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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Radioactive Test Platform Demonstrating the Direct Feed Low-Activity Waste Process at Hanford - 18628

The Hanford Site is implementing a phased process for treating liquid legacy radioactive waste. The process is called the Direct Feed Low-Activity Waste process (DFLAW). The DFLAW includes a filtration and ion-exchange unit in the Low-Activity Waste Pretreatment System (LAWPS) that removes solids and cesium. This pretreat feed is then sent to a vitrification facility where glass-forming solids are added and the waste is turned into glass in a melter. The melter off-gas liquid condensate is evaporated in the Effluent Management Facility (EMF), with the concentrate recycled back into the feed. A laboratory-scale test platform has been built to test various aspects of the DFLAW process using samples of actual radioactive waste. The current platform represents four of the major unit operations: filtration, ion-exchange, vitrification, and off-gas condensation. A lab scale EMF is being designed to supplement the current platform. The platform is expandable to add any or all unit operations. The radioactive test platform has three missions. The first is simply to demonstrate the DFLAW process with real waste. A second objective is to address flowsheet gaps and opportunities, some of which require testing with real waste. Some gaps and opportunities include addressing the fate of mercury in the flowsheet, evaluating salt and aluminum precipitation, and proving new glass formulations with real waste. A third objective of the radioactive test platform is to demonstrate or evaluate future operations and alternative flowsheets. Some of the future operations evaluated will be the life expectancy of the resin in real waste, resin sampling, evaluation of how waste affects melter throughput, and grouting various primary and secondary waste streams. The radioactive test platform is therefore a crucial tool for verifying and optimizing the DFLAW flowsheet at Hanford with real waste. (authors)

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

Remote Retrieval, Disruption and Processing of Legacy Nuclear Waste - 16179

The UK National Nuclear Laboratory (NNL) is currently undertaking, in conjunction with Sellafield Ltd, a series of high profile trials of robotic equipment on a full scale test facility at their Workington Laboratory in England. These trials are designed to underpin and substantiate the processes proposed for the remote retrieval, disruption and processing of legacy Miscellaneous Beta Gamma Waste (MBGW) from a number of different donor plants on the Sellafield site, including some of the older fuel storage ponds and storage silos. These processes will be conducted inside the new Box Encapsulation Plant (BEP) which is due to be commissioned at Sellafield by 2020. These processes will subsequently enable Sellafield Ltd to safely recover legacy Intermediate Level Waste (ILW) from the older storage facilities on site and process it into a safer form which is more suitable for long term storage and then ultimately final disposal in a Geological Disposal Facility. As part of this legacy waste processing within the BEP plant, MBGW waste items will have to be remotely removed from a variety of different types of import skips. This waste will then be transferred, using robots, into the export container, which is a 3 m{sup 3} box liner, where the processed waste will be flood grouted to produce an encapsulated waste package. As part of this processing, any MBGW items consisting of sealed containers will have to be remotely disrupted before being grout encapsulated. This disruption is in order to (a) release any trapped liquors or gases from the containers and (b) ensure full grout infiltration of the container contents, thus minimising the amount of free liquors and voids within the final waste product. The NNL are conducting this technical work to confirm the viability of the proposed concept, processes and equipment. The developed solution consists of a linear layout of processing equipment with a pair of heavy duty, standard industrial, 'commercially off the shelf' (COTS) robots positioned around a Waste Handling Table. These robots have a 500 kg payload capacity and are operated remotely to handle all of the various types of MBGW waste, including the heavy and bulky items. The robots can be operated in manual tele-operation mode as well as pre-programmed automatic mode and have a range of primary and secondary tooling in order to perform the waste processing and disruption tasks as well as undertake general house-keeping and cleaning duties within the cell. The main tools include two types of hydraulic grab for the main waste recovery and handling tasks and a hydraulic shear for the waste disruption activities, but a number of other tools have also been designed and built for ancillary activities, such as a set of general purpose gripper jaws, drilling tool, bolt removal tool, pump tool, spray tool and a number of different housekeeping tools etc. The robot trials were started in November 2014, and have attracted a significant degree of high level stakeholder interest, including the UK Nuclear Decommissioning Authority (NDA), Office for Nuclear Regulation (ONR), Environment Agency (EA) and Radioactive Waste Management Limited (RWML) witnessing a number of the critical demonstration trials. Robots of the type selected are readily available, robust, reliable and easy to use. The trials have been very successful in proving the retrieval, processing and disruption operations, reducing the burden in the legacy waste facilities. Utilising robots to perform these tasks will lower the probability of producing non-compliant waste packages. In addition the robots can increase the consolidation of incoming waste skips into outgoing 3 m{sup 3} box liners, significantly increasing packing efficiencies and reducing the overall number of export containers produced thus lowering the decommissioning and waste storage/disposal costs. The trials have also enabled an assessment of the optimum way to operate the robots, as well as their reliability and maintainability issues. Engineering work associated with recovering failed robots and removing them from a typical Sellafield plant environment and then subsequently replacing them with functioning robots has also been undertaken and further work is now planned. The success of the trials, completed safely to a very challenging timescale, has enabled the use of robots in BEP to be endorsed as a viable option by the key stakeholders. Following a robust decision making process, this was a clear cut decision. The technical development and testing of the robot technology is currently continuing in order to investigate areas such as robot reliability, failure recoveries, human factors, ease of waste identification, optimisation of throughput rates, nuclearisation modifications etc. (authors)

