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At least 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Hydrogenolysis Versus Hydrocracking for Polyolefin Upcycling

Global plastic production has reached 413.8 million metric tons in 2024 and is forecasted to surpass 1.2 billion metric tons by 2050. Polyolefins, mainly polyethylene (PE) and polypropylene (PP), dominate single-use packaging and account for approximately 55% of global plastic waste. The chemical inertness that makes these materials desirable for commercial applications also renders them persistent in the environment. Current recycling technologies have proven to be insufficient to divert plastic waste from landfills or environmental loss due to technical limitations and poor economic incentives. Conventional mechanical recycling is a form of downcycling, in which the polymer remelting process results in products with inferior material properties and reduced market value. Pyrolysis, a thermochemical route used for the chemical recycling of plastic wastes into refinery feedstock, requires severe conditions, typically 400–600 °C in an oxygen-free environment. The high temperature drives up energy costs and produces a wide range of poorly defined products, including undesirable light gases and heavy tars that pose challenges for downstream processing. As an alternative, catalytic chemical recycling offers a promising route for converting waste polyolefins back into value-added hydrocarbons—such as fuels, lubricant base oils, and other chemical feedstocks—at temperatures typically below 300 °C. This opinion article focuses on two dominant pathways in heterogeneous catalysis that are used to cleave the C–C bonds of polyolefins: hydrogenolysis and hydrocracking. Both catalytic pathways have the potential to selectively convert polyolefin waste into valuable fuels and chemical feedstocks under mild conditions. Notably, while hydrocracking primarily yields branched hydrocarbon products, hydrogenolysis predominantly produces linear hydrocarbons. Here, we compare the mechanisms and catalyst designs for hydrocracking and hydrogenolysis, analyze critical technical challenges from catalyst stability to process engineering, and provide an outlook on how these complementary pathways can be used to repurpose plastic waste into valuable products.

Zhang, Ruoxi [Iowa State Univ., Ames, IA (United S↗
DOE OSTI2020

Reduced Neutralization Feasibility Study for H-Canyon Accelerated Basin De- Inventory (ABD) Program

An alternative approach to Spent Nuclear Fuel (SNF) and Nuclear Material Processing was developed for future H-Canyon (HCAN) and L-Area operations that involves a paradigm shift from current HCAN, Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The alternative, referred to as Accelerated Basin De-inventory (ABD), requires that all Domestic and Foreign Research Reactor SNF currently at the Savannah River Site (SRS) will be dissolved, stored, and then transferred to CSTF without recovery of Highly Enriched Uranium (HEU). Concentrated nitric acid is utilized to dissolve aluminum spent nuclear fuel (ASNF) in HCAN. The vessels and piping in HCAN are fabricated from 304L stainless steel and are ideally suited to handle the acidic waste stream. However, as the waste is transferred to the CSTF and DWPF, it will contact the carbon steel waste tanks in CSTF. In order to prevent corrosion of the carbon steel, the acidic waste is neutralized (i.e., pH adjusted over 11) by the addition of sodium hydroxide (NaOH). The NaOH is added until the final solution contains 1.2 M excess -OH. Additionally, if the waste is not neutralized to a pH greater than 11, then aluminum hydroxide (Al(OH) 3 ) would form, and solids may form in the piping as it is transferred to CSTF. This document presents an analysis of the influence of reducing the excess caustic that is added to the neutralization tanks on the corrosion protection scheme primarily for the CSTF waste tanks. The implications to HCAN and DWPF were also assessed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Enduring Transuranic Liquid Waste Volume Forecast and Constituent Evaluation (FY20)

