Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Forms”

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 1,585 records · Page 88
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Combustion Synthesis of Transuranic-doped Ceramics for Nuclear Waste Immobilization

Utilizing the devised SCS technique, crystalline transuranic-doped zirconate pyrochlores were formed for the first time without the need for a subsequent heating step. Typically, materials rapidly synthesized in a single-step self-sustaining exothermic reaction require further heating to induce crystallization and purify the product. The need for this secondary processing step can be attributed to factors such as an improper oxidizer to fuel ratio and the utilization of strong complexing compounds as fuels (i.e., reducing agents), all of which hinder complete combustion. The efficacy of the developed process was demonstrated by fabricating crystalline plutonium-doped Gd2Zr2O7 from solutions containing metal nitrates and urea in a properly tuned ratio

Burton-Allen, Janiya D.↗
DOE OSTI2019

Methods and systems for capturing and storing carbon dioxide

Methods and systems for capturing and storing carbon dioxide are disclosed. In some embodiments, the methods include the following: mixing materials including magnesium or calcium with one or more acids and chelating agents to form a magnesium or calcium-rich solvent; using the organic acids derived from biogenic wastes as acids or chelating agents; generating carbonate ions by reacting a gas including carbon dioxide with a carbonic anhydrase biocatalyst; reacting the solvent with the carbonate ions to form magnesium or calcium carbonates; recycling a solution containing the biocatalyst after forming magnesium or calcium carbonates for re-use in the generating step; using the magnesium and calcium carbonates as carbon neutral filler materials and using the silica product as green filler materials or inexpensive absorbents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Innovative Process for Comprehensive Treatment of Liquid Radioactive Waste - 12551

This paper presents the results of research activities aimed at creation of a principally new LRW distilling treatment method. The new process is based on the instantaneous evaporation method widely used in distillation units. The main difference of the proposed process is that the vapor condensation is conducted without using heat exchangers in practically ideal mode by way of direct contacting in a vapor-liquid system. This process is conducted in a specially designed ejector unit in supersonic mode. Further recuperation of excess heat of vaporization is carried out in a standard heat exchanger. Such an arrangement of the process, together with use of the barometric height principle, allows to carry out LRW evaporation under low temperatures, which enables to use excess heat from NPS for heating initial LRW. Thermal calculations and model experiments have revealed that, in this case, the expenditure of energy for LRW treatment by distilling will not exceed 3 kilowatt-hour/m{sup 3}, which is comparable with the reverse-osmosis desalination method. Besides, the proposed devices are 4 to 5 times less metal-intensive than standard evaporation units. These devices are also characterized by versatility. Experiments have revealed that the new method can be used for evaporation of practically any types of LRW, including those containing a considerable amount of oil products. Owing to arrangement of the evaporation process at low temperatures, the new devices are not sensitive to 'scale formation'. This is why, they can be used for concentrating brines of up to 500-600 g/l. New types of such evaporating devices can be required both for LRW treatment processes at nuclear-power plants under design and for treating 'non-standard' LRW with complex physicochemical and radionuclide composition resulting from the disaster at the Fukushima I Nuclear Power Plant.) As a result of accidents at nuclear energy objects, as it has recently happened at NPP 'Fukushima-1', personnel faces the necessity to take emergency measures and to use marine water for cooling of reactor zone in contravention of the technological regulations. In these cases significant amount of liquid radioactive wastes of complex physicochemical composition is being generated, the purification of which by traditional methods is close to impossible. According to the practice of elimination of the accident after-effects at NPP 'Fukushima' there are still no technical means for the efficient purification of liquid radioactive wastes of complex composition like marine water from radionuclides. Therefore development of state-of-the-art highly efficient facilities capable of fast and safe purification of big amounts of liquid radioactive wastes of complex physicochemical composition from radionuclides turns to be utterly topical problem. Cesium radionuclides, being extremely dangerous for the environment, present over 90% of total radioactivity contained in liquid radioactive wastes left as a result of accidents at nuclear power objects. For the purpose of radiation accidents aftereffects liquidation VNIIHT proposes to create a plant for LRW reprocessing, consisting of 4 major technological modules: Module of LRW pretreatment to remove mechanical and organic impurities including oil products; Module of sorption purification of LWR by means of selective inorganic sorbents; Module of reverse osmotic purification and desalination; Module of deep evaporation of LRW concentrates. The first free modules are based on completed technological and designing concepts implemented by VNIIHT in the framework of LLRW Project in the period of 2000-2001 in Russia for comprehensive treatment of LWR of atomic fleet. These industrial plants proved to be highly efficient and secure during their long operation life. Module of deep evaporation is a new technological development. It will ensure conduction of evaporation and purification of LRW of different physicochemical composition, including those containing hardness salts, resulted in generation of LRW concentrate 300-600 g/l. The method is based on utilization of supersonic ejector for intensification of thermal physic processes and performance of evaporation in brine recycling mode. All proposed technological solutions are totally based on patented Russian developments. Proposed work will allow to construct modular plants, which will be totally prepared for efficient purification of any types of liquid radioactive wastes from radionuclides in case of force majeure. According to proposed scheme concentration level of cesium radionuclides in safe-for-storage form will make up not less than 5000. With respect to purification from cesium radionuclides of liquid radioactive wastes stored at NPP 'Fukushima' about 10 t of inorganic sorbents, loaded in 160 protective filter-containers, will be required for solving this problem. The amount of secondary wastes will be reduced approximately in 5 times in comparison with traditional schemes, applied in purification of secondary LRW of Fukushima-1 by Areva (France) and Kurion (USA) companies. All units of modular plants will be constructed and manufactured as totally automated, providing their twenty-four-hour safe operation. Modular design will ensure efficiency and let optimize the costs of secondary LRW treatment. In order to ensure off-line operation in emergency conditions the plant should be equipped with auxiliary modules: energy and ventilation ones. Under normal conditions these modules can be stored in 'mothballed' condition at special warehouses under the authority of federal bodies. It will be reasonable to choose required transport facilities, the most suitable for transportation of modules to target destination beforehand, using vessel classification list.

