Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “WASTES”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes.

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward–Clyde in 1983 in which waste packages are assembled into “strings” and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

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

Solid Secondary Waste Immobilization in Cementitious Waste Forms at the Hanford Site - 19081

During operations of the Hanford Tank Waste Treatment and Immobilization Plant, solid secondary wastes (SSW) will be generated. These materials include HEPA filters, granular activated carbon (GAC), and silver mordenite (AgM). The planned disposal pathway for SSW is through solidification (blending) or encapsulation in a cementitious waste form and then disposal on site at the Hanford Site's Integrated Disposal Facility. Performance assessment evaluations of SSW forms have been performed; however, limited experimental data on simulated SSW is available as input. This paper will cover efforts to fill this gap in experimental data through testing of simulated SSW materials. Two grout formulations were used in the tests: 1) Hanford grout mix 5, a mix of 25 wt% ordinary Portland cement (OPC) and 75 wt% fly ash (FA); and 2) the Cast Stone dry blend of 47 wt% blast furnace slag, 45 wt% FA, and 8 wt% OPC. The sorption (Kd) of potential contaminants (iodide, iodate, technetium) to the grouts and SSW materials was investigated in simulated grout pore water. Waste forms containing GAC and AgM were prepared using the two grout formulations and characterized using scanning electron microscopy energy dispersive X-ray spectroscopy and X-ray diffraction. Acquiring this information will further support the candidacy of cementitious waste forms to be used for disposal of SSW at Hanford. (authors)

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

A Study into the Benefits of Using Discrete Event Simulation in TRU and Mixed Low-Level Waste (MLLW) Disposal at Advanced Mixed Waste Treatment Plant (AMWTP) - 19536

As part of the 1995 Settlement agreement reached between the Idaho State Government and the Idaho National laboratory, it was agreed that the approximately 65,000 cubic metres of waste stored at the AMWTP site was to be shipped out of the state to the waste isolation pilot plant (WIPP) in New Mexico or a similar disposal facility by the end of December 2018. In June 2016, the operation of AMWTP was taken over by Fluor Idaho, a subsidiary of Fluor Inc. To improve their capability of processing all of the varied waste contained on the extensive site, Fluor Idaho requested the assistance of DBD Ltd. We were tasked with producing an all-encompassing operational research (OR) model/discrete event simulation (DES) model of the site, which is able to produce live end-date predictions based on current inventory and waste tracking against milestones. The DES model, is capable of predicting years worth of plant operation in a matter of minutes including the variation caused by such complexities as resource/staff availability, equipment breakdown and maintenance, prioritisation of different waste types, and varying shift patterns of both staff and facilities. The DES model can be run from its own bespoke dashboard, that can be customised to suit the audience it is being used to inform, producing clear key performance indicator (KPI) graphics and high-level overviews for senior management, down to drum by drum breakdowns of the predicted feed plans for each individual facility on the AMWTP site for a work planner to analyse and compare to the current operations. Simulation can produce these types of outputs for very low cost and low risk, as it avoids experimentation to find potential alternatives to the status-quo that could result in loss of production or time. These methods can also be used to produce cost benefit analyses for suggested processing options feeding into optioneering studies and cost reduction strategies. As a tool for plant optimisation and tracking, discrete event simulation can provide detailed results quickly and reliably. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (approximately 1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology.

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

Forward Osmosis-based System for Treatment of Waste Water Generated During Energy Production Using Waste Carbon Dioxide and Waste Heat. Final Report

The installation of COHO at a power plant will reduce the volume of wastewater, reduce the energy costs to treat wastewater, expand the capabilities of water treatment all while reducing the cost of carbon capture from flue gas. In Phase IIB studies, Porifera explored the use of a next-generation COHO system to synergistically capture carbon dioxide from flue gas and treat wastewater at power plants. We found that the next generation system benefits from improved operational efficiency and safety. In addition, we performed a technoeconomic analysis of the system and found it to save significant cost compared to conventional technologies.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2018

A U-bearing composite waste form for electrochemical processing wastes

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.

