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.     

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.     

Determination of the leaching rate of radionuclide 134Cs from the solidified radioactive wastes in Syrian Portland cement and cement–microsilica matrixes

The suitability of Syrian Portland cement for disposal of solidified low-level radioactive waste was assessed by measuring the leaching rate of ¹³⁴Cs. In ordinary cement concrete, a leaching rate of 1.309 × 10^?3 g/cm² per day was measured. Mixing this concrete with microsilica reduced significantly the leaching rate to 3.106 × 10^?4 g/cm² per day for 1% mixing, and to 9.645 × 10^?5 g/cm² per day for 3% mixing. It was also found that the application of a latex paint reduced these leaching rates by about 10%. These results, along with mechanical strength tests (under radiation exposure, high temperature, long water immersion and freeze–thaw cycling) indicate that Syrian Portland cement is suited for the disposal of low-level radioactive waste.     

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     

Effect of curing time on the fraction of137Cs leached from cement matrix

To assess the safety of disposal of radioactive waste material in concrete, curing conditions and time of leaching radionuclide¹³⁷Cs have been studied. Leaching tests in concrete were carried out in accordance with a method recommended by IAEA¹.     

Effect of container adsorption on the210Pb/210Bi secular equilibrium relationship in spiked aqueous solutions

The radioactive boric acid wastes generated from the nuclear power plants have been solidified with cement. One of the properties required for solidified radioactive wastes is resistance to leaching. The leachability of¹³⁷Cs from solidified waste specimens prepared by various formulation has been studied according to the proposed ANS 16.1 Standard Leach Test. For the evaluation of radiological safety in the solidified radioactive waste disposal, the leachability indices have been determined from the leaching test data. The results have turned out to be 5.97–7.60, depending on formulation and solidifying matrix.     

Leaching kinetics of <Superscript>54</Superscript>Mn and <Superscript>89</Superscript>Sr radionuclides fixed in cement­vermiculite composition

Summary Leach characteristics of ⁵⁴Mn and ⁸⁵Sr radionuclides from ordinary Portland cement have been studied using International Atomic Energy’s (IAEA) standard leach method. The retardation factors, K_F, and coefficients of distribution, k_d, have been determined using a simplified mathematical model for analyzing the migration of radionuclides. The lowest leaching values after 60 days were achieved in samples with 5% of vermiculite. Results presented in this paper are the examples of results obtained in a 10 year mortar and concrete testing project, which will influence the design of the engineered trench system for a future central Serbian radioactive waste storage center.     

A radioanalytical phantom for assessing the efficacy of electrokinetic decontamination of entrained radioactivity within concrete media

A phantom for the evaluation of electrokinetic remediation of radioactive species from water saturated concrete is described. The phantom has been designed to be a general analogue for environments where structural concrete is saturated by radioactive aqueous solutions and where electrokinetic remediation may be deployed. It is also a specific analogue for the walls of storage ponds for legacy spend nuclear fuel pins where the pond water comprises a large volume of hazardous active waste that may penetrate the pond wall. The fabricated phantom was evaluated using a fixed electrical potential to monitor the rate of cationic transport of K⁺ through concrete samples of different thickness (20 and 35 mm respectively). Results of the evaluation show K⁺ diffusion coefficients of 5.20 × 10^?13 and 7.61 × 10^?13 m² s^?1 for the 20 mm and 35 mm samples, consistent with those seen in literature for the transport of caesium through concrete of similar thickness. The phantom offers a means of experimental validation of computational electrokinetic models as well as providing a basis to test the effects of electrode material on ionic transport rates, to interrogate the effects of pH on all components of the system, and as a basis for instruction, education and training in nuclear decommissioning and waste treatment.     

Leaching studies of137Cs from ion-exchange resin incorporated in cement or bitumen

Leaching of¹³⁷Cs from ion-exchange resin incorporated in cement and bitumen composite using two methods based on theoretical equations has been developed. These were: Method I, diffusion equation derived for a plane source model¹ and Method II, empirical model employing a polynominal equation^2,3. Results presented in this paper are examples of data obtained in a cement and bitumen testing project which will influence the design of a future radioactive waste storage center.     

Migration of radiocesium in a zeolite-solution system

Some properties of natural zeolite from Zaloka gorica, Slovenia, Yugoslavia were investigated to establish its applicability in solidification and for storage purposes of radioactive waste. It was found that this material is a rather good sorber for cesium, with a capacity of about 0.6 meq g^–1. The migration rate of Cs⁺ in a system ion exchanger-aqueous solution was investigated and correlated with the sorption behaviour of cesium in a system. The leaching rate of cesium from concrete containing zeolite was measured.     