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

Energy Conversion Loop: Process Heat Applications of Hybrid Energy Systems

Climate change due to greenhouse gas (GHG) emissions has prompted a shift from fossil fuel energy to other sources for electricity production such as nuclear, solar or wind. These alternatives since no GHG is produced, and no fossil fuels are burned. Both nuclear and other renewable energy sources are recognized as reliable technologies in the United States and around the world. However, they all have a unique set of challenges. Two of the most common renewable energy sources are solar and wind. A particular shortcoming of these two options is the variability of the energy produced; that is they are highly dependent on weather. Many times wind and/or solar power systems are not able to meet the electric demand during peak times. Replacing fossil fuels with nuclear energy is an option (non fossil fuel) for clean energy production. Two primary challenges with using nuclear energy are; wasted heat energy and the lack of load following capabilities. First, heat energy is wasted in the production stream from steam that is created while cooling the core of a nuclear reactor. This steam is initially used in turbines where the electricity is produced. However, once the steam passes through the turbines it is cooled, re-condensed to water, and reused. Large amounts of heat energy are released in the form of heat to the environment (either to the atmosphere or as waste water) and go unused. Secondly, nuclear energy is primarily a baseload power producer - the power output does not fluctuate with power demands, or is not load following. A nuclear hybrid energy system (NHES) may be a solution to the two issues mentions above. Traditional hybrid energy systems (HES) are a combination of two or more energy producers working together to create different energy commodities such as fuel or electricity. To create a NHES, normal use of the reactor for electricity production is combined with renewable power sources to meet the peak power demand of the grid. When electricity demand is high the renewable source and NPP are used in conjunction to meet the demand. When the electricity demand is low the renewable power is used to produce electricity while the NPP operates at full capacity. Rather than sending the excess heat to the cooling towers the excess energy created by the NPP (heat) is diverted to another industrial process.There is no need to decrease the power coming out of the NPP. This research proposal addresses harvesting unused thermal energy from a NPP during normal operations and in times of low power demand.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Chemical Disposition of Plutonium in Hanford Site Tank Wastes

