Study on a separation method of radionuclides (Ba, Sr) from LiCl salt wastes generated from the electroreduction process of spent nuclear fuel

In this paper, a separation method of radionuclides (Ba, Sr) from LiCl salt wastes generated from the electroreduction process of spent nuclear fuel was studied to recover pure LiCl salts and reduce radioactive wastes. The method consisted of chemical conversion process of BaCl₂ and SrCl₂ in LiCl molten salts by using lithium compounds and vacuum distillation process of LiCl salts. In the chemical conversion, BaCl₂ and SrCl₂ in LiCl molten salts were mainly converted into (Ba,Sr)CO₃ or (Ba,Sr)SO₄. Contents of Ba and Sr in LiCl salts recovered from the vacuum distillation process were equal to about 0.01 of initial concentrations of Ba and Sr in LiCl molten salts. These results will be utilized to recycle the LiCl salt wastes.     

Removal behavior of Cs from molten salt by using zeolitic materials

Radioactive molten salt generated from a pyrochemical process to separate reusable U and TRU elements is one of problematic wastes to manage for a final disposal. For the minimization of final waste, it is desirable to selectively remove radionuclides from the waste salts. In this paper, structural change of some zeolites in a series of molten salt systems and its removal behavior of CsCl was investigated. Zeolite-4A(LTA) was transformed into LiAlSiO₄ and Li-sodalite with the mol-fraction of LiCl in LiCl–KCl system at 650 °C while it was not changed in NaCl–KCl at 750 °C, regardless of mol-fraction of metal chloride. Other commercial zeolite with specific structure (FAU) had the same trends on the structural stability in molten salt system. From the Cs removal experiments, the decomposed zeolitic materials in molten salt lost their removal ability of Cs. In conclusion, a new selective material or method should be investigated or developed for obtaining the validity on the separation of group I and II radionuclides from a molten waste salt because the zeolite 4A is unstable in the LiCl system and it also showed a low capacity in the LiCl–KCl phase. This paper gives basic information on the removal of radionuclides from molten systems by using zeolitic materials.     

Development of the Self-doping Porous Carbon and Its Application in Supercapacitor Electrode

The massive discharge of biomass wastes not only causes waste of resources, but also pollutes the environment. Therefore, converting biomass wastes into carbon materials is an effective way to solve the above problems. Here, using biomass waste pig nails as raw materials and K₂CO₃ as chemical activators, the N-doped porous carbon(KPNC) is prepared by direct pyrolysis. As an electrode for supercapacitors, the electrochemical tests of KPNCs showed that they exhibited good electrochemical performance and excellent cycling stability. When the current density is 0.2 A/g, the specific capacitance is up to 344.6 F/g. Moreover, it still maintains 97.6% initial capacitance retention after 2000 cycles at a high current density of 5 A/g. Above exceptional electrochemical performances may be ascribed to an appropriate porous structure(S_micro/S_total=80.31%, V_micro/V_total=76.19%), high nitrogen contents(4.44%, atomic fraction), oxygen contents(9.13%, atomic fraction) as well as small internal resistance. The above experimental results show that the conversion of pig nails to porous carbon can reduce the waste of resources and alleviate environmental pollution.     

Lithium Recovery from Radioactive Molten Salt Wastes by Electrolysis

In order to determine the operating conditions of an electrolyzer to recover lithium metal from molten salt wastes composed of LiCl, Li₂O, Cs₂O, and SrO, electrolytic reduction experiments have been carried out in a single compartment electrochemical reactor with a mono-polar connection. All the combinative experiments were conducted in an argon atmospheric glove box, and each applied potential-current value was synchronously measured and analyzed in aspects of the preferentially recovering probability of lithium in mixed phases. The effect of the electrode surface area on the current was also observed. Based on our experimental results compared with electrochemical thermodynamic evaluation, it is revealed that Li₂O can be preferentially reduced to lithium by controlled LiCl concentration and applied potential.     

