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.     

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.     

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 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.     

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.     

Determination of Cs and Cs in environmental samples: A review

Radionuclides of caesium are environmentally important since they are formed as significant high yield fission products (¹³⁵Cs and ¹³⁷Cs) and activation products (¹³⁴Cs and ¹³⁶Cs) during nuclear fission. They originate from a range of nuclear activities such as weapons testing, nuclear reprocessing and nuclear fuel cycle discharges and nuclear accidents. Whilst ¹³⁷Cs, ¹³⁴Cs and ¹³⁶Cs are routinely measurable at high sensitivity by gamma spectrometry, routine detection of long-lived ¹³⁵Cs by radiometric methods is challenging. This measurement is, however, important given its significance in long-term nuclear waste storage and disposal. Furthermore, the ¹³⁵Cs/¹³⁷Cs ratio varies with reactor, weapon and fuel type, and accurate measurement of this ratio can therefore be used as a forensic tool in identifying the source(s) of nuclear contamination. The shorter-lived activation products ¹³⁴Cs and ¹³⁶Cs have a limited application but provide useful early information on fuel irradiation history and have importance in health physics.     

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).     

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.     

Removal of fission products from mixed solvents using zeolites

The solvents tributylphosphate (TBP) and odourless kerosene (OK) are widely used to extract plutonium(IV) and uranium(V) in nuclear processing plants. Although these solvents are chosen because of their low affinity for fission products it is, nevertheless, of value to study ways of removing the small quantities of fission products, such as isotopes of Cs, Sr and Ru, which can be present in the solvents. This paper describes simple distribution coefficients (K _d) measurements made to test the ability of the aluminosilicate zeolites to remove¹⁰⁶Ru from TBP, OK, and their mixtures.     

The study on radioactive contamination in foodstuffs imported from Japan after the Fukushima accident

During the Fukushima nuclear accident, a large number of radionuclides were released, causing direct radioactive contamination to the environment and the ecosystem, and eventually indirect contamination to the foodstuffs. The Radiation Monitoring Group, Office of Atoms for Peace (OAP) in cooperation with the Food and Drug Administration (FDA), has been monitoring the foodstuffs imported from Japan, especially the Fukushima and nearby areas. This monitoring program covered the potentially contaminated foodstuffs. The measurements were mainly conducted using Gamma spectroscopy systems in order to analyze the radioactive concentration of ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs fission products. From 442 samples collected between after the nuclear accident to the end of September 2012, 28 of contaminated food products of fission isotopes have been found with the contamination ranges of 0.63–15.25 Bq/kg fresh, 1.45–44.70 Bq/kg fresh, and 0.45–51.10 Bq/kg fresh) for ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs, respectively. However, the contamination levels are still well under the Thailand standard of food contaminated with radionuclide and the relevant international standards. Finally, the research also investigated the relationship between the times the samples were collected and the contamination levels.     

Adsorption of 137Cs on titanate nanostructures

Various types of sodium and potassium titanate nanostructures (nanotubes, nanofibers, nanoribbons, nanwires) were synthesized and characterized by X-ray diffraction, SEM and TEM, as well BET and BJH methods. Adsorption of radiotracer ¹³⁷Cs⁺ ions from aqueous solutions on synthesized titanate nanostructures was investigated in batch technique as a function of contact time, concentration of sodium ions and pH of the solutions. It was found that among the studied nanostructures nanotubes shows the highest selectivity for ¹³⁷Cs, which is related to a zeolitic character of Cs⁺ adsorption. The efficient adsorption of ¹³⁷Cs was obtained in Na⁺ solutions with concentration below 10^?2 M, at pH 7–9 and in contact time above 2 h. Moreover, nanotubes have the higher specific surface area than other nanostructures, which results in better availability of ion exchange groups and high ion exchange capacity. These properties of nanotubes indicate that they may be used for adsorption of ¹³⁷Cs from various types of nuclear wastes.     

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).     

Isolation of radiocesium from a fission product mixture

A method of¹³⁷Cs isolation from strongly, acidic solutions of fission products is described, in which vanadyl ferrocyanide is used as a selective ion exchanger for cesium. The effects of the acidity of medium and the carrier concentration on the quantitative yield of separation have been studied and convenient conditions have been found for¹³⁷Cs isolation from the solution of fission products formed after irradiating uranium with neutrons.     

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.     

Uncertainty of decay heat calculations originating from errors in the nuclear data and the yields of individual fission products

The calculation of the abundance pattern of the fission products with due account taken of feeding from the fission of²³⁵U,²³⁸U, and²³⁹Pu, from the decay of parent nuclei, from neutron capture, and from delayed-neutron emission is described. By means of the abundances and the average beta and gamma energies the decay heat in nuclear fuel is evaluated along with its error derived from the uncertainties of fission yields and nuclear properties of the individual fission products.     

Actinides and Long-Lived Radionuclides in Tissues of the Japanese Population: Summary of the Past 20-Year Studies

Radioactive fallout constitutes the major source of contamination of the environment with fission products. Our primary interest was in selected fission products, such as ¹³¹I, ⁸⁹Sr, ⁹⁰Sr, and ¹³⁷Cs, and neutron activation products, such as ³H and ¹⁴C. Plutonium-239,240, ²⁴¹Am and ⁹⁰Tc are generated from nuclear tests, and they are important by-products of nuclear industries. Polonium-210, ²¹⁰Pb and ²³²Th, ²³⁰Th and ²²⁸Th occur widely in nature. These radionuclides enter the human body through inhalation and the ingestion through food and water. These nuclides may cause radiation doses to certain organs of the body. Assessment of the resulting health hazards is an essential public health activity, which demands reliable techniques for the assay of the various radionuclides in man and his environment. In this paper, we present the accumulation of radionuclides from man-made sources and primordial radionuclides in various tissues of the Japanese population. The studies were performed at the Department of Public Health, Akita University School of Medicine, during the periods from 1973 to 1995.     

