Leaching behavior of137Cs in cement

To assess the safety of disposal of a radioactive waste-cement composite, the leaching of¹³⁷Cs from a waste composite into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA.¹ The leachability was measured as a function of bentonite clay to cement ratio. The fraction of¹³⁷-Cs leached from a specimen of Portland cement is 0.03–0.13 at a leaching time of 400 d. Results presented in this paper are examples of data obtained in a 10 y mortar and concrete testing project, which will influence the design of the engineered trench system for a future Yugoslav radioactive waste storage center.^2,3     

Sorption properties of chemically treated clinoptilolites with respect to Cs and Co

The sorption ability of clinoptilolite treated by NaOH solution has been studied. The distribution coefficients of¹³⁷Cs and⁶⁰Co were studied by the radiochemical method. The variation of distribution coefficients with pH was studied. Also the influence of competitive ions on distribution coefficients were investigated. The total cation exchange capacity and XRD analysis were also measured.     

Radioactive waste treatment using natural Syrian bentonite

The uptake of¹³⁷Cs and⁹⁰Sr/⁹⁰Y onto Syrian bentonite has been studed, using batch and column tracer techniques in order to investigate its utilization for aqueous radioactive effluent treatment. Parameters influencing the percent uptake considered and studied in this work are: nuclide concentration, pH of the aqueous phase, heat treatment and particle size. Leaching experiments using natural sea and ground waters were carried out on bentonite/cement composites. They demonstrated the effectiveness of calcination and cement containment.     

Mathematical modeling of physico-chemical characteristics of cement-waste composites in radioactive waste management

An optimization of mortar (as matrix), improved with bentonite clay, used for immobilization of radionuclides⁶⁰Co,¹³⁷Cs,⁸⁵Sr and⁵⁴Mn, is presented. A relatively simple mathematical model is given, which permits minimization of leach rate and permeability and maximization of compressive strength. An optimal solution, based on experimental data, is given. These results will be used for a future Yugoslav radioactive waste storing center.     

Cesium exchange reaction on natural and modified clinoptilolite zeolites

Cesium cation exchange reaction with K, Na, Ca and Mg ions on natural and modified clinoptilolite has been studied. Batch cation-exchange experiments were performed by placing 0.5 g of clinoptilolite into 10 ml or 20 ml of 1·10⁻³M CsCl solution for differing times. Two type deposits of clinoptilolite zeolites from, Nižny Hrabovec (NH), Slovakia and Metaxades (MX), Greece were used for ion-exchange study. The distribution coefficient (K _d ) and sorption capacity (Γ) were evaluated. For the determination of K, Na, Ca and Mg isotachophoresis method, the most common cations in exchange reaction was used. Cesium sorption was studied using ¹³⁷Cs tracer and measured by γ-spectrometry.     

Distribution characteristics of <Superscript>14</Superscript>C and <Superscript>3</Superscript>H in spent resins from the Canada deuterium uranium-pressurized heavy water reactors (CANDU-PHWRs) of Korea

It is regarded that the spent resins from the water purification systems of moderator (MOD) and the primary coolant of the Canada deuterium uranium-pressurized heavy water reactor (CANDU-PHWR) are a unique waste, owing to their high ¹⁴C and gamma-emitting nuclides. In this work, ¹⁴C and ³H contents, anion and cation fractions and the predominant gamma-emitting nuclides of the spent resins from 4 units of CANDUPHWRs, were investigated. Also the chemical species of ¹⁴C of the spent resins were determined. For a simultaneous separation of ¹⁴C and ³H from the spent resins, the wet oxidation-16 wt% H₂SO₄ stripping process was utilized. The ¹⁴C and ³H activity concentration range of the spent resins of the nuclear power plant (NPP), 4 units of all CANDU-PHWR types, was 2.48E5 Bq/g ∼5.33E6 Bq/g, 1.29E5 Bq/g and ∼2.33E5 Bq/g, respectively. Among the analyzed spent resins, the highest ¹⁴C and ³H activity concentration was detected in units 4 and 3, respectively. It was found that more than 92% of the ¹⁴C activity concentration was retained on the anion resin and the predominant chemical species was inorganic ¹⁴C. It was revealed that the anion resin fraction of the spent resins from unit 1 and unit 2, was about 40% and that of unit 3 and unit 4 was around 60%. More than 80% of the total gamma-radioactivity concentration was associated with the cation fraction of the spent resin. The predominant gamma-emitting nuclide of the spent resin for unit 2 was ¹³⁷Cs, a fission product, and that for unit 4 was ⁶⁰Co, a corrosion product.     

