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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
NASA NTRS1978

Glass matrix composites. I - Graphite fiber reinforced glass

An experimental program is described in which graphite fibers of Hercules HMS and HTS, Thornel 300, and Celanese DG-12 were used to reinforce, both uniaxially and biaxially, borosilicate pyrex glass. Composite flexural strength distribution, strength as a function of test temperature, fracture toughness and oxidative stability were determined and shown to be primarily a function of fiber type and the quality of fiber-matrix bond formed during composite fabrication. It is demonstrated that the graphite fiber reinforced glass system offers unique possibilities as a high performance structural material.

Prewo, K. M.↗
DOE OSTI2022

Microstructural examination of interactions between chromia-based refractory and nuclear glass in a melter

This study shows the effects of nuclear waste glass production on Monofrax K-3 refractory corrosion. A continuously fed research-scale melter containing an Fe- and Ni-rich simulated nuclear waste feed with borosilicate glass-forming chemicals was cyclically melted at 1150°C and idled at 1050°C for a total of 11 weeks. Chemical maps using scanning electron microscopy show the interactions between the quenched melt and the refractory. Nanoscale X-ray-computed tomography was used for a three-dimensional visualization of certain parts of the interface. Unreacted K-3 consists of primarily corundum (Al,Cr) 2 O 3 and spinel (Fe 2+ ,Mg)(Al,Cr) 2 O 4 interlocking crystalline phases. Some of the Cr from the refractory interacts with the Ni and Fe from the melt to form a reaction layer comprising (Ni,Fe 2+ )(Cr,Fe 3+ ) 2 O 4 spinel crystals. Simultaneously, melt components (Na,Si) infiltrate into the refractory. This interaction proceeds at the expense of the integrity of the refractory structure. Intact refractory grains (e.g., (Al,Cr) 2 O 3 ) as well as the reaction layer itself can lose mechanical integrity and spall off into the melt, especially near the top of the melter. As the reaction layer can be a protective boundary for the refractory against further melt infiltration, a reduction in the reaction layer thickness allows an increase in refractory corrosion.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Hydrothermal Alteration of Glass from Underground Nuclear Tests: Formation and Transport of Pu-clay Colloids at the Nevada National Security Site

The testing of nuclear weapons at the Nevada National Security Site (NNSS), formerly the Nevada Test Site (NTS), has led to the deposition of substantial quantities of plutonium into the environment. Approximately 2.8 metric tons (3.1×10 4 TBq) of Pu were deposited in the NNSS subsurface as a result of underground nuclear testing. While 3 H is the most abundant anthropogenic radionuclide deposited in the NNSS subsurface (4.7×10 6 TBq), plutonium is the most abundant from a molar standpoint. The only radioactive elements in greater molar abundance are the naturally occurring K, Th, and U isotopes. 239 Pu and 240 Pu represent the majority of alpha-emitting Pu isotopes. The extreme temperatures associated with underground nuclear tests and the refractory nature of Pu results in most of the Pu (98%) being sequestered in melted rock, referred to as nuclear melt glass (Iaea, 1998). As a result, Pu release to groundwater is controlled, in large part, by the leaching (or dissolution) of nuclear melt glass over time. The factors affecting glass dissolution rates have been studied extensively. The dissolution of Pu-containing borosilicate nuclear waste glasses at 90ºC has been shown to lead to the formation of dioctahedral smectite colloids. Colloid-facilitated transport of Pu at the NNSS has been observed. Recent groundwater samples collected from a number of contaminated wells have yielded a wide range of Pu concentrations from 0.00022 to 2.0 Bq/L. While Pu concentrations tend to fall below the Maximum Contaminant Level (MCL) established by the Environmental Protection Agency (EPA) for drinking water (0.56 Bq/L), we do not yet understand what factors limit the Pu concentration or its transport behavior. To quantify the upper limit of Pu concentrations produced as a result of melt glass dissolution and determine the nature of colloids and Pu associations, we performed a 3 year nuclear melt glass dissolution experiment across a range of temperatures (25-200 °C) that represent hydrothermal conditions representative of the underground nuclear test cavities (when groundwater has re-saturated the nuclear melt glass and glass dissolution occurs). Colloid loads and Pu concentrations were monitored along with the mineralogy of both the colloids and the secondary mineral phases. The intent was to establish an upper limit for Pu concentrations at the NNSS, provide context regarding the Pu concentrations observed at the NNSS to date and the Pu concentrations that may be observed in the future. The results provide a conceptual model for the risks posed by Pu migration at the NNSS.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Study of lithium borosilicotitanate glasses with LiCl