Nuclear decay studies of fission-product nuclides using an on-line mass separation technique

An isotope-separator-on-line (ISOL) system has been developed at the Idaho National Engineering Laboratory to enable a wide variety of nuclear decay studies to be made for fission-product radionuclides. The system is unique in that it utilizes the spontaneous fission source,²⁵²Cf, as the source of fission-product radioactivity. Fission products are transported to the ion source of the mass separator by the helium gas-jet technique. Mass-separated beams of previously unattainable rare-earth nuclides are produced with this system because of the higher yield of fission products with A>150, relative to that for thermal-neutron fission of²³⁵U, and the use of a relatively efficient ion source. Recent decay studies reported here include systematic measurements of rare-earth nuclide half-lives and comparison of them to theoretical prediction, a decay scheme investigation for¹⁵⁴Nd, and -strength function measurements for¹⁴⁰Cs.     

Distribution and behavior of some radionuclides associated with the Trinity nuclear test

The activities of ¹³³Ba, ¹³⁷Cs, ¹⁵²Eu, ¹⁵⁴Eu, ¹⁵⁵Eu, ²³⁹Pu, and ²⁴¹Am were determined by gamma spectroscopy on the largest sample set (n = 49) of bulk trinitite to date. The range in activity for all isotopes is large. For example, the activity of ²⁴¹Am (normalized to the time of detonation) ranges between 1 and 42 Bq/g. Comparison of activities for isotopes derived from the device, ²⁴¹Am versus ¹³⁷Cs, ¹⁵⁵Eu, and ²³⁹Pu, indicate positive trends. Correlations were not observed between the activities of the soil-derived activation products ¹⁵²Eu and ¹⁵⁴Eu and the radioisotopes from the device. The calculated ratio of fission products (¹⁵⁵Eu/¹³⁷Cs) is 0.012 ± .006 (1σ, n = 3), which is lower than predicted for the thermal neutron-induced fission of ²³⁹Pu (~0.03). This discrepancy may be attributed to the spontaneous fission of the natural U tamper resulting in mixing between fission products from ²³⁹Pu and ²³⁵U. The spatial distribution of the trinitite samples relative to ground zero has been modeled based on the activity of ¹⁵²Eu. The calculated distances do not correlate with any of the activities for the radioisotopes investigated here, and suggest a relatively homogeneous distribution. However, trinitite samples with the highest activities for ¹³⁷Cs, ²³⁹Pu, and ²⁴¹Am yield the shortest calculated distances of 50–60 m away from ground zero.     

The development of radioactive glass surrogates for fallout debris

The production of glass that emulates fallout is desired by the nuclear forensics community for training and measurement exercises. The composition of nuclear fallout is complex, with widely varying isotopic compositions (Fahey et al., Proc Natl Acad Sci USA 107(47):20207–20212, 2010; Bellucci et al., Anal Chem 85:7588–7593, 2013; Wallace et al., J Radioanal Nucl Chem, 2013; Belloni et al., J Environ Radioact 102:852–862, 2011; Freiling, Science 139:1058–1059, 1963; Science 133:1991–1999, 1961; Bunney and Sam Government Report: Naval Ordinance Laboratory, White Oak, 1971). As the gaseous cloud traverses from hotter to cooler regions of the atmosphere, the processes of condensation and nucleation entrain environmental materials, vaporized nuclear materials and fission products. The elemental and isotopic composition of the fission products is altered due to chemical fractionation (i.e. the fission product composition that would be expected from fission of the original nuclear material is altered by differences in condensation rates of the elements); the fallout may be enriched or depleted in volatile or refractory fission products. This paper describes preliminary work to synthesize, irradiate and fractionate the fission product content of irradiated particulate glass using a thermal distillation 2 h after irradiation. The glass was synthesized using a solution-based polymerization of tetraethyl orthosilicate. (Izrael, Radioactive fallout after nuclear explosions and accidents, 2002) Uranium was incorporated into the glass particulate at trace concentrations during polymerization. The particulate was subjected to a short thermal neutron irradiation then heated to 1,273 K approximately 2 h after the end of irradiation. Fission products of ^{133, 134, 135}I, ^{132, 134}Te, ¹³⁵Xe, ¹³⁸Cs and ^{91, 92}Sr were observed to be distilled from the particulate. The results of these preliminary studies are discussed.     

A complementary cesium coolant concentration ratio for localizing defective PWR fuel rods during reactor operation

Rough techniques for pinpointing defective fuel pins during actual reactor operation were developed for nuclear power plants. These techniques are based on various fission product concentration ratios. Here, a new cesium concentration ratio,¹³⁴Cs/¹³⁶Cs, was tested in combination with the more usual cesium ratio¹³⁴Cs/¹³⁷Cs. This new cesium ratio confirmed the conclusions drawn from the ratio¹³⁴Cs/¹³⁷Cs and provided some additional information on the location of the defective fuel rods. Application of this second cesium ratio improves the reliability of the rough localization method.