Diffusion of 137Cs in compacted bentonite: Effect of pH and concentration

The effect of pH and concentration on the diffusion of ¹³⁷Cs in the compacted bentonite is studied with capillary method at the ionic strength of 0.1M NaClO₄. The apparent diffusion coefficient increases with increasing concentrations if the sorption of radionuclides is largely dependent on the radionuclide concentrations. The apparent diffusion coefficient decreases with increasing pH because most of the radionuclide sorption on the bentonite increases with increasing pH. The interlaminary space contributes significantly to the radionuclide diffusion and sorption in compacted bentonite. The relationship of the apparent diffusion coefficient and the effective diffusion coefficient of ¹³⁷Cs is also discussed.     

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.     

The Use of Natural Zeolites for Radioactive Waste Treatment: Studies on Leaching from Zeolite/Cement Composites

Samples of the natural zeolites chabazite, clinoptilolite and a clinoptilolite-rich tuff, were loaded with the isotope ¹³⁷Cs. Composites of these labeled materials were made with cement and blast furnace slag. Standard leaching experiments were carried out with synthetic sea, ground and "pond" waters, as well as distilled water. Rates of leaching were calculated and compared to similar systems.     

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.     

Sorption characteristics of137Cs onto clay minerals: Effect of mineral structure and ionic strength

The sorption behavior of¹³⁷Cs onto kaolinite, bentonite, illite, and zeolite was studied at different ionic strengths of Na⁺, K⁺, Ca²⁺. A significant effect of ionic strengths on the sorption has been observed. Clay minerals with 21 structure (bentonite, illite) showed much higher sorption than that of 11 structure (kaolinite). Zeolite showed high selectivity for¹³⁷Cs sorption. Sorption behavior of¹³⁷Cs on clay minerals can be explained by their surface charge characteristics originated from mineral structure.     

Comparative interpretation of leach rate of137Cs and60Co from cement matrix

A technique for leaching¹³⁷Cs and⁶⁰Co from a cement matrix, using three methods based on theoretical equations, has been developed. These were: Method I, diffusion equation derived for a plane source model, Method II, rate equation for diffusion coupled with a first-order reaction. The leaching data were also analyzed by an empirical method employing a polynomial equation, Method III. Results presented in this paper are examples of data obtained in a cement testing project which will influence the design of the future radioactive waste storage center.     

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.     

Comparison of thermal ionization mass spectrometry and Multiple Collector Inductively Coupled Plasma Mass Spectrometry for cesium isotope ratio measurements

In the nuclear domain, precise and accurate isotopic composition determination of elements in spent nuclear fuels is mandatory to validate neutron calculation codes and for nuclear waste disposal. The present study presents the results obtained on Cs isotope ratio by mass spectrometric measurements. Natural cesium is monoisotopic (¹³³Cs) whereas cesium in spent fuels has 4 isotopes (¹³³Cs, ¹³⁴Cs, ¹³⁵Cs, and ¹³⁷Cs). As no standard reference material is available to evaluate the accuracy of Cs isotopic measurements, a comparison of cesium isotopic composition in spent nuclear fuels has been performed between Thermal Ionization Mass Spectrometry (TIMS) and a new method involving Multiple Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICPMS) measurements. For TIMS measurements, isotopic fractionation has been evaluated by studying the behavior of cesium isotope ratios (¹³³Cs/¹³⁷Cs and ¹³⁵Cs/¹³⁷Cs) during the analyses. For MC-ICPMS measurements, the mass bias effects have been corrected with an external mass bias correction using elements (Eu and Sb) close to cesium masses. The results obtained by the two techniques show good agreement: relative difference on ¹³³Cs/¹³⁷Cs and ¹³⁵Cs/¹³⁷Cs ratios for two nuclear samples, analyzed after chemical separation, ranges from 0.2% to 0.5% depending on the choice of reference value for mass bias correction by MC-ICPMS. Finally the quantification of the ¹³⁵Cs/²³⁸U ratio by the isotope dilution technique is presented in the case of a MOx (mixed oxide) spent fuel sample. Evaluation of the global uncertainties shows that this ratio could be defined at an uncertainty of 0.5% (k = 2). The intercomparison between two independent mass spectrometric techniques is fundamental for the evaluation of uncertainty when no isotopic standard is available.     

Characterization of modified clinoptilolite

Samples of clinoptilolite were modified using insoluble hexacyanoferrate from aqueous solution. The modified samples were characterized by elemental analysis, powder X-ray diffraction, solid state NMR and vibrational spectroscopy. The sorption properties of modified clinoptilolite were studied, too. Higher affinity for¹³⁷Cs sorption in comparison with the natural clinoptilolite has been proved.     

