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DOE OSTI2012

Auto shredder residue recycling: Mechanical separation and pyrolysis

Highlights: Black-Right-Pointing-Pointer In this work, we exploited mechanical separation and pyrolysis to recycle ASR. Black-Right-Pointing-Pointer Pyrolysis of the floating organic fraction is promising in reaching ELV Directive targets. Black-Right-Pointing-Pointer Zeolite catalyst improve pyrolysis oil and gas yield. - Abstract: sets a goal of 85% material recycling from end-of-life vehicles (ELVs) by the end of 2015. The current ELV recycling rate is around 80%, while the remaining waste is called automotive shredder residue (ASR), or car fluff. In Europe, this is mainly landfilled because it is extremely heterogeneous and often polluted with car fluids. Despite technical difficulties, in the coming years it will be necessary to recover materials from car fluff in order to meet the ELV Directive requirement. This study deals with ASR pretreatment and pyrolysis, and aims to determine whether the ELV material recycling target may be achieved by car fluff mechanical separation followed by pyrolysis with a bench scale reactor. Results show that flotation followed by pyrolysis of the light, organic fraction may be a suitable ASR recycling technique if the oil can be further refined and used as a chemical. Moreover, metals are liberated during thermal cracking and can be easily separated from the pyrolysis char, amounting to roughly 5% in mass. Lastly, pyrolysis can be a good starting point from a 'waste-to-chemicals' perspective, but further research should be done with a focus on oil and gas refining, in order both to make products suitable for the chemical industry and to render the whole recycling process economically feasible.

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DOE OSTI2011

SULFATE SOLUBILITY LIMIT VERIFICATION FOR DWPF SLUDGE BATCH 7B

The objective of this study was to determine a sulfate solubility limit in glass for Sludge Batch 7b (SB7b). The SB7b composition projection provided by Savannah River Remediation (SRR) on May 25, 2011 was used as the basis for formulating glass compositions to determine the sulfate limit. Additions of Na{sub 2}O to the projected sludge composition were made by the Savannah River National Laboratory (SRNL) due to uncertainty in the final concentration of Na{sub 2}O for SB7b, which is dependent on washing effectiveness and the potential need to add NaOH to ensure an acceptable projected operating window. Additions of 4, 6, and 8 wt % Na{sub 2}O were made to the nominal May 25, 2011 composition projection. An updated SB7b composition projection was received from SRR on August 4, 2011. Due to compositional similarities, no additional experimental work using the August 4, 2011 compositions was considered to be necessary for this study. Both Frit 418 and Frit 702 were included in this study. The targeted sulfate (SO{sub 4}{sup 2-}) concentrations of the study glasses were selected within the range of 0.6 to 0.9 wt % in glass. A total of 52 glass compositions were selected based on the compositional variables of Na{sub 2}O addition, Actinide Removal Process (ARP) stream addition, waste loading, frit composition, and sulfate concentration. The glasses were batched, melted, and characterized following SRNL procedures. Visual observations were recorded for each glass after it cooled and used as in indicator of sulfur retention. Representative samples of each of the glasses fabricated were subjected to chemical analysis to determine whether the targeted compositions were met, as well as to determine the quantity of sulfate that was retained after melting. In general, the measured composition data showed that there were only minor issues in meeting the targeted compositions for the study glasses, and the measured sulfate concentrations for each study glass were within 10% of the targeted values. The results for the SB7b glasses fabricated with Frit 418 showed an apparent trend of increasing sulfate retention with increasing Na{sub 2}O additions to the 5/25/11 sludge projection. This trend appears contradictory to other recent studies of sulfate retention in Defense Waste Processing Facility (DWPF) type glasses. Additional apparent contradictions to this trend were found in the data collected in the present study. Overall, the results for the SB7b sulfate study glasses with Frit 418 and the 5/25/11 projection with Na{sub 2}O additions showed that subtle changes in this complex glass composition impacted the degree of sulfate retention. These results do however provide confidence that a 0.6 wt % sulfate limit in glass is warranted for Frit 418 with the SB7b compositions evaluated in this study. The results for the SB7b glasses fabricated with Frit 702 are consistent with those of the previous SB7a study in that Frit 702 allowed for higher sulfate retention as compared to Frit 418 for the same sludge compositions. It is recommended that the DWPF implement a sulfate concentration limit of 0.6 wt % in glass for SB7b processing with Frit 418. If a higher than projected sulfate concentration is measured when SB7b processing begins (i.e., if a sulfate concentration higher than 0.6 wt % becomes necessary to achieve targeted waste loadings), DWPF should consider a transition to Frit 702. The sulfate limit could likely be raised to 0.8 wt % by transitioning to this frit. However, if DWPF considers transitioning from Frit 418 to Frit 702, additional glasses should be fabricated to confirm this higher limit due to the issues with incorrect B{sub 2}O{sub 3} concentrations for some of the glasses made with Frit 702 in this study. There are several factors other than sulfate retention that must also be carefully considered prior to changing frit compositions.