Chen, X. (ORCID:0000000332734431)↗
DOE OSTI2021

Predicting zeolites’ stability during the corrosion of nuclear waste immobilization glasses: Comparison with glass corrosion experiments

During the long-term corrosion of nuclear waste glasses under nuclear waste disposal conditions, the precipitation of zeolitic phases has been linked to a delayed acceleration in glass corrosion (known as “Stage III”). Hence, predicting the thermodynamic propensity for zeolites to form upon the dissolution of nuclear waste glasses is key to ensure their long-term performance. Here, we compile a unified, internally-consistent thermodynamic database “clay20” to estimate the stability of clay and feldspar phases relevant to nuclear waste immobilization glasses, including beidellite(Mg, Ca, Na, K), kaolinite, montmorillonite(Mg, Ca, Na, K), nontronite(Mg, Ca, Na, K), saponite(Ca, Na, K), and albite. Based on this, we report a geochemical modeling method allowing us to predict the stability of secondary phases (including zeolites, calcium–silicate–hydrate gels, and clays) upon the dissolution of nuclear waste immobilization glasses. We show that this approach offers a realistic description of the stability of the secondary phases forming during the dissolution of two archetypical model nuclear glasses (namely, the International Simple Glass, ISG, and WVUTh-203) under conditions relevant to nuclear waste disposal (T = 90°C, p = 1 bar) as a function of pH. We find that the formation of silica and clay secondary phases is thermodynamically favored at low pH (pH < 10), whereas, in contrast, zeolite (analcime) and calcium–silicate–hydrate phases are favored at high pH (pH > 10.5). This suggests that thermodynamics (i.e., not solely kinetics) plays a key role in determining the range of solution pH wherein stage III corrosion may occur, i.e., when zeolite formation is favored.

Zhen-Wu, Bi Yun↗
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 OSTI2021

Melting rate correlation with batch properties and melter operating conditions during conversion of nuclear waste melter feeds to glasses

The rate of conversion of nuclear waste melter feed to glass is affected by the selection of melter feed materials and by melter design and operation. The melting rate correlation (MRC) is an equation that relates the glass production rate with two types of variables: (1) feed and melt properties: conversion heat, cold-cap bottom temperature, and glass melt viscosity; and (2) melter design and operation parameters: melter geometry, melter operating temperature, and gas bubbling rate. The MRC shows good agreement for an extended melting-rate data set of high-level waste (HLW) melter feeds and a data set generated for low-activity waste (LAW) melter feeds. Laboratory observation of heated melter feed samples is often used to assess the cold-cap bottom temperature of HLW melter feeds (moderately foaming feeds), but this technique appears inadequate for LAW melter feeds (vigorously foaming feeds). For LAW feeds, an adequate assessment of the cold-cap bottom temperature was achieved using evolved gas analysis, which allows identification of the collapse of primary foam for oxidized feeds. This assessment shows that the cold-cap bottom temperature for vigorously foaming LAW feeds is higher than that for moderately foaming HLW feeds. When the results of MRC are compared, LAW feeds are generally less sensitive to the bubbling rate and melt viscosity, and more sensitive to the cold-cap bottom temperature than HLW feeds. The MRC qualifies as a promising tool to support the selection of melter feed materials and melter operating conditions, which is determined from expensive independent scaled melter experiments, and sophisticated mathematical models.

Lee, Seung Min↗
DOE OSTI2020

Soil Sample Plan in Support of the Site 300 Explosives Waste Treatment Facility Ecological Risk Assessment Pursuant to the Hazardous Waste Facility Permit, Part V, Special Condition #14 a-f. August 2020

The Department of Toxic Substances Control issued the Final Hazardous Waste Facility Permit to Lawrence Livermore National Laboratory, Site 300 in 2017. As part of the continual process to ensure waste treatment emissions are not impacting ecological receptors, LLNL analyzed soil samples in the vicinity of the Explosive Waste Treatment Facility (EWTF) Burn Units (i.e., Burn Cage and Burn Pan) and the Detonation Pad. Soil samples were analyzed by EPA Solid Waste 846 Methods for furans, explosives, metals, and semivolatile compounds. In all samples, furans, explosives, and semi-volatile organic compounds were not detected. Metals were detected at concentrations below background levels. This soil sample plan will sample the same four downwind areas of the Burn Units and the same downwind areas of the Detonation, as well as including four new sample locations. The four new locations will be located to the northeast and southwest of the Burn Units and Detonation (as shown in Figure 3). The constituents to be analyzed are in Tables 2 and 3. Perchlorate was not identified as a constituent of concern in the 2007 Soil Sampling Plan. However, for this plan, all soil samples will be analyzed for perchlorate. After completion of the sampling event, a soil sample report will be submitted to DTSC in accordance with requirements specified in the Hazardous Waste Facility Permit, Part V, Special Conditions 16 a-f. This proposed soil sampling plan will be fully implemented no later than one year after receiving DTSC approval, in accordance with the Hazardous Waste Facility Permit, Part V, Special Condition 15.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Soil Sample Workplan in Support of the Site 300 Explosives Waste Treatment Facility Ecological Risk Assessment Pursuant to the Hazardous Waste Facility Permit (Part V, Special Condition #14 a-f)

