Technetium mainly forms during artificial nuclear fission; it exists primarily as TcO4(-) in nuclear waste, and it is among the most hazardous radiation-derived contaminants because of its long ...half-life (t1/2 = 2.13 × 10(5) years) and environmental mobility. The high water solubility of TcO4(-) (11.3 mol L(-1) at 20 °C) and its ability to readily migrate within the upper layer of the Earth's crust make it particularly hazardous. Several types of materials, namely resins, molecular complexes, layered double hydroxides, and pure inorganic and metal-organic materials, have been shown to be capable of capturing TcO4(-) (or other oxoanions) from solution. In this review, we give a brief description about the types of materials that have been used to capture TcO4(-) and closely related oxyanions so far and discuss the possibility of using metal-organic frameworks (MOFs) as next-generation ion-exchange materials for the stated application. In particular, with the advent of ultra-stable MOF materials, in conjunction with their chemical tunability, MOFs can be applied to capture these oxyanions under real-life conditions.
Efficient and cost‐effective removal of radioactive pertechnetate anions from nuclear waste is a key challenge to mitigate long‐term nuclear waste storage issues. Traditional materials such as resins ...and layered double hydroxides (LDHs) were evaluated for their pertechnetate or perrhenate (the non‐radioactive surrogate) removal capacity, but there is room for improvement in terms of capacity, selectivity and kinetics. A series of functionalized hierarchical porous frameworks were evaluated for their perrhenate removal capacity in the presence of other competing anions.
Going nuclear: Functionalized hierarchical porous frameworks were used for the efficient removal of perrhenate from aqueous solution (see figure).
The kinetics of the feed-to-glass conversion affects the waste vitrification rate in an electric glass melter. The primary area of interest in this conversion process is the cold cap, a layer of ...reacting feed on top of the molten glass. The work presented here provides an experimental determination of the temperature distribution within the cold cap. Because direct measurement of the temperature field within the cold cap is impracticable, an indirect method was developed in which the textural features in a laboratory-made cold cap with a simulated high-level waste feed were mapped as a function of position using optical microscopy, scanning electron microscopy, energy dispersive spectroscopy, and X-ray diffraction. The temperature distribution within the cold cap was established by correlating microstructures of cold-cap regions with heat-treated feed samples of nearly identical structures at known temperatures. This temperature profile was compared with a mathematically simulated profile generated by a cold-cap model that has been developed to assess the rate of glass production in a melter.
The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the ...present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm by mass, added as KReO(4) (Re(7+)) were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be <10 ppm (by mass), these Re results implied that the solubility should not be a limiting factor in processing radioactive wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.
The feed‐to‐glass conversion, which comprises complex chemical reactions and phase transitions, occurs in the cold cap during nuclear waste vitrification. To investigate the conversion process, we ...analyzed heat‐treated samples of a simulated high‐level waste feed using X‐ray diffraction, electron probe microanalysis, leaching tests, and residual anion analysis. Feed dehydration, gas evolution, and borate phase formation occurred at temperatures below 700°C before the emerging glass‐forming melt was completely connected. Above 700°C, intermediate aluminosilicate phases and quartz particles gradually dissolved in the continuous borosilicate melt, which expanded with transient foam. Knowledge of the chemistry and physics of feed‐to‐glass conversion will help us control the conversion path by changing the melter feed makeup to maximize the glass production rate.
Volatile loss of radioactive technetium-99 (99Tc) to off-gas is a major challenge when vitrifying low-activity waste (LAW) at the U.S. Department of Energy's Hanford Site in Washington State. We ...investigated the partitioning and incorporation of rhenium (Re) (a nonradioactive surrogate for 99Tc) into the glass melt during crucible melting of two simulated LAW feeds that have exhibited a large difference in 99mTc/Re retention in glass from small-scale melter tests. Each feed was prepared from a simulated liquid LAW and additives (boric acid, silica sand, etc.). The as-mixed slurry feeds were dried at 105°C and heated to 600–1100°C at 5K/min. The dried feeds and heat-treated samples were leached with deionized water for 10min at room temperature followed by 24-h leaching at 80°C. Chemical compositions of the resulting solutions and insoluble solids were analyzed. Volume expansion measurements and X-ray diffraction (XRD) analyses were performed on dried feeds and heat-treated samples to characterize the progress of feed-to-glass conversion reactions. We found that incorporation of Re into the glass melt was virtually completed during the major feed-to-glass conversion reactions that occurred at ≤700°C. The results of our study suggest that the different compositions of the salt phases formed during early stages of melting at ≤700°C are responsible for the large difference in Re incorporation into the glass melt in these two feeds.
•We studied incorporation of Re (99Tc surrogate) into low-activity waste glass melt.•Two glass feeds with different Tc retention in glass from melter tests were used.•Leaching and chemical analyses were performed to examine Re partitioning.•We propose the source of large difference in Re retention for these two glass feeds.•Different salt compositions formed during melting are responsible for the difference.
The rate of glass production during vitrification in an all‐electrical melter greatly impacts the cost and schedule of nuclear waste treatment and immobilization. The feed is charged to the melter on ...the top of the molten glass, where it forms a layer of reacting and melting material, called the cold cap. During the final stages of the batch‐to‐glass conversion process, gases evolved from reactions produce primary foam, the growth and collapse of which controls the glass production rate. The mathematical model of the cold cap was revised to include functional representation of primary foam behavior and to account for the dry cold cap surface. The melting rate is computed as a response to the dependence of the primary foam collapse temperature on the heating rate and melter operating conditions, including the effect of bubbling on the cold cap bottom and top surface temperatures. The simulation results are in good agreement with experimental data from laboratory‐scale and pilot‐scale melter studies. The cold cap model will become part of the full three‐dimensional mathematical model of the waste glass melter.
Pyroprocessing is are processing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while ...avoiding wet reprocessing. Pyroprocessing UNF with a LiCl–KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the glass matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.
•We investigated crystallization in borosilicate glasses containing rare earth oxides.•New crystallinity and durability data are shown for glasses proposed in the literature.•Both liquidus temperature and chemical durability increased as the waste loading increased.
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