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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors

The overall summary of the waste management envelope discussed in this report is represented by the diagram shown in Figure S1. The streams that are included in this discussion include the following from reactor systems in which chlorine-based salts with and without processing and fluorine-based salts with and without processing are being considered: 1) Off-gas streams (e.g., particulates, aerosols, reactive gases, 3 H, water, residual halides, N 2 , O 2 , noble gases); 2) Salt waste streams (e.g., chloride- or fluoride-based); 3) Separated salt streams (e.g., 7 Li recycle, 37 Cl recycle, eutectic salts, fission products); 4) Metal waste streams (e.g., reactor components, plenums); 5) Carbon waste streams (e.g., graphite, SiC); 6) Decommissioning and decontaminating (D&D) waste streams; 7) Operating waste streams

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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.

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DOE OSTI2019

Evaluation of the current state of knowledge for thermolysis of organics within SRS waste forming Volatile Organic Compounds (VOCs)

The 2017 report Flammable Gas Generation Mechanisms for High Level Liquid Waste Facilities, X-ESRG- 00062 Rev. 1, assessing mechanisms for flammable gas generation acknowledged the presence of flammable organics, concluding that available data indicated their concentrations pose a minor contribution (e.g., <5%) to the composite lower flammable limit. Savannah River Remediation (SRR) chartered the Savannah River National Laboratory (SRNL) to perform a more detailed assessment of available and emerging data to assess the current state of knowledge for formation of Volatile Organic Compounds (VOCs) from thermolysis of organics in Savannah River Site (SRS) caustic tank waste. Based on the assessment as detailed herein, SRNL concludes that sufficient evidence exists to demonstrate formation of methane, in addition to hydrogen, does occur during thermolysis of SRS high level waste containing organics under select conditions. This assessment cannot reliably quantify the amount of methane that forms. Nevertheless, the potential rates are sufficient – methane generation rates of approximately 30% and 100% those of hydrogen generation rates for Tank 38 radioactive waste and a high boiling point (HBP) simulant waste, respectively – to advise that ongoing program efforts be enhanced to include additional analysis for off-gas generation for concentration and screening measurements for other possible flammable species.

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DOE OSTI2014

Thermophysical Properties of Multiphase Borosilicate Glass-Ceramic Waste Forms

Multiphase borosilicate glass-ceramics represent one candidate to contain radioactive nuclear waste separated from used nuclear fuel. In this work, the thermophysical properties from room temperature to 1273 K were investigated for four different borosilicate glass-ceramic compositions containing waste loadings from 42 to 60 wt% to determine the sensitivity of these properties to waste loading, as-fabricated microstructure, and potential evolutions in microstructure brought about by temperature transients. The thermal expansion, specific heat capacity, thermal diffusivity, and thermal conductivity are presented. The impact of increasing waste loading is shown to have a small but measurable effect on the thermophysical properties between the four compositions, contrasted to a much greater impact observed when transitioning from predominantly crystalline to amorphous systems. Thermal cycling below 1273 K was not found to measurably impact the thermophysical properties of the compositions investigated here.

Glass Ceramic↗
DOE OSTI2016

Materials for Tc Capture to Increase Tc Retention in Glass Waste Form

99Technetium is a long-lived fission product found in the tank waste at the Hanford site in Washington State. In its heptavalent species, it is volatile at the temperatures used in Hanford Tank Waste Treatment and Immobilization Plant vitrification melters, and thus is challenging to incorporate into waste glass. In order to decrease volatility and thereby increase retention, technetium can be converted into more thermally stable species. Several mineral phases, such as spinel, are able to incorporate tetravalent technetium in a chemically durable and thermally stable lattice, and these hosts may promote the decreased volatility that is desired. In order to be usefully implemented, there must be a synthetic rout to these phases that is compatible with both technetium chemistry and current Hanford Tank Waste Treatment and Immobilization Plant design. Synthetic routes for spinel and other potential host phases are examined.

spinel↗
DOE OSTI2017

Sub-Grid Model Improving Shrinking Core Simulations of Redox-Sensitive Technetium Release From Slag Cement Waste Forms - 17295

