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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Facile incorporation of technetium into magnetite, magnesioferrite, and hematite by formation of ferrous nitrate in situ : precursors to iron oxide nuclear waste forms

The fission product, 99 Tc, presents significant challenges to the long-term disposal of nuclear waste due to its long half-life, high fission yield, and to the environmental mobility of pertechnetate (TcO 4 – ), the stable Tc species in aerobic environments. Here, the migration of 99 Tc from disposal sites can potentially be prevented by incorporating it into durable waste forms based on environmentally stable minerals.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Novel Zoned Waste-forms for High-Priority Radionuclide Waste Streams

This report describes the potential of a novel class of materials—α-ZrW 2 O 8 , Zr 2 WP 2 O 12 , and related compounds that contract upon amorphization as possible radionuclide waste-forms. The proposed ceramic waste-forms would consist of zoned grains, or sintered ceramics with center- loaded radionuclides and barren shells. Radiation-induced amorphization would result in core shrinkage but would not fracture the shells or overgrowths, maintaining isolation of the radionuclide. In this report, we have described synthesis techniques to produce phase-pure forms of the materials, and how to fully densify those materials. Structural models for the materials were developed and validated using DFPT approaches, and radionuclide substitution was evaluated; U(IV), Pu(IV), Tc(IV) and Tc(VII) all readily substitute into the material structures. MD modeling indicated that strain associated with radiation-induced amorphization would not affect the integrity of surrounding crystalline materials, and these results were validated via ion beam experimental studies. Finally, we have evaluated the leach rates of the barren materials, as determined by batch and flow-through reactor experiments. ZrW 2 O 8 leaches rapidly, releasing tungstate while Zr is retained as a solid oxide or hydroxide. Tungsten release rates remain elevated over time and are highly sensitive to contact times, suggesting that this material will not be an effective waste-form. Conversely, tungsten releases rates from Zr 2 WP 2 O 12 rapidly drop, show little dependence on short-term changes in fluid contact time, and in over time, become tied to P release rates. The results presented here suggest that this material may be a viable waste-form for some hard-to-handle radionuclides such as Pu and Tc.

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DOE OSTI2017

Aqueous Synthesis of Technetium-Doped Titanium Dioxide by Direct Oxidation of Titanium Powder, a Precursor for Ceramic Nuclear Waste Forms

Technetium-99 (Tc) is a problematic fission product that complicates the long-term disposal of nuclear waste due to its long half-life, high fission yield, and the environmental mobility of pertechnetate, its stable form in aerobic environments. One approach to preventing Tc contamination is through incorporation into durable waste forms based on weathering-resistant minerals such as rutile (titanium dioxide). Here, the incorporation of technetium into titanium dioxide by means of simple, aqueous chemistry - direct oxidation of titanium powder in the presence of ammonium fluoride - is achieved. X-ray absorption fine structure spectroscopy and diffuse reflectance spectroscopy indicate that Tc(IV) replaces Ti(IV) within the structure. Rather than being incorporated as isolated Tc(IV) ions, Tc is present as pairs of edge-sharing Tc(IV) octahedra similar to molecular Tc(IV) complexes such as [(H 2 EDTA)TcIV](μ-O) 2 . Technetium-doped TiO 2 was suspended in deionized water under aerobic conditions, and the Tc leached under these conditions was followed for 8 months. The normalized release rate of Tc (LR Tc ) from the TiO 2 particles is low (3 × 10 -6 g m -2 d -1 ), which illustrates the potential utility of TiO 2 as waste form. However, the small size of the as-prepared TiO 2 nanoparticles results in an estimated retention of Tc of 10 4 years, which is only a fraction of the half-life of Tc (2.1 × 10 5 years).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Effects of composition and canister centerline cooling on microstructure, phase distribution, and chemical durability of dehalogenated iron phosphate waste forms

This paper discusses the effects of composition and canister centerline cooling (CCC) on the microstructures, phase distribution, and chemical durability of dehalogenated iron phosphate waste forms starting from chloride-based salt simulants including KCl (Simple-1), NdCl3 (Simple-2), SrCl2 (Simple-3), LaCl3 (Simple-4), LiCl (Simple-5), CsCl (Simple-6), as well as a complex simulant containing LiCl, KCl, NaCl, CsI, SrCl2, CeCl3, and NdCl3 called ERV2. The simulants represent salt constituents present in salt wastes from electrorefiners used for electrochemical reprocessing. These salts are reacted with ammonium dihydrogen phosphate (ADP) resulting in a NH4Cl (dechlorination) byproduct that is captured and phosphate product containing oxides of the salt cations. This product is further stabilized using Fe2O3 through vitrification at higher temperatures. The results presented herein describe the properties of CCC-treated materials following dechlorination and vitrification including phase identification and quantification, elemental distributions, and chemical durabilities. The chemical durability in the CCC-cooled samples are lower durability than quenched materials for the ERV2 samples. The phases formed upon CCC cooling are very complex compositionally and microstructurally.

iron phosphate glass, iron phosphate waste forms, ↗
DOE OSTI2017

Enhanced 99 Tc retention in glass waste form using Tc(IV)-incorporated Fe minerals