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NASA NTRS2012

Curved Waveguide Based Nuclear Fission for Small, Lightweight Reactors

The focus of the presented work is on the creation of a system of grazing incidence, supermirror waveguides for the capture and reuse of fission sourced neutrons. Within research reactors, neutron guides are a well known tool for directing neutrons from the confined and hazardous central core to a more accessible testing or measurement location. Typical neutron guides have rectangular, hollow cross sections, which are crafted as thin, mirrored waveguides plated with metal (commonly nickel). Under glancing angles with incoming neutrons, these waveguides can achieve nearly lossless transport of neutrons to distant instruments. Furthermore, recent developments have created supermirror surfaces which can accommodate neutron grazing angles up to four times as steep as nickel. A completed system will form an enclosing ring or spherical resonator system to a coupled neutron source for the purpose of capturing and reusing free neutrons to sustain and/or accelerate fission. While grazing incidence mirrors are a known method of directing and safely using neutrons, no method has been disclosed for capture and reuse of neutrons or sustainment of fission using a circular waveguide structure. The presented work is in the process of fabricating a functional, highly curved, neutron supermirror using known methods of Ni-Ti layering capable of achieving incident reflection angles up to four times steeper than nickel alone. Parallel work is analytically investigating future geometries, mirror compositions, and sources for enabling sustained fission with applicability to the propulsion and energy goals of NASA and other agencies. Should research into this concept prove feasible, it would lead to development of a high energy density, low mass power source potentially capable of sustaining fission with a fraction of the standard critical mass for a given material and a broadening of feasible materials due to reduced rates of release, absorption, and non-fission for neutrons. This advance could be applied to direct propulsion through guided fission products or as a secondary energy source for high impulse electric propulsion. It would help meet national needs for highly efficient energy sources with limited dependence on fossil fuels or conflict materials, and it would improve the use of low grade fissile materials which would help reduce national stockpiles and waste.

Coker, Robert↗
DOE OSTI2018

FY18 Status Report: SNL Research into Stress Corrosion Cracking of SNF Interim Storage Canisters

This progress report describes work done in FY18 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). The work focuses on stress corrosion cracking (SCC), the only mechanism by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY18 continued several studies initiated in FY17 that are aimed at refining the understanding of the chemical and physical environment on canister surfaces, and evaluating the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL canister environment work focused on evaluating the stability of sea-salt deliquescent brines on the heated canister surface; an additional opportunity to analyze dusts sampled from an inservice spent nuclear fuel storage canister also arose. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments. The scope of these efforts targets near-marine Independent Spent Fuel Storage Installation environments which are generally considered to be most aggressive for pitting and SCC. Work to define the chemical and physical environment that could develop on storage canister surfaces in near-marine environments included experiments to evaluate the thermal stability of magnesium chloride brines, representative of the first brines to form when sea-salts deliquesce, with the specific goal of understanding and interpreting results of sea-salt and magnesium chloride corrosion experiments carried out under accelerated conditions. The experiments showed that magnesium chloride brines, and by extension, low RH sea-salt deliquescent brines, are not stable at elevated temperatures, losing chloride via degassing of HC1 and conversion to Mg-hydroxychlorides and carbonates. The experiments were carried out on an inert substrate to eliminate the effects of corrosion reactions, simulating brine stabilities in the absence of, or prior to, corrosion. Moreover, analysis of salts recovered from actively corroding metal samples shows that corrosion also supports or drives conversion of magnesium chloride or sea-salt brines to less deliquescent salts. This process has significant implications on corrosion, as the secondary phases are less deliquescent than magnesium chloride; the conversion reaction results in decreases in brine volume, and potentially results in brine dry-out. The deliquescence properties of these reaction products will be a topic of active research in FY19.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Microwave-Assisted dry reforming of toluene as a model tar compound using low-cost iron catalyst for syngas clean-up

Gasification of waste feedstock such as biomass, waste plastics suffers from high tar yields during hydrogen-rich syngas production. The presence of tars result in lower quality syngas, lower syngas yields, reactor blockage, reactor down time, and costly maintenance. Therefore, removal of tar from syngas during gasification is essential. Catalytic reforming of tars via in-situ syngas cleanup is an effective way of mitigating tars. This work explores the possibility of microwave-assisted catalytic dry reforming of tars for syngas production. Further, due to its chemical complexity, toluene, which is one of the main tar constituents, could be used as a model tar compound. Toluene dry reforming was studied using the Fe/Al 2 O 3 catalyst under CO 2 under the temperature range of 400–700 °C. The toluene reforming reaction was conducted using microwave and conventional thermal reactors. Under microwave irradiation, CO 2 and toluene conversions are boosted to 80% at 500 °C. Hydrogen and carbon monoxide yields were approximately five times and ten times higher in the microwave reactor at 500 °C, respectively, compared to the productions obtained in the conventional fixed-bed reactor at 700 °C. Filamentous carbon was also produced as a valuable side product to improve the economy of this process and such value-added carbon was only observed in the microwave reactor. Three reaction pathways were observed during microwave reaction: the toluene decomposition produces an initial hydrogen and carbon deposit on the catalyst; the formation of methane and benzene suggests toluene hydrodemethylation as a secondary reaction; and toluene hydrogenolysis forms light alkanes such as methane, and through reforming reaction under CO 2 to syngas.