Melting and vaporization of salts in a U-LiCl-Li2O system

Summary The electrochemical reduction of uranium oxide in the treatment of spent nuclear fuel requires a characterization of the LiCl-Li₂O salt used as a reaction medium. Physical properties, melting and vaporization are important for the application of the salt and thus they have been investigated by differential scanning calorimetry (DSC) and thermogravimetry (TG), respectively. Experimental data suggest LiCl and Li₂O compound formations, leading to a melting point depression of the LiCl and a co-vaporization of the LiCl-Li₂O salt.     

A comparative study using liquid scintillation counting and X-ray spectrometry to determine 55Fe in radioactive wastes

The radionuclide ⁵⁵Fe was determined in samples of radioactive wastes from the water cleanup system of the IEA-R1 nuclear research reactor. In order to validate the results, the ⁵⁵Fe activity concentration was measured in eight waste samples and in six simulated samples containing the most important interfering radionuclides. A simple method was employed to separate and purify ⁵⁵Fe from other radionuclides present in these samples, combining co-precipitation with ammonium hydroxide and purification with anionic ion-exchange resin, which enables ⁵⁵Fe to be quantified either by liquid scintillation counting (LSC) or by X-ray spectrometry using a low-energy germanium spectrometer (LEGe). Both measurement methods were used so that the separation and purification process could be confirmed by comparison of spectra with and without the utilization of anionic ion-exchange resin. Activity and interferences were compared in the results obtained from LSC and LEGe measurement methods.     

Simultaneous solidification of two catalyst wastes and their effect on the early stages of cement hydration

Two catalyst wastes (RNi and RAI) from polyol production were considered as hazardous, due to their respective high concentration of nickel and aluminum contents. This article presents the study, done to avoid environmental impacts, of the simultaneous solidification/stabilization of both catalyst wastes with type II Portland cement (CP) by non-conventional differential thermal analysis (NCDTA). This technique allows one to monitor the initial stages of cement hydration to evaluate the accelerating and/or retarding effects on the process due to the presence of the wastes and to identify the steps where the changes occur. Pastes with water/cement ratio equal to 0.5 were prepared, into which different amounts of each waste were added. NCDTA has the same basic principle of Differential Thermal Analysis (DTA), but differs in the fact that there is no external heating or cooling system as in the case of DTA. The thermal effects of the cement paste hydration with and without waste presence were evaluated from the energy released during the process in real time by acquiring the temperature data of the sample and reference using thermistors with 0.03 °C resolution, coupled to an analog–digital interface. In the early stages of cement hydration retarding and accelerating effects occur, respectively due to RNi and RAl presence, with significant thermal effects. During the simultaneous use of the two waste catalysts for their stabilization process by solidification in cement, there is a synergic resulting effect, which allows better hydration operating conditions than when each waste is solidified separately. Thermogravimetric (TG) and derivative thermogravimetric analysis (DTG) of 4 and 24 h pastes allow a quantitative information about the main cement hydrated phases and confirm the same accelerating or retarding effects due to the presence of wastes indicated from respective NCDTA curves.     

Nuclear Waste Storage in Gel-Derived Materials

For long life nuclear wastes (essentially actinides) research is in progress to propose new matrices with an improvement of the chemical durability. High silica content glasses present high chemical durability, and we describe a process to prepare silica glass embedding the nuclear waste. Porous silica (gel) is used as a host matrix for nuclear waste. Neodymium oxide and cerium oxide are used to simulate the actinide oxides. The gel is soaked in a solution containing the simulant in nitrate salt form.We investigate the effect of the porous network features (pore size distribution) on the ability of the material to soak up the simulating salts. A new kind of gel (composite aerogel) is proposed owing to its large permeability. After drying and nitrate decomposition, the composite material is fully sintered trapping the nuclear waste.The glass-ceramic materials (silica glass + simulating oxide) have been synthesized with simulant content close to 20 weight percent. We show that the 2 simulant oxides behave differently because of their ability to form silicate phases. The chemical durability of the glass-ceramics is investigated. The large improvement of the chemical durability (60 times) compared to the classical borosilicate glasses shows that this containment process can be a suitable way to confine actinides and long life nuclear wastes.     

