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The importance of glass and glass-ceramic nuclear waste forms has been reaffirmed in recent years by the growing interest in nuclear power as a reliable energy source to meet the requirements of technologies such as artificial intelligence. Waste processing schemes for the disposal of halide-containing wastes will be essential for the advancement of nuclear technologies such as non-aqueous fuel reprocessing. Phosphate-based dechlorination and subsequent vitrification of radioactive salt waste into an iron phosphate waste form has been identified as a potential processing scheme for electrochemical processing waste. The impact of H3PO4-based dechlorination of complex salt mixtures on the vitrification processmore » and structure of the final iron phosphate waste form has not yet been investigated. In this work, iron phosphate glass-ceramics were made from simulant salt waste (48LiCl-33KCl-19NaCl mol%) dechlorinated with the H3PO4-based method. The glass forming region was compared to that of traditionally prepared Na2O-Fe2O3-P2O5 systems. For a candidate glass-forming composition, the processing scheme presented here was determined to favor Fe3+ species. The O/P molar ratio was consistent for the candidate composition when dechlorinated at 400°C and 600°C in air and argon environments, indicating glass network connectivity was maintained despite variations in processing parameters. The results presented here validate processing schemes requiring iron-phosphate waste form synthesis following H3PO4-based dechlorination.« less

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

The U.S. Department of Energy Hanford Site in Washington State is in the process of commissioning the Waste Treatment and Immobilization Plant to process a portion of the 54 million gallons of radioactive and chemical waste from cold war weapon production. Technologies for the capture of volatile species of concern are still being assessed, and new methods and materials are developed as operational and flowsheet mission risks are identified. One such area still being assessed is the abatement efficacy of the Carbon Adsorber units to retain gaseous mercury and ¹²⁹I released during processing. It is challenging to predict the mercurymore » chemistry due to the variability of the feed, and different methods/materials are required for the capture of gaseous Hg⁰ and Hg^II compounds. In this study, the feasibility of using developmental carbon foam (CF) sorbents for the capture of iodine and mercury was assessed using static and dynamic flow testing and compared against a commercially available sorbent, BATII-37. Both CF and CF functionalized with bismuth particles (CF-Bi) chemisorbed iodine, and CF-Bi had similar mercury capture performance to BATII-37 in dynamic flow tests. While species loading concentrations were measurable, limitations in achieving a mass balance prevented a full evaluation of capture efficacy. Nonetheless, the results serve as an important first step in demonstrating the potential for simultaneous iodine and mercury capture.« 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

The development of nuclear waste forms for both existing and future nuclear wastes is critical to ensuring global environmental safety. This study focuses on waste management from molten salt reactors, where fuel exists in a salt form and could be processed in real time for the removal of neutron poisons such as xenon isotopes (e.g., ¹³⁵Xe) and rare earth elements (REEs, e.g., ¹⁴⁹Sm). To ensure safe, stable, and long-term disposal in geological repositories, REEs must be incorporated into a durable waste form. Iron-phosphate glasses are a promising candidate due to their low melting points, high chemical durability, and their abilitymore » to incorporate high concentrations of REEs. In this study, we successfully prepared iron-phosphate glass waste forms with high Nd loadings (up to 37 mass %) in batch sizes ranging from small (23 g) to large (1600 g). The resulting materials contained up to 75 mass % NdPO₄, contributing to their mechanical resilience and exceptional chemical durability. These findings highlight the potential of iron-phosphate glasses as high-efficiency, chemically durable waste forms and demonstrate the successful transition from theoretical design to scaled-up production.« less

The increasing reliance on nuclear energy as a significant low-carbon power source necessitates effective solutions for managing radioactive emissions. This study introduces a novel application of MXene nanohybrids, specifically silver-MXene (Ag-Ti₃C₂T_x), as an effective sorbent for radioiodine off-gas capture at an operating temperature of 150 °C. Through comprehensive material characterization, including X-ray diffraction, scanning and transmission electron microscopies, energy-dispersive X-ray spectroscopy, Raman spectroscopy, thermogravimetric analysis, inductively coupled plasma optical emission spectroscopy, and gas sorption analyses, the successful loading of Ag nanoparticles onto Ti₃C₂T_x is confirmed and the subsequent formation of AgI upon iodine capture. The results demonstrate that Ag-Ti₃C₂T_x exhibitsmore » superior iodine uptake compared to traditional silver-based sorbents such as silver mordenite zeolite (AgZ) and silver-functionalized silica aerogel (AgAero). The Ag-Ti₃C₂T_x achieves an iodine loading of 946 mg g⁻¹, significantly outperforming AgZ (131 mg g⁻¹). These findings highlight the potential of Ag-Ti₃C₂T_x as a highly efficient, thermally stable sorbent for radioiodine capture, and potentially addressing key limitations of existing materials.« less

