Glassy Waste Forms Research Articles

Research on calcination thermal decomposition process of simulated high-level liquid waste based on two-step vitrification

The two-step vitrification process involves converting high-level waste from a liquid to a solid state through calcination, followed by melting the calcines with glass matrix material to form glassy wasteforms. The calcination process is a critical stage in the vitrification process. This study investigated the thermal decomposition performance and phase evolution of power reactor simulated high-level waste during the calcination process. As the calcination temperature increased from 100 to 900 °C, the resultant waste calcines showed increasingly darker color and smaller general particle sizes. Below 600 °C, the waste experienced thermal decomposition of nitrate and formate compounds, leading to an approximate 40 wt% weight reduction. At 600 °C, the waste's weight remained constant, and crystalline phases such as ZrxM1-xO2 (M = Ce, Pr, Nd), BaMoO4, CeO2, and GdxCe1-xO2 were observed.

Multiphase synergistic immobilization of complex trialkyl phosphine oxide end-waste into an iron-containing aluminosilicate glass-ceramic

The objective of this study was to address the challenges associated with complex TRPO waste, by utilizing a natural aluminosilicate material to produce glass-ceramic waste forms. When the simulated waste content was below 30 wt.%, glassy waste forms were successfully obtained. Ce and Fe played crucial roles in the formation of Si–O–Ce bonds and [FeO4]-tetrahedra in the glass network, which effectively immobilized other waste elements. However, when the waste content exceeded 30 wt.%, the waste was incorporated into feldspar, iron-manganese crystals, fluorite ceramic, and glass. This combination of ceramic and glass matrices synergistically immobilized the waste, resulting in excellent mechanical performance and chemical durability. The leaching rates of LRCe and LRNd were remarkably low around ∼10−6 to 10−7 g m−2 d−1, after 42 d. Furthermore, the study also investigated the role of multi-valence elements, such as Ce, Fe, and Mn, in the formation of iron-containing aluminosilicate glass-ceramics. The findings offer a novel approach to effectively immobilize complex nuclear waste.

Preparation and characterization of glassy waste forms based on SrF2-Fe2O3-PbO/Bi2O3-P2O5 system

Glassy waste forms based on SrF2-Fe2O3-PbO/Bi2O3-P2O5 system were prepared by using a one-step melting quench method. Glass forming ability, actual composition, structure and thermal stability of the samples were studied by using X-Ray fluorescence (XRF) spectroscopy, X-Ray diffractometry (XRD), infrared spectroscopy (IR), dilatometry and differential thermal analysis (DTA). Water resistance of the samples was evaluated by using dissolution rate and Product Consistency Test (PCT) methods. Melting temperature of the simulated 90SrF2 waste forms was reduced to 1000–1050 °C and their melting time was shortened to 0.5 h. Water resistance of the waste forms met the requirements of national industrial standard EJ1186–2005, while their thermal stability was much higher than those of the phosphate waste forms vitrified with >20 wt.% halide wastes reported in the open literature. It is concluded that waste forms with mixed structure of orthophosphate and pyrophosphate glasses have high water resistance and high thermal stability.

The fate of Si and Fe while nuclear glass alters with steel and clay

The French concept developed to dispose high-level radioactive waste in geological repository relies on glassy waste forms, isolated from the claystone host rock by steel containers. Understanding interactions between glass and surrounding materials is key for assessing the performance of a such system. Here, isotopically tagged SON68 glass, steel and claystone were studied through an integrated mockup conducted at 50 °C for 2.5 years. Post-mortem analyses were performed from nanometric to millimetric scales using TEM, STXM, ToF-SIMS and SEM techniques. The glass alteration layer consisted of a crystallized Fe-rich smectite mineral, close to nontronite, supporting a dissolution/reprecipitation controlling mechanism for glass alteration. The mean glass dissolution rate ranged between 1.6 × 10−2 g m−2 d−1 to 3.0 × 10−2 g m−2 d−1, a value only 3–5 times lower than the initial dissolution rate. Thermodynamic calculations highlighted a competition between nontronite and protective gel, explaining why in the present conditions the formation of a protective layer is prevented.