Leaching of uranium,radium and thorium from vertisol soil by ground water

Shallow land burial is routinely used for the disposal of low-level radioactive waste. Natural processes causing leaching of radionuclides can lead to contamination of surrounding ground water and soil by the radionuclides. The comparative leachability of radionuclides U_(nat), ²²⁶Ra, ²²⁸Ra and Th_(nat) from the soil of a radioactive waste disposal site, by ground water was evaluated. The probability of leaching was obtained in the following order Ra (≈77%) > U (≈40%) > Th (≈20%). Observed ratios (OR) were calculated to correlate leachability of radionuclides to that of major cations Ca²⁺ and Mg²⁺. The leaching of the radionuclides was seen to be dependent on Ca²⁺ and SO₄²⁻ leached from the soil. This study provides sitespecific leachability of radionuclides, that can be used as indicator of the tendency for migration or retention in soil. It can play an important role during an unforeseen accident like breach of containment at the waste disposal site leading to contamination of soil and ground water and causing hazard to public via drinking water route.     

Corrosion of simulated nuclear waste glass

Glasses are suitable host matrices for the immobilization of high-level radioactive wastes. The corrosion behaviour of nuclear waste glass in water is of considerable importance, since a potential route for returning of radionuclides to the biosphere is their leaching from the waste form into groundwater and subsequent transport by the groundwater to the surface. In this study, the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilicate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates (g·cm^–2·day^–1), as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na₂O, 10.39% B₂O₃, 45.31% SiO₂, 13.42% ZnO, 6.61% TiO₂ and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed.     

Effect of neutralizing agent content on137Cs leaching from solidified boric acid waste products

For the preservation of environment from radioactive contamination, one of the properties necessary for solidified radioactive wastes is resistance to the release of radionuclides due to leaching by ground water after land disposal. In order to seek the optimum solidifying formulation for cement solidified boric acid concentrate from pressurized water reactor, a variety of specimens were prepared, varying the content of neutralizing agents and neutralization methods. For the leaching measurements, the ANS 16.1 Standard Leach Test was carried out for the specimens which had been mixed with small concentration of¹³⁷Cs to estimate the leachability index. According to our investigation, the optimum contents of neutralizing agents for neutralization of boric acid radioactive waste were determined in view of resistance ability to leaching. Eventually the leachability indices have turned out to be 5.807.91 depending on the formulations.     

The MEDINA facility for the assay of the chemotoxic inventory of radioactive waste packages

A new analytical facility called MEDINA (Multi Element Determination based on Instrumental Neutron Activation) and based on Prompt- and Delayed-Gamma-Neutron-Activation-Analysis (P&DGNAA) using a 14 MeV neutron generator is developed for the determination of non-radioactive elements and substances in 200-l radioactive waste drums. The qualitative performance of MEDINA is studied by measuring (1) the gamma-ray spectra for the empty chamber, (2) for an empty 200-l steel drum in presence or not of ⁶⁰Co and ¹³⁷Cs sources and (3) for a 200-l steel drum filled with concrete in order to determine the level of interferences induced by activation products of the facility components and to investigate further optimization steps to improve the measurements sensitivity. The measurement carried out for the empty drum with the radioactive sources show the possibility to characterize radioactive waste packages containing both ⁶⁰Co and ¹³⁷Cs with activities ranging between 20 and 80 MBq at least.     

Use of radiometric emanation method in the characterization of anthropogenic glass analogue for vitrification of nuclear waste

Anthropogenic analogues can be used as one of the sources of information about long-term behaviour of engineered barriers, used for geological disposal of radioactive waste. The radiometric emanation method, based on the measurement of radon release from the solid samples, was used in the study of the alteration of anthropogenic glass analogues for radioactive vitrified waste. The samples were labeled at their surface by a solution containing ²²⁸Th and ²²⁴Ra,serving as source of ²²⁰Rn nuclides. The radon ²²⁰Rn nuclides were introduced in the uranium glass samples due to the recoil energy of the α-decay of ²²⁸Th and ²²⁴Ra nuclides. The man-made uranium glass samples from a locality in the Czech Republic were used. The measurement, based on the release of radon atoms from the samples, was applied to characterize the radon diffusion permeability and microstructure changes of weathered and non-weathered uranium glass samples during heat treatments. The radon diffusion characteristics and microstructure stability of the uranium glass analogue samples were evaluated by using a mathematical model.     