This report examines the chemical disposition of plutonium (Pu) in Hanford Site tank wastes, by itself and in its observed and potential interactions with the neutron absorbers aluminum (Al), cadmium (Cd), chromium (Cr), iron (Fe), manganese (Mn), nickel (Ni), and sodium (Na). Consideration also is given to the interactions of plutonium with uranium (U). No consideration of the disposition of uranium itself as an element with fissile isotopes is considered except tangentially with respect to its interaction as an absorber for plutonium. The report begins with a brief review of Hanford Site plutonium processes, examining the various means used to recover plutonium from irradiated fuel and from scrap, and also examines the intermediate processing of plutonium to prepare useful chemical forms. The paper provides an overview of Hanford tank defined-waste–type compositions and some calculations of the ratios of plutonium to absorber elements in these waste types and in individual waste analyses. These assessments are based on Hanford tank waste inventory data derived from separately published, expert assessments of tank disposal records, process flowsheets, and chemical/radiochemical analyses. This work also investigates the distribution and expected speciation of plutonium in tank waste solution and solid phases. For the solid phases, both pure plutonium compounds and plutonium interactions with absorber elements are considered. These assessments of plutonium chemistry are based largely on analyses of idealized or simulated tank waste or strongly alkaline systems. The very limited information available on plutonium behavior, disposition, and speciation in genuine tank waste also is discussed. The assessments show that plutonium coprecipitates strongly with chromium, iron, manganese and uranium absorbers. Plutonium’s chemical interactions with aluminum, nickel, and sodium are minimal to non-existent. Credit for neutronic interaction of plutonium with these absorbers occurs only if they are physically proximal in solution or the plutonium present in the solid phase is intimately mixed with compounds or solutions of these absorbers. No information on the potential chemical interaction of plutonium with cadmium was found in the technical literature. Definitive evidence of sorption or adsorption of plutonium onto various solid phases from strongly alkaline media is less clear-cut, perhaps owing to fewer studies and to some well-attributed tests run under conditions exceeding the very low solubility of plutonium. The several studies that are well-founded show that only about half of the plutonium is adsorbed from waste solutions onto sludge solid phases. The organic complexants found in many Hanford tank waste solutions seem to decrease plutonium uptake onto solids. A number of studies show plutonium sorbs effectively onto sodium titanate. Finally, this report presents findings describing the behavior of plutonium vis-à-vis other elements during sludge dissolution in nitric acid based on Hanford tank waste experience gained by lab-scale tests, chemical and radiochemical sample characterization, and full-scale processing in preparation for strontium-90 recovery from PUREX sludges.

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

Crystal structures of a dodecameric multicopper oxidase from Marinithermus hydrothermalis

We report multicopper oxidases (MCOs) represent a diverse family of enzymes that catalyze the oxidation of either an organic or a metal substrate with concomitant reduction of dioxygen to water. These enzymes contain variable numbers of cupredoxin domains, two, three or six per subunit, and rely on four copper ions, a single type I copper and three additional copper ions organized in a trinuclear cluster (TNC), with one type II and two type III copper ions, to catalyze the reaction. Here, two crystal structures and the enzymatic characterization of Marinithermus hydrothermalis MCO, a two-domain enzyme, are reported. This enzyme decolorizes Congo Red dye at 70°C in the presence of high halide concentrations and may therefore be useful in the detoxification of industrial waste that contains dyes. In two distinct crystal structures, MhMCO forms the trimers seen in other two-domain MCOs, but differs from these enzymes in that four trimers interact to create a dodecamer. This dodecamer of MhMCO forms a closed ball-like structure and has implications for the sequestration of bound divalent metal ions as well as substrate accessibility. In each subunit of the dodecameric structures, a Trp residue, Trp351, located between the type I and TNC sites exists in two distinct conformations, consistent with a potential role in facilitating electron transfer in the enzyme.

59 BASIC BIOLOGICAL SCIENCES↗
NASA NTRS1976

The dilemma of future electric power demand

Dim prospects are held out for continued exponential growth rates in the generation and use of electric power worldwide and in many local areas, extrapolating from current trends and common projections. While acknowledging the unique advantages of electric power use, the article points out the 30% level of efficiency in conversion of other forms of energy to electrical energy, with huge volumes of wasted energy plus thermal pollution. Even growth at a fixed rate is found problematical, with future needs exhausting water and land resources and fossil fuels in time. Alleviation of the situation by shifting much of the load to still unproven technologies is considered cautiously

Wu, Y.-C.↗
DOE OSTI2016

Shipping Grossly Dewatered resins in the RT-100 Type B Transportation Cask - 16068