10 SYNTHETIC FUELS↗
DOE OSTI2017

Assessing the costs of municipal solid waste treatment technologies in developing Asian countries

Highlights: • Municipal waste treatment costs are assessed for the first time in developing Asia. • A database with primary and secondary data from existing plants was prepared. • Capital and operational expenditures patterns were analyzed and discussed. • Incineration identified by large margin as the costliest approach for treating waste. • Context specific circumstances are key to explain deviations from the norm. - Abstract: The management of municipal solid waste (MSW) is one of the main costs incurred by local authorities in developing countries. According to some estimates, these costs can account for up to 50% of city government budgets. It is therefore of importance that policymakers, urban planners and practitioners have an adequate understanding of what these costs consist of, from collection to final waste disposal. This article focuses on a specific stage of the MSW value chain, the treatment of waste, and it aims to identify cost patterns associated with the implementation and operation of waste treatment approaches in developing Asian countries. An analysis of the capital (CAPEX) and operational expenditures (OPEX) of a number of facilities located in countries of the region was conducted based on a database gathering nearly 100 projects and which served as basis for assessing four technology categories: composting, anaerobic digestion (AD), thermal treatment, and the production of refuse-derived fuel (RDF). Among these, it was found that the least costly to invest, as a function of the capacity to process waste, are composting facilities, with an average CAPEX per ton of 21,493 USD{sub 2015}/ton. Conversely, at the upper end featured incineration plants, with an average CAPEX of 81,880 USD{sub 2015}/ton, with this treatment approach ranking by and large as the most capital intensive of the four categories assessed. OPEX figures of the plants, normalized and analyzed in the form of OPEX/ton, were also found to be higher for incineration than for biological treatment methods, although on this component differences amongst the technology groups were less pronounced than those observed for CAPEX. While the results indicated the existence of distinct cost implications for available treatment approaches in the developing Asian context, the analysis also underscored the importance of understanding the local context as a means to properly identify the cost structure of each specific plant. Moreover, even though CAPEX and OPEX figures are important elements to assess the costs of a waste treatment system, these should not be considered on a standalone basis for decision making purposes. In complement to this internal cost dimension, the broader impacts – to the economy, society and the environment – resulting from the adoption of a certain treatment approach should be properly understood and, ideally, measured and expressed in monetary terms.

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

Multimodal imaging of oil-in-water bilgewater emulsion and biofilms

Bilgewater emulsions, consisting of oil and surfactant mixtures in water, affect many aquatic species. Thus development of methods and techniques to treat or mitigate the formation and undesired consequences of shipboard emulsions are urgently needed. This work presents new results of the evolution of the bilgewaer chemical and physical properties using in situ imaging. A Navy O/W emulsion consisting of three oils and a detergent mixture was used as the synthetic bilgewater model [1, 2]. Synthetic oil-in-water (O/W) emulsion droplets were seled in a vacuum compatible microfluidic sample holder and imaged using a suite of tools including scanning electron microscopy (SEM), time-of-flight secondary ion mass spectrometry (ToF-SIMS), confocal laser scanning microscopy, and optical microscopy, highlighting the advantage of multiscale analysis and its applications in waste reduction and treatment in the environment. The transferrable and vacuum compatible microfluidic interface, System for Analysis at the Liquid Vacuum Interface (SALVI), was used in this study. SALVI enabled surface analysis of liquids and liquid-solid interactions using ToF-SIMS and SEM [3, 4]. Its detection window is an aperture of 2 ?m in diameter open to vacuum, permitting direct detection of the liquid surface. Liquid is withheld by surface tension within the aperture. The interface is composed of a silicon nitride membrane and polydimethylsiloxane microchannel [3, 4]. A variety of samples including complex liquid mixtures, ionic liquids, single mammalian cells, live biofilms, and solid-electrolyte interface have been analyzed using in situ imaging [5, 6]. The feasibility of using SALVI for in situ SEM imaging of nanoparticles and colloids in liquid was recently demonstrated [7]. This paper shows new findings of multimodal imaging and analysis of synthesized emulsion particles in water. Figure 1 depicts a schematic of multimodal in situ imaging of bilgewater emulsion using SEM and ToF-SIMS. Furthermore, we study the interactions of microbes and bilgewater emulsion. Microbes play an important role in the ocean. We hypothesize that microbes release organics that can act as surfactants that affect bilgewater formation or weakening. We present the first systematic study of emulsions and biofilms and investigate the effects of biofilms on bilgewater emulsions. Pseudomonas was selected as a model strain. Biofilms were cultured in a microchannel to allow culture [8]. Once a thick layer of biofilms was formed, the medium solution was changed to a mixture consisting of 50 % bilgewater emulsion. Dispersed biofilms were collected at 24 hrs. and 48 hrs. after emulsions were introduced into the channel. Bilgewater emulsions, biofilms, and mixtures of bilgewater emulsions and biofilms were analyzed using multiple in situ and ex situ techniques as described earlier. Our findings indicate that biofilms change the chemical makeup of the emulsion surface compositions and emulsion droplet size distribution, confirming the hypothesis that extracellular polymeric substance (EPS) related components released from biofilms can function as surfactants and change the oil-in-water interfaces.