The effect of LiCl addition on the properties of Li{sub 2}O: B{sub 2}O{sub 3}: SiO{sub 2}: TiO{sub 2} glasses has been studied. It has been observed that the ionic conductivity increases by about half an order with LiCl addition. The decrease in glass transition temperature T{sub g} and increase in the molar volume with LiCl addition have good correlation with conductivity results. The observed increase in density of glasses has been explained on the basis of heavier Cl{sup -} ion which is accommodated in the interstices in the glass network.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Predicting initial dissolution rates using structural features from molecular dynamics simulations

Predicting chemical durability of glass materials is important for various applications from daily life such as drink glass and kitchen ware to advanced technologies such as nuclear waste disposal and biomedicine. In this work, we explored prediction of initial dissolution rate through structural features from molecular dynamics (MD) simulations for a wide range of glass compositions (total 28) including borosilicate and aluminosilicate glasses, ZrO 2 -containing and V 2 O 5 -containing boroaluminosilicate glasses. The initial dissolution rates (r 0 ) measured experimentally at 90 °C with varying solution conditions were correlated with structural features (e.g., polyhedral linkages and non-bridging oxygen species) obtained from MD simulations, either from this study or from literature. Since hydrolysis of the glass network through breaking of the network former linkages (e.g., Si-O-Si, Si-O-Al, etc.) is a critical step of network glass dissolution, the statistics of these linkages obtained from MD were correlated to r0 through linear regression, where the coefficient of determination (R 2 ) and root mean square error are found to be 0.949 and 0.681, respectively. This model was compared and discussed with existing models developed by various approaches including machine learning, the kinetic rate equation, topological constraint theory, and other descriptors from MD simulations. The discussion provides insights on future model improvements to predict glass dissolution. In addition, the impact of V 2 O 5 on the glass dissolution was examined in detail, implicating that the impact is not the same across all glass compositions and test conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Evaluation of GlassNet for physics-informed machine learning of glass stability and glass-forming ability

Glassy materials form the basis of many modern applications, including nuclear waste immobilization, touch-screen displays, and optical fibers, and also hold great potential for future medical and environmental applications. However, their structural complexity and large composition space make design and optimization challenging for certain applications. Of particular importance for glass processing and design is an estimate of a given composition's glass-forming ability (GFA). However, there remain many open questions regarding the underlying physical mechanisms of glass formation, especially in oxide glasses. It is apparent that a proxy for GFA would be highly useful in glass processing and design, but identifying such a surrogate property has proven itself to be difficult. While glass stability (GS) parameters have historically been used as a GFA surrogate, recent research has demonstrated that most of these parameters are not accurate predictors of the GFA of oxide glasses. Here, in this work, we explore the application of an open-source pre-trained neural network model, GlassNet, that can predict the characteristic temperatures necessary to compute GS with reasonable performance and assess the feasibility of using these physics-informed machine learning (PIML)-predicted GS parameters to estimate GFA. In doing so, we track the uncertainties at each step of the computation—from the original ML prediction errors to the compounding of errors during GS estimation, and finally to the final estimation of GFA. While GlassNet exhibits reasonable accuracy on all individual properties, we observe a large compounding of error in the combination of these individual predictions for the PIML prediction of GS, finding that random forest models offer similar accuracy to GlassNet. We also break down the performance of GlassNet on different glass families and find that the error in GS prediction is correlated with the error in crystallization peak temperature prediction. Lastly, we utilize this finding to assess the relationship between top-performing GS parameters and GFA for two ternary glass systems: sodium borosilicate and sodium iron phosphate glasses. We conclude that to obtain true ML predictive capability of GFA, significantly more data needs to be collected.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change. The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.