Ion-exchange resins are used in the purification of coolant water in boiling water reactor nuclear power plants. The spent resins are a form of low-level or intermediate-level radioactive waste. These resins may have a relatively high specific activity level and are normally transported to a disposal facility within a secondary container placed in a Type B transportation cask. During transportation, hydrogen gas may build up in the cask due to radiolysis of the hydrogenous material in the contents. In the licensing process, a Type B cask is carefully analyzed to ensure that the maximum amount of hydrogen gas generated does not exceed 5%. The calculation to determine the mole fraction of hydrogen gas generated is a function of many variables, including the properties of the contents and the volume ratios of the contents. Because water generates hydrogen gas when it undergoes radiolysis, the amount of water present in the waste is normally limited. Typically, Type B transportation casks are allowed to carry dewatered resins, meaning less than 1% free-standing water by volume. To ensure compliance with this limitation, the dewatering process requires specialized equipment to pump the water from the secondary container. In an effort to enhance the usability of the RT-100 Type B transportation cask, Robatel Technologies re-analyzed the hydrogen gas generation calculation for contents including 'grossly dewatered' resins. Grossly dewatered resins are defined as resins containing less than 20% free-standing water by volume. This higher limit is made possible by a more thorough hydrogen gas generation calculation, which ensures that the flammability limit is not reached during shipment. In July 2015, The USNRC issued Revision 1 of the Certificate of Compliance for the Robatel Technologies RT-100 Type B transportation cask. This revision allows the cask to be used for grossly dewatered resins. With the revised Certificate of Compliance, the RT-100 is the first of its kind whose authorized contents include greater than 1% free liquids by volume. As a result of this change, the cask user can meet the 20% maximum free-standing water volume with a much simpler gravity-driven dewatering process. It is no longer necessary to have a complex dewatering system to ensure compliance with regulatory requirements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Exploring the speciation of actinide salts in the presence of contaminants related to molten salt reactors

Molten salt reactors (MSR) are up-and-coming Generation IV nuclear reactors with either coolant and/or radioactive fuel in molten salt form. Due to their improved safety, efficiency, affordability and convenient waste processing system, these reactors are considered a superior alternative to conventional ones. However, prior to commercial implementation, it is necessary to have an extensive understanding of the reactions occurring in the MSR. The safety parameters of the MSR are determined by the rheological properties such as density and viscosity of the molten salt and these properties are governed by the local structure of the molten salts. Due to the high operational temperatures of MSR's, corrosion plays an important role. Because of their influence, the container corrosion products become part of the fuel salt matrix. In this work, we are investigating how these corrosion products, e.g., Mo, Ni, Cr, would impact the speciation of the actinide metal centers. For the initial studies, lanthanides are used as a surrogate for actinides. Lanthanide chloride salts (LnCl3) are mixed with transition metals and excess of alkali or alkali earth metal salts or salt mixtures (LiCl, NaCl, KCl). These are added to an alumina crucible, placed in a quartz tube, and sealed under a vacuum. This is heated up to around 1000 °C and slowly cooled to a temperature above the melting point of the salt and at that temperature, the reaction is taken out of the furnace. The excess salt is removed, and the resulting product is analyzed using X-ray diffraction techniques, scanning electron microscopy, and spectroscopic methods such as Raman, UV-Visible, and FTIR spectroscopy.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
NASA NTRS2005

Documentation: No Substitute for Communication

SO WHAT IS AN RFI? IT WAS ONE OF THE FIRST THINGS I learned about back when I started my project management career with my first large construction firm. I learned how to use these forms as a convenient and effective means of documenting the many legitimate clarifications needed on a major project. However, like most other young engineers, I also learned to use the RFI as a weapon in the ongoing battle between owners. or their designer and the construction contractors. Recently, our project team has done a few simple things to greatly reduce the waste and frustration that comes from this type of battle. The RFI form can be a great tool if used properly, and I certainly don t recommend that they be eliminated entirely. The RFI form was created to document the many clarifications that are commonly required on projects. Typically, the contractor uses the top half of the form to clarify-or request permission to vary from-the contract documents. The bottom half of the form is used to record the answer. But this seemingly simple process is plagued by a number of problems. From the contractor s perspective, RFIs are needed to secure information that should have been in the contract documents in the first place. The missing information keeps their crews from working effectively, and it makes hitting already demanding cost and schedule targets even more difficult. Owners, or their design firms, often view the RFI as a means of harassment. Both sides of the issue have legitimate complaints, and both sides cause most of their own pain.