bilgewater, multimodal imaging, biofilms↗
DOE OSTI2016

Groundwater Remediation in a Flood plain Aquifer at Shiprock, New Mexico - 16097

The Shiprock, New Mexico, Disposal Site is the location of a former uranium and vanadium ore-processing mill that operated from 1954 to 1968. During milling years, tailings leachate and other mill-related liquids migrated into underlying sediments, resulting in the contamination of groundwater beneath the river terrace containing the mill and a nearby flood plain alluvial aquifer adjacent to the San Juan River. The US DOE Office of Legacy Management manages the site and is implementing the current groundwater remedy for the flood plain area, consisting of natural flushing supplemented by extraction of alluvial groundwater to accelerate reduction in contaminant concentrations. The alluvial aquifer contains elevated concentrations of sulfate, uranium, and nitrate derived largely from the historical eastward migration of groundwater through terrace alluvium and underlying bedrock to the flood plain. The bedrock formation beneath the entire Shiprock site is Mancos shale, which forms an escarpment that separates the terrace from the flood plain. Contamination in the flood plain alluvial aquifer is influenced by a dynamic flow system that changes seasonally and from year to year. Simulations with a flood plain flow model based on monitored groundwater levels and coincident river-surface elevations show that background groundwater flow, unaffected by remediation pumping, is characterized by distinct hydraulic conditions for three separate periods each year. During months of low river flow in winter and early spring, river losses to the aquifer create a large hypo-rheic zone in the south half of the flood plain. Simultaneously, recharge of continually flowing surface water emptying onto the north half of the flood plain from Bob Lee Wash creates divergent, radial flow from the wash outlet and diverts northwestward-migrating contaminant plumes to the northeast where they discharge to the river. Subsequent bank storage processes in May and June from high snow melt runoff change background flow directions, increasing the spread of contamination in the alluvium. From mid-summer through fall, river flows are typically similar in value to those observed prior to the onset of snow melt, but groundwater elevations in the aquifer are noticeably lower than water levels observed in winter and early spring and may induce river losses to the aquifer. Several different contaminant sources have impacted the alluvial aquifer. Much of the contamination originated as tailings leachate and raffinate wastewater that infiltrated the terrace subsurface and eventually reached the flood plain at the escarpment. Additional sources of contamination in the aquifer included discharge of mill effluent to Bob Lee Wash and a pond at the base of the escarpment, and aqueous mobilization of solid-phase contamination in the form of windblown deposits and contaminants adsorbed to alluvial sediments. The degree to which remnant, secondary-source processes, such as leaching of solid-phase deposits and discharge of terrace groundwater across the escarpment, contribute to existing contaminant plumes is uncertain. The flood plain remediation system consists of two groundwater extraction wells, pumping from two horizontal wells placed in trenches near the base of the escarpment, a sump collecting discharges from seeps in the escarpment wall, and discharge of collected water to an evaporation pond on the terrace. Flow-model simulations indicate that most contaminated areas are captured by remediation pumping, resulting in several zones of freshwater between the river and the escarpment. Despite the highly transient nature of groundwater flow beneath the flood plain and the varied contaminant sources, initiation of remediation pumping in 2003 and coordinated operation of the system's various components have successfully removed large amounts of contaminant mass and helped prevent discharge of contaminated groundwater to the river. The percentage reductions of dissolved sulfate, uranium, and nitrate mass in the aquifer estimated for the first 9 years of remediation range from 40% to 72%. (authors)

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

Initial Assessment of TeF 6 Adsorption

Research and development that supports the management of off-gases from nuclear fuel reprocessing has historically been focused on the off-gas streams that arise from aqueous reprocessing technology. However, as Gen-IV reactor development pathways move toward deployment, alternative spent nuclear fuel (SNF) processing and disposition pathways have been considered more actively. This work is focused upon aspects of fluoride volatility (FV) processing. The versatility of this method for deployment against multiple types of spent fuel encourages the continued advancement of both the primary separations processes and the secondary processes, including waste treatment, material control and accountability, and engineering designs. Recent work noted that TeF 6 , the most highly volatile fluorinated compound produced during FV processing, did not have a clear abatement technology recommended in the literature. Thus, this work performed scoping tests on activated alumina and copper shot to assess whether they could be used to remove TeF 6 from gas streams that bear F 2 . Previous work on this topic was not well described in the literature and was not performed with excess F 2 in the stream as would be typical of spent fuel processing via FV. To support an understanding of the concentration of TeF 6 contacting the adsorbent beds, a series of preliminary testing identified the TeF 6 production rate and the equilibrium concentration of TeF 6 in the gas stream contacting the adsorbent. Excess F 2 was determined to not affect the ability of activated alumina to remove TeF 6 from the gas stream quickly and completely. A determination of whether excess F 2 affected the distribution depth of Te in the sorbent bed is still pending analysis of the used sorbent. Literature suggests that when activated alumina is near saturation, TeF 6 could migrate from the sorbent bed. Future testing should investigate this possibility. Copper metal did not adsorb TeF 6 in the presence of F 2 across the sorbent temperature range of 50 to 335°C. F 2 was fully removed by the copper bed. Although preliminary thermodynamics would indicate adsorption to be energetically favorable, other factors that impact adsorption (e.g., slow kinetics, excess fluorine on the copper surface, an unfavorable transition state) are likely preventing the adsorption of TeF 6 at an easily measurable rate. The work also investigated TeF 6 production from multiple forms of Te in the temperature range of 100 to 250°C. No previous study had assessed the initial reaction rates for this process. A carefully designed study allowed determination of the activation energy for the production of TeF 6 from Te metal. The substantial amount of data collected during this study merits analysis beyond what is described here. A more in-depth kinetic analysis will be pursued. Additional analytical results will provide the ability to benchmark adsorption coefficients for TeF 6 , understand the distribution of TeF 6 within the alumina bed, and better understand the effect of F 2 partial pressure on Te fluorination. The data from this report, as supplemented by these additional analyses, will be submitted to a peer-reviewed journal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Engineered Barrier System R&D and International Collaborations – LANL (FY20): Spent Fuel and Waste Disposition

