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).     

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.     

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.     

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.     

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.     

Adsorption of cesium and strontium on natrified bentonites

The influence of chemical activation–natrification of bentonites on adsorption of Cs and Sr was studied with regards to utilization of bentonites for depositing high-level radioactive waste and spent nuclear fuel. Bentonite samples from three Slovak deposits in three different grain-size (15, 45 and 250 μm), natural and natrified forms (Na-bentonites); under various experimental conditions, such as contact time, adsorbent and adsorbate concentration have been studied. When comparing the Na-bentonites and their natural analogues, the highest adsorbed Cs and Sr amounts were reached on the natrified samples. After the Sr adsorption a drop in the pH equilibrium value was observed together with the increase of the initial Sr concentration. A disadvantage of the natrified bentonite forms is formation of colloid particles. After 2 h of phase mixing a gentle turbidity was observed as well as formation of a gel-like form. The above findings were confirmed by observing the particle distribution in dry and wet dispersion and centrifugation at two different speeds. Natrification as a technological process of bentonite quality improvement cannot be applied when constructing a long-term repository for high-level radioactive waste and spent nuclear fuel. The main problem of natrification is a technological process which leads to a significant pH increase. Alkaline environment in combination with the K presence and increased temperature in the vicinity of radio-active waste can lead to a rapid illitization of smectite and loss of the original adsorption qualities. Moreover, sodium additions are a significant point of uncertainty since it is not possible to state what amount of Na enters the interlayer space and what amount stays in the inter-partition space.     

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.     

Thermal analysis of reference repository for RBMK-1500 spent nuclear fuel in crystalline rocks

Total amount of RBMK-1500 type spent nuclear fuel (SNF) in Lithuania is approximately 22 thousands of fuel assemblies. All these assemblies should be stored for 50 years and then disposed of. International consensus prevails that SNF and long-lived high-level radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. Disposed nuclear waste induces a number of coupled thermo-hydro-mechanical processes around the repository. Thermal analysis of a deep geological repository could provide temperature distribution which is required for the repository’s design and for evaluation of thermal integrity of the engineered barriers. One of the most critical parameters for the repository is peak temperature at the outer surface of the canister. This temperature cannot exceed 100 °C; otherwise, unfavourable groundwater chemistry can adversely affect chemical stability of the engineered barriers. Thermal behaviour of the conceptual Lithuanian repository for RBMK-1500 type SNF in crystalline rocks was modelled using numerical codes ANSYS FLUENT and COMPASS. Very similar temperature distributions around the disposal canister were determined using both modelling tools. The modelling results revealed the importance of coupled heat and hydrodynamical processes for peak temperature in the engineered barriers, whereas the impact of mechanical processes evaluation was insignificant. It was also determined that peak temperature at the outer surface of the disposal canister does not exceed the permitted 100 °C.     

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).     

Possibility of Various Types SNF Reprocessing at the PA Mayak exampled with AMB SNF

For the purpose of reprocessing of irradiated nuclear fuel from the water-cooled graphite-moderated pressure-tube reactor named AMB from decomissioned Russian “Atom Peaceful Big”, modernization of the process flow-sheet of the RT-1 plant is being carried out at PA Mayak with participation of FSUE KRI and VNIINM. A particular AMB SNF feature is extremely broad range of fuel compounds with the main ones being the uranium-molybdenum metal, uranium oxide and uranium carbide compositions usually dispersed in magnesium or calcium. Wide range of fuel compositions required to amend SNF dissolution, extraction processing, evaporation of high-level radioactive wastes and vitrification of high-level radioactive wastes. The above set of laboratory research was completed with dynamic tests using samples of AMB from the water-cooled graphite-moderated pressure-tube reactor. Tests have shown the possibility of processing the entire range of AMB SNF at the radiochemical plant RT-1 plant of the PA Mayak. Thus, the ability of the RT-1 plant to process different fuel compositions, including the long-term research reactor fuel have been proved experimentally.     

Pre-concentration and determination of 90Sr in radioactive wastes using solid phase extraction techniques

This paper describes the use of IBC??s AnaLig^®Sr-01 molecular recognition technology product to effectively and selectively pre-concentrate, separate and recover strontium from radioactive waste samples. The use and effectiveness of AnaLig^®Sr-01 gel was successfully validated by analysis of International Atomic Energy Agency (IAEA 375) reference soil and National Physical Laboratory (NPL)?CHigh Alpha?CBeta (2003) liquid sample. The second part of this paper focuses on analysis of radioactive waste samples from nuclear power plant A1 Jaslovske Bohunice in Slovak Republic (NPP A1).     