The “FY11 Radioactive Liquid Waste (RLW) Influent Boundary Condition Assessment and Forecast” report provided influent volume forecasts that were critical in establishing the design capacity for the new Radioactive Liquid Waste Transuranic (RLW-TRU) capability. The design for the new RLW-TRU capability was paused as the final design was being completed and prior to the transition of the new management and operations contract for Los Alamos National Laboratory (LANL). Recently, NNSA has directed the project to restart and complete the design effort. As part of the restart process, LANL was asked to validate or document any material changes in the influent volumes or constituents that formed the design basis for the new capability. This analysis updates the transuranic liquid waste influents volumes, incorporating the changes in the LANL program of record since the FY11 assessment, and includes a cursory validation of influent constituents. The results of the FY19 update to influent volume forecasts shows: 1) the influent volume of 29,000 L should continue to be used as the project design-basis; 2) There have been no material changes in the operational processes that would change the constituents of concern from either a treatment or material compatibility perspective.

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DOE OSTI2012

The Transmutation of Nuclear Waste in the Two-Zone Subcritical System Driven by High- Intensity Neutron Generator - 12098

The main problems of transmutation of high-level radioactive waste (minor actinides and long-lived fission products) are considered in our work. The range of radioactive waste of nuclear power is analyzed. The conditions under which the transmutation of radioactive waste will be most effective are analyzed too. The modeling results of a transmutation of the main radioactive isotopes are presented and discussed. The transmutation of minor actinides and long-lived fission products are modeled in our work (minor actinides - Np-237, Am-241, Am-242, Am-243, Cm-244, Cm-245; long-lived fission products - I-129, Tc-99). The two-zone subcritical system is calculated with help of different neutron-physical codes (MCNP, Scale, Montebarn, Origen). The ENDF/B-VI nuclear data library used in above calculations. Thus, radioactive wastes can be divided into two main groups that need to be transmuted. The minor actinides form the first group and the long-lived fission products form the second one. For the purpose of effective transmutation these isotopes must be extracted from the spent nuclear fuel with the help of either PUREX technology or pyrometallurgical technology. The two-zone reactor system with fast and thermal regions is more effective for nuclear waste transmutation than the one-zone reactor. Modeling results show that nearly all radioactive wastes can be transmuted in the two-zone subcritical system driven by a high-intensity neutron generator with the external neutron source strength of 1.10{sup 13} n/sec. Obviously, transmutation rate will increase with a rise of the external neutron source strength. From the results above we can also see that the initial loading of radioactive isotopes into the reactor system should exceed by mass those isotopes that are finally produced. (authors)

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DOE OSTI2015

Regulatory inspection of the provision of disposability assessment and waste packaging advice in the UK to provide confidence that higher activity waste stored on licensed sites will be suitable for safe handling, transport and disposal - 15088

Radioactive Waste Management Limited (RWM), the developer and future operator of a geological disposal facility (GDF) in the United Kingdom, operates a disposability assessment process to minimise the risk that current conditioning and packaging of higher activity wastes (HAW) will result in packages incompatible with geological disposal in the future. Specialists in nuclear and environmental safety from the Office for Nuclear Regulation (ONR) and the Environment Agency (EA) in 2013/14 completed an inspection of the RWM's process of disposability assessment. The inspection gave the regulators confidence that RWM has in place a systematic process to provide an important means of reducing risks associated with waste packaging in advance of an operational GDF and that the process of disposability assessment is providing the necessary information for licensees to assure future HAW disposal to the GDF. This work forms part of our continuing programme to review RWM's work related to geological disposal of higher activity radioactive waste (HAW). (authors)

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DOE OSTI2013

Absorption characteristics of lithium bromide (LiBr) solution constrained by superhydrophobic nanofibrous structures

An experimental study on absorption characteristics of water vapor into a thin lithium bromide (LiBr) solution flow is presented. The LiBr solution flow is constrained between a superhydrophobic vapor permeable wall and a solid surface that removes the heat of absorption. As opposed to conventional falling film absorbers, in this configuration, the solution film thickness and velocity can be controlled independently to enhance the absorption rate. The effects of water vapor pressure, cooling surface temperature, solution film thickness, and solution flow velocity on the absorption rate are studied. An absorption rate of approximately 0.006 kg/m(2) s was measured at a LiBr solution channel thickness and flow velocity of 100 mu m and 5 mm/s, respectively. The absorption rate increased linearly with the water vapor driving potential at the test conditions of this study. It was demonstrated that decreasing the solution film thickness and increasing the solution velocity enhance the absorption rate. The high absorption rate and the inherently compact form of the proposed,absorber facilitate development of compact small-scale waste heat or solar-thermal driven cooling systems. Published by Elsevier Ltd.

