste disposal on surface and ground water quality in their vicinity using various radiometric and radioanalytical methods. The impact of nuclear power plant Temelin on activation concentrations and fission products in hydrosphere, including tritium, was detected. The annual average tritium concentrations in Vltava River correspond to the previously calculated estimates for average and minimal quaranteed flow rates. The concentrations histories of ⁹⁰Sr and ¹³⁷Cs in surface water show a decreasing trend. This trend was not influenced by the nuclear power plant pilot operation. In the case of tritium, a concentration increase trend has been already observed since the startup of pilot operation. An attempt has made interpreting the sorption and diffusion data for radionuclides of cesium, strontium and tritium and technetium as representatives of multivalent elements. Sorption and diffusion data of ¹³⁷Cs and ⁹⁰Sr in contact with natural sorbent bentonite lead to the conclusion that both diffusion and batch sorption experiments can be simulated by an exchange model. Sorption of technetium on various bentonites plus additives materials is described. Retention of technetium on these solid phases is driven by sorption of reduced form of technetium Tc(IV).     

Solvent Extraction Systems for Separation of An(III) and Ln(III): Overview of Static and Dynamic Tests

Moscow University Chemistry Bulletin - The management of high-level radioactive waste generated during the reprocessing of spent nuclear fuel in the PUREX process is among of the main problems of...     

Adsorption of La(III) onto GMZ bentonite: effect of contact time,bentonite content,pH value and ionic strength

Bentonite has been studied extensively because of its strong adsorption capacity. A local Na-bentonite named GMZ bentonite, collected from Gaomiaozi County (Inner Mongolia, China), was selected as the first choice of buffer/backfill material for the high-level radioactive waste repository in China. In this research, the adsorption of La (ΙΙΙ) onto GMZ bentonite was performed as a function of contact time, pH, solid content and metal ion concentrations by using the batch experiments. The results indicate that the adsorption of La (III) on GMZ bentonite achieves equilibration quickly and the kinetic adsorption follows the pseudo-second-order model; the adsorption of La (III) on the adsorbent is strongly dependent on pH and solid content, the adsorption process follows Langmuir isotherm. The equilibrium batch experiment data demonstrate that GMZ bentonite is effective adsorbent for the removal of La (III) from aqueous solution with the maximum adsorption capacity of 26.8 mg g⁻¹ under the given experimental conditions.     

Performance of immobilized moss in the removal of 137Cs and 90Sr from actual low-level radioactive waste solutions

The efficiency of immobilized moss as a bio-sorbent for the removal of ¹³⁷Cs and ⁹⁰Sr radionuclides from actual low-level radioactive waste (LLW) solutions was investigated. Preliminary batch experiments with the moss (Funaria hygrometrica) for the sorption of Cs and Sr have shown a pH dependent binding trend from pH 1–13, with maximum binding between pH 5–10. Time dependence of the batch studies showed that a contact time of 30 minutes was sufficient to reach equilibrium. Column experiments for the sorption of Cs and Sr by moss after immobilizing in polymer silica matrix demonstrated that the sorbent is capable of removing considerable amounts of Cs and Sr from actual LLW solutions under constant flow conditions. The adsorption capacity was estimated to be 8.5 mg/g for Cs and 15 mg/g for Sr. These sorbed metal ions from the column could be leached out using 0.20M nitric acid. The regenerated sorbent exhibited relatively the same initial binding capacity of both Cs and Sr even after 3 cycles of reuse. This revised version was published online in July 2006 with corrections to the Cover Date.     

Characterization of MX-80 bentonite and its sorption of radionickel in the presence of humic and fulvic acids

MX-80 bentonite is considered as one of the best backfill materials for high-level radioactive nuclear waste. Herein, the bentonite is characterized by using XRD and FTIR techniques. Sorption of radionickel to MX-80 bentonite in the presence/absence of humic acid (HA) or fulvic acid (FA) as a function of pH is investigated. The results indicate that the presence of HA or FA decreases the sorption of Ni²⁺ obviously. The different experimental processes do not affect the sorption of nickel to FA/HA bound bentonite. The sorption of Ni²⁺ on FA/HA-bound bentonite decreases with the increasing FA/HA content in the systems. The mechanism of nickel sorption is also discussed in detail.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Fe（Ⅱ）-矿物对亚硒酸的还原作用

79Se是核反应过程中产生的长半衰期（T1/2=2.95×105年）裂片产物核素,具有化学和辐射双重毒性,是高放废物地质处置中重点关注的几个放射性核素之一.硒的溶液化学性质比较特殊,以高价态形式存在的硒酸和亚硒酸,几乎不受溶解度控制,并难以被黏土或花岗岩等处置库围岩介质所吸附,迁移性极强,而当其以低价态（0、–I、–II）形式存在时,易形成固体沉淀.因此将高价态的硒还原生成沉淀是阻滞79Se迁移的最主要方式.从另一方面讲,Fe（II）-矿物是自然界中普遍存在的具有还原性的物质.本文对近年来国内外在Fe（II）-矿物还原亚硒酸方面开展的研究工作进行了一个回顾,并从热力学的角度,对这些矿物还原亚硒酸的可行性及控制产物生成的因素进行了分析.     

Implementation of Americium Separation from a PUREX Raffinate

Recovering of minor actinides from spent nuclear fuel is investigated for heterogeneous recycling in Generation-IV reactors. After plutonium, americium is the main contributor to residual heat of long term radioactive waste which determines waste density within geological repository. Selective americium separation (EXAm process) by liquid-liquid extraction from a PUREX raffinate was studied. Two experiments were performed in ATALANTE facilities. The first test, on surrogate solution, validated the americium extraction performance. The second trial was carried out in the high-level shielded process line from a genuine PUREX raffinate. Tools used to manage the process are introduced to show performance process achievement.     

Sorption of strontium on Slovak bentonites

Sorption of Sr on five Slovak bentonites of deposits has been studied with the use of batch technique. In the experiments there have been used natural, chemically modified and irradiated samples, in three different kinds of grain size. The pH influence on sorption of strontium on bentonites, pH change after sorption and influence of competitive ions have been studied. Distribution ratios have been determined for bentonite–strontium solution system as a function of contact time, pH and sorbate concentration. The data have been interpreted in term of Langmuir isotherm. The uptake of Sr has been rapid and the sorption of strontium has increased by increasing pH. The percentage sorption has decreased with increasing metal concentrations. The pH value after sorption for the natrificated forms of bentonite starts already in the alkaline area and moves to the higher values. For the natural bentonites the values occur in the neutral or in the acidic area. Sorption of Sr has been suppressed by presence of competitive cations as follows: Ba²⁺ > Ca²⁺ > Mg²⁺ > NH₄ ⁺ > K⁺ > Na⁺. By sorption on natrificated samples colloidal particles and pH value increase have been formed. The bentonite exposure as a result of interaction of γ-rays has led to expansion of the specific surface, increasing of the sorption capacity and to the change in the solubility of the clay materials.