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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Densification and Immobilization of AgI-Containing Iodine Waste Forms Using Spark Plasma Sintering

Here in this study, porous Ag-xerogel (Ag-Xero), Ag-faujasite (Ag-FAU) zeolite, and Ag-mordenite (Ag-MOR) zeolite sorbents were loaded with iodine gas [I 2 (g)] under saturated conditions at 150 °C for 24 h, followed by densification and consolidation into monolithic waste forms using spark plasma sintering (SPS). For Ag-Xero materials, SPS pellets were made with as-loaded samples, while others were made with preheated (PH; 500 °C for 2 h) samples to help with densification. SPS processing was conducted at 50 MPa under different temperatures (T = 200–800 °C) for different times (t = 0.5–30 min), where eleven AgI-Xero samples, five AgI-FAU, and two AgI-MOR separate samples were produced. The primary goal was to look for the optimum processing parameters for each material to yield pellets with high iodine retentions, high densities, and low porosities while preventing AgI decomposition. The Ag-Xero showed the highest iodine loadings (qe = 470 mg g –1 ) compared to Ag-FAU (qe = 368 mg g –1 ) and Ag-MOR (qe = 108 mg g –1 ). Measured iodine concentrations were the highest in AgI-Xero pellets without PH, followed by AgI-Xero with PH, AgI-FAU, and then AgI-MOR. Silver utilization (I/Ag on a mol % basis) values were in the order of AgI-MOR ≈ AgI-Xero (no PH) > AgI-Xero (PH) > AgI-FAU. Chemical durabilities of SPS-densified AgI-Xero (PH) pellets were very favorable, with lower releases than SPS pellets made from AgI-Xero samples without PH. These results show promise for iodine waste form production.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Reassessment of Dose Contributions from the M-Area Glass Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater (GW) and inadvertent human intruder (IHI) screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in U-235 being brought back into the list of radionuclides requiring inventory limits within the M-Area Glass SWF. In addition, while reviewing the earlier vadose zone (VZ) transport analysis efforts, an input deck error was observed in the U-234G VZ transport simulation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Small-scale Assessment of Simplified Ceramic Waste Form Processing

The feasibility of directly processing glass-bonded sodalite ceramic waste form materials by heating a mixture of Zeolite 4A, chloride salt, and sodium borosilicate binder glass was demonstrated by generating laboratory-scale materials. This direct processing method eliminates pre-setting the moisture content of the Zeolite 4A, eliminates the step of occluding salt in the prepared Zeolite 4A prior to processing, and can be conducted using larger particle sizes of crushed Zeolite 4A and crushed borosilicate glass than those called for in the current method. These simplifications are expected to facilitate material transfer and handling in a hot cell or controlled atmosphere environment and be more readily implemented at large scale than the current method. Most materials made to assess these simplifications were directly processed at 925 °C for two hours in an argon atmosphere glovebox, but one material was processed at 925 °C for four hours and one material was processed at 880 °C for two hours. Sodalite was generated and became microencapsulated by the binder glass in all materials. The microstructures were uniform throughout each product, were similar in all products, and were similar to the microstructures of materials made previously using the pressureless consolidation or hot isostatic pressing methods. Various formulations showed the efficiency of sodalite generation was not sensitive to the salt-to-Zeolite 4A ratio or salt-to-glass mass ratio, although a greater relative mass of glass is required to encapsulate sodalite generated from large particles of aggregated Zeolite 4A. The upper limit of salt loadings that can be effectively processed remains to be determined. The effectiveness of direct processing provided new insights into the conversion mechanism. The salt was likely dissolved into the glass that transported NaCl into the Zeolite 4A aggregates and sodalite was generated in situ as NaCl migrated from the outside of the aggregate inward. Other salt cations (e.g., potassium, strontium, cesium, and probably lithium) remain dissolved in the glass encapsulating the zeolite/sodalite domains. When the glass solidifies during cooling, small halite inclusion phases form in glass within sodalite domains and large mixed salt inclusions form in glass surrounding the sodalite due to the low solubility of chloride in the (solid) glass. Other salt cations were oxidized during processing and formed inclusions in the bulk glass (e.g., neodymium). Initial degradation tests show the dissolution behavior of directly processed CWF (SCWF) materials is similar to CWF materials made using different methods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Toward the performance assessment of advanced nuclear waste forms: temperature dependence of lanthanide borosilicate glass dissolution