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DOE OSTI2018

Activating Effect of Aluminum Chloride in the Preparation of Sulfur Concrete from Sulfur and Silica

Using physicochemical characterization techniques and quantum-chemical calculations, we have demonstrated that aluminum chloride has an activating effect on the chemical interaction between sulfur and silica. As a result, we have developed a low-waste process for producing sulfur concrete with good physicomechanical and functional properties.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Sorbmatech{sup R}Cs, an Innovative Highly Selective Cs Sorbent for Wastewater Decontamination - 20065

The treatment of radioactive wastewater from the operating nuclear industry (energy production, reprocessing of spent fuel, decommissioning of shut-down facilities, etc.) or in case of incidental or accidental situation (such as Fukushima Daichii NPP accident) is an important issue requiring constant improvement of the technologies and techniques used to prevent radioactive releases into the environment and/or to reduce the production of secondary waste. CEA, Orano and CTI are collaborating since 2014, with the support of the French government's program 'investments for the future', to develop Sorbmatech{sup R}Cs, an innovative and patented selective sorbent (Grandjean and Barre 2014) to deal with the removal of radioactive {sup 137}Cs, one of the most hazardous isotope to be managed in nuclear waste, highly efficient either in agitated reactor or in column process. Sorbmatech{sup R}Cs sorbent consists of mesoporous silica particles loaded with potassium-copper hexacyanoferrate nanoparticles (KCu-HCF). This peculiar structure enables much faster sorption kinetics compared to denser crystalline structures, such as Silicotitanate or Zeolite sorbents, whose drawback consists in hindering the diffusion process through the material, and whose larger sorbent particles slow down the ion exchange process. The equilibrium of Sorbmatech{sup R}Cs is reached in few minutes, even at high flow rates in a fixed-bed column process (up to 10 m.h{sup -1} Darcy velocity) with low drop pressure. Moreover, Sorbmatech{sup R}Cs can be used over a wide pH range from 2 to 10 and has a very high selectivity towards {sup 137}Cs even in brine solution (saline, mineral or sea water). This new sorbent also has a K{sub d} higher than 10{sup 5} mL.g{sup -1} depending on the salinity of the wastewater, enabling a highly efficient removal of Cs trace in radioactive wastewater compared to Cs selective sorbent such as Zeolites and Silicotitanates. Its radiological resistance enables the treatment of high, medium and low activity wastewater. The CEA-Orano-CTI/ALSYS partnership leads to the development of a simple and low-cost synthetic route, which demonstrated large-scale manufacturing of the product on a batch of 100 kg. Two products have been developed and are ready for short run market introduction so far with grain size well adapted either for column processes or agitated reactors.. Finally, the industrial implementation has been studied and largely benefits from the properties of the sorbent. Indeed, its high sorption kinetics enable high flow rates resulting in a decrease of operation times, unit sizes, and volumes of secondary waste produced. Moreover, Sorbmatech{sup R}Cs is designed to be combined into a full and integrated solution, ranging from preliminary optimization of operating conditions through numerical dimensioning model, industrial implementation and treatment of the wastewater (through cartridge or preconditioned columns systems developed thanks to Orano's extensive expertise), to long-term stabilization of generated waste (under progress). (authors)