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.     

Reducing waste contamination from animal-processing plants by anaerobic thermophilic fermentation and by flesh fly digestion

There is currently no market in Israel for the large amounts of waste from fish- and poultry-processing plants. Therefore, this waste is incinerated, as part of the measures to prevent the spread of pathogens. Anaerobic methanogenic thermophilic fermentation (AMTF) of wastes from the cattle-slaughtering industry was examined previously, as an effective system to treat pathogenic bacteria, and in this article, we discuss a combined method of digestion by thermophilic anaerobic bacteria and by flesh flies, as a means of waste treatment. The AMTF process was applied to the wastes on a laboratory scale, and digestion by rearing of flesh fly (Phaenicia sericata) and housefly (Musca domestica) larvae on the untreated raw material was done on a small scale and showed remarkable weight conversion to larvae. The yield from degradation of poultry waste by flesh fly was 22.47% (SD = 3.89) and that from fish waste degradation was 35.34% (SD = 12.42), which is significantly higher than that from rearing houseflies on a regular rearing medium. Bacterial contents before and after thermophilic anaerobic digestion, as well as the changes in the chemical composition of the components during the rearing of larvae, were also examined.     

The Intermediate Level Liquid Molybdenum-99 Waste Treatment Process at the Australian Nuclear Science and Technology Organisation

ANSTO manufactures ⁹⁹Mo for radiopharmaceutical use. Alkaline Intermediate Level Liquid Waste (ILLW) from this process, plus legacy acidic waste, are planned to be treated by converting both wastes into stable, solid, waste forms with oxide-basis loadings of 25-35 wt% and 30-50 wt%, respectively. The hot-cell plant design utilises the same unit process steps to treat both wastes. Hot-Isostatic Pressing (HIP) is employed to consolidate the processed waste and achieve substantial waste volume reductions compared to a cementation option. In this paper an overview of the treatment process and selected waste forms for ANSTO's ⁹⁹Mo production ILLW is given.     

Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO₃) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.     

Effective Removal of Cesium and Strontium from Radioactive Wastes Using Chemical Treatment Followed by Ultra Filtration

Chemical treatment assumes an important role in the management of radioactive wastes as it is a simple technique and offers advantage in terms of handling of wastes thereby reducing the risk of mansievert exposure. Low level wastes (LLW) and intermediate level wastes (ILW) are generated in various facets of nuclear fuel cycle and have various chemical composition. A systematic study was carried out by using copper ferrocyanide and calcium phosphate precipitation methods for the removal of cesium and strontium, respectively. The supernatants were subjected to ultra filtration (UF) using a membrane having a pore size of 0.2 m. The decontamination factors (DF) at 2 and 24-hour intervals with and without UF were estimated. The DF obtained was in the range of 200–300 for cesium and 200 for strontium with LLW solution which has chemical characteristics similar to ground water. Two hours of settling is adequate for strontium before UF. In case of cesium there is no much change in the DF values by UF. However, the UF has helped in the solid — liquid separation as the flocks of copper ferrocyanide precipitate are feathery in nature. The effect of ionic strength and the presence of TBP on the removal efficiency of cesium and strontium have also been studied. DF are observed to be a function of ionic strength and are low in deionized water, in salt solutions containing 1 to 4M sodium nitrate and also in solutions of ILW. However, increasing the chemical dosing to two times of normal plant dosing has yielded a DF of about 200 for sodium nitrate solutions with respect to cesium removal. When the concentration of ammonium nitrate in the waste exceeds 0.1M, the DF reduces. Entrained TBP as well as soluble TBP reduces the removal efficiency of cesium. This paper deals with the experimental data and mechanism of the processes involved in the removal of cesium and strontium.     

