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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Radionuclide Incorporation and Long Term Performance of Apatite Waste Forms

This project aims to combines state-of-the-art experimental and characterization techniques with atomistic simulations based on density functional theory (DFT) and molecular dynamics (MD) simulations. With an initial focus on long-lived I-129 and other radionuclides such as Cs, Sr in apatite structure, specific research objectives include the atomic scale understanding of: (1) incorporation behavior of the radionuclides and their effects on the crystal chemistry and phase stability; (2) stability and microstructure evolution of designed waste forms under coupled temperature and radiation environments; (3) incorporation and migration energetics of radionuclides and release behaviors as probed by DFT and molecular dynamics (MD) simulations; and (4) chemical durability as measured in dissolution experiments for long term performance evaluation and model validation.

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DOE OSTI2019

Performance Metric for Cementitious Waste Form Inventory Release in the Integrated Disposal Facility

This report describes work supporting Phase 1 of the 2018 Department of Energy (DOE) Office of River Protection’s (ORP) Grand Challenge recipient proposal “Cementitious Immobilization of Treated LAW and Ancillary Benefits to Risk Reduction in Cross Site Transfer”. Phase 1 was to establish a performance metric for grout leachability against which laboratory leach data (as observed diffusivities) can be assed, followed by a comparison of historical test data against this metric. Historically, advancements in the cementitious (grout) waste form performance have been made over the last 40 years, with the majority of progress achieved within the past 8 years as part of the Washington River Protection Solutions, LLC testing program. This report presents a summary of that effort relevant to the Hanford Site.

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DOE OSTI2014

Deep Borehole Disposal Research: Geological Data Evaluation Alternative Waste Forms and Borehole Seals

This report documents deep borehole disposal research during FY2014, as directed by U.S. Department of Energy (DOE) Used Fuel Disposition (UFD) Campaign. These research efforts are principally directed at advancing the deep borehole disposal project to the implementation of a full-scale Research, Development, and Demonstration (RD&D) project. Activities of particular relevance to this goal include evaluation of guidelines for selection of a deep borehole field test site, analyses of deep borehole disposal of alternative waste forms, and technical planning for borehole seals research.

Arnold, Bill 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.

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DOE OSTI2015

Heat of Hydration of Low Activity Cementitious Waste Forms

During the curing of secondary waste grout, the hydraulic materials in the dry mix react exothermally with the water in the secondary low-activity waste (LAW). The heat released, called the heat of hydration, can be measured using a TAM Air Isothermal Calorimeter. By holding temperature constant in the instrument, the heat of hydration during the curing process can be determined. This will provide information that can be used in the design of a waste solidification facility. At the Savannah River National Laboratory (SRNL), the heat of hydration and other physical properties are being collected on grout prepared using three simulants of liquid secondary waste generated at the Hanford Site. From this study it was found that both the simulant and dry mix each had an effect on the heat of hydration. It was also concluded that the higher the cement content in the dry materials mix, the greater the heat of hydration during the curing of grout.

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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.

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DOE OSTI2014

Advanced Waste Form and Melter Development for Treatment of Troublesome High-level Wastes

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe 2 O 3 (with higher Al 2 O 3 ). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.

Marra, James↗
DOE OSTI2015

Advanced Waste Form and Melter Development for Treatment of Troublesome High-level Wastes

A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

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DOE OSTI2017

Solid Secondary Waste Form Testing Activities for the Integrated Disposal Facility Performance Assessment Support - 17567

Solid secondary waste (SSW) that will be generated at the Waste Treatment and Immobilization Plant (WTP) and from other activities at the Hanford site is destined for disposal at the Integrated Disposal Facility (IDF). Construction of the IDF was undertaken based on a Performance Assessment (PA) performed in 2001. In 2016, a SSW data package was prepared to support the development of an updated PA (2017 IDF PA). There are many inputs required to support the IDF PA. Data needs were identified based on experience from the previous PAs and from the development of conceptual models and modeling approaches to be applied for the 2017 PA. Interactions with the IDF PA team and the inventory data package developer helped to focus detailed efforts on a few key waste streams, contaminants of concern, and disposal configurations that are expected to be the primary contributors to the dose resulting from SSW disposed at the IDF. The key waste streams include: High Efficiency Particulate Air (HEPA) filters, ion exchange resins, granular activated carbon, and Ag-mordenite. The key parameters addressed in the 2016 data package include: saturated hydraulic conductivity, moisture characteristic curves, effective diffusion coefficient, distribution coefficient and porosity. The IDF PA team also identified a set of key contaminants that are expected to be the primary concern for the PA calculations (Tc-99, I-129, Cs-137, Sr-90, uranium isotopes (and total uranium), chromium, mercury and nitrate). These contaminants were the focus of the SSW data package. Specific formulations had not been identified for cementitious materials that will be used to encapsulate or stabilize SSW, and no IDF-specific experiments had been conducted to obtain the data for the PA. Thus, for the 2016 SSW data package, recommended property values were provided for a range of representative candidate materials based on a review of existing literature. However, starting in FY17, IDF-specific experiments will be performed for Hanford SSW expected to contribute most significantly to the dose for the IDF. This paper will summarize the information obtained for the 2016 data package and the plans for further testing to refine key inputs for the IDF PA. (authors)

