Determination of alpha-emitting isotopes in radioactive wastes

An analytical procedure was developed for the alpha-spectrometric determination of uranium (²³⁸U, ²³⁵U, ²³⁴U) and transuranium isotopes (^239+240Pu, ²⁴⁴Cm) in liquid radioactive wastes (sludges, evaporation residues) of low and medium activity from the VVER-440 reactor of the nuclear power plant in Paks, and in waste waters to be released into the environment. Radioactive sludge samples were separated to a liquid phase and a wet suspension (solid) phase by centrifuging, and the two phases were treated and analyzed separately, in different ways. A sample preparation procedure based on chemical separation was worked out for the isolation of the alpha-emitting components of radioactive evaporation residues which were saturated with boric acid. To the separate determination of the low activity alpha-emitting isotopes appearing in waste waters to be released into the environment, a microvawe concentration technique was applied. The accuracy of the results obtained using the above chemical separation and alpha-spectrometry was tested in the framework of the international U. S. DOE EML Quality Assessment Program.     

Endlagerung radioaktiver Abfälle

Shales, granites and rock salt are currently under investigation as host rocks for radioactive waste. With respect to heat‐producing waste (spent fuel, high‐active waste) these rock types comprise contrasting mechanical and chemical behavior. The differences are due to the respective geological formations: Shales form by slow accumulation of fine‐grained minerals from seawater with subsequent compaction and diagenesis; crystallization of deep‐seated magmas at 700 to 850°C is the process that generates granitic rocks in the upper 20 km of the earth's continental crust; rock salt is a chemical sediment which forms by precipitation of chloride and sulfate minerals from seawater evaporation in shallow marine basins under arid conditions. The extent of chemical reactions between granitic rocks and migrating saline fluids upon canister‐induced heating is quite small. However, thermally induced reactions between sheet silicate minerals in shales may result in a gradual loss of adsorption capacities for released radionuclides. Canister‐induced temperature gain in rock salt results in increasing creep rates which lead to an enhanced enclosure process. Great care has to be taken in the selection of salt formations as host rocks with respect to brines; depending on their composition and temperature brines might react with e.g. potash‐seams.     

Direct burn-up determination of fast reactor mixed oxide (MOX) fuel by preferential evaporation of interfering elements

Spent fuel of uranium–plutonium mixed oxide (MOX) from sodium cooled Fast Breeder Test Reactor (FBTR) was analyzed for at.% burn-up by preferential evaporation method. A sequential pattern of analysis of fission monitor Nd and heavy elements, U and Pu provided an un-interfered isotopic composition. Concentrations of individual elements were determined by isotopic dilution mass spectrometry. The proposed method provides at.% burn-up with an uncertainty of about 4% (compared to ASTM method), is less time consuming, does not involve any chemical separation, reduction in radioactive waste and substantial reduction in the radiation exposure to analyst.     

Radioactive waste treatment products studied by Mössbauer spectroscopy

Mössbauer effect was used for the characterization of the radioactive waste treatment products, e.g. precipitate formed during the treatment of LLAW (Low Level Active Waste) using iron compounds and their conditioned matrix obtained by cementation.     

Adsorption of hazardous ions from radioactive waste on chelating cloth filter

A cloth filter was synthesized by grafting of acrylonitrile/methacylic acid (AN/MAA ≈80%/20% molar ratio) onto cotton cloth using a radiation-induced technique followed by amidoximation reaction. The fate of adsorption of radionuclide (e.g. U(VI)) on chelating cloth filter (CCF) from radioactive waste was investigated. The adsorption ability of the CCF increases as pH increases from 6 to 10. The predominant composition of the resulting complex was determined. A chemical adsorption mechanism was confirmed by examining the relationships between the adsorbed amount of radionuclide and the contact time.     