Treatment of phosphogypsum waste using suitable organic extractants

Phosphogypsum (PG) is a residue of the phosphate fertilizer industry that has relatively high concentrations of harmful radioactive materials. The reduction in concentration of the radionuclides from PG was investigated. The removal process is based on leaching of radionuclides using suitable organic extractants. The studied radionuclides were ²²⁶Ra, ²¹⁰Pb, ²³⁸U and ⁴⁰K. The factors affect the leaching process such as type of leaching materials, contact time, concentration of the desired solvent, liquid to solid ratio, and temperature were studied. Based on the experimental results, about 71.1, 76.4, 62.4, and 75.7% of ²²⁶Ra, ²¹⁰Pb, ²³⁸U and ⁴⁰K respectively were successfully removed from the PG. The reduction in the concentration of radionuclides was accompanied by reduction in the concentration of rare earth elements (∑REE) equals to 69.8%. Using the desired organic extractant under optimum conditions for treatment of the PG waste leads to obtain a decontaminated product that can be safely used in many industrial applications.     

Carbon sorbents for immobilization of tritium-containing waste oils

Emission of tritium from the pump oil-carbon sorbent-cement compositions (where the carbon sorbent is thermally expanded graphite or charcoal fabricated by thermal or radiation evaporation of a cellulose material) was studied. The compositions are characterized by a high hold-up ability with respect to oil components, when the matrix is contacted with water. The apparent diffusion coefficients of tritium are equal to 10⁻¹⁵–10⁻¹⁶ m² s⁻¹. The compositions are found to be resistant to radiation and can be recommended for solidification of radioactive waste oils.     

Synthetic calcium aluminosilicate monolith: V. Change in the pore structure in hydration and carbonization

The physicochemical properties of synthetic calcium aluminosilicate (SCAS) monoliths produced from fly ash, limestone, and sand in a three-stage process (filtration combustion with superadiabatic heating, fine grinding, and pressing) were studied. It was found that hydration and carbonization in a SCAS monolith during long hardening under natural (laboratory) conditions lead to perfection of the structure of pores, which improves its physicochemical properties. The presence of unreacted β-Ca₂SiO₄ in the SCAS monolith throughout the hardening period ensures its high immobilizing properties under the action of the hydrosphere on the matrix containing hazardous (including radioactive) wastes because of calcium hydrosilicate gel formation, which decreases the pore space volume. Examples were given for determining the dependence of the total rate of leaching of SCAS monoliths by deionized water at 90°C on the treatment time (MCC-1 test). The rate of leaching of a SCAS-MRW monolith (where MRW is model radioactive waste of closed nuclear fuel cycle) was found to be 6.7 × 10^?7, 7.2 × 10^?7, and 8.3 × 10^?7 g cm^?2 day^?1 at MRW contents of 10, 20, and 30 wt %, respectively. The possibility of integrated solutions of some environmental problems using energy- and resource-saving technologies was considered.     

Analysis of 99Tc in the radioactive liquid waste after extraction into suitable solvent

⁹⁹Tc is one of the long lived fission product with high fission yield. From radioactive waste management point of view it is very much essential to evaluate the concentration of technetium in the radioactive liquid waste in order to finalise the treatment process to extract/isolate it from the stream which is discharged to the environment. For the estimation of ⁹⁹Tc in the radioactive liquid waste stream, extraction of the stable complex of technetium-tetraphenyl arsonium chloride (TPAC) into chloroform followed by beta counting was studied. Various parameters like pH, time of equilibration, concentration of TPAC in chloroform, use of other solvent for extraction as well as interference of various other radionuclides present in the waste were also studied. The radioactive liquid waste being handled in plant contains high concentrations of salts in the form of sodium nitrate. Hence effect of salt concentration on the percentage extraction was also evaluated. The extraction behavior does not dependent on change in the pH of the solution. Almost 99.5% extraction was observed in the pH range of 1?C13.0. High concentration of salt is affecting the extraction. However, this can be taken care by diluting the radioactive waste. It takes almost 90?min time for maximum extraction. Presence of radionuclides like ¹³⁷Cs, ⁹⁰Sr are not interfering the extraction of ⁹⁹Tc. However, ¹⁰⁶Ru is getting slightly extracted along with ⁹⁹Tc. The error due to ¹⁰⁶Ru can be eliminated by taking gamma spectrum and deducting the activity from the total beta activity to get ⁹⁹Tc activity. Nitrobenzene can be used for extraction of Tc?CTPAC complex in place of chloroform.     

Structural modifications in iron hydrated oxides induced by foreign cation

The aqueous radioactive wastes are complex chemical systems as consequence of the variety of the contained cations. The chemical systems precipitated from FeSO₄ and FeCl₃ simulating the aqueous radioactive waste treatment have been investigated by Mössbauer spectroscopy. Mg²⁺, Ca²⁺ and Ba²⁺ were used as foreign cations, the calcium being among the main components of aqueous radioactive waste. The results showed the ion radius influence varying from Mg²⁺ to Ba²⁺ in the case of the precipitation from FeSO₄. For the precipitation from ferric chloride the influences are not so evident.