The United States Department of Energy’s Spent Fuel and Waste Disposition (SFWD) program is investigating the design and safety function of generic nuclear geologic repositories in a variety of geologic settings (salt, argillite, and crystalline rock). Different configurations and loadings of spent nuclear fuel and waste within disposal canisters are also being investigated, some of which have the potential to generate repository temperatures higher than previously considered (i.e., temperatures >100ºC) by foreign and domestic concepts. This report expands on engineered barrier material stability in a high temperature crystalline rock repository through high temperature hydrothermal experiments. Experiments were designed to develop engineered barrier system (EBS) concepts in a hightemperature crystalline environment in 1) bentonite-Grimsel Granodiorite interactions, 2) bentonite-cement reactions, and 3) interaction between waste canister materials and bentonite. Experiment results are applied to understanding long-term repository performance in terms of radionuclide isolation. One hydrothermal experiment was completed in the rocking autoclaves at LANL in FY-20: IEBS-6 (Grimsel Granodiorite + Wyoming bentonite + cured ordinary Portland cement + Grimsel Granodiorite synthetic groundwater, 250ºC/150 bar, 8 weeks). Several other experiments were planned but were delayed due to the COVID-19 pause in laboratory work. In addition, some characterization of the reaction products of experiments conducted in FY-20 was prevented by COVID-19. Mineral phase chemistry was not measured via electron microprobe analyses for IEBS-6. Quantitative X-ray diffraction results from IEBS-6 and HBT-1 were not completed. The missing analyses and discussion of the results will be included in next year’s report. New characterization that was completed includes scanning electron microscopy of reaction products from IEBS-6, quantitative X-ray diffraction results are presented from IEBS-1 through IBES-5, preliminary scanning electron microscope images and chemical analyses for IEBS-6, X-ray diffraction of the clay fraction from all experiments, measurement/imaging of mineral growth on the surface of steel coupons. Major observations pertaining to bentonite stability in a Grimsel Granodiorite environment include the stability of Na-montmorillonite at 250°C, the formation of trace CSH phases, and the formation of bentonite colloids on experiment cooling. The addition of a cured chip of Portland cement to the bentonite-Grimsel system results in slightly higher pH values and the formation of diverse secondary mineral phases that were not observed in the previous experiments (e.g., analcime, garronite, CSH phases). The new characterization efforts related to the interaction of stainless-steel coupons and bentonite clay focused on thickness and mineralogy of phases that formed at the steel surface. In Wyoming bentonite + Grimsel Granodiorite systems, newly formed minerals at the bentonitesteel coupon interface included alteration of the outermost steel edge to Fe,Cr-oxide phases, followed by Fe-rich phyllosilicates (Fe-saponite, chlorite) and interbedded Fe,Cr,Ni-sulfide phases (pentlandite). Hydrothermal experiments were completed to assess uranium-carbonate complexation at conditions relevant to high-temperature disposal. Autoclave solubility experiments were conducted at 150 to 250ºC with a range of carbonate and uranium concentrations. The experiment results were characterized via situ UV-Visible spectroscopy and synchrotron-based in situ XAS techniques. Results show a significant decrease in the stability of uranyl-carbonate complexes at temperatures above 100ºC along with the precipitation of uranium oxides. Further, at T > 200ºC, results show that uranyl-hydroxyl complexes control solubility of uranium instead of the previously predicted uranyl-carbonate species. These results are significant for understanding the mobility of uranium in the EBS, which will likely contain carbonate-rich fluids. International research efforts focused on three main areas: 1) participation in international conferences, 2) building collaborations with foreign repository programs, and 3) the initiation of an experimental program to complement the full-scale HotBENT test at the Grimsel test site. This experiment included Wyoming bentonite + low carbon steel + Grimsel Granodiorite synthetic groundwater and was run at the planned maximum temperature of the HotBENT test (200°C). Complete characterization of reaction products was hindered by disruptions to laboratory work but will be reported in the next FY. The experimental results obtained in FY-20 continue to document the wide-ranging effects of bulk composition and pressure-temperature conditions in the mineralogical and geochemical evolution of a high-temperature repository environment. Concepts developed will be used to inform models of long-term material stability in a generic crystalline rock-hosted repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Alteration of Granite Surface with the Formation of Calcium-Silicate-Hydrate by Alkaline Groundwater - 19165