Concentration of radioactive components in liquid low-level radioactive waste by membrane distillation

The paper addresses some aspects of liquid low-level radioactive waste (LLLW) purification. Since the volume of liquid low-level wastes is usually large and the activity is rather low, the radioactive substances separated from the non-active portion have to be concentrated into the small volume for subsequent conditioning and disposal. The need for the improvement of decontamination and minimisation of the costs have led to new specific methods being under examination and development. The method proposed in the paper is membrane distillation. The experimental work described below supports the statement that membrane distillation can be an attractive alternative for liquid radioactive waste treatment. The advantages of membrane distillation over the other processes commonly used for the processing of LLLW are discussed in the paper.     

Treatment of radioactive liquid organic waste using bacteria community

Waste management plays an important role in radioactive waste volume reduction as well as lowering disposal costs and minimizing the environment-detrimental impact. The employment of biomass in the removal of heavy metals and radioisotopes has a significant potential in liquid waste treatment. The aim of this study is to evaluate the radioactive waste treatment by using three different bacterial communities (BL, BS, and SS) isolated from impacted areas, removing radioisotopes and organic compounds. The best results were obtained in the BS and BL community, isolated from the soil and a lake of a uranium mine, respectively. BS community was able to remove 92% of the uranium and degraded 80% of tributyl phosphate and 70% of the ethyl acetate in 20 days of experiments. BL community removed 81% of the uranium and degraded nearly 60% of the TBP and 70% of the ethyl acetate. SS community collected from the sediment of São Sebastião channel removed 76% of the uranium and 80% of the TBP and 70% of the ethyl acetate. Both americium and cesium were removed by all communities. In addition, the BS community showed to be more resistant to radioactive liquid waste than the other communities. These results indicated that the BS community is the most viable for the treatment of large volumes of radioactive liquid organic waste.     

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.     

Europium removal with a mineral mixtre

Natural mixture containing mostly minerals of iron, sillicon, magnesium, aluminium and calcium was exploited for the decontamination study of europium radionuclides from aqueous radioactive waste solutions. The physicochemical conditions, such as shaking and equilibration time, nature of hydrogen ions, pH, temperature, concentrations of adsorbate and adsorbent were experimentally determined. This study showed quantitative adsorption beyond pH 7 and under optimized conditions, up to 33 g of the adsorbate can be rapidly removed from radioactive effluents using only 1 kg of the mineral mixture (MM). Desorption study of the solidified radioactive waste product reveals no significant loss (< 0.01% month), indicating MM as an effective material for removal of radioactive europium and storing it in solid form over a long period of time.     

Sorption of europium on TiO2 and cement at high Ph in the presence of organic ligands

Summary Organic substances present in radioactive waste lower the sorption of metal ions at the high pH in cement matrices and, hence, enhance their possible migration. The aim of this study was to develop a method to compare organic substances or their degradation products with respect to what extent they affect metal sorption. Batch sorption studies were performed with cement or TiO₂ as solid phase and Eu(III) as a model element for trivalent lanthanides and actinides at pH 12.5 (representative of a cement waste matrix during the first approximately 100,000 years). Different kinds of ligands were studied in a broad concentration range, e.g., organic acids, cement additives, cleaning agents and degradation products from ion-exchange resin.     

The effects of hazardous ions adsorption on the morphological and chemical properties of reactive cloth filter

Reactive cloth filter is fabricated by grafting of acrylonitrile/methacrylic acid onto cotton cloth, using mutual irradiation technique and the subsequent amidoximation of the reactive intermediate nitrile groups. The incorporation of the amidoxime/carboxyl groups was confirmed by different techniques. The effect of the hazardous ions chelation from radioactive waste on the morphological and chemical structure was studied. The cloth filter possessed good morphological and chemical stability suitable for practical use. The fabricated cloth filter can be used for low-level radioactive waste treatments.     