Numerical simulation of Tc transport in reducing cementitious materials undergoing oxidation is challenging because the effective solid-solute partition coefficient (K{sub d}) varies greatly between the reduced and oxidized states, assuming the Tc mass loading far exceeds that required to reach solubility in the aqueous phase, as is typical in waste disposal. In a conventional finite-volume formulation, the local aqueous concentration increases orders of magnitude when a computational cell switches from a reducing to oxidizing condition, which is an artifact of numerical discretization. In principal such artifacts can be reduced to an insignificant level by refining the computational grid. However, this approach is not feasible in typical facility scale applications due to computer runtime and storage limitations. In this paper an alternative sub-grid Tc release model is derived to approximately capture redox processes occurring at scales smaller than the computational mesh. This alternative approach introduces minimal additional computational expense, while significantly reducing numerical artifacts in the form of Tc concentration / flux spikes. The method is illustrated for two simple geometries and the Savannah River Site Saltstone waste Disposal Facility. The sub-grid model is shown to significantly reduce artificial spikes in Tc-99 release, resulting in more physically realistic numerical simulations. (authors)

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DOE OSTI2017

Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

The U.S. Department of Energy (DOE) in DOE O 435.1 Chg. 1, Radioactive Waste Management, requires the preparation and maintenance of a composite analysis (CA). The primary purpose of the CA is to provide a reasonable expectation that the primary public dose limit is not likely to be exceeded by multiple source terms that may significantly interact with plumes originating at a low-level waste disposal facility. The CA is used to facilitate planning and land use decisions that help assure disposal facility authorization will not result in long-term compliance problems; or, to determine management alternatives, corrective actions or assessment needs, if potential problems are identified.

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DOE OSTI2018

Evaluation of the Current State of Knowledge for Thermolysis of Organics within SRS Waste Forming Volatile Organic Compounds (VOCs)

The 2017 report Flammable Gas Generation Mechanisms for High Level Liquid Waste Facilities , X-ESRG- 00062 Rev. 1, assessing mechanisms for flammable gas generation acknowledged the presence of flammable organics, concluding that available data indicated their concentrations pose a minor contribution (e.g., <5%) to the composite lower flammable limit. Savannah River Remediation (SRR) chartered the Savannah River National Laboratory (SRNL) to perform a more detailed assessment of available and emerging data to assess the current state of knowledge for formation of Volatile Organic Compounds (VOCs) from thermolysis of organics in Savannah River Site (SRS) caustic tank waste.

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DOE OSTI2019

Tailored Perovskite Waste Forms for Plutonium Trapping

Perovskite ceramics have been extensively studied as host matrixes for radionuclide entrapment for nuclear waste disposal. As an expansion of these investigations, cerium, neodymium, and plutonium were incorporated into a perovskite phase, ACu 3 FeTi 3 O 12 (A = Nd, Ce, Pu), using sol–gel methods under oxidizing and reducing atmospheres. The targeted materials contained varying levels of Ce 3+ and Nd 3+ on the A site, yielding potential compositions of Nd 1–x Ce x Cu 3 FeTi 3 O 12 (x = 0, 0.1, 0.2, 0.3, 0.4, 0.8). However, interrogation of these materials shows that the maximum Ce 3+ loading is achieved near x ≈ 0.2. A single composition with plutonium was targeted, Nd 0.9 Pu 0.1 Cu 3 FeTi 3 O 12 , in order to properly model more realistic loading levels for a repository-destined material. These compounds were characterized using powder X-ray diffraction with Rietveld refinements of the structures and by a variety of spectroscopic techniques. Finally, the data suggest that, in order to achieve Pu 3+ substitution onto the A sites in the Nd 0.9 Pu 0.1 Cu 3 FeTi 3 O 12 , a reducing atmosphere must be employed. Otherwise, the redox activity of plutonium results in substitution onto multiple sites in the material as well as the formation of secondary phases such as TiO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. - The sealed HIP container avoids the release of any radionuclides. - The outside of the waste package is not contaminated. - The HIP process time is shorter than the HVP process time. The isostatic press avoids anisotropic density distributions. - Simple filling of the HIP container has advantages over the filling of an axial die. (authors)

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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.

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DOE OSTI2013

Solution-based approaches for making high-density sodalite waste forms to immobilize spent electrochemical salts

Three different solution-based approaches were taken to make sodalite minerals as a host for a mixed salt simulating the waste generated during the electrochemical separations process of nuclear fuel reprocessing that contains alkali, alkaline earth, and lanthanide chlorides plus trace iodine and actinides. All of the approaches included an aqueous solution of mixed chlorides (simulated waste) but the other reactants varied: (1) Al(OH)3 + NaOH + CS, (2) NaAlO2 + CS, and (3) Al2Si2O7 + NaOH, (CS = colloidal silica). The products were dried, ground, pressed into pellets, fired (650–950 °C), and characterized. Both 5 and 10 mass% of a Si-Na-B glass binder were introduced at different stages in the process. Route (2) proved the most successful at producing high sodalite fractions (up to 100%) with minimal glass binder additions and showed high consolidation potential (up to 91.4% of theoretical density). Detailed comparisons are provided of the results.