Isfahani, RN↗
DOE OSTI2020

DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗
NASA NTRS1987

Transient radiative cooling of a droplet-filled layer

A proposed lightweight radiator system for waste heat dissipation in space would eject streams of coolant in the form of small, hot liquid droplets. The droplets would lose radiative energy by direct exposure to the very low-temperature environment of space, and would then be collected for reuse. The cooling behavior of a layer composed of many small droplets was studied by numerical solution of the radiative integral equations. Since there is mutual interference for radiative energy dissipation, an array droplet will cool more slowly than if each drop is exposed individually. Since liquid metal droplets may be used, the study includes results for conditions with high scattering. For optically thin regions, especially with high scattering, the temperature distribution is sufficiently uniform that the cooling can be computed using the approximation of a constant layer emittance. For optically thick layers starting at uniform temperature, the temperature distributions become nonuniform with time. It was found that the cooling process goes through a starting transient; a constant emittance condition is then achieved where the emittance is lower than that for a layer at uniform temperature.

Siegel, R.↗
NASA NTRS2007

Core-Cutoff Tool

A tool makes a cut perpendicular to the cylindrical axis of a core hole at a predetermined depth to free the core at that depth. The tool does not damage the surrounding material from which the core was cut, and it operates within the core-hole kerf. Coring usually begins with use of a hole saw or a hollow cylindrical abrasive cutting tool to make an annular hole that leaves the core (sometimes called the plug ) in place. In this approach to coring as practiced heretofore, the core is removed forcibly in a manner chosen to shear the core, preferably at or near the greatest depth of the core hole. Unfortunately, such forcible removal often damages both the core and the surrounding material (see Figure 1). In an alternative prior approach, especially applicable to toxic or fragile material, a core is formed and freed by means of milling operations that generate much material waste. In contrast, the present tool eliminates the damage associated with the hole-saw approach and reduces the extent of milling operations (and, hence, reduces the waste) associated with the milling approach. The present tool (see Figure 2) includes an inner sleeve and an outer sleeve and resembles the hollow cylindrical tool used to cut the core hole. The sleeves are thin enough that this tool fits within the kerf of the core hole. The inner sleeve is attached to a shaft that, in turn, can be attached to a drill motor or handle for turning the tool. This tool also includes a cutting wire attached to the distal ends of both sleeves. The cutting wire is long enough that with sufficient relative rotation of the inner and outer sleeves, the wire can cut all the way to the center of the core. The tool is inserted in the kerf until its distal end is seated at the full depth. The inner sleeve is then turned. During turning, frictional drag on the outer core pulls the cutting wire into contact with the core. The cutting force of the wire against the core increases with the tension in the wire and, hence, with the frictional drag acting on the outer sleeve. As the wire cuts toward the center of the core, the inner sleeve rotates farther with respect to the outer sleeve. Once the wire has cut to the center of the core, the tool and the core can be removed from the hole. The proper choice of cutting wire depends on the properties of the core material. For a sufficiently soft core material, a nonmetallic monofilament can be used. For a rubber-like core material, a metal wire can be used. For a harder core material, it is necessary to use an abrasive wire, and the efficiency of the tool can be increased greatly by vacuuming away the particles generated during cutting. For a core material that can readily be melted or otherwise cut by use of heat, it could be preferable to use an electrically heated cutting wire. In such a case, electric current can be supplied to the cutting wire, from an electrically isolated source, via rotating contact rings mounted on the sleeves.