Influence of long-term aqueous leaching of irradiated graphite on surface properties and behavior of radionuclides

AbstractSamples of real irradiated (i-)graphite from bushings of RBMK reactor after several types of decontamination treatments were put in contact with aqueous solutions modelling underground water of Nizhnekansky massif (Russia) equilibrated with bentonite for periods up to 1.5 years. Leach rates of radionuclides and evolution of graphite surface morphology and oxidation state were monitored using nuclear spectroscopy, SEM and XPS. After the experiment, less than one third of surface carbons remains unoxidised. Extensive precipitation of secondary phases (alumosilicates, carbonates) was observed on some samples and is possibly correlated with surface oxidation extent. The leach rates of dose-forming 14C and 36Cl are comparable with the leach rates of radionuclides in glassy waste forms. According to the current study i-graphite could be regarded as a waste form, which is suitable for near-surface disposal.

Research on chemical durability of iron phosphate glass wasteforms vitrifying SrF2 and CeF3

A part of the radioactive waste produced in the Thorium Molten Salt Reactor (TMSR) will be present as fluorides. Because of some characters of fluoride, it is undesirable for immobilizing the waste as other usual oxide waste in borosilicate glass that often produces immiscibility. However, iron phosphate glasses have been made with up to 30 wt% SrF2 and 20 wt% CeF3, and about 20 wt% SrF2 and CeF3 combined. The iron phosphate glass matrix mixed with fluoride waste could be melted at temperatures between 1050 °C and 1200 °C in only 30–60 min, its melting temperature is lower than that of borosilicate glass. Powder X-ray diffraction (XRD) showed that the glassy wasteforms that contains more than 20% wt% SrF2 and CeF3 mixture, or more than 20 wt% CeF3 would start crystallizing. The product consistency test (PCT) was used to identify the chemical durability of the glass. Glass powder (−100 to +200 mesh) after washing was soaked in 10 ml of deionized water at 90 °C for 7 days. The normalized elemental mass release of elements Sr, Ce and F were calculated at the level of 10−3 g/m2, through the concentration of element measured after PCT, that was equivalent to or better than that of borosilicate glass wasteforms. Iron phosphate glasses are concluded to be a practical alternative for vitrifying the radioactive waste of TMSR.

Compositional Dependence of Solubility/Retention of Molybdenum Oxides in Aluminoborosilicate-Based Model Nuclear Waste Glasses.

Molybdenum oxides are an integral component of the high-level waste streams being generated from the nuclear reactors in several countries. Although borosilicate glass has been chosen as the baseline waste form by most of the countries to immobilize these waste streams, molybdate oxyanions (MoO42-) exhibit very low solubility (∼1 mol %) in these glass matrices. In the past three to four decades, several studies describing the compositional and structural dependence of molybdate anions in borosilicate and aluminoborosilicate glasses have been reported in the literature, providing a basis for our understanding of fundamental science that governs the solubility and retention of these species in the nuclear waste glasses. However, there are still several open questions that need to be answered to gain an in-depth understanding of the mechanisms that control the solubility and retention of these oxyanions in glassy waste forms. This article is focused on finding answers to two such questions: (1) What are the solubility and retention limits of MoO3 in aluminoborosilicate glasses as a function of chemical composition? (2) Why is there a considerable increase in the solubility of MoO3 with incorporation of rare-earth oxides (for example, Nd2O3) in aluminoborosilicate glasses? Accordingly, three different series of aluminoborosilicate glasses (compositional complexity being added in a tiered approach) with varying MoO3 concentrations have been synthesized and characterized for their ability to accommodate molybdate ions in their structure (solubility) and as a glass-ceramic (retention). The contradictory viewpoints (between different research groups) pertaining to the impact of rare-earth cations on the structure of aluminoborosilicate glasses are discussed, and their implications on the solubility of MoO3 in these glasses are evaluated. A novel hypothesis explaining the mechanism governing the solubility of MoO3 in rare-earth containing aluminoborosilicate glasses has been proposed.