The vitrification of high-level waste (HLW) by heating a mixture of glass-forming chemicals (GFCs) with the waste can be improved using a constrained optimization problem. This study explores how different uncertainty propagation (UP) methods implemented with the optimization process can affect the glass formulation of nuclear waste glasses. UP is the effort of propagating uncertain inputs through a system to understand and quantify output distributions. Uncertainty intervals are crafted from output distributions to inform the optimization algorithm. UP is often implemented with Monte Carlo (MC) sampling for large nonlinear systems, which can be difficult to implement within a constrained optimizationmore » algorithm that requires derivative information. Other UP methods often used for optimization under uncertainty (OUU) can be designed to work within an established constrained optimization framework. Methods of UP are evaluated in this study including iterative sampling approaches, first-order approximations, and surrogate modeling with machine learning (ML). A method of dimensional reduction based on global sensitivity analysis is introduced to support the UP methods for the large dimensionality of the problem. Analytical UP methods able to achieve similar optimums 10 times faster than the baseline MC approach, and produce 93.9% similar output distributions are reported.« 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

Dual-sorbent systems are proposed for radioiodine management with a regenerated primary bed for multiple cycles of use in complex conditions and a secondary bed for disposal with higher waste loadings. Sorbent approaches for the effective capture of gaseous radioiodine (isotopes ¹²⁹I and ¹³¹I) produced from a range of nuclear processes have been studied for over half a century. (1−5) Whether or not a sorbent (e.g., molecular sieve) is required to physically screen/trap or chemically bind a radionuclide of interest through chemisorption, the complexity of the gas stream has a large impact on the performance (e.g., loading capacity, selectivity) and activemore » life of a sorbent bed. (3) Silver mordenite (AgZ), the U.S. Department of Energy baseline sorbent for radioiodine capture from nuclear processes, performs well within acidic conditions and at elevated temperatures (6) and can be consolidated into a chemically durable waste form for long-term disposal. (7,8) However, new sorbents are being sought because optimal capture performance of AgZ significantly decreases in dynamic oxidizing environments with competing species, and it is expensive and it contains Ag (a toxic metal). (9) Until a new sorbent is found to replace AgZ, the regeneration and reuse of AgZ is an attractive alternative to a single-use primary sorbent bed. In this regard, a primary sorbent could be designed for enhanced capture in complex gas streams and the ability to be regenerated for reuse. Here, a secondary sorbent could then be tailored for maximum iodine loading in the gas stream and chemical durability within a disposal facility.« less

Four promising iodine “getter” materials (Ag, Cu, Bi, and Sn) for radioiodine capture were assessed in their pure metal-iodide (MI_x) pelletized forms to compare relative chemical durabilities. To study chemical durability, commercial MI_x compounds of AgI, BiI₃, BiOI, CuI, and SnI₄ were converted to dense monolithic pellets using spark plasma sintering. Semidynamic leach testing in the form of modified ASTM C1308 tests was then performed on the pellets in two different forms including unmounted (as-pressed) specimens (i.e., “U”) and epoxy-mounted specimens (i.e., “M”) with polished surfaces. The chemical durability results and sample characterizations showed that three of the five MI_xmore » compounds tested (i.e., AgI, CuI, and BiOI) displayed moderate to high leach resistances. Further, the remaining two MI_x compounds (i.e., BiI₃ and SnI₄), which are both desirable iodine waste forms due to their high iodine loading capacities, readily decomposed during leach testing, indicated by crystallographic changes in the specimens as well as large amounts of iodine detected in the leachate solutions. The instabilities of BiI₃ and SnI₄ raise uncertainties for using the base metals/cations (i.e., Bi⁰/Bi³⁺ and Sn⁰/Sn⁴⁺, respectively) as viable getters for radioiodine capture due to likely poor waste form chemical durabilities after capture and consolidation into waste forms.« less