Gheen, Darrell↗
DOE OSTI2015

Rankine cycle waste heat recovery system

A waste heat recovery (WHR) system connects a working fluid to fluid passages formed in an engine block and/or a cylinder head of an internal combustion engine, forming an engine heat exchanger. The fluid passages are formed near high temperature areas of the engine, subjecting the working fluid to sufficient heat energy to vaporize the working fluid while the working fluid advantageously cools the engine block and/or cylinder head, improving fuel efficiency. The location of the engine heat exchanger downstream from an EGR boiler and upstream from an exhaust heat exchanger provides an optimal position of the engine heat exchanger with respect to the thermodynamic cycle of the WHR system, giving priority to cooling of EGR gas. The configuration of valves in the WHR system provides the ability to select a plurality of parallel flow paths for optimal operation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2020

Electrolytically Assisted Surface Decontamination (EASD{sup TM}) for POCO Operations - 20282

NNL in collaboration with C-Tech Innovation Ltd and Sellafield Ltd has been exploring innovative technologies to enable a significant reduction of radiological hazards within facilities during the Post Operational Clean Out (POCO) phase of a nuclear plant's life cycle. Reducing the hazard by effectively decontaminating plants in-situ during POCO delivers huge cost reductions for future decommissioning operations. These cost reduction benefits are realized by reducing the number and complexity of remote operations as well as lowering the long-term waste disposal costs. Whilst chemical decontamination can achieve the desired level of decontamination, applying aggressive chemical reagents is hazardous, potentially difficult to control and requires there to be complimentary effluent treatment and waste routes. This work has aimed to develop flexible and controllable decontamination processes which could be operated without the additional complexity and issues associated with chemical decontamination. The processes needed to be relatively fast and effective to minimize the time operators would spend in an active area. In addition, there was a driver to produce decontamination methodologies which generated a secondary waste compatible with current routes and which takes advantage of current waste capacity. The strategy therefore was to come up with a solution that works with, rather than against, the science and engineering behind process plants at Sellafield. Electrolytically Assisted Surface Decontamination (EASD{sup TM}) is an innovative electrochemical decontamination process (developed by NNL and C-Tech Innovation Ltd) that can remove activity from contaminated metal in very short time periods. Application of the patented electric waveform to the surface when treating contaminated metal with nitric acid has been shown to significantly enhance the decontamination performance. When compared to proposed baseline washout procedures at Sellafield, the only change to the process is the applied electrical waveform which is temporary and controllable. The current causes dissolution of the metal surface being decontaminated which leads to activity transferring from the plant item into the nitric acid effluent stream. This innovation has the potential to transform the POCO process and allow a nitric acid based washout approach to deliver POCO quickly and cost effectively within the existing infrastructure. Several devices have now been designed that incorporate this EASD{sup TM} technology and enable the decontamination process to be applied to a range items commonly identified as being contaminated during nuclear decommissioning programmes such as pipework, tanks and hotspots of walls/floors. Inactive and active laboratory-scale testing has been completed using both simulated and 'real-life' contaminated (low level waste) metallic items retrieved from different nuclear sites. Active trials demonstrated that contaminated items could be treated to free release levels within minutes. Engineering scale tests are currently being performed, with guidance from Sellafield's system engineers and plant managers, to provide the necessary re-assurance the designed devices could be successfully deployed in an on-plant scenario. The final stage of this development work is to carry out an active demonstration of the decontamination technology on Thorp at Sellafield. This paper aims to highlight progress made to date with the EASD{sup TM} technology and, more specifically, the development of an in-situ decontamination device for deployment within radioactive pipework. (authors)

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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↗
NASA NTRS1978

Non-equilibrium radiation nuclear reactor

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

Thom, K.↗
DOE OSTI2013

Effect of Feeding Rate on the Cold Cap Configuration in a Laboratory-Scale Melter