Riley, Brian J. (ORCID:0000000277456730)↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Intrinsic dosimetry of glass containers: A potential interrogation tool for nuclear forensics and waste management

Intrinsic dosimetry is the method of measuring total absorbed dose received by the walls of a container holding radioactive material. By considering the total absorbed dose received by a container in tandem with the physical characteristics of the radioactive material housed within that container, this method has the potential to provide law enforcement officials or waste management personnel with enhanced pathway information regarding the history of the container and its radioactive contents. We report the latest in a series of experiments designed to validate and demonstrate this newly developed tool. Thermoluminescence (TL) dosimetry was used to measure dose effects on the raw stock material of borosilicate container glass with time up to 60 days after gamma ray, x-ray, beta particle or ultraviolet irradiations at doses from 0.15 to 20 Gy. Two peaks were identified in the TL glow curve when irradiated with 60Co, a relatively unstable peak around 120°C and a more stable peak around 225°C. Signal strength of both peaks decayed with time. The minimum measurable dose using this technique was measured to be 0.15 Gy, which is roughly equivalent to a 24 hr irradiation at 1 cm from a 50 ng source of 60Co. After fading, this dose would still be observable approximately 1 year post-irradiation. Similar doses with beta particles and x-rays produced TL glow curves of lower intensity. The TL glow curves were separated into five peaks centered at approximately 120, 160, 225, 300, and 340°C. Differences in TL glow curve shape and intensity were observed for the glasses from different geographical origins. These differences can be explained by changes in the intensities of the five peaks. EPR results showed production of BOHC and E’ silica centers.

Clark, Richard A.↗
DOE OSTI2012

Control of growth and the processes of energy transfer from CdSe quantum dots for Nd{sup 3+} ions in a vitreous system: Thermal annealing time

The authors report clear evidence of radiative and nonradiative energy transfer from CdSe quantum dots to Nd{sup 3+} ions in a glass system synthesized by melting method. An efficient control of energy transfer can be obtained by controlling the mean radius of the CdSe quantum dots. Increase of about 100% of the near infrared Nd{sup 3+} emission ({sup 4}F{sub 3/2}{yields}{sup 4}I{sub 9/2}) was observed as a function of thermal annealing time.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2014

Toward Understanding the Effect of Low-Activity Waste Glass Composition on Sulfur Solubility

The concentration of sulfur in nuclear waste glass melter feed must be maintained below the point where salt accumulates on the melt surface. The allowable concentrations may range from 0.37 to over 2.05 weight percent (of SO3 on a calcined oxide basis). If the amount of sulfur exceeds its tolerance level a molten salt will accumulate and upset melter operations and potentially shorten melter useful life. Therefore relatively conservative limits have been placed on sulfur loading in melter feed which in-turn significantly impacts the amount of glass that will be produced, in particular at the Hanford site. Crucible-scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. A model was developed to predict the solubility of SO3 in glass based on 312 individual glass compositions. This model was shown to well represent the data, accounting for over 80% of the variation in data and was well validated. The model was also found to accurately predict the tolerance for sulfur in melter feed based on 19 scaled melter tests. The model is appropriate for control of waste glass processing which includes uncertainty quantification. The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that most increase sulfur solubility are Li2O > V2O5 ≈ TiO2 < CaO < P2O5 ≈ ZnO. The components that most decrease sulfur solubility are Cl > Cr2O3 > SiO2 ≈ ZrO2 > Al2O3.

Borosilicate↗
DOE OSTI2012

Lanthanum and neodymium solubility in simplified SiO{sub 2}-B{sub 2}O{sub 3}-Na{sub 2}O-Al{sub 2}O{sub 3}-CaO high level waste glass