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DOE OSTI2020

DECOVALEX-2019 (Task G 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. This document is the final report of Task G which was proposed by the Swedish Radiation Safety Authority (SSM) and coordinated by geomecon GmbH, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. Task G in DECOVALEX-2019 is primarily concerned with the evolution of transmissivity throughout the lifetime of a repository for radioactive waste, and in particular of spent nuclear fuel, in sparsely fractured and competent crystalline rock masses, and the representation of these processes via numerical simulation. The numerical simulations were to be evaluated for their suitability, and to be validated. Over the course of the work presented here, not only were transmissivity changes and the best approaches to simulate those change examined, but also strategies were to be developed on how to monitor transmissivity change. Guidelines on what a control program and a monitoring system for a repository should include and be able to measure, and how this can be done in practice, was a key objective of the research project.

58 GEOSCIENCES↗

DOE OSTI2019

Development of Alternative Vacuum Pumps for Tritium Service

In tritium process systems, vacuum pumps are typically used to evacuate volumes and piping, as well as transfer gas to other parts of the process. This was done using the combination of an all-metal scroll pump with a metal bellows backing pump. The all-metal scroll pump, manufactured by Normetex, has been unavailable since 2012, and efforts continue to find a suitable replacement. The main obstacle is finding a pump that has no oils or polymer components, which degrade when exposed to tritium and introduce corrosive and/or hazardous impurities into the process. Since turbomolecular pumps are used in tritium processing, it is thought that pumps similar to the turbomolecular pumps would be of interest. The pumps would need to operate at lower rotational speeds to handle higher pressures and flow rates, which is best suited for Molecular Drag Pumps (MDP). To determine this, the vacuum pump characteristics must be determined for the MDP. This report details pump characteristics using an MDP backed by a Metal Bellows (Met-Bel) pump under static and flow conditions for various gases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

COMSOL Multiphysics Model for HLW Canister Filling

The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al 2 O 3 and Na 2 O can contribute to nepheline (generally NaAlSiO 4 ) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered. This level of precision is considered acceptable for the scoping nature of the model and the subsequent laboratory glass studies Using the model, two additional glass pouring cycles were conducted. Representative thermocouple data were plotted to show the variations between the two cycles. This provides preliminary data that will be used in laboratory experiments to determine the potential for controlling nepheline crystallization in glass by varying the glass pouring conditions.

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DOE OSTI2019

Feasibility of Using Steel Drums for the Confinement of Burned Low-Level Radioactive Waste Ashes

The selection of suitable materials to construct a facility for the final disposal of radioactive wastes (repository) is a challenge in materials science, since these facilities must be designed to provide confinement to the radionuclides for long periods of time. In the case of the low-level radioactive waste (LLRW) repository, the confinement of the species must be guaranteed for more than 300 years. In this context, the evaluation of the proposed materials must be performed under accelerated laboratory tests. One type of LLRW consists of ashes that result from the incineration of solid combustible radioactive waste. In Argentina, one of the proposed projects is to place these ashes inside steel drums and immobilized them by a cementation process. The issue to be resolved is whether the steel drums will undergo internal corrosion depending on the presence of certain deleterious species. The aim of the present work is to evaluate the corrosion susceptibility of steel drums containing simulated cemented incineration ashes. To this purpose, a special type of specimen was designed and the corrosion process was followed up for almost 4 years. It was found that, although the deleterious effect of chloride is undeniable, corrosion rates are not as high as to compromise the integrity of the steel drums for a design period of 300 years.