Determination of <Superscript>129</Superscript>I using volatilization method and liquid scintillation spectrometry

A simple and rapid separation method for ¹²⁹I determination in radioactive waste samples was developed. Suitable conditions for iodine volatilization were tested. Iodine was trapped in 1.5 mol L^?1 NaOH and precipitated as PdI₂·H₂O by addition of PdCl₂ with recoveries higher than 80%. The method was applied for analysis of contaminated soil, radioactive sludge, evaporator concentrate and heterogeneous waste samples from nuclear power plants in Slovak Republic. ¹²⁹I was measured on liquid scintillation counter TRI CARB 2900 TR using Ultima Gold AB scintillation cocktail.     

Hydrometallurgical Reprocessing of BREST-OD-300 Mixed Uranium-plutonium Nuclear Fuel

The duration of external fuel cycle of BREST-OD-300 reactor with mixed U-Pu nitride fuel (MNIT) including hydrometallurgical reprocessing should not exceed 3 years. An average burnup of the fuel should be 6% of heavy metal (HM) with the potential increase up to 10% HM. Therefore, the technology should provide the reprocessing of spent nuclear fuel (SNF) after less than 2 years cooling time and with fissile materials (FM) content of 10 – 15%. Pellets technology has been chosen for the MNIT fuel production. That means necessity to receive the recycled actinides oxides of high purification coefficient (∼ 10⁶). Currently on a laboratory scale, the following process stages have been tested on the real products: actinide oxides production and rare-earth and trans-plutonium elements separation. Moreover, on a pilot scale the process of high level radioactive waste (HLW) and intermediate level radioactive waste (ILW) concentration by evaporation has been tested, as well as the Am-Cm separation. In 2015, the design of the MNIT SNF reprocessing facility has been started, placed at the JSC Siberian Chemical Plant site as a part of the pilot demonstration power complex (PDPC) with BREST-OD-300 reactor. MNIT SNF reprocessing plant (RP) should be put in operation after 2020.     

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.     

Sorption of technetium and rhenium on natural sorbents under aerobic conditions

The uptake of technetium on bentonite materials has been studied from the point of view of the characterization of long-term radioactive elements behavior in nuclear waste repository. It is generally known that bentonite materials show an excellent cation-exchange capacity and on the other hand a poor uptake of anions. The aim of our research has been to find out the conditions suitable for technetium sorption on selected bentonite under oxidizing conditions. The influence of the addition of different materials (e.g., activated carbon, graphite, Fe²⁺, Fe) with bentonite, the effect of solid:aqueous phase ratio and a pH value on the percentage of technetium uptake and on the K _d values were tested.This revised version was published online in November 2005 with corrections to the Cover Date.     

Interaction of <Superscript>99</Superscript>Tc onto bentonite and influence of Fe(II) and (Ca,Mg) ion exchange in its interlayers

⁹⁹Tc is a redox active radionuclide, which is present as contaminant at a number of sites where nuclear fuel cycle operations have been carried out. The aim of our research was study of bentonite interaction with Fe and Fe²⁺ cation and these influences on the migration behavior. Radioanalytical and analytical methods were used for the concentration and chemical forms of ⁹⁹Tc determination. It was found that Fe²⁺–Ca²⁺, Mg²⁺ ion exchange should be considered as a process of reaction among the corrosion products of carbon steel container and bentonite in the environment of radioactive waste repository.     

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.     

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...     

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).     

Monitoring of bentonite pore water with a probe based on solid-state microsensors

Repositories for the disposal of radioactive waste generally rely on a multi-barrier system to isolate the waste from the biosphere. This multi-barrier system typically comprises the natural geological barrier provided by the repository host rock and its surroundings and an engineered barrier system (EBS). Bentonite is being studied as an appropriated porous material for an EBS to prevent or delay the release and transport of radionuclides towards biosphere. The study of pore water chemistry within bentonite barriers will permit to understand the transport phenomena of radionuclides and obtain a database of the bentonite-water interaction processes. In this work, the measurement of some chemical parameters in bentonite pore water using solid-state microsensors is proposed. Those sensors are well suited for this application since in situ measurements are feasible and they are robust enough for the long periods of time that monitoring is needed in an EBS. A probe containing an ISFET (ion sensitive field effect transistor) for measuring pH, and platinum microelectrodes for measuring conductivity and redox potential was developed, together with the required instrumentation, to study the chemical changes in a test cell with compacted bentonite. Response features of the sensors’ probe and instrumentation performance in synthetic samples with compositions similar to those present in bentonite barriers are reported. Measurements of sensors stability in a test cell are also presented.