Lanthanide borosilicate (LaBS) glasses are among the most promising waste forms for the immobilization of high-level radioactive waste generated from advanced nuclear fuel cycles. However, the temperature dependence of their dissolution kinetics remains poorly understood and constrained, limiting the integration of these materials into established performance assessment models. Here, we investigate the dissolution behavior of the legacy AmCm2-19 LaBS glass and the benchmark alkali aluminoborosilicate ISG-1 in deionized water between 50 °C and 250 °C using ASTM C1285 (Product Consistency Test-B) protocols. For AmCm2-19 LaBS glass, normalized elemental release rates for boron and silicon increase with temperature before plateauing near 150 °C, consistent with solubility-limited behavior. From data obtained at 50 °C and 100 °C, Arrhenius analysis yields activation energies of E a (B) = 24.8 ± 0.3 kJ mol⁻¹ and E a (Si) = 14.4 ± 0.2 kJ mol⁻¹, similar or slightly lower than those previously reported for two other compositions of LaBS glasses. No secondary phases or alteration layers were detected by SEM-EDX or pXRD. These results establish one of the first temperature-dependent kinetic datasets for LaBS glass dissolution, providing quantitative parameters to inform mechanistic corrosion models and predictive simulations of glass degradation in geological disposal environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Crystalline versus Glassy Nature of Iron Phosphate Waste Forms Subjected to Different Slow Cooling Curves

This report provides experimental details and associated results on the elemental distributions (compositions), crystalline/amorphous phase distributions, and microstructures for iron phosphate reference materials (DPF5-336) made with different cooling curves and starting from different temperatures (i.e., 1050°C and 1200°C). The primary goal of this work was to evaluate the properties of the DPF5- 336 reference material after it was melted and cooled at different rates to simulate the cooling profile of different types of canisters. Crystal fractions across the SCC#2, SCC#3, and SCC#5 samples were all similar ranging in crystal content of 35.66–41.14 mass% with amorphous fractions being the balance (64.34–58.86 mass%, respectively). This shows that the heat treatment profile did not seems to greatly affect either the total crystal content or the phase distributions. The quenched sample was almost completely amorphous (99.55 mass%) with very small peaks identified as Li 3 Fe 2 (PO 4 ) 3 whereas the SCC-treated materials showed peaks for what appear to be ten separate phases present in various concentrations. Another goal of these experiments was to see how the phase distribution affected chemical durability of these waste forms, but that information is being collected at Argonne National Laboratory and will be published in a separate report

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Chemical durability and corrosion-induced microstructure evolution of compositionally complex titanate pyrochlore waste forms with uranium incorporation

Chemical durability of compositionally-complex rare-earth titanate pyrochlore solid solutions with or without uranium incorporation was investigated by semi-dynamic leaching testing. A Ti-enriched amorphous passivation film formed on the surface due to the surface alteration-reorganization mechanism with preferential release of the weakly oxygen-bonded rare earth elements. Elemental release was found to be strongly correlated with chemical disorder, with a negligible correlation with entropy. The release rate of the uranium gradually decreased with increasing numbers of components, implying a possibility of designing and optimizing waste form performance with enhanced chemical durability by controlling composition complexity and chemical disorder for effective nuclear waste management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Synthesis and characterization of super occluded LiCl-KCl in zeolite-4A as a chloride salt waste form intermediate