Effect of neodymium oxide on the solubility of MoO 3 in an aluminoborosilicate glass

High molybdenum and rare earth concentrations in soda–lime aluminoborosilicate glasses may lead to the crystallization of molybdenum-rich phases such as alkali and alkaline-earth molybdates (Na 2MoO 4, CaMoO 4) and also rare earth (RE)-rich phases such as apatite (Ca 2RE 8(SiO 4) 6O 2) during melt cooling that must be controlled particularly during the preparation of highly radioactive nuclear glassy wasteforms. To understand the effect of neodymium addition (from 0 to 16 wt.% Nd 2O 3) on the phase separation and the crystallization tendency of molydbate phases and on the structure of a Mo-bearing nuclear glass belonging to the SiO 2–B 2O 3–Al 2O 3–Na 2O–CaO–MoO 3 system, crystallization and structural studies have been performed by X-ray diffraction, scanning electron microscopy, electron microprobe analysis, Raman and optical absorption spectroscopies. The results obtained show that the addition of an increasing amount of Nd 2O 3 induces a significant increase of the solubility of molybdenum in the glass, characterized by a decrease of the phase separation and of the crystallization tendency of molybdate phases. The increase of chemical disorder in the structure of Mo-bearing glasses when Nd 2O 3 is added – and more precisely in the depolymerized regions where Nd 3+ cations and [MoO 4] 2− entities are located – could be at the origin of the evolution of the molybdenum solubility in the glass.

Glassy Wasteforms for Nuclear Waste Immobilization

Glassy wasteforms currently being used for high-level radioactive waste (HLW) as well as for low- and intermediate-level radioactive waste (LILW) immobilization are discussed and their most important parameters are examined, along with a brief description of waste vitrification technology currently used worldwide. Recent developments in advanced nuclear wasteforms are described such as polyphase glass composite materials (GCMs) with higher versatility and waste loading. Aqueous performance of glassy materials is analyzed with a detailed analysis of the role of ion exchange and hydrolysis, and performance of irradiated glasses.

Structural investigations of borosilicate glasses containing MoO 3 by MAS NMR and Raman spectroscopies

High molybdenum concentration in glass compositions may lead to alkali and alkaline-earth molybdates crystallization during melt cooling that must be controlled particularly during the preparation of highly radioactive nuclear glassy waste forms. To understand the effect of molybdenum addition on the structure of a simplified nuclear glass and to know how composition changes can affect molybdates crystallization tendency, the structure of two glass series belonging to the SiO 2–B 2O 3–Na 2O–CaO–MoO 3 system was studied by 29Si, 11B, 23Na MAS NMR and Raman spectroscopies by increasing MoO 3 or B 2O 3 concentrations. Increasing MoO 3 amount induced an increase of the silicate network reticulation but no significant effect was observed on the proportion of BO 4 - units and on the distribution of Na + cations in glass structure. By increasing B 2O 3 concentration, a strong evolution of the distribution of Na + cations was observed that could explain the evolution of the nature of molybdate crystals (CaMoO 4 or Na 2MoO 4) formed during melt cooling.

Leaching of glass components and surrogate nuclides from glassy waste forms for radioactive incineration ash

The applicability of vitrification technology to treat radioactive incineration ash was studied, especially in terms of leaching characteristics, by using several glassy waste forms which are fabricated with simulated incineration ash and base-glass at different mixing ratios. The ISO leaching test has been conducted for 820 days. Two semi-empirical models were applied to find out the dominant leaching mechanism of glass elements. Dissolution associated with diffusion was the dominant leaching mechanism and the elemental leaching characteristic depended upon its solubility in water. A theoretical leaching prediction model was applied to observe the long-term leaching behavior of major glass elements and surrogate nuclides. Diffusion coefficients and dissolution rate constants, the main parameters in the long-term prediction model, of glass elements and surrogates were obtained using short- and long-term experimental data. The model was found to be useful in predicting the long-term behavior of such elements in order to assess the stability of glassy waste forms.

