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Phosphate-based reagents are being considered by the U.S. Department of Energy (DOE) Office of Nuclear Energy to process halide salt-based nuclear wastes for stabilization prior to disposal. As evidenced by the Experimental Breeder Reactor-II (EBR-II) project, electrochemical processing (pyroprocessing) can be employed to recover uranium and other actinides for reintegration into the nuclear fuel cycle from metallic fuels. The resultant salt-based wastes generated from electrochemical processing of EBR-II fuel contains fission products within a LiCl–KCl eutectic salt that necessitate appropriate disposal. This paper provides an overview of recent efforts to support halide-based salt waste treatment for disposition, as well asmore » a basis for comparison with other related efforts in salt waste treatment through salt partitioning initiatives. The U.S. DOE has selected a phosphate waste form reference material for further investigation and longer-term studies.« less

Only a handful of high-halide salt waste forms have been demonstrated for vitrification-based immobilization strategies for halide-salt nuclear waste streams (e.g., pyroprocessing wastes, molten salt reactor wastes) and they all have low waste loading potential and most have low chemical durabilities for high-alkali streams. An alternative approach to direct salt immobilization is salt partitioning prior to waste form fabrication and one option for partitioning is halide removal (called dehalogenation). Removing the halogen fraction through dehalogenation can significantly reduce the waste volume required for disposal in the primary waste form. Furthermore, when dehalogenation is performed using organic acids, the dehalogenation reagentmore » can decompose during high-temperature vitrification, reducing waste loading limitations in the waste form. In the current work, different organic acids (i.e., oxalic, formic, acetic, oxamic, and citric) were evaluated for dehalogenation efficiency of a simple chloride salt simulant (7.19% LaCl₃, 53.77% LiCl, and 39.04% KCl, by mole) and a more complex chloride salt simulant called ERV3 (electrorefiner version 3) at 150 °C–300 °C and using H⁺/Cl^–molar ratios of 1:1, 2:1, and 3:1. Additionally, a borosilicate glass waste form called TARS (or the average of refined specifications) was formulated, produced, and characterized for dehalogenated ERV3.« less

Here, waste forms containing glassy and crystalline phosphate and silicate phases were produced to immobilize salt waste simulants from pyroprocessing and characterized by using Raman spectroscopy, Mössbauer spectroscopy, X-ray diffraction, scanning electron microscopy, heat capacity, and chemical durability measurements. In this work, a phosphosilicate waste form is presented to leverage the benefits of both borosilicate glasses and iron phosphate glasses. To improve waste loading, prior to immobilization, salt simulants were successfully dechlorinated using ammonium dihydrogen phosphate, mixed with a borosilicate frit (5–30 wt %) and Fe₂O₃, and vitrified. Additions of 2.5–15 wt % borosilicate glass (NBS3) improved normalized release ratesmore » for Cs relative to iron-phosphates without NBS3, resulting in chemical durabilities similar to high-level waste borosilicate glass reference materials. The release rates of the alkalis (i.e., Li, Na, K, Cs) were the lowest with the addition of 5 wt % NBS3. Although Sr was not specifically targeted in this study, evidence exists that it preferentially partitioned with Si to form an amorphous droplet phase within the iron phosphate glass matrix.« less

This Review summarizes a range of analytical techniques that can be used to detect, quantify, and/or distinguish between isotopes of iodine (e.g., long-lived ¹²⁹I, short-lived ¹³¹I, stable ¹²⁷I). One reason this is of interest is that understanding potential radioiodine release from nuclear processes is crucial to prevent environmental contamination and to protect human health as it can incorporate into the thyroid leading to cancer. It is also of interest for evaluating iodine retention performances of next-generation iodine off-gas capture materials and long-term waste forms for immobilizing radioiodine for disposal in geologic repositories. Depending upon the form of iodine (e.g., molecules,more » elemental, and ionic) and the matter state (i.e., solid, liquid, and gaseous), the available options can vary. In addition, several other key parameters vary between the methods discussed herein, including the destructive vs nondestructive nature of the measurement process (including in situ vs ex situ measurement options), the analytical data collection times, and the amount of sample required for analysis.« less

Here this paper reviews key properties, applications, and examples of ceramic-metal composites (cermets), and metal matrix composites (MMC) with emphasis on their applicability as waste forms for immobilizing nuclear waste. While the literature is mature for vitrified and cementitious radioactive waste forms, cermet materials have not received adequate attention as potential candidates for immobilizing nuclear waste. To promulgate this effort, this review connects cermet and MMC design, such as hardened tools, with the chemistry of radioactive waste streams. A discussion on certification and qualification standards for cermet waste forms, literature gaps, and “how-to” sections on cermet processing techniques. Key parametersmore » discussed include thermal conductivity, chemical durability, and waste loading, as well as examples for metal-containing waste forms. As cermet waste forms gain momentum within the community, this review paper aims ensure the end-of-life nuclear fuel cycle is addressed from a materials and waste disposition perspective.« less