The Department of Toxic Substances Control issued the Final Hazardous Waste Facility Permit to Lawrence Livermore National Laboratory, Site 300 in 2017. As part of the continual process to ensure waste treatment emissions are not impacting ecological receptors, LLNL analyzed soil samples in the vicinity of the Explosive Waste Treatment Facility (EWTF) Burn Units (i.e., Burn Cage and Burn Pan) and the Detonation Pad. Soil samples were analyzed by EPA Solid Waste 846 Methods for furans, explosives, metals, and semi-volatile compounds. In all samples, furans, explosives, and semi-volatile organic compounds were not detected. Metals were detected at concentrations below background levels. This soil sample plan will sample the same four downwind areas of the Burn Units and the same downwind areas of the Detonation, as well as including four new sample locations. The four new locations will be located to the northeast and southwest of the Burn Units and Detonation (as shown in Figure 3). The constituents to be analyzed are in Tables 7 and 8. Perchlorate was not identified as a constituent of concern in the 2007 Soil Sampling Plan. However, for this plan, all soil samples will be analyzed for perchlorate. Nitrate and all CAM17 metals have been added to the sampling plan to comply with DTSC comments received in correspondence dated June 4, 2021. After completion of the sampling event, a soil sample report will be submitted to DTSC in accordance with requirements specified in the Hazardous Waste Facility Permit, Part V, Special Conditions 16 a-f. This proposed soil sampling plan will be fully implemented no later than one year after receiving DTSC approval, in accordance with the Hazardous Waste Facility Permit, Part V, Special Condition 15.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
NASA NTRS1973

Study of extraterrestrial disposal of radioactive wastes. Part 3: Preliminary feasibility screening study of space disposal of the actinide radioactive wastes with 1 percent and 0.1 percent fission product contamination

A preliminary study was conducted of the feasibility of space disposal of the actinide class of radioactive waste material. This waste was assumed to contain 1 and 0.1 percent residual fission products, since it may not be feasible to completely separate the actinides. The actinides are a small fraction of the total waste but they remain radioactive much longer than the other wastes and must be isolated from human encounter for tens of thousands of years. Results indicate that space disposal is promising but more study is required, particularly in the area of safety. The minimum cost of space transportation would increase the consumer electric utility bill by the order of 1 percent for earth escape and 3 percent for solar escape. The waste package in this phase of the study was designed for normal operating conditions only; the design of next phase of the study will include provisions for accident safety. The number of shuttle launches per year required to dispose of all U.S. generated actinide waste with 0.1 percent residual fission products varies between 3 and 15 in 1985 and between 25 and 110 by 2000. The lower values assume earth escape (solar orbit) and the higher values are for escape from the solar system.

Hyland, R. E.↗
NASA NTRS1992

Distribution of human waste samples in relation to sizing waste processing in space

Human waste processing for closed ecological life support systems (CELSS) in space requires that there be an accurate knowledge of the quantity of wastes produced. Because initial CELSS will be handling relatively few individuals, it is important to know the variation that exists in the production of wastes rather than relying upon mean values that could result in undersizing equipment for a specific crew. On the other hand, because of the costs of orbiting equipment, it is important to design the equipment with a minimum of excess capacity because of the weight that extra capacity represents. A considerable quantity of information that had been independently gathered on waste production was examined in order to obtain estimates of equipment sizing requirements for handling waste loads from crews of 2 to 20 individuals. The recommended design for a crew of 8 should hold 34.5 liters per day (4315 ml/person/day) for urine and stool water and a little more than 1.25 kg per day (154 g/person/day) of human waste solids and sanitary supplies.

Parker, Dick↗