Under the hyper alkaline Ca-rich condition caused by cementitious materials used for the geological disposal system, host rocks containing various silicate minerals will alter with the dissolution of the minerals, and the expansion of flow paths may accelerate the migration of radionuclides. On the other hand, the formation of secondary minerals under such condition will contribute to the confinement for radionuclides. As a typical secondary mineral, we focused on calcium-silicate-hydrate (C-S-H). In particular, this study examined the C-S-H formation on the granite surface under the hyper-alkaline Ca-rich condition by the immersion tests and the estimation based on the film theory. In the immersion tests, granite chips immersed in a Ca(OH)2 or a NaOH solution adjusted pH to 12.3 over 60 days. From the results, the Si concentration in the Ca(OH)2 solution was much lower than that in the NaOH solution, and the Ca concentration decreased in the case of Ca(OH)2 solution. These indicates the C-S-H formation around the granite surface. Furthermore, assuming that the C-S-H layer was formed on the granite surface, the depth point of the C-S-H layer, where C-S-H was formed by reacting Ca ions with silicic acid leached from granite, was estimated by the film theory. The estimation result showed the C-S-H layer would grow with the C-S-H formation at the surface contacting granite and C-S-H. These results in this study suggest that micro flow-paths in the host rocks will be clogged by the continuous growth of the C-S-H layer on the surface of silicate minerals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1995

Incineration as a Method for Resource Recovery from Inedible Biomass in a Controlled Ecological Life Support System

Resource recovery from waste streams in a space habitat is essential to minimize the resupply burden and achieve self-sufficiency. In a Controlled Ecological Life Support System (CELSS) human wastes and inedible biomass will represent significant sources of secondary raw materials necessary for support of crop plant production (carbon, water, and inorganic plant nutrients). Incineration, pyrolysis, and water extraction have been investigated as candidate processes for recovery of these important resources from inedible biomass in a CELSS. During incineration CO2 is produced by oxidation of the organic components and this product can be directly utilized by plants. Water is concomitantly produced, requiring only a phase change for recovery. Recovery of inorganics is more difficult, requiring solubilization of the incinerator ash. The process of incineration followed by water solubilization of ash resulted in loss of 35% of the inorganics originally present in the biomass. Losses were attributed to volatilization (8%) and non-water-soluble ash (27%). All of the ash remaining following incineration could be solubilized with acid, with losses resulting from volatilization only. The recovery for individual elements varied. Elemental retention in the ash ranged from 100% of that present in the biomass for Ca, P, Mg, Na, and Si to 10% for Zn. The greatest water solubility was observed for potassium with recovery of approximately 77% of that present in the straw. Potassium represented 80% of the inorganic constituents in the wheat straw, and because of slightly greater solubility made up 86% of the water-soluble ash. Following incineration of inedible biomass from wheat, 65% of the inorganics originally present in the straw were recovered by water solubilization and 92% recovered by acid solubilization. Recovery of resources is more complex for pyrolysis and water extraction. Recovery of carbon, a resource of greater mass than the inorganic component of biomass, is more difficult following pyrolysis and water extraction of biomass. In both cases, additional processors would be required to provide products equivalent to those resulting from incineration alone. The carbon, water, and organic resources of inedible biomass are effectively separated and output in usable forms through incineration.

Bubenheim, David L.↗
DOE OSTI2019

Fundamental Study on Sorption Behavior of Europium to Mica Flakes - 19116

It is well known that biotite exists with flake form in actual underground environment, dominating sorption of cationic radionuclides in a geological disposal system consisting of crystalline rock. The behavior of sorption of radionuclides will differ depending on the difference in shape, while most of the previous studies have used powder biotite. This study focused on the influence of flake structure of phyllosilicate minerals such as biotite on the diffusion and sorption processes of radionuclides, examining the sorption behaviors of Eu on biotite, phlogopite and muscovite in flake form. These flakes (5 mm x 6 mm) were prepared for the sorption experiments and the initial concentrations of the Eu ions were set to 1 mM. The sorption experiments were conducted by mixing 0.4 g of biotite flakes and 4 mL at pH 5 for 7 days. As a result, the sorption amount of Eu increased little by little with internal diffusion in all minerals, respectively. The diffusion process of Eu into the layer structure could be also observed by secondary ion-microprobe mass spectrometry (SIMS) analysis. The apparent diffusion coefficient of Eu into each flake was estimated by fitting a 2-D diffusion model to the sorption data. The estimated diffusion values were clearly smaller than that of matrix diffusion in granite. Hence it was suggested that the so-called Kd-model, which is a local equilibrium model, is not always applicable to the estimation of the retardation effect (i.e., migration rate) of radionuclides in rock matrix. In other words, the migration rate might depend also on the diffusion process of radionuclide in the inside of phyllosilicate minerals in addition of the macro diffusion process in rock matrix. Besides, sorption amount and diffusion coefficient of biotite, phlogopite and muscovite were largely different. Not only the tetrahedral structure adjacent to the potassium but also the octahedral structure might affect the restraint of the interlayer ion (K ions) and the sorption behavior. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Independent Confirmatory Survey Summary and Results for the Enrico Fermi Atomic Power Plant, Unit 1, Newport, Michigan (Final Report No. 2)