Electroreduction of Dy(III) assisted by Zn and its co-deposition with Zn(II) in LiCl–KCl molten salt

To recover dysprosium (Dy) from LiCl–KCl molten salt, the electrochemical mechanism of Dy(III) on liquid Zn electrode and co-deposition of Dy(III) and Zn(II) on W electrode were studied using electrochemical methods. Cyclic voltammetry results demonstrated that the redox process of Dy on liquid Zn electrode is reversible and controlled by diffusion. Reverse chronopotentiograms showed that the transition time ratio of reduction and oxidation is ~3:1, revealing the redox of Dy on liquid Zn electrode is a kind of soluble–soluble system: Dy(III) + 3e⁻ = (Dy–Zn)_solution. The half-wave potential of Dy(III) was almost constant with the increase in scanning rate. The electrochemical separation of metallic Dy from the molten salt was performed using constant potential electrolysis, and the product characterized using X-ray diffraction and scanning electron microscopy–energy-dispersive X-ray spectroscopy was the thermodynamic unstable compound DyZn₅. Also, the co-deposition mechanism of Dy(III) and Zn(II) was explored, indicating that Dy(III) could deposit on pre-deposited Zn and form Dy–Zn compounds: Zn(II) + 2e⁻ = Zn and xDy(III) + yZn + 3xe⁻ = Dy_xZn_y. Moreover, the effect of Dy(III) concentration on the formation of Dy–Zn compounds was investigated. The redox peak currents corresponding to different Dy–Zn compounds changed with the increase in Dy(III) concentration. The co-deposition of Dy(III) and Zn(II) was performed using constant current electrolysis at diverse Dy(III) concentrations. The different Dy–Zn compounds were produced by controlling Dy(III) concentration.     

Interaction of uranium(VI) with phthalic acid

Phthalic acid, a ubiquitous organic compound found in soil, water, and in domestic and nuclear wastes can affect the mobility and bioavailability of metals and radionuclides. We examined the complexation of uranium with phthalic acid by potentiometric titration, electrospray ionization-mass spectroscopy (ESI-MS), and extended X-ray absorption fine structure (EXAFS) analysis. Potentiometric titration of a 1:1 U/phthalic acid indicated uranyl ion bonding with both carboxylate groups of phthalic acid; above pH 5 the uranyl ion underwent hydrolysis with one hydroxyl group coordinated to the inner-sphere of uranium. In the presence of excess phthalic acid, ESI-MS analysis revealed the formation of both 1:1 and 1:2 U/phthalic acid complexes. EXAFS studies confirmed the mononuclear biligand 1:2 U/phthalic acid complex as the predominant form. These results show that phthalates can form soluble stable complexes with uranium and may affect its mobility.     

Origin of Surface Reconstruction in Lattice Oxygen Oxidation Mechanism Based-Transition Metal Oxides: A Spontaneous Chemical Process

A fundamental understanding of surface reconstruction process is pivotal to developing highly efficient lattice oxygen oxidation mechanism (LOM) based electrocatalysts. Traditionally, the surface reconstruction in LOM based metal oxides is believed as an irreversible oxygen redox behavior, due to the much slower rate of OH⁻ refilling than that of oxygen vacancy formation. Here, we found that the surface reconstruction in LOM based metal oxides is a spontaneous chemical reaction process, instead of an electrochemical reaction process. During the chemical process, the lattice oxygen atoms were attacked by adsorbed water molecules, leading to the formation of hydroxide ions (OH⁻). Subsequently, the metal-site soluble atoms leached from the oxygen-deficient surface. This work also suggests that the enhancement of surface hydrophilicity could accelerate the surface reconstruction process. Hence, such a finding could add a new layer for the understanding of surface reconstruction mechanism.     