Separation of europium, cobalt and zinc on zirconium tungstate ion exchanger

Summary Amorphous zirconium tungstate inorganic ion-exchanger has been prepared under optimum conditions and characterized by IR, X-ray and thermal analysis. Surface area and capacity are determined. It has characteristic IR absorption peaks at 3242, 1628, 955, 868 and 432 cm^-1 and is thermally stable up to 450 °C. Its surface area was 16 m²/g with an exchange capacity of 0.541 meq/g. The sorption of radioactive europium from different media at ambient temperature by the zirconium tungstate (ZW) exchanger has been studied. The aim was to optimize the conditions for sorbing Eu from radioactive waste and cleaned the ZW from for recycling. The effect of contact time, metal concentration, pH and temperature has been measured. Percentual uptake of Eu(III) reaches 95% for HCl at pH 4 and increases with temperature indicating an endothermic sorption process. Uptake of Zn(II) and Co(II) on ZW from acetic acid was found to be 42% and 24% for cobalt and zinc, respectively. Desorption after saturation and the effect of other radioactive ions on the percentual uptake of Eu on ZW were investigated. A solution of 3M HCl releases 90%, 25% and 13% of the loaded Eu(III), Co(II) and Zn(II), respectively.     

Chemiluminescent tags in immunoassays

Although radioimmunoassay procedures have a number of advantages they do pose problems, especially with regard to the disposal of radioactive waste. Alternative methods, such as chemiluminescence, enzyme, fluorescence, or spin immunoassays, have been tested with the aim of replacing radioactive labels without loss of sensitivity, precision and accuracy.     

Opportunities and challenges of high-pressure ion exchange chromatography for nuclide separation and enrichment

With the rapid development of the nuclear industry, more-stringent requirements are proposed for high-level radioactive waste liquid treatment and the enrichment of isotope products. High-pressure ion exchange chromatography has been widely accepted for the fine separation of elements and nuclides due to its advantages, such as high efficiency, environmental friendliness, ease of operation, and feasibility for large-scale industrial applications. Here, we summarized the evolution of high-pressure ion exchange chromatography and the relevant research progress in ion exchange equilibrium and related separation technology. The prospects for application of high-pressure ion exchange chromatography to rare earth elements, actinide elements and isotope separation were discussed. High-pressure ion exchange chromatography represents a promising strategy for the extraction of rare earth elements and actinide elements from high-level radioactive waste liquid, as well as being an effective method for the automated production of high purity isotope products with great environmental benefits.     

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.     

Production of low and high specific activity <Superscript>64</Superscript>Cu in a reactor

Production of radioactive scandium by irradiating natural titanium metal in Pakistan Research Reactor-1 was evaluated. The production rate of ⁴⁷Sc and other radioactive scandium was estimated. High specific activity ⁴⁷Sc can be produced by irradiating enriched ⁴⁷Ti in sufficient quantities needed for therapeutic applications. A new separation technique based on column chromatography was developed. Neutron irradiated titanium was dissolved in hydrofluoric acid, which was evaporated and taken in distilled water. The resulting solution was loaded on silica gel column. The radioactive scandium comes out first and the inactive titanium is removed with 2 M HCl. More than 95% radioactive scandium was recovered, while chemical impurity of titanium determined by optical emission spectroscopy was less than 0.01 μg/mL in final product.     

Separation of 99Tc from low level radioactive liquid waste using hollow fiber supported liquid membrane: optimisation and performance evaluation

Journal of Radioanalytical and Nuclear Chemistry - Low level radioactive liquid waste (LLW) obtained after treatment of Intermediate level radioactive waste (ILW) is highly alkaline and rich in...     

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.     

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.     

Waste minimization in the Los Alamos Medical Radioisotope Program

Since the mid-1970s the Los Alamos Medical Radioisotope Program has been irradiating target materials to produce and recover radioisotopes for applications in medicine, environmental science, biology, physics, materials research, and other disciplines where radiotracers find utility. By necessity, the chemical processing of targets and the isolation of radioisotopes generates radioactive waste materials. In recent years there have been federal mandates requiring us to discontinue the use of hazardous materials and to minimize radioactive waste volumes. As a result, substantial waste reduction measures have been introduced at the irradiation facility, in processing approaches, and even in the ways the product isotopes are supplied to users.