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DOE OSTI2011

Final Report Start-Up and Commissioning Tests on the DuraMelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants

This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Atom Probe Tomography Investigation of Clustering in Model P 2 O 5 -Doped Borosilicate Glasses for Nuclear Waste Vitrification

Atom probe tomography (APT) has been utilized to investigate the microstructure of two model borosilicate glasses designed to understand the solubility limits of phosphorous pentoxide (P 2 O 5 ). This component is found in certain high-level radioactive defence wastes destined for vitrification, where phase separation can potentially lead to a number of issues relating to the processing of the glass and its long-term chemical and structural stability. The development of suitable focused ion beam (FIB)-preparation routes and APT analysis conditions were initially determined for the model glasses, before examining their detailed microstructures. In a 3.0 mol% P 2 O 5 -doped glass, both visual inspection and sensitive statistical analysis of the APT data show homogeneous microstructures, while raising the content to 4.0 mol% initiates the formation of phosphorus-enriched nanoscale precipitates. This study confirms the expected inhomogeneities and phase separation of these glasses and offers routes to characterizing these at near-atomic scale resolution using APT.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Using polymerization, glass structure, and quasicrystalline theory to produce high level radioactive borosilicate glass remotely: a 20+ year legacy

Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in borosilicate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt a highly variable waste with some glass forming additives such as SiO 2 and B 2 O 3 in the form of a premelted frit and pour the molten mixture into a stainless steel canister. Seal the canister before moisture can enter the canister (10’ tall by 2’ in diameter) so the canister does not corrode from the inside out. Glass has also become widely used for HLW is that due to the fact that the short range order (SRO) and medium range order (MRO) found in the structure of glass atomistically bonds the radionuclides and hazardous species in the waste. The SRO and MRO have also been found to govern the melt properties such as viscosity and resistivity of the melt and the crystallization potential and solubility of certain species. Furthermore, the molecular structure of the glass also controls the glass durability, i.e. the contaminant/radionuclide release, by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to HLW waste variability. Nuclear waste glasses melt between 1050-1150°C which minimizes the volatility of radioactive components such as 99 Tc, 137 Cs, and 129 I. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models were developed based on the molecular structure of glass, polymerization theory of glass, and quasicrystalline theory of glass crystallization. These models create a glass which is durable, pourable, and processable with 95% accuracy without knowing from batch to batch what the composition of the waste coming out of the storage tanks will be. These models have operated the Savannah River Site Defense Waste Processing Facility (SRS DWPF), which is the world’s largest HLW Joule heated ceramic melter, since 1996. This unique “feed forward” process control, which qualifies the durability, pourability, and processability of the waste plus glass additive mixture before it enters the melter, has enabled ~8000 tons of HLW glass and 4242 canisters to be produced since 1996 with only one melter replacement.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Connecting particle interactions to agglomerate morphology and rheology of boehmite nanocrystal suspensions

Rheology imposes significant challenges on processing of complex suspensions such as nuclear waste slurries at the Hanford and Savannah River sites. Understanding rheology connecting to microstructures and underlying particle interactions in complex slurries is therefore important for both fundamental knowledge and practical applications. Here, we use suspensions of aluminum oxyhydroxide minerals in the form of boehmite as an analog of the radioactive waste slurry to gain physical insights on the correlation between particle interactions, microstructures, and slurry rheology. Specifically, we use a combination of Couette rheometry and small-angle scattering techniques (independently and simultaneously) to understand how the slurry microstructure changes under flow and how these structural changes manifest themselves in the bulk rheology of the suspensions. Our experiments show that the boehmite slurries are thixotropic, with the rheology and structure of the suspensions changing with increasing exposure to flow. In the slurries, particle aggregates begin as loose, system-spanning clusters, but exposure to moderate shear rates causes the aggregates to irreversibly consolidate into denser clusters of finite size. The microstructural changes directly influence the rheological properties of the slurries such as viscosity and viscoelasticity. More importantly, our study shows that solution pH affects the amount of structural rearrangement and the kinetics of the rearrangement process, with an increase in pH leading to faster and more dramatic changes in the bulk rheology. Such dynamic microstructural changes and resultant rheology are understood via correlations between particle interactions and strength of particle network, coupled with surface chemistry and anisotropic nature of particle interactions. Nearly identical structural changes are also observed in Poiseuille flow geometries, implying that the observed changes are relevant in the pipe flow conditions present during waste processing.