Sorption of Cs(I) and Sr(II) on a mixture of bentonite and magnetite using SCM + IExM: A parametric study

Summary The sorption of Cs(I) and Sr(II) on bentonite and magnetite was experimentally studied and numerically simulated using surface complexation (SCM) and ion-exchange (IExM) models. The empirical system consisted of: (1) synthetic granitic water with a given ionic strength (0.1 or 0.01 NaNO₃), (2) radionuclides studied (10^-6M CsCl or SrCl₂ ^. 6H₂O spiked with ¹³⁷Cs or ⁸⁵Sr), and (3) bentonite pre-treated with the aim to remove carbonates, and magnetite as a representative of steel canister corrosion products. The parametric study covered, e.g., the influence of pH, bentonite to magnetite ratio and volume-liquid ratio on the values of selectivity coefficients and K_d.     

Pick up of cesium and cobalt activity by oxide sludge in steam generator

A pinhole developed near the tube sheet in the steam generator (SG) tube of a pressurised heavy water reactor (PHWR) caused leak of primary coolant containing radioactive contaminants (¹³⁷Cs, ¹³⁴Cs and ⁶⁰Co) to the shell side. The sludge collected from the tube sheet region was found to have adsorbed these radionuclides at the high temperature (230-240 °C) that prevailed in the SG. An attempt has been made to evaluate the quantity of activity retained in the various oxide phases that constitute the sludge and their mode of pick-up. The sludge was characterized by XRD and XPS, which showed the presence of various oxides of iron, copper and nickel along with the silicates of calcium, magnesium and aluminum. Gamma-spectrometry of the sludge confirmed the presence of ¹³⁷Cs¹³⁴Cs and ⁶⁰Co to an extent of 7.6, 1.3 and 0.9 mCi/g of sludge, respectively. Selective dissolution in various EDTA based formulations and equilibration with nitric acid and magnesium chloride solutions helped to understand the quantity of activity adsorbed by various constituents of the sludge. It was concluded that a major portion of cesium was picked up by a reversible ion exchange process on various oxide constituents and about 10% by an irreversible specific adsorption process on insoluble silicates. Also, it was proved that ⁶⁰Co was specifically adsorbed over the oxides of iron and nickel.     

The effect of the synthesis method on the parameters of pore structure and selectivity of ferrocyanide sorbents based on natural minerals

Ferrocyanide sorbents were obtained via thin-layer and surface modification of natural clinoptilolite and marl. The effect of modification method on surface characteristics of these sorbents and their selectivity for cesium was studied. It was shown that the modification resulted in an increase of selectivity of modified ferrocyanide sorbents to cesium as compared with the natural clinoptilolite in presence of Na⁺, as well as in an increase of cesium distribution coefficients in presence of K⁺. The nickel–potassium ferrocyanide based on the clinoptilolite showed the highest selectivity for cesium at sodium concentrations of 10^?4—2 mol L^?1: cesium distribution coefficient was lg K _d = 4.5 ± 0.4 L kg^?1 and cesium/sodium separation factor was α(Cs/Na) = 250. In the presence of NH₄ ⁺, all modified sorbents showed approximately equal selectivity for ¹³⁷Cs. Probable applications of the sorbents were suggested.     

Diffusion of Sr, Cs, Co and I in argillaceous rock as studied by radiotracers</p> </p>

Summary Diffusions of⁸⁵Sr,¹³⁷Cs,⁶⁰Co and¹²⁵I radionuclides have been studied in borecore samples from Boda siltstone/claystone formation (BCF) under ambient and in situ conditions. In-diffusion (⁸⁵Sr,¹³⁷Cs,⁶⁰Co and¹²⁵I) and through-diffusion measurements (¹²⁵I) were performed at ambient conditions, and for iodine, in-diffusion measurements were also carried out at in situ conditions (100 bar, 50 °C). In the case of cationic species carrier-containing solutions were also applied.⁶⁰Co was detected only in the first slice of borecore at each sample, while¹³⁷Cs was detected also in the first-, second- and third slices according to the concentration-increase of inactive carrier. Among the investigated cations,⁸⁵Sr exhibited the fastest diffusion rate with 2.7-6.0 ^. 10^-12m²/s apparent diffusivity values. In the course of in-diffusion measurements 4.7 ^. 10^-11 m ² /s, during through-diffusion investigations 1.4-1.6 ^. 10^-12m²/s and at in situ conditions 5.0-8.0 ^. 10^-12 m²/s apparent diffusivities were obtained for¹²⁵I.Modest sorption of¹²⁵I can also be deduced from the results.</p> </p>     

Comparative interpretation of leaching137Cs from radwaste sludge immobilized in cement

This paper presents results from studies made to determine the leachability of¹³⁷Cs from immobilized evaporator sludge from a Pressurized Water Reactor with cement. Leaching of¹³⁷Cs from cement matrix using three methods based on theoretical equations has been developed. These were: Method I, diffusion equation derived for a plane source model, Method II, rate equation for diffusion coupled with a first-order reaction. The leaching data were also analyzed by an empirical method employing a polynomial equation —Method III. Results presented in this paper are examples of data obtained in a cement testing project which will influence the design of a future radioactive waste storage center.