Strickland, John↗
DOE OSTI2018

Rheology of simulated radioactive waste slurry and cold cap during vitrification

During the vitrification of radioactive waste in a Joule-heated melter, aqueous melter feed slurry forms a cold cap, a reacting and melting material, which floats on the surface of the molten glass. The rheological behavior of the feed affects cold cap formation and shape, and is vital for modeling the feed-to-melt conversion process. Here, we used slurry feed simulant and fast-dried slurry solids representing the cold cap to investigate the rheological behavior of the feed as it transforms into glass. Both low-temperature and high-temperature rheometry were performed and a new scheme was applied to estimate the feed viscosity. This study shows that the conversion advances in four sequential stages that form distinct regions in the cold cap: (i) a fast-spreading boiling slurry from which water evaporates, (ii) a porous solid region (viscosity > 10 8 Pa s) containing reacting solids and molten salts, (iii) a plastic region in which glass-forming melt connects the refractory solids (~10 8 to ~10 6 Pa s), and (iv) a viscous foam layer in which the viscosity drops from ~10 5 to ~10 1 Pa s. The implications for the mathematical modeling of the cold cap are discussed.

glass melting↗
DOE OSTI2017

From Dealing with Legacy Waste to Avoiding It - 17432

When dealing with radioactive waste management and its regulation, there are various national approaches. At the same time there is a growing tendency towards consideration of fundamental principles and standards thus forming the underlying basics in detailing the customized regulatory framework for specific applications. Dealing with legacy waste poses some particular problems as at the time of its production current safety principles were sometimes not well established. Considering today's standards this frequently requires remedial actions. Usually, this involves considerable effort and cost as well as radiation exposure for the involved personnel. Resolving legacy waste related problems not only benefits from an established regulatory framework but allows conclusions on how deal with legacy waste problems in general. The most important lessons learned from such operations might be how to prevent the production of legacy waste in the first place, thus avoiding further problems. By reflecting on practical applications from a number of countries the presentation is set to carve out the crucial recommendations from internationally recognized standards that give a substantial dividend in terms of not only time and budget but most of all safety and radiation exposure when properly implemented. The scope of the examples given ranges from dealing with legacy waste to avoiding it and includes recommendations towards essential features of best practice as seen from the perspective of an independent expert organization. (authors)

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

Applying understanding of likelihood and consequences to demonstrate post-closure criticality safety of a geological disposal facility - 14377

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls (e.g. fissile material limits and/or water barriers) will ensure that waste packages are sub-critical. These controls, plus the containment provided by the engineered barrier system of a geological disposal facility (GDF) will also ensure that criticality is prevented for a long time after disposal. However, over post-closure timescales conditions in the GDF will evolve, and therefore underpinning knowledge is required to demonstrate: under what conditions criticality could occur; what the likelihood of such conditions developing is; and what the consequences would be if a post-closure criticality did occur. These requirements are set out in the UK environment agencies' Guidance on Requirements for Authorisation of a GDF. This paper discusses the disposal of Pressurised Water Reactor fuel, as an example of a waste type that may require disposal. It briefly selects relevant research to combine and summarise results from three recent RWM research projects. The first 'Likelihood of Criticality' considered under what conditions post-closure criticality could occur, whilst a second 'Modelling of Consequences of Hypothetical Criticality' focused on understanding the local physical consequences (temperature rise, power and rock response) of hypothetical 'what-if' post-closure criticality. The paper then explains how this understanding is applied to make safety arguments as to why post-closure criticality is not a significant concern. This is achieved via a third, stream of work that has considered the effects of a potential criticality on pathways that might give rise to a post-closure risk in the future, such as transport of radionuclides in groundwater. Collectively RWMs post-closure criticality safety work allows us to conclude in our generic environmental safety case that a post closure criticality is a low likelihood event, and even if one of these low likelihood events was to occur its consequences would be insignificant to GDF performance. This paper also serves to give context to two companion UK post-closure criticality safety papers which communicate further information on our modelling development and more detailed results underpinning our demonstration of post-closure criticality safety. (authors)

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

Approaches to consider covers and liners in a low-level waste disposal facility performance assessment - 15300