Weston, Javen S.↗

DOE OSTI2017

Coupled THMC models for bentonite in an argillite repository for nuclear waste: Illitization and its effect on swelling stress under high temperature

Subsurface manipulations such as those expected from the disposal of heat-emanating radioactive waste in deep repositories can induce strongly coupled Thermal (T), hydrological (H), mechanical (M) and chemical (C) processes. Adequate coupled THMC models are highly desirable or even indispensable for performance assessment of such repositories, for examples for the analysis of bentonite-based engineered barrier system (EBS) surrounding the emplaced waste. In this study, we present coupled THMC model simulations of a generic nuclear waste repository in an argillite with a bentonite-based buffer. The objective is to evaluate the chemical changes in the EBS bentonite and their effect on mechanical behavior under high temperature, attempting to shed light on whether EBS bentonite can sustain temperatures higher than 100 °C without significant impact on barrier's performance. Two scenarios were simulated for comparison: a case in which the temperature near the waste canister peaks at 200 °C and a case in which the temperature at the same spot culminate with about 100 °C. Simulations for a generic case with Kunigel-VI bentonite as backfill and Opalinus Clay as host rock were conducted for 1000 years and reported in the previous study (Zheng et al., 2015). In this paper, simulations for 100,000 years have been done for two types of bentonite-based buffer materials: Kunigel-VI and FEBEX bentonite. This enables us to evaluate how different types of bentonite behave in terms of the illitization and its impact on swelling stress and whether we can generalize these results to support decision making. The simulations show the occurrence of illitization in the bentonite buffer and the enhancement of illitization under high temperature. However, FEBEX bentonite undergoes less illitization mainly due to the higher ion concentration in pore water and the lower content of K-feldspar in the bentonite mineral composition. Moreover, the reduction of swelling stress by chemical changes is more pronounced for Kunigel-VI bentonite than for FEBEX bentonite. Furthermore, the results of our model simulations suggest that an argillite repository with a bentonite-based EBS that is similar to FEBEX bentonite could sustain temperatures much higher than 100 °C as far as illitization concerns. Model results also reveal that illitization is stabilized after about 2000 years in bentonite near the waste package, but continues in bentonite near the bentonite-argillite interface, which manifests the strong effect of geochemical interaction between EBS bentonite and host rock on long term illitization in bentonite.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