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in December 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

PFAS Removal by Ion Exchange Resins: Background and Knowledge Gaps with Respect to the Hanford Site

Per- and polyfluoroalkyl substances (PFAS) have been a rising concern for the past two decades, with the United States Department of Defense and Environmental Protection Agency investing millions of dollars in research into remediation and clean-up technologies. Due to the environmental persistence, toxicity, biological uptake, and ongoing changes in both federal and state regulatory space, understanding the fate and transport of PFAS compounds has been of growing concern to the US Department of Energy (DOE). The DOE’s Hanford Site is investigating historical use of PFAS and will be doing site characterization for PFAS. Thus, PFAS have not yet been identified as a contaminant concern in regulatory documents. Based on historical records that mention the discharge of aqueous film-forming foam containing PFAS and having on-site fire stations (a risk factor for PFAS contamination), it seems likely that environmental releases of PFAS may have occurred. Pump and treat (P&T) remediation is the selected remedy for multiple groundwater contaminant plumes at Hanford. These P&T systems use ion exchange (IX) as a component of aboveground treatment, with the specific resins depending on the target contaminants. There is potential that these IX resins may be able to remove PFAS from groundwater, but investigation is needed to understand affinity/selectivity and removal capacity given the groundwater composition and the operating conditions. This report provides background on PFAS uses and chemistry, then provides a review of IX resin applications for PFAS, identifying knowledge gaps. Recommendations are provided regarding research needed to address knowledge gaps and acquire information needed to propose IX as a future PFAS remediation technology at the Hanford Site, as well as other U.S. Department of Energy sites. Generally, PFAS compounds are fluorinated substances that contain at least one fully fluorinated methyl or methylene carbon – with a few noted exceptions, any chemical with at least a perfluorinated methyl group (–CF3) or a perfluorinated methylene group (–CF2–) is a PFAS. These chemical compounds are characterized as non-biodegradable, non-reactive, non-photolytic, and hydrolysis resistant. This makes them highly recalcitrant within the environment, however polyfluoroalkyl materials are less recalcitrant as the carbon chains contain C–H bonds which are more easily broken than carbon – fluorine (C–F) bonds. The backbone carbon structures are commonly punctuated with a head group, the most well-known of them are perfluorooctanesulfonic acid and perfluorooctanoic acid, which possess a sulfonate and a carboxylate group, respectively. IX resins are marketed for the removal of PFAS from water systems and industrial water, however, the mechanism of removal is not as well understood as for anion or cation removal. A better understanding of the mechanism of removal would enable the development of IX resins that have improved specificity for PFAS removal. Four knowledge gaps were identified: 1) the effect of dissolved ions on the IX resin PFAS removal effectiveness, 2) the effect of additional primary contaminants of concern (PCOCs) or secondary contaminants of concern (SCOCs) on the effectiveness of PFAS via IX resin, 3) the mechanisms of PFAS removal from water, and 4) practical solutions to IX resin regeneration and waste disposal.

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

Phase Equilibria and Thermochemistry of Advanced Fuels: Modeling Burnup Behavior (Final Report)

Achieving the goal of developing advanced fuel concepts that meet the DOE objectives of being robust, demonstrating high performance, and are more tolerant of accident conditions than current fuel systems will require a thorough understanding of the thermophysical and thermochemical properties of the constituent materials. Non-oxide fuel systems are being explored under the Advanced Fuels Program that hold significant promise for improved performance and accident tolerance, including the uranium silicide-based system considered in the current work. Prospective cladding materials currently considered that contribute to improved accident tolerance include silicon carbide composites and ferritic alloys (Fe-Cr-Al base compositions). Thus, the effort developed thermochemical models and values, supported with targeted experiments, to evaluate the ferritic alloy and silicon carbide composite cladding systems in contrast to current zirconium alloy cladding. The developed detailed understanding will serve to aid in relatively early screening of candidate systems to avoid wasted effort, guide development of new fuel forms, and to provide a basis for predicting and modeling fuel performance. Major deliverables for the project included: Thermochemical assessment and models of phases in the U-Si and U-Si-N systems; thermochemical evaluation supported by experimental measurements of fuel-cladding interactions of silicide fuel with baseline zirconium, ceramic composite, and ferritic alloy cladding; thermochemical assessment and models of phases supported by experimental measurements for silicide fuel with key fission products provided in a dataset and reported in refereed publications. Within the project a significantly refined U-Si phase diagram was developed and reported that now includes homogeneity ranges for key phases, such at the U 3 Si 2 proposed fuel phase, and settles issues with regard to uncertainty in the stability of some phases. Computational efforts together with key experiments has determined phase formation in interactions between U 3 Si 2 and Zircaloy-4 cladding material, a ferritic FeCrAlY alloy of interest as an advanced cladding material, and silicon carbide, also of interest as a fiber-reinforced composite cladding. As expected, very significant reactions occur between U3Si2 and Zircaloy-4, with much less interaction at higher temperatures for the ferritic alloy, and finally interactions with SiC only in the region of contact. A potentially major issue is the stability of U 3 Si 2 fuel that has undergone significant burnup. The result is the loss of the uranium metal, liberating silicon, and the formation of concomitant fission product elements that either dissolve in the U 3 Si 2 phase or form independent, and possibly silicide phases. A combination of experimental determinations of phase formation of U 3 Si 2 reacted with representative fission products yttrium, gadolinium, cerium, zirconium, and molybdenum and first principles calculations has helped understand the fuel chemistry. The behavior of the U 3 Si 2 phase and the partitioning of silicon to possible fission product phases with burnup was thus determined, with significant dissolution in U 3 Si 2 of cerium, gadolinium, zirconium, and plutonium predicted along with independent phase formation of a U-Mo-Si ternary phase, yttrium silicide, and elemental selenium. It can be concluded that at significant burnup there will be a very minor amount of the U 3 Si 2 fuel phase that will decompose to a lower silicide or a uranium alloy as silicon preferentially forms a secondary phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Research and Technology Development Activities to Address the DOE-EM Environmental Mercury Challenge - 17497