electrochemical↗
DOE OSTI2018

Neutron Measurements from Boron Carbide Addition to a Plutonium Waste Form

The Department of Energy (DOE) has identified the Dilute and Dispose approach as the preferred method for the disposition of surplus plutonium materials. Dilute and Dispose requires surplus plutonium to be converted to an oxide product and then dry blended with a multi-component adulterant mixture to reduce the Safeguards and Security attractiveness level of the materials. The resulting product will be discarded in a geologic repository at the Waste Isolation Pilot Plant (WIPP).

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DOE OSTI2019

Metal-Organic-Framework Glasses as Rad Contaminant Sequesters and Nuclear Waste Forms

Remediation of Tc remains an unresolved problem at SRS and other DOE sites. The objective of this project was to develop novel metal organic framework (MOF) glasses for radioactive contaminant sequestration and stabilization from aqueous media. During FY19, we synthesized, characterized and evaluated cetyl trimethylammonium bromide (CTAB)-functionalized and F – , Cl – , I – , CF 3 SO 3 – -exchanged MIL-101. MIL-101- CTAB was demonstrated to have high ReO 4 – removal capacity (~130 mg/g sorbent) from artificial groundwater. Re chemical speciation and binding mechanism in MIL-101-CTAB were also studied by synchrotron X-ray absorption spectroscopy. ReO 4 – was demonstrated as being in the pore structure with slightly larger Re-0 bond distances than those in NaReO 4 . In addition, a new Ni-TIPA MOF was demonstrated to be very stable, selective and effective for TcO 4 – removal from the SRS tank waste stream (~90% removal). This research may provide a highly applicable platform for solving critical DOE and industrial problems related to nuclear environmental stewardship and nuclear power production.

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DOE OSTI2020

Metal-Organic Framework Glasses as Rad Contaminant Sequesters and Nuclear Waste Forms

Remediation of Tc remains an unresolved problem at SRS and other DOE sites. The objective of this project was to develop novel metal organic framework (MOF) glasses for radioactive contaminant sequestration and stabilization from aqueous media. During FY20, we synthesized, characterized and evaluated additional cetyltrimethylcimmonium bromide (CTAB)-functionalized and F - , Cl - , I - , CF 3 SO 3 - exchanged MIL-101-Cr samples. MIL-10l-Cr-NO 3 -CTAB was demonstrated to have high ReO 4 - removed capacity (139 mg/g sorbent) from artificial groundwater (AGW). Re chemical specicition and binding mechanism on MIL-101-CTAB were also studied by synchrotron X-ray absorption spectroscopy. ReO 4 - was demonstrated as being in the pore structure with slightly larger Re-O bond distances than those in NciReO 4 and binding with the positively charged sites of CTAB. In addition, a new Ni-TIPA MOF was demonstrated to be very stable, selective and effective for TcO 4 - removed from the SRS tank waste stream (~90% removed). The Ni-TIPA MOF sample containing ReO 4 - was prepared and shipped/planned for vitrification studies by a high-pressure technique. This research may provide a highly applicable platform for solving critical DOE and industrial problems related to nuclear environmental stewardship and nuclear power production.

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DOE OSTI2018

Metal-Organic-Framework Glasses as Rad Contaminant Sequesters and Nuclear Waste Forms

Remediation of Tc remains an unresolved challenge at SRS and other DOE sites. The objective of this project was to develop novel metal organic framework (MOF) glasses for radioactive contaminant sequestration and stabilization from aqueous media. We synthesized, characterized and evaluated ten MOFs of known crystal structure and chemistry for screening MOFs for TcO 4 - and ReO 4 - sequestration. MIL-101(Cr) was most effective for TcO 4 - and ReO 4 - sequestration from artificial groundwater at equilibration pH values with the removal capacity of 0.076 mg/g for Tc and 0.197 mg/g for Re. MIL-101(Fe), SLUG-22 and UiO- 66- NO 2 were moderately effective under similar conditions (pH < 8). These results established a solid foundation for further development of MIL-101(Cr) and new MOFs to increase TcO 4 - removal capacity and improve MOF structural stability in FY19.This research may provide a highly applicable platform for solving critical DOE and industrial problems related to nuclear environmental stewardship and nuclear power production.

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