DECOVALEX-2023: Task E Final Report

This is the Task E final report for DECOVALEX-2023. Task E is focused on understanding thermal, two-phase hydrological, and mechanical (TH2M) processes, especially related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and essentially non-porous in the far field (away from excavations). Investigation of the short-term (days to years) near-field (centimeters to tens of meters) behavior of salt is important for radioactive waste disposal because this early period strongly controls the amount of brine in a salt repository. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment. The main test case used in Task E is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. The Task was divided into a series of Steps. Step 0 was an introduction to processes in salt, that included matching historical unheated brine inflow data from boreholes at WIPP and matching temperature observations during BATS heater test 1a. Step 1 included validation of models against a thermo-poroelastic analytical solution relevant to heated boreholes in salt, and two-phase flow around an excavation in salt. Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the BATS 1a heater test. There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. The key learning points from Task E have been: • Heat conduction through salt typically requires non-linear thermal conductivity (as a function of temperature), but most models do a good job matching observations, given appropriate adjustments to the applied power and some thermocouple locations. • Thermal pressurization requires coupled thermal-hydrological-mechanical (THM) responses that consider the thermal expansion of the fluid and solid phases. • Initialization of two-phase flow models around a borehole or excavation in salt are more realistically represented as “wetting up”, rather than “drying down” (i.e., the initial state after excavation is mostly dry, rather than mostly wet). • The BATS 1a heater test includes a significant release of brine after the end of heating, which requires a large increase in permeability to recreate. Task E has been a great learning experience for all the teams involved, and feedback from the modeling teams has led to changes in the design of follow-on BATS experiments, which are now ongoing underground at WIPP. There was a balance throughout the task between freedom to model phenomena how each team saw fit, and prescriptiveness in problem design to bring the modeling teams closer together to allow attribution of smaller differences between models to different modeling choices. The modeling approaches seem to go through two phases: an early phase of discovery or testing, and a later phase of refinement and improvement. In future modeling efforts, different field data could be used (e.g., BATS 2) and more time should be included in the processes for teams to make multiple model refinement or even significant changes to their conceptual model or setup, based on lessons learned from the modeling exercise.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Synthesis of Bismuth Functionalized Grapheme Oxide for Radioactive Iodine Gas Removal - 16114

Released radioactive iodine from nuclear reactors and nuclear severe accidents, such as those experienced at Chernobyl and Fukushima, is one of the most hazardous materials and can affect human health. Uptake of radioactive iodine from gaseous and aqueous waste, consisting of mainly organic forms such as methyl iodide and elemental iodine after reacting with organic matter in scrubber process, is a recent concern during nuclear power plant operation. In general, Triethylenediamine (TEDA) or KI-impregnated activated carbons, silver-loaded silica aerogels, and silver-loaded zeolite are used to remove radioactive iodine compounds. However, previous research has shown TEDA and KI-impregnated activated carbons to have some limitations for use in responding to severe accidents or in pressurized-water reactors, because they are easily burned up and the iodine removal efficiency of silver-loaded sorbents is greatly decreased at high temperatures. To improve the capability of synthesized sorbent material used in iodine gas removal, highly efficient sorbents for methyl and elemental iodine removal were prepared through the functionalization of graphene oxide (GO) with bismuth. GO was synthesized from natural graphite powder (99.95%) by the modified Hummers method, and then mixed with Bi(NO{sub 3}){sub 3}.5H{sub 2}O. After developing a sorbent for the removal of iodine, the synthesized sorbent was characterized using x-ray diffraction, a scanning electron microscope/energy dispersive spectrometer, and thermogravimetric analysis/differential scanning calorimetry, and the surface area of sorbents was determined using the BET-N{sub 2} adsorption method. Methyl iodine batch experiments were conducted for 7 days and elemental iodine gas adsorption experiments were conducted at high temperature (190 deg. C) using a carrier gas (air). The final concentration of iodine in batch and off-gas scrubber solutions (0.1 M NaOH) was analyzed using inductively coupled plasma mass spectrometry. The breakthrough curve and decontamination factor values were also investigated using bismuth-functionalized GO and the results were compared with Ag-zeolite, Sn-apatite, and activated carbon. (authors)