Human activities have altered trace metal distributions globally. This is especially true for the trace metal mercury (Hg), a pervasive global pollutant that can be methylated to form highly toxic methylmercury (MeHg), which bio-accumulates in aquatic food webs, endangering humans and other biota. Currently there are more than 3,000 mercury-contaminated sites identified worldwide and the United Nations Environment Programme has recently highlighted the risk of this contamination to human health. The Oak Ridge Reservation (ORR) represents an example of one of these mercury-contaminated sites. Unlike other contaminants-metals, radionuclides, and organic solvents-that impact the Department of Energy Office of Environmental Management (DOE-EM) cleanup program at the ORR and other DOE sites, mercury has several unique characteristics that make environmental remediation of the Y-12 National Security Complex one of the most formidable challenges ever encountered. These distinctive physicochemical properties for mercury include the following: it is a liquid at ambient temperature and pressure; it is the only metal that bio-magnifies; and it is the only contaminant transported as a cation, as a dissolved or gaseous elemental metal (similar to an organic solvent), or as both a cation and a dissolved or gaseous elemental metal under environmental conditions. Because of these complexities, implementing cost effective and sustainable solutions that reduce mercury flux from various primary and secondary contamination sources will require linking basic science understanding and applied research advancements into Oak Ridge Office of Environmental Management's (OREM) cleanup process. Currently, DOE is investing in mercury-related research through a variety of programs, including the Office of Science-sponsored Critical Interfaces Science Focus Area, EM headquarters-sponsored Applied Field Research Initiative, OREM-sponsored Lower East Fork Poplar Creek (LEFPC) Mercury Technology Development Program, Small Business Innovative Research (SBIR), and EM's Minority Serving Institutions Partnership Program. Collectively, these multi-institutional and multidisciplinary programs are generating new tools, knowledge, and remediation approaches that will enable efficient cleanup of mercury contaminated systems locally and globally. In this talk we will highlight the progress made to date in addressing key knowledge gaps required to solve this watershed-scale conundrum. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Estimation of Sorption Behavior of Europium(III) Using Biotite Flakes - 13272

The interaction of biotite and Eu(III) (europium (III)) was examined by using secondary ion-microprobe mass spectrometer (SIMS), fluorescence emission spectrum and decay behavior of fluorescence emission spectrum in addition to the time-changes of Eu(III) and potassium ions concentrations in a solution, using the flake form samples. The results of SIMS showed that the intensity of Eu was gradually decreasing with depth, while the intensity of Eu in the case shaken for 30 days exceeded that in the case for 1 day. Furthermore, the spatial distribution of Eu(III) and potassium ions in the flake of biotite suggested that Eu ions diffuse mainly from the edges of biotite flake, while Eu ions can slightly diffuse through some small cracks existing on the flake surface far from the edges. Besides, the elution amount of potassium from the biotite flakes into a solution was proportional to the sorption amount of Eu(III). The changes nearly revealed ion exchange between these ions, while muscovite flake sample did not show such ion exchange reaction. In addition, from the time-change of Eu(III) concentration, an apparent diffusion coefficient was estimated to be 8.0x10{sup -12} m{sup 2}/s, by using two-dimensional diffusion model coupled with a film between the solid phase and the liquid phase. Furthermore, the fluorescent intensity decreased with the shaking (contacting) time. This means that Eu(III) gradually diffuses into the inside of biotite edges of the biotite flakes, after the sorption of Eu(III) in the edges. This tendency was observed also in the powder samples. The observed fluorescence decay (at 592 nm in wave length) showed almost similar curve in any samples, indicating a certain sorption form of Eu(III) onto the edges of the biotite flakes. These results mentioned above suggest that the diffusion processes through internal layer in biotite mainly control the sorption behavior of multivalent ions. Such diffusion processes affect the retardation-effects on fracture surfaces in the rock matrix, depending on the fluid flow velocity of groundwater. That is, a more reliable model considering the mass transfer in the internal layer of biotite may be required to estimate the sorption behavior of RNs with biotite which controls the whole sorption behavior of granite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