Technetium (99Tc) immobilization by doping into iron oxide mineral phases may alleviate the problems with Tc volatility during vitrification of nuclear waste. Reduced Tc, Tc(IV), substitutes for Fe(III) in the crystal structure by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation of Fe oxide minerals. Two Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were prepared and tested for Tc retention in glass melt samples at temperatures between 600 – 1,000 oC. After being cooled, the solid glass specimens prepared at different temperatures were analyzed for Tc oxidation state using Tc K-edge XANES. In most samples, Tc was partially oxidized from Tc(IV) to Tc(VII) as the melt temperature increased. However, Tc retention in glass melt samples prepared using Tc-incorporated Fe minerals were moderately higher than in glass prepared using KTcO4 because of limited and delayed Tc volatilization.

Um, Wooyong↗
DOE OSTI2017

Getters for improved technetium containment in cementitious waste forms

A cementitious waste form, Cast Stone, is a possible candidate technology for the immobilization of low activity nuclear waste (LAW) at the Hanford site. This paper focuses on the addition of getter materials to Cast Stone that can sequester Tc from the LAW, and in turn, lower Tc release from the Cast Stone. Two getters which produce different products upon sequestering Tc from LAW were tested: Sn(II) apatite (Sn-A) that removes Tc as a Tc(IV)-oxide and potassium metal sulfide (KMS-2) that removes Tc as a Tc(IV)-sulfide species, allowing for a comparison of stability of the form of Tc upon entering the waste form. The Cast Stone with KMS-2 getter had the best performance with addition equivalent to ~0.08 wt% of the total waste form mass. The observed diffusion (D obs ) of Tc decreased from 4.6 ± 0.2 × 10 -12 cm 2 /s for Cast Stone that did not contain a getter to 5.4 ± 0.4 × 10 -13 cm 2 /s for KMS-2 containing Cast Stone. Finally, it was found that Tc-sulfide species are more stable against re-oxidation within getter containing Cast Stone compared with Tc-oxide and is the origin of the decrease in Tc D obs when using the KMS-2.

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DOE OSTI2023

Aluminophosphate Waste Forms for Immobilizing Cations from Electrochemical Salt Wastes

This report provides experimental details and results of evaluating aluminophosphate waste forms for treating and immobilizing the salt cations from salt wastes generated during electrochemical reprocessing of used nuclear fuel. In the waste form process for these materials, chloride salt streams are reacted with NH 4 H 2 PO 4 , the chlorine is removed from the salts and driven off as NH 4 Cl (a solid condensate that can be captured), and then the product can be vitrified in conjunction with glass-forming chemicals (e.g., Fe 2 O 3 , Al 2 O 3 ) to create a high-durability waste form. This study was initiated with some literature review on aluminophosphates containing high alkali oxide content and some of this information is summarized in this report. Following literature review, three new samples were synthesized where two contained Fe 2 O 3 +Al 2 O 3 (i.e., samples G3 and G5 ) and one was only Al 2 O 3 (Fe 2 O 3 -free) (i.e., sample G6 ). In addition to these samples, G1 was also made, which is the baseline reference waste form referred to as DPF5-336 (made without Al 2 O 3 ). Samples G1, G3, G5, and G6 had phases of Li 3 Fe 2 (PO 4 ) 3 (likely), monazite (below XRD detection limits), AlPO 4 , and AlPO 4 , respectively. Characterizations on these materials included optical images, scanning electron microscopy, energy dispersive X-ray spectroscopy, and X-ray diffraction. These samples were shipped to Argonne National Laboratory for chemical durability testing. Depending on how the samples perform in these tests, an additional phase of aluminophosphate formulations could be designed and tested. This report completes the milestone M4FT-23PN030104041 with details provided in Appendix B.

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DOE OSTI2015

Test plan for formulation and evaluation of grouted waste forms with shine process wastes

The objective of this experimental project is to demonstrate that waste streams generated during the production of Mo99 by the SHINE Medical Technologies (SHINE) process can be immobilized in cement-based grouted waste forms having physical, chemical, and radiological stabilities that meet regulatory requirements for handling, storage, transport, and disposal.

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DOE OSTI2025

FY24 Development of Improved Grout Waste Forms for Alternative Low Activity Waste Treatment

Since the WTP Low Activity Waste (LAW) Vitrification Facility was not designed to process the entire inventory of Hanford LAW, up to half of the retrieved Hanford LAW will require supplemental immobilization. Immobilizing LAW in a cementitious waste form known as Cast Stone has been investigated as a possible candidate supplemental immobilization technology. In FY21, Washington River Protection Solutions, LLC (WRPS) tasked Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA) to perform testing to evaluate methods for reducing the release of COCs, particularly nitrate, 99Tc, and 129I, from cementitious waste forms made from aqueous LAW derived from Hanford Tank Waste. FY22 work built on the FY21 results and further developed formulations while targeting higher waste loadings. The objective of this work was to perform laboratory-scale testing to further refine the most promising formulation(s) that were identified in the FY23 work. The goal of the refinement was to further reduce the release rates for 99Tc, Cr, 129I, and nitrate while maintaining workability of the fresh grout, and to increase waste loading.