Lanthanum and neodymium incorporation in simplified high level waste glasses has been investigated for SiO{sub 2}B{sub 2}O{sub 3}Na{sub 2}OAl{sub 2}O{sub 3}CaO compositions quenched from 1200 degrees C, for varying La /(La + Nd) (atomic) and increasing rare-earth oxides contents. In this system and beyond the solubility limit, rare-earths (RE) elements are reported to form apatite phases with the general formula Ca{sub 2}RE{sub 8}(SiO{sub 4}){sub 6}O{sub 2}. In the current study, speciation of these trivalent RE{sup 3+} cations in both amorphous network and crystal phases was determined from X-ray diffraction, scanning electron microscopy, optical absorption at 10 K, Raman spectroscopy, and electron probe microanalysis. It appeared that RE{sub 2}O{sub 3} solubility was higher for La -rich formulations than for Nd -rich ones and that an increase in the RE oxide content reduces the connectivity of the network building units through formation of non-bridging oxygens at the expense of the oxygen bridges. This depolymerization of the glass network did not affect neodymium environment which consisted in silicate tetrahedra. The composition of the apatite crystals was found to be affected by the La /(La + Nd) of the parent glass and deviation from the ideal composition (Ca{sub 2}RE{sub 8}(SiO{sub 4}){sub 6}O{sub 2}) occurred in the neodymium end of the system. It thus appears that both RE{sub 2}O{sub 3} solubility and crystal composition are strongly dependent on the type and crystal chemistry of the RE elements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Single-component-at-a-time variation study for glass-ceramic waste forms

Here, a 51-sample composition variation study was performed on glass-ceramic waste forms for a raffinate waste stream from aqueous reprocessing of used nuclear fuel containing high fractions of Mo, alkalis, alkaline earths (AEs), and rare earths (REs). The study was designed with a single-component-at-a-time variation approach off a centroid composition. The components that were varied included Al, B, Ca, Li, Mo, Na, REs, Si, Zr, and Others (containing minor components). Data analysis included crystallization curves, microstructure, and phase compositions. A number of components (i.e., Li 2 O, B 2 O 3 , REO x , MoO 3 , Na 2 O, and ZrO 2 ) significantly impacted the concentration and chemistry of phases, especially the primary phases of oxyapatite [i.e., Ca 2 RE 8 (SiO 4 ) 6 O 2 ] and powellite (i.e., AEMoO 4 ), precipitated in the slow-cool heat-treated waste forms; minor phases included cerianite [i.e., Ce x Zr (1-x) O 2 ], Ba-molybdate [i.e., Ba(Gd 0.67 Mo 0.33 O 3 ], noble metals, pollucite (i.e., CsAlSiO 4 ), and RE-borosilicate (i.e., RE 3 BSi 2 O 10 ).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Volatile species of technetium and rhenium during waste vitrification

Volatile loss of technetium (Tc) during vitrification of low-activity wastes is a technical challenge for treating and immobilizing the large volumes of radioactive and hazardous wastes stored at the U.S. Department of Energy's Hanford Site. There are various research efforts being pursued to develop technologies that can be implemented for cost effective management of Tc, including studies to understand the behavior of Tc during vitrification, with the goal of eventually increasing Tc retention in glass. Furthermore, one of these studies has focused on identifying the form or species of Tc and Re (surrogate for Tc) that evolve during the waste-to-glass conversion process. This information is important for understanding the mechanism of Tc volatilization. In this paper, available information collected from the literature is critically evaluated to clarify the volatile species of Tc and Re and, more specifically, whether they volatilize as alkali pertechnetate and perrhenate or as technetium and rhenium oxides after decomposition of alkali pertechnetate and perrhenate. The evaluated data ranged from mass spectrometric identification of species volatilized from pure and binary alkali pertechnetate and perrhenate salts to structural and chemical analyses of volatilized materials during crucible melting and scaled melter processing of simulated wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Micro-fluidics and integrated optics glass sensor for in-line micro-probing of nuclear samples

We study the miniaturization of Thermal Lens Spectrometry (TLS) towards Lab-on-chip integration in order to reduce the volume of fluid assays in nuclear process control. TLS is of great interest in this context since it combines the advantages of optical detection methods with an inherent suitability for small-scale samples. After validating the experimental principle in a classical thermal lens crossed-beam setup, we show the integration of a Young-interferometer with a microcapillary on a glass substrate, reducing the necessary sample size to 400 nl. The interferometer translates the photo-thermally induced refractive index change in the fluid to a phase shift of the fringe pattern, which can then be detected by a camera. Measurements of Co(II) in ethanol yield a detection limit of c = 5 x 10{sup -4} M for the crossed-beam setup and c = 6 x 10{sup -3} M for the integrated sensor. At an interaction length of 10 m, it detects a minimum absorbance of K = 1.2 x 10{sup -4} in a probed volume of 14 pl. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