NON-COMBUSTIBLE WASTE VITRIFICATION WITH PLASMA TORCH MELTER

Non-combustible radioactive wastes generated from Nuclear Power Plants (NPPs) are composed of concrete, glass, asbestos, metal, sand, soil, spent filters, etc. The melting tests for concrete, glass, sand, and spent filters were carried out using a 60 kW plasma torch system. The surrogate wastes were prepared for the tests. Non-radioactive Co and Cs were added to the surrogates in order to simulate the radioactive waste. Several kinds of surrogate prepared by their own mixture or by single waste were melted with the plasma torch system to produce glassy waste forms. The characteristics of glassy waste forms were examined for the volume reduction factor (VRF) and the leach rate. The VRFs were estimated through the density measurement of the surrogates and the glassy waste forms, and were turned out to be 1.2–2.4. The EPA (Environmental Protection Agency) Toxicity Characteristic Leaching Procedure (TCLP) was used to determine the leach resistance for As, Ba, Hg, Pb, Cd, Cr, Se, Co, and Cs. The leaching index was calculated using the total content of each element in both the waste forms and the leachant. The TCLP tests resulted in that the leach rates for all elements except Co and Cs were lower than those of the Universal Treatment Standard (UTS) limits. There were no UTS limits for Co and Cs, and their leach rate & index from the experiments were resulted in around 10 times higher than those of other elements.

00/03532 Leaching characteristics of glassy waste forms containing two different incineration ashes

00/03532 Leaching characteristics of glassy waste forms containing two different incineration ashes

Leaching characteristics of glassy waste forms containing two different incineration ashes

Several glassy waste forms are fabricated from two different incineration ashes, respectively, at different mixing ratios of ashes to base glass. Incineration ashes used in the experiments are hazardous waste (HW) incineration ash from a paint factory and simulated radioactive waste (RW) incineration ash from a nuclear power plant. To assess the chemical durability, which is known to be the most important factor in the waste form integrity, of glassy waste forms, long-term leaching tests such as MCC-4S (for HW glassy waste form) and ISO (for RW glassy waste form) methods were conducted at 70°C for over 500 days. By observing the initial leach ratio of glass components, it was found that silicon and calcium uniformly leached out after the early leaching of sodium and boron. For sodium and boron, diffusion is a dominant leaching mechanism, while dissolution associated with diffusion affects the leaching of silicon and calcium. Like glass constituents, leaching characteristics of hazardous metals and surrogate nuclides are also dependent upon solubility.

Immobilization of CsCl and SrF 2 in iron phosphate glass

A large part of the radioactive 137Cs and 90Sr stored at the Hanford site is present as CsCl and SrF 2. Attempts to incorporate halides in borosilicate glass often produces immiscibility which is undesirable for immobilizing nuclear materials. Conversely, iron phosphate glasses have been made with up to 26 and 31 mol% CsCl and SrF 2, respectively, and up to 34 mol% CsCl and SrF 2 combined. These compositions are melted at temperatures as low as 950°C for 2 h and form fluid melts. The low melting temperature of the iron phosphate wasteforms reduces the probability that radioactive 137Cs will vaporize from the melt, which becomes increasingly likely at the higher temperature (⩾1150° C) used to melt borosilicate glasses. Analysis of the glasses produced in this study show that very little, if any, Cs and Sr was volatilized from the batches, although the majority of the halide is released during melting. All of the iron phosphate glassy wasteforms in the current study had a total ion release in distilled water of approximately 10 mg/l which is about one tenth that measured for the approved reference material (ARM-1), a reference borosilicate glass. The dissolution rate of the iron phosphate wasteforms in distilled water decreased with increasing iron content.