Electrochemical reduction of UO2 to U in a LiCl–KCl-Li2O molten salt

An electrochemical reduction of UO₂ to U in a LiCl–KCl-Li₂O molten salt has been investigated in this study. A diagram showing equilibrium potentials (relative to Cl₂/Cl^?) plotted versus the negative logarithms of oxide-ion activity (pO^2?) was constructed. The crushed UO₂ pellets in the cathode basket of an electrolytic reducer were successfully reduced to U. The reduction of UO₂ is proved to proceed mainly through chemical reaction with in situ generated Li and K at the cathode. The control of cathode potential is essential to prevent the deposition and subsequent vaporization of K metal at the cathode for the applications of a LiCl–KCl-Li₂O molten salt as an electrolyte for the metal production from its oxide sources.     

Fabrication and physical properties of lanthanide oxide glass wasteform for the immobilization of lanthanide oxide wastes generated from pyrochemical process

The pyrochemical process, which uses a dry method to recycle used nuclear fuel generates waste LiCl–KCl salt containing radioactive lanthanide elements. To reuse LiCl–KCl salt, the lanthanide elements are separated through a precipitation method promoted by oxygen sparging and the separated fission product of lanthanide oxide should be fabricated into durable wasteforms sustainable for several 1,000 years to store in a final geological repository. Herein, we report the fabrication of a borosilicate glass based wasteform with a glass matrix of SiO₂–Al₂O₃–B₂O₃ having a high waste loading of 50 wt% lanthanide oxide. Th physical properties of four kinds of wasteforms having a different lanthanide oxide waste composition were evaluated. To investigate the long-term physical stability of each sample having 50 wt% lanthanide oxide waste loading, time–temperature–transformation (TTT) test was conducted at 500 and 700 °C for 60 and 180 h, and the physical properties were evaluated after each TTT test.     

Synthesis and Characterization of Poly(ethylene terephthalate)/Modified Lithium‐Aluminum‐Layered Double Hydroxide Nanocomposites Prepared Using In‐situ Polymerization

Layered double hydroxides are a type of layered stacked compound, which can be intercalated with organic‐molecule modifiers. An ion‐exchange process for layered double hydroxide (LDH) was used to intercalate water‐soluble sulfanilic acid salt (SAS) and dimethyl 5‐sulfoisopthalate (DMSI) into lithium aluminum layered double hydroxides (LiAl LDHs). In this work, a hydrothermal process was used to modify LiAl LDHs, and the modified LiAl LDHs were treated with either SAS or DMSI through an ion‐exchange process and were then intercalated using bis‐hydroxyethylene terephthalate (BHET). The results indicate that the modified LiAl LDHs improved the interlayer compatibility between the PET and LiAl LDH layers; thus, enabling the oligomer molecules to more easily enter the gallery of the LiAl LDH layers so that polymer chains could be included between the LDH layers during polymerization of the matrix. The better barrier, mechanical properties, and thermal stability of these new types of PET nanocomposites are discussed.     

Chemical speciation of radionuclides migrating in groundwaters

In order to more accurately predict the rates and mechanisms of radionuclide migration from lowlevel waste disposal facilities via groundwater transport, ongoing studies are being conducted at field sites at Chalk River Laboratories to identify and characterize the chemical speciation of mobile, long-lived radionuclides migrating in groundwaters. Large-volume water sampling techniques are being utilized to separate and concentrate radionuclides into particulate, cationic, anionic, and nonionic chemical forms. Most radionuclides are migrating as soluble, anionic species which appear to be predominately organoradionuclide complexes. Laboratory studies utilizing anion exchange chromatography have separated several anionically complexed radionuclides, e.g.,⁶⁰Co and¹⁰⁶Ru, into a number of specific compounds or groups of compounds. Large-volume ultra-filtration experiments have shown that significant fractions of the radionuclides are being transported in these groundwaters in the form of macromolecules having molecular weights ranging from less than 3,000 to 100,000.