77 NANOSCIENCE AND NANOTECHNOLOGY↗

DOE OSTI2015

Radiological data consolidation and 3D modeling of a building - 15473

A portion of a building at an industrial facility that conducted monazite sands-processing operations in the 1950's under contract to the Atomic Energy Commission (AEC) became contaminated. Subsequently the U.S. Department of Energy designated the facility for remedial action under the Formerly Used Sites Remedial Action Program (FUSRAP). Remedial investigation, remedial actions and post-remediation surveys have been conducted in portions of the building. Radiological data for the interior of the building were compiled, evaluated and presented in a three dimensional (3-D) model (the model) created by AMEC Foster Wheeler in order to identify areas that meet the Record of Decision (ROD) cleanup criteria, and identity areas that may require remedial action or may require additional investigation due to data gaps. The data from different phases of investigation and remediation cover different portions of the building and in some cases overlap. In addition, the data were collected over a period of more than ten years, for different objectives, and thus the instruments and methods for establishing efficiency were not necessarily consistent. The data consolidation process included determining the relative quality of data points to include in the model and the development of rationale associated with assigning a quality level (i.e. ranking) to each data point. The rationale for ranking data quality included, as the primary determining factors, chronological order of the data (most recent was ranked as more reliable), certainty of the sample locations within the building, and analytical data quality. A 3-D model of the building was created using a laser point cloud scan which was then converted into a 3-D design model. Sample locations were then entered into the model and linked with the analytical database to allow display of the radiological results in the model. The visual model allows both strategic and tactical decisions to be made, depending on the level of detail examined. The evaluation process allowed maximum use of the historical data set, rather than requiring older data to be disregarded, and provided at a minimum a qualitative sense of data quality/confidence in the data. The (authors)process also highlights the importance of collecting high quality data from the very beginning with a consistent locational coordinate system, and the need for a good visualization model in order to assess contaminant locations during remedial action or post-remediation surveys.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Waste Feed Delivery Planning at Hanford - 13232

The Integrated Waste Feed Delivery Plan (IWFDP) describes how waste feed will be delivered to the Waste Treatment and Immobilization Plant (WTP) to safely and efficiently accomplish the River Protection Project (RPP) mission. The IWFDP, which is integrated with the Baseline Case operating scenario, is comprised of three volumes. Volume 1 - Process Strategy provides an overview of waste feed delivery (WFD) and describes how the WFD system will be used to prepare and deliver feed to the WTP based on the equipment configuration and functional capabilities of the WFD system. Volume 2 - Campaign Plan describes the plans for the first eight campaigns for delivery to the WTP, evaluates projected feed for systematic issues, projects 242-A Evaporator campaigns, and evaluates double-shell tank (DST) space and availability of contingency feed. Volume 3 - Project Plan identifies the scope and timing of the DST and infrastructure upgrade projects necessary to feed the WTP, and coordinates over 30 projectized projects and operational activities that comprise the needed WFD upgrades. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Project Plan, Status, and Lessons Learned for the LANL 3,706 m{sup 3} TRU Waste Campaign - 13085

The Los Alamos National Laboratory (LANL) is currently engaged in a campaign to disposition 3,706 m{sup 3} of transuranic (TRU) waste stored above grade at its Technical Area 54 (TA-54) Area G waste management facility before June 30, 2014. This campaign includes complete removal of all non-cemented above-grade TRU waste that was in storage on October 1, 2011, and is defined as 3,706 m{sup 3} of material. TRU waste containers were placed into storage up to 40 years ago, and most of the older containers must be remediated to address compliance issues before the waste can be characterized, certified as meeting the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), and shipped for disposition. More than half of the remaining TRU waste volume stored above grade is contained within oversize boxes that contain waste items that must be repackaged or size reduced. Facilities and major types of equipment needed to remediate and characterize the TRU waste inventory include two additional oversize box processing lines that are being brought into service as Nuclear Hazard Category III facilities in fiscal year (FY) 2013. Multiple work shifts are scheduled for most remediation lines in FY 2013. An integrated risk-based project management schedule for all disposition activities has been developed that is based on a 'Solution Package' approach. Inventories of containers that have issues in common were compiled into about 15 waste categories and about 75 'Solution Packages' that identify all of the activities needed to disposition the inventory of TRU waste in storage. Scheduled activities include all precursor activities to begin remediation, remediation processing, characterization and certification to the WIPP WAC, and shipping of containers to WIPP. Other industrial processing practices that have been adopted to improve efficiency include staging of containers for remediation, characterization, and shipping; establishment of a transportation center; and load management practices for transportation payloads. Progress and accomplishments during FY 2012 are reviewed, and plans for FY 2013 are presented in some detail. Lessons learned on adoption of industrial processing practices are also discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