Here, this paper reports the hygroscopic properties of eutectic LiCl-KCl after absorption into zeolite-4A, up to salt loadings of 75 wt%. Samples of the salt occluded zeolite were hydrated in a humidity chamber at constant temperature and relative humidity for up to 100 h. At up to 45 wt% salt loading, the un-occluded phase of salt consisted primarily of NaCl, which forms when the Na + ions present in the zeolite framework exchange with Li + and K + ions from the eutectic LiCl-KCl. This results in minimal water absorption and corrosion of contacted stainless steel. At greater than 45 wt% salt loading, water absorption and corrosion progressively worsened. The mixture has a significant amount of excess LiCl-KCl, making it highly hygroscopic. This study reveals an option for the intermediate treatment of waste salt from spent nuclear fuel electrorefiners that could facilitate it to be stored in a non-inert atmosphere for extended periods of time before final conversion into a permanent waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Analysis of NuCycle® Process Waste Streams and Identification of Candidate Waste Forms (CRADA 719) Abstract

The objective of this project is to support collaboration between PNNL and Curio to address technical gaps in the immobilization of waste streams generated by the NuCycle® process, thereby improving the overall viability of the technology. This work will be carried out in two phases, described below. Phase 1 – Waste Management Study The first phase consists of a comprehensive waste management study focused on the waste streams produced during the NuCycle® process, including those containing long lived fission products (LLFPs, e.g., I 129) and transuranic elements (TRUs). This phase involves compiling waste compositions and characteristics based on recently completed NuCycle® testing and modeling efforts. Established waste forms and processing methods will be identified for streams where they are applicable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Phase field-volumetric lattice Boltzmann model of ion uptake in porous nuclear waste form materials under continuous flow

The flow field within the mesopores of sorbent particles plays a crucial role in radionuclide diffusion and ion uptake kinetics, thus, impacting the overall performance of porous nuclear waste form materials. To fundamentally understand the influence of microstructures and material properties on the radionuclide absorption and retention processes requires a coupled multi-physics model that considers the advection and diffusion within the flow field, the reaction at liquid-solid interfaces, and finally, the solid-state diffusion within a complex nanoporous medium. Here, this study employs the volumetric lattice Boltzmann method (VLBM) to accurately and efficiently calculate the steady state velocity field inside the mesopores of sorbent particles. The obtained velocity field is then utilized to calculate the advection of ions in the steady flow. A phase field (PF) model of ion uptake is used to describe the reaction occurring at the solid-liquid interface and diffusion inside the porous medium. The integrated PF-VLBM model is verified in terms of the mass conservation and numerical efficiency and validated qualitatively with experimental observation data. Then, it is applied to study the influence of thermodynamic and kinetic properties, as well as flow field conditions on the ion uptake kinetics. The numerical results demonstrate that the ion uptake kinetics in porous particles has three distinct stages, which is in agreement with the observations in continuous flow experiments. In the first stage, the kinetics is predominantly controlled by the flow field and ion diffusivity in the liquid phase. The kinetics in the second stage is primarily governed by ion diffusivity in the solid phase. In the third stage the system reaches a dynamic equilibrium with a net zero uptake flux at the interface. It is also found that porous structures significantly affect the efficiency and capacity of ion uptake. The simulation results can help to understand the physics behind the observed ion uptake kinetics in experiments and to facilitate the development of constitutive equations that can account for heterogeneous microstructures in engineering performance codes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Impact of oxidation and carbonation on the release rates of iodine, selenium, technetium, and nitrogen from a cementitious waste form