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DOE OSTI2016

Validation of the Modelling of Corrosion Driven Expansion in Grouted Waste-forms - 16125

One of the current preferred solutions for the disposal of Intermediate Level radioactive waste in the UK is grout encapsulation inside steel boxes for long term storage. These boxes are then to be entombed in a Geological Disposal Facility once one is available. However, both prior to and during geological disposal, waste contents can react with the water present in the encapsulant grout and begin to corrode. This corrosion product occupies a larger volume (by several hundred percent) than the original material and the corrosion occurs over the course of decades and centuries, long after the encapsulant grout has set and hardened. The expanding waste then creates stresses in the encapsulant causing both it and the surrounding box to deform and, in extreme circumstances, and after long times, for the structural integrity of the box to fail. This behaviour needs to be quantified before the box can be approved for use as any excessive deformation could lead to a loss of containment or, since the boxes are stacked on top of each other, to a significant accident in the stores. Due to the long timescales involved, physical testing of the phenomenon in full scale is impossible. Instead, a combination of chemical modelling and finite element analysis is used to predict the extent of the corrosion in the waste and its effect of the integrity of the box. There are a number of difficulties in modelling the corrosion-driven expansion and subsequent deformation of the boxes. Although the corrosion occurs over a very long time scale, the event is highly non-linear and dynamic. All of the material models, including the steel and grout are non-linear. The deformation is in the order of 10s of centimetres, well beyond the static elastic range. Finally, the components must be allowed to move relative to each other which involve a lot of part to part contacts. This means that simple static modelling, which would solve quite quickly, cannot be used. Instead a dynamic analysis code, LS-DYNA, has been used to expand the product over time to simulate the corrosion-driven expansion. This software is generally used for modelling car crashes or scenarios where waste containers have been dropped. The analyses therefore include a significant inertial component. This paper will detail the NNL's approach to modelling this long-term, quasi-static phenomenon in the dynamic, impact oriented software code LS-DYNA. It will cover the problems inherent in this approach, the errors that it creates as well as the tests and solutions that NNL worked on in order to overcome them and be confident that the solution could be validated. This work will ultimately be used to determine the maximum inventory which can be stored in each box based on the waste content and the potential extent of expansion as well as any modification that could be made to the box design or the processing streams that may improve the storage solution. (authors)

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DOE OSTI2016

Spent Fuel Reprocessing Development and Usage of Fission Products as Starting Precursor for Ceramic Waste-Forms - 16498

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile isotopes and waste preparing for disposal. Reprocessing is a series of chemical operations that separates plutonium and uranium from other constituents in nuclear waste. Indeed, the plutonium separated can be used to fuel reactors, but also to make nuclear weapons. Since reprocessing operations are far more expensive than using uranium fuel and disposing of the spent fuel directly, this process is not readily considered. Without doubts the reprocessing price of $1000 per kilogram of heavy metal (kgHM), like uranium must be less by at least 2-3 times compared to directly deposing the spent fuel. And some more simple technology (in comparison with PUREX process) must be developed, for example, technology based on precipitation processes. One of the options to change reprocessing technology is the idea to use certain fission products occurring in spent fuel in considerable amount and capable to form insoluble substances as suitable materials for long-term storage or final disposal of radioactive wastes. The development of new materials for HLW immobilization are intended to reduce disposal cost. The present work was aimed at the search for the simple engineering decisions that would provide opportunity of fission products usage as starting precursor for ceramic waste-forms. The results of experiments on synthesis of Pd-based metallo-ceramic compositions are presented in this paper. (authors)

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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.