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DOE OSTI2022

Impact of carbonation on leaching of constituents from a cementitious waste form for treatment of low activity waste at the DOE Hanford site

Carbonation can be a major aging process during disposal of alkaline cementitious waste forms and can impact constituent leaching by changes in material alkalinity, pore structure, and controlling mineral phases. The effect of carbonation on the leaching of major and trace constituents from Cast Stone, a cementitious waste form developed to treat high salt content low activity waste, was studied through a combination of leaching experiments and reactive transport simulations. Diffusive transport of constituents in the waste form was evaluated using reactive transport modeling of diffusion-controlled leaching test results and a geochemical speciation model derived from pH-dependent leaching. Comparisons between Cast Stone materials aged under nitrogen, air, and 2% carbon dioxide in nitrogen showed that carbonation impacts solubility, physical retention and observed diffusivity of major and trace constituents. Carbonation under 2% CO 2 decreased the diffusion-controlled leaching of chromium by two orders of magnitude. Finally, modeling results suggest that carbonation may also decrease solubility of technetium while changes to microstructure by carbonation increases effective diffusivity of constituents in Cast Stone.

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DOE OSTI2017

The Effect of Ettringite on Cementitious Waste Form Development for Low- Activity Waste - 17170

The current Cast Stone formulation, with 8% ordinary Portland cement (OPC), 45% fly ash (FA), and 47% blast furnace slag (BFS), proposed for cementitious waste forms (CWFs) at the Hanford site has been shown to be inadequate for secondary liquid wastes sent to the Effluent Treatment Facility (ETF). This is due to the high sulfate concentrations and non-alkaline conditions present in the ETF waste stream that lead to the late formation of ettringite, [Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26(H{sub 2}O)], which can consequently cause undesirable swelling and cracks in the CWF that facilitate {sup 99}Tc release. To overcome this challenge, hydrated lime (HL), Ca(OH){sub 2}, can be used instead of FA to provide the Ca necessary to sequester sulfate early in the cast stone curing process through early formation of ettringite. Early ettringite formation minimizes the risk of localized disruptive action or harmful stress that lead to undesirable swelling and cracking of the CWF. Furthermore, early formation reduces the permeability of the CWF, further reducing the risk of {sup 99}Tc release and increasing structural stability. Two liquid waste simulants, 1) WTP off-gas condensate (WTP) and 2) Effluent Management Facility (EMF) process condensate combined with low activity waste (LAW) caustic scrubber (labeled EMF/CS), were prepared and tested for ettringite formation effects on HL-containing CWF development and {sup 99}Tc leachability. After curing for 28 days, X-ray diffraction analysis revealed that ettringite was formed in the early stages of curing for the CWFs prepared using both simulants. However, because of the higher sulfate concentration in the EMF/CS simulant and limited Ca and Al sources, some sulfate did not form ettringite during the early stages of curing, resulting in significant monolith swelling and cracking during subsequent leaching tests. Despite observable cracking in monoliths prepared with the EMF/CS simulant, low {sup 99}Tc release from HL-based waste forms was measured in later leaching periods tested by the EPA Method 1315 in both simulants. Ettringite-{sup 99}Tc co-precipitation batch experiments support a possible chemical sequestration mechanism that is hypothesized to contribute to the observed low {sup 99}Tc leach rates, which are likely the result of an additional chemical reaction that sequesters {sup 99}Tc rather than simply physical factors such as reduced porosity and permeability (or tortuosity). Overall these findings quantitatively support the use of HL to overcome the deleterious effects of sulfate-rich conditions on CWFs through early formation of ettringite and provide evidence for {sup 99}Tc release rates that fall within low risk levels reported by the United States Department of Energy (US DOE). (authors)

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DOE OSTI2013

Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

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DOE OSTI2017

Corrosion Behavior and Microstructure Influence of Glass-Ceramic Nuclear Waste Forms

Glass ceramic waste forms present a potentially viable technology for the long term immobilization and disposal of liquid nuclear wastes. Through control of chemistry during fabrication, such waste forms can have designed secondary crystalline phases within a borosilicate glass matrix. In this work, a glass ceramic containing powellite and oxyapatite secondary phases was tested for its corrosion properties in dilute conditions using single pass flow through testing (SPFT). Three glass ceramic samples were prepared using different cooling rates to produce samples with varying microstructure sizes. In testing at 90 °C in buffered pH 7 and pH 9 solutions, it was found that increasing pH and decreasing microstructure size (resulting from rapid cooling during fabrication) both led to a reduction in overall corrosion rate. The phases of the glass ceramic were found, using a combination of solutions analysis, SEM and AFM, to corrode preferably in the order of powellite > bulk glass matrix > oxyapatite.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2015

Impact of Zr Content on Radionuclide Release Rates in Metal Waste Forms

Reference alloy waste forms (RAW) underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates. The waste forms tested at SRNL were variations of the RAW-3 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Zr:U ratio on the release rates of actinides and Tc-99. The test solutions were air saturated alkaline and acidic brines, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements, 2) hybrid potentiostatic hold/exposure measurements, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.

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