Evaluation of the long-term retention mechanisms and potential release rates for the primary constituents of potential concern (COPCs) (i.e., Tc, I, Se, and nitrate) is necessary to determine if Cast Stone, a radioactive waste form, can meet performance objectives under near-surface disposal scenarios. In this work, a mineral and parameter set accounting for the solubility of I and Se in Cast Stone was developed based on pH-dependent and monolithic diffusion leaching test results, to extend a geochemical speciation model previously developed. The impact of oxidation and carbonation as environmental aging processes on the retention properties of Cast Stone for primary COPCs was systematically estimated. Physically, the effective diffusion coefficients of 4 COPCs in Cast Stone were increased after carbonation and/or oxidation, reflecting an increase in permeability to diffusion. Chemically, i) pH & pe conditions in the original Cast Stone were favorable for the stabilization of Tc, but not for I, Se, and N; ii) oxidation (with/without carbonation) of Cast Stone changed the pe & pH conditions to be detrimental for Tc stabilization; and iii) carbonation (with/without oxidation) of Cast Stone modified the pH & pe conditions to be beneficial for the stabilization of I (in system with Ag added) and Se.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Process optimization of caustic scrubber and iodine-129 immobilization in sodalite-based waste forms

Caustic scrubbers are proposed to remove vaporized iodine from nuclear fuel reprocessing plants. Resulting caustic slurries of radioactive iodine ( 129 I) and other anions can be captured and immobilized into the cage structures of sodalite-type minerals. Here, mixed anion sodalites were hydrothermally synthesized from a simulated caustic scrubber solution containing hydroxide/water, carbonate, chlorine, bromine, iodine, nitrate, and nitrite ions with added kaolinite. Experiments were conducted with twenty-five different sets of process conditions. Resulting powders were characterized by quantitative X-ray diffraction (XRD) and multi-functional thermal analysis, including gravimetry, calorimetry, and mass spectrometry. Using statistical analysis, process variables were tailored to maximize conversion of kaolinite into cage silicates, immobilize caustic scrubber anions, and minimize amorphous content. Further, chemical analysis showed the incorporation of all the targeted ions into the solids. Detailed investigation of a batch characterized after washing (creating secondary waste) and one without washing but subsequent calcination is given. This work contributes to defining the constraints required for subsequent consolidation process steps of the waste form to avoid volatilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Effect of charge and anisotropic diffusivity on ion exchange kinetics in nuclear waste form materials

Motifs of radionuclide absorption particles (absorbers or fillers) for the nuclear waste treatment usually have crystal structures with nano-sized tunnels for fast ion diffusion. On the other hand, the accumulation of ions generates an electric field which affects the ion diffusion. Therefore, the strong anisotropic kinetic properties and the electric field affect ion exchange kinetics, especially in an assembly of these particles which may block the fast diffusion tunnels for each other. In this work, we developed a mesoscale model to describe the effect of anisotropic kinetic properties and electric field on ion exchange kinetics. With the model we simulated the effect of anisotropic diffusivity, electrochemical potential, particle morphology, size and aggregation on ion exchange kinetics during batch experiments. Here, tt is found that 1) ion exchange results in charge accumulation inside the particle because different ions have different diffusivities and different diffusion driving forces. The electric neutrality is not preserved; 2) the electric field speeds up the slower diffusion ions (either uptake or release) and slows down the faster diffusion ions to reduce charge accumulation and maintains charge neutrality; 3) the ion exchange becomes very slow at the later stages because a) diffusion driving force (electrochemical gradient, Cs + concentration in the solution and Na + inside the particle) continuously decreases, as observed in batch experiments; and b) the diffusivity decreases due to the Cs + concentration dependence of ion diffusivity. This can explain the observed experimental phenomena that the radionuclide capacity usually cannot be reached; 4) for needle-like particle and their clusters, both surface area and diffusion distance along the direction with the fastest diffusivities have significant impact on the ion exchange kinetics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The corrosion behavior of borosilicate glass in the presence of cementitious waste forms