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DOE OSTI2015

Development of Tc(IV)-Incorporated Fe Minerals to Enhance 99Tc Retention in Glass Waste Form

Iron minerals have been considered to be good hosts for Tc immobilization because the Tc(IV) ion substitutes for Fe(III) in the crystal structure of the Fe oxide due to similarities in (1) cation size [Tc(IV) = 78.5 pm ; Fe(III) = 69 or 78.5 pm], (2) metal-oxygen interatomic distance (Tc—O = 0.199 nm, Fe—O = 0.203 nm), (3) number of coordinating oxygen atoms (both 6-fold coordinated), and (4) the redox potential (Eh=ca. +20 mV at pH = 7) for a redox couple between Tc(VII)/Tc(IV) and Fe(III)/Fe(II). Magnetite, maghemite, and trevorite are iron oxide minerals and all belong to spinel mineral group. Laboratory testing shows that Tc can be removed from aqueous waste solutions by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation with iron oxide minerals during recrystallization of Fe(OH)2(s) used as an initial solid precursor. X-ray absorption near edge structure (XANES) spectroscopy confirmed that Tc was in the +4 oxidation state in final Tc-Fe minerals. The Tc-incorporated Fe minerals were also tested for Tc retention in glass melts at different temperatures between 600 – 1,000 oC in a furnace. After being cooled in air, the solid glass specimens collected at different temperatures were analyzed for Tc oxidation state using XANES and Tc retention using liquid scintillation counting (LSC). Even though Tc(IV) started to reoxidize at 600 oC, Tc retention in the final glass specimen prepared with Tc-incorporated Fe mineral even at high temperatures was at least two times higher than glass prepared with KTcO4 salt. Higher Tc retention in glass is considered to result from limited and delayed Tc volatilization process due to Fe mineral encapsulation for Tc. Therefore, the results showing the presence of Tc(IV) in the Fe mineral structure indicate strong possibility to enhance Tc retention in borosilicate glass as well as to reduce the remediation costs at the Hanford Site.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2015

Development of Tc(IV)-incorporated Fe minerals to enhance {sup 99}Tc retention in glass waste form - 15239

Iron minerals have been considered to be good hosts for Tc immobilization because the Tc(IV) ion substitutes for Fe(III) in the crystal structure of the Fe oxide due to similarities in (1) cation size [Tc(IV) = 78.5 pm; Fe(III) = 69 or 78.5 pm], (2) metal-oxygen interatomic distance (Tc-O = 0.199 nm, Fe-O = 0.203 nm), (3) number of coordinating oxygen atoms (both 6-fold coordinated), and (4) the redox potential (Eh=ca. +20 mV at pH = 7) for a redox couple between Tc(VII)/Tc(IV) and Fe(III)/Fe(II). Magnetite, maghemite, and trevorite are iron oxide minerals and all belong to spinel mineral group. Laboratory testing shows that Tc can be removed from aqueous waste solutions by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation with iron oxide minerals during recrystallization of Fe(OH){sub 2}(s) used as an initial solid precursor. X-ray absorption near edge structure (XANES) spectroscopy confirmed that Tc was in the +4 oxidation state in final Tc-Fe minerals. The Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were also tested for Tc retention in glass melts at different temperatures between 600 - 1,000 deg.C in a furnace. After being cooled in air, the solid glass specimens collected at different temperatures were analyzed for Tc oxidation state using XANES and Tc retention using liquid scintillation counting (LSC). Although Tc(IV) started to re-oxidize at 600 deg.C, Tc retention in the final glass specimen prepared with Tc-incorporated Fe mineral at high temperatures was at least two times higher than glass prepared with KTcO{sub 4} salt. Higher Tc retention in glass is considered to result from limited and delayed Tc volatilization process due to Fe mineral encapsulation for Tc. Therefore, the results showing the presence of Tc(IV) in the Fe mineral structure indicate strong possibility to enhance Tc retention in borosilicate glass as well as to reduce the remediation costs at the Hanford Site. (authors)

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DOE OSTI2016

Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. Erratum

Please insert erratum page in report PNNL-14805. This erratum serves to correct the K g (Glass apparent equilibrium constant based on the SiO 2 (aq) activity) values affected by a calculation error for glasses LAWA44, LAWB45 and LAWC22. The error was first identified by Papathanassiu et al. (2011). The K g values were corrected by Freedman et al. (2015) by multiplying the erroneous K g values by approx. 3.4. As indicated in Freedman et al. (2015), this 3.4 factor originates from the erroneous use of the molecular weight of H 4 SiO 4 (96.114 g·mol –1 ) instead of the molecular weight for Si (28.086 g·mol –1 ).

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