High level waste melter feed is converted into glass in a joule heated melter, where it forms a floating layer of reacting feed, called the cold cap. After the glass-forming phase becomes connected, evolving gases produce bubbles that form a foam layer under the cold cap. The bubbles coalesce into cavities that escape around the edges of the cold cap. The foam layer insulates the cold cap from the heat transferred from the molten glass below. More information is needed about the formation and behavior of the foam layer to control, limit and possibly avoid foaming, thus allowing for a higher rate of melting. The cold cap behavior was investigated in a laboratory scale assembly with a sealed silica-glass crucible. A high alumina waste simulant was fed into the crucible and the feed charging rate was varied from 3 to 7 mL min-1. After a fixed amount of time (35 min), feed charging was stopped and the crucible was removed from the furnace and quenched on a copper block to preserve the structure of the cold cap and foam during cooling. During the rapid quenching, thermal cracking of the glass and cold cap allowed it to be broken up into sections for analysis. The effect of the charging rate on the height, area and volume of the cold cap was determined. The size of the bubbles collected in the foam layer under the cold cap increased as the cold cap expanded. Under the cold cap, the bubbles coalesced into oblong cavities. These cavities allowed the evolved gases to escape around the edges of the cold cap through the molten glass into the melter plenum.

Dixon, Derek R.↗
DOE OSTI2016

Migration Rates of Some Cations in Unsaturated Layer around the Near-Surface Disposal Facilities of Radioactive Wastes - 16116

In the porous layer around the near-surface underground, the intermittent recharge process of rainwater forms a layer perfectly not saturated with liquid phase. Here, such a layer is simply called 'unsaturated layer'. In this study, the retardation effects of radionuclide migration were examined under an unsaturated condition by using a column packed with silica sand particles. As tracer ions, sodium (Na), europium (Eu) and cesium (Cs) were selected for estimating the retardation effects under both the saturated and unsaturated condition. Particularly, Cs is one of the key elements included in the contaminated soil due to Fukushima Daiichi Nuclear Power Plant Accident. These tracer ions were injected into column with the condition of constant flow rate or constant pressure flow. To quantify the influence of the unsaturated condition on the migration of radionuclides, retardation coefficient (R{sub d}) and Peclet number (P{sub e}) were estimated by applying advection diffusion equation to the experimental results. As a result, the conflicting effects on the retardation effect in the unsaturated layer were suggested; one is the decrease in the retardation effect by interrupting the sorption of ions with gas phase, the other is the increase in the retardation effect by the clogging with gas phase in flow paths. For the assessment of the radionuclide migration in the unsaturated layer, it is important to clarify such complicated behaviors with considering key parameters related to the migration of ions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Environmental Geochemistry and Acid Mine Drainage Evaluation of an Abandoned Coal Waste Pile at the Alborz-Sharghi Coal Washing Plant, NE Iran

In this paper, an abandoned waste coal pile, which is resulted from Alborz-Sharghi coal washing plant, NE of Iran was mineralogically and geochemically characterized to evaluate pyrite oxidation, acid mine drainage (AMD) generation, and trace element mobility. After digging ten trenches and vertical sampling, a quantitative method including the atomic absorption test, and the quality-based methods including optical study were carried out for determination of pyrite fractions in the waste pile. The geochemical results revealed that the fraction of remaining pyrite increased with depth, indicating that pyrite oxidation is limited to the shallower depths of the pile which were confirmed by variations of sulfate, pH, EC, and carbonate with depth of the pile. To evaluate the trend of trace elements and mineralogical constituents of the waste particles, the samples were analyzed by using XRD, ICP-MS, and ICP-OES methods. The results showed the secondary and neutralizing minerals comprising gypsum have been formed below the oxidation zone. Besides, positive values of net neutralization potential indicated that AMD generation has not taken in the waste pile. In addition, variations of trace elements with depth reveal that Pb and Zn exhibited increasing trends from pile surface toward the bottom sampling trenches while another of them such as Cu and Ni had decreasing trends with increasing depth of the waste pile.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

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