Borosilicate glasses are widely used for radioactive waste disposal due to their ability to incorporate a variety of contaminants and radionuclides while exhibiting high durability in various disposal scenarios. This research evaluated the dissolution of borosilicate glass using both single-pass-flow-through (ASTM C1662-18) and product consistency test (ASTM C1285-21) methods with different solutions, including a cementitious-contacted water (called grout-contacted, GC, from this point) and solutions with varying levels of dissolved cementitious species such as Si, Ca, Al. Here, the results indicated that the presence of Ca plays a crucial role in suppressing glass corrosion, as evidenced by the slower normalized dissolution rates, which were one order of magnitude lower for boron and two orders of magnitude lower for rhenium, observed in both Ca-amended and GC solutions compared to the pH 12 buffer solution. This effect is attributed to the formation of a dense, low-porosity, and strongly bonded calcium silicate hydrate (CSH) layer on the glass surface, which implies that a glass corrosion process is influenced by ion exchange involving alkali ions Na + , K + , Ca 2+ , and hydrogen-containing species. A small number of glass particles treated in the GC solution showed minor corrosion pits in the form of shallow craters with an average diameter of approximately 500 μm. This observation is correlated with a significant reduction, 2000 to 3000 times lower, in the cumulative volume of glass pores, indicating that smaller pore voids were “sealed” in the presence of Ca 2+ ions, likely attributed to the formation of CSH precipitation or other corrosion products such as calcium carbonate saturated from the grout solution. These findings suggest that the presence of dissolved Ca in the GC solution can slow down the dissolution of borosilicate glass, contrary to the expected trend of higher dissolution rates resulting from exposure to high alkaline and thus higher pH solutions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Towards informatics-driven design of nuclear waste forms

Informatics-driven approaches, such as machine learning and sequential experimental design, have shown the potential to drastically impact next-generation materials discovery and design.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

A Review of Advanced Characterization Methods and Techniques for Amorphous and Poorly Crystalline Materials Applicable to Cementitious Waste Forms

This review covers advanced characterization methods of cementitious materials. The particular focus is to understand cementitious compositions with amorphous content because conventional techniques such as X-ray diffraction are more applicable to materials with long range order. Also of interest is advanced characterization that provides deeper insights into properties such as porosity. There are five groups of characterization methods discussed in this review which are: electron beam/microscopy, X-ray, neutron diffraction and scattering, nuclear magnetic resonance, and optical techniques. The techniques can be further grouped by the categories of nanoscale structural understanding (transmission electron microscopy and pair distribution function), microscale structural understanding (focused ion beam and x-ray tomography), microscopic elemental distributions (energy dispersive spectroscopy and X-ray fluorescence), and chemical binding information (electron energy loss spectroscopy, x-ray absorption spectroscopy, nuclear magnetic resonance, Raman spectroscopy, and infrared spectroscopy), with advantages and disadvantages of each method discussed. The recommendation for further research is to begin with techniques that are easier to prepare samples for, measure, and analyze, such as the optical techniques of Raman and infrared spectroscopy. If more information is needed, pursue collaborations, other techniques, or proposals for beam time at the synchrotron sources.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Replacement of Fly Ash with Natural Pozzolans in Cementitious Waste Forms: Progress Report

The Department of Energy Environmental Management (DOE EM) Office is currently responsible for treating radioactive and mixed waste, environmental restoration, and closing contaminated tanks and facilities resulting from radioactive isotope production for defense applications during the Cold War. Cementitious reagents are the most widely used materials for (1) chemically stabilizing and encapsulating radionuclides and hazardous metals contaminated with radionuclides and Resource Conservation and Recovery Act (RCRA) hazardous metals and (2) solidifying radioactive and mixed wastewater for final disposal. Cementitious grouts and flowable concretes are also the most widely used materials for DOE waste tank and facility closures.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Preparation of iodine-bearing waste forms for hot isostatic pressing

Oak Ridge National Laboratory (ORNL) has prepared a new set of iodine-bearing sorbents. The sorbent materials used for this activity were supplied by IONEX (NexSorb-99) or C&CS (AC6120-7%). Metal canisters were previously prepared by ORNL in 2022. They are 2 in. long by 1.5 in. in diameter and are made of 316 stainless steel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