On-site disposal cells are in use and being considered at several US DOE sites as the final disposition for large amounts of waste associated with cleanup of contaminated areas and facilities. These disposal cells are typically regulated by States and/or the USEPA in addition to having to comply with requirements in DOE Order 435.1, Radioactive Waste Management. The US DOE-EM Office of Site Restoration formed a working group to foster improved communication and sharing of information for personnel associated with these Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) disposal cells and work towards more consistent assumptions, as appropriate, for technical and policy considerations related to performance and risk assessments in support of a Record of Decision and Disposal Authorization Statement. One task completed by the working group addressed approaches for considering the performance of covers and liners/leachate collections systems in the context of a performance assessment (PA). A document has been prepared which provides recommendations for a general approach to address covers and liners/leachate collection systems in a PA and how to integrate assessments with defense-in-depth considerations such as design, operations and waste acceptance criteria to address uncertainties. Specific information and references are provided for details needed to address the evolution of individual components of cover and liner/leachate collection systems. This information is then synthesized into recommendations for best practices for cover and liner system design and examples of approaches to address the performance of covers and liners as part of a performance assessment of the disposal system. (authors)

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

Cold-cap formation from a slurry feed during nuclear waste vitrification

The time-temperature history of waste slurry feed during melting strongly affects various kinetic processes, such as reaction or dissolution rates, and, consequently, the melting rate. To analyze the time-temperature history of the feed in a cold-cap during nuclear waste vitrification, this work focuses on understanding how the main coldcap body forms, how the aqueous feed slurry interacts with the cold-cap, and estimating the cold-cap heat conductivity. To simulate the conditions during cold-cap formation, samples were prepared by rapid water evaporation from slurry feed. After water evaporated from the sample, a fresh slurry with tracer was poured onto the dry sample. X-ray fluorescence was then used to investigate the degree of penetration and/or mixing between the incoming slurry and the original sample. We show that the slurry does not interact or mix with the previously dry cold-cap crust, but that water-soluble components concentrate at the bottom of the boiling pools, where most of the water evaporates. Further, using the rate of water evaporation, the heat conductivity of the cold-cap was calculated from the measured temperature profile. The resulting conductivity is significantly higher than the values obtained in previous studies. We discuss the implications of the results for the cold-cap formation in the melter and for the development of the cold-cap mathematical model.

Hujova, Miroslava↗
DOE OSTI2015

Numerical study of radiation effect on the municipal solid waste combustion characteristics inside an incinerator

Highlights: • A 3-D model for the MSW incinerator with preheated air was developed. • Gas radiative properties were obtained from a statistical narrow-band model. • Non-gray body radiation model can provide more accurate simulation results. - Abstract: Due to its advantages of high degree volume reduction, relatively stable residue, and energy reclamation, incineration becomes one of the best choices for Municipal Solid Waste (MSW) disposal. However, detailed measurements of temperature and gas species inside a furnace are difficulty by conventional experimental techniques. Therefore, numerical simulation of MSW incineration in the packed bed and gas flow field was applied. In this work, a three dimensional (3-D) model of incinerator system, including flow, heat transfer, detailed chemical mechanisms, and non-gray gas models, was developed. Radiation from the furnace wall and the flame formed above the bed is of importance for drying and igniting the waste. The preheated air with high temperature is used for the MSW combustion. Under the conditions of high temperature and high pressure, MSW combustion produces a variety of radiating gases. The wavelength-depend radiative properties of flame adopted in non-gray radiation model were obtained from a statistical narrow-band model. The influence of radiative heat transfer on temperature, flow field is researched by adiabatic model (without considering radiation), gray radiation model, and non-gray radiation model. The simulation results show that taking into account the non-gray radiation is essential.

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

Polyacrylonitrile Composites of Ag–Al–Si–O Aerogels and Xerogels as Iodine and Iodide Sorbents

Mechanically robust engineered sorbents were constructed by embedding Ag-loaded aerogels or xerogels in a polyacrylonitrile (PAN) matrix at 30 or 50 mass% loadings of gel in PAN. Here, the process for making the base gels as well as the composite materials is described herein, a technology that is very flexible for making hybrid sorbents using various inorganic or organic active sorbents. The PAN-gel composite materials were evaluated for iodine uptake in the form of gaseous iodine [I 2(g) ] at room temperature and at 150°C as well as iodide (I – ) at room temperature from simulated liquid nuclear wastes. The composites show high I 2(g) uptake and high I – uptake (= 93 mass%) in simulated liquid secondary waste spiked with 5 mg L -1 I – . These results demonstrate the utility of such a composite form for iodine removal from contaminated gas and liquid streams.

36 MATERIALS SCIENCE↗