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DOE OSTI2013

Solidification of spent ion exchange resins into the SIAL matrix at the Dukovany NPP, Czech Republic

Based on the decision of the State Office for Nuclear Safety, the Dukovany NPP has been obliged to secure the efficient capacities for the disposal of spent ion exchange resins. Therefore, in September 2010, based on the contract with supplier company AMEC Nuclear Slovakia s.r.o. has begun with pumping and treatment of ion exchange resins from the storage tank 0TW30B02, situated in the auxiliary building. The SIAL{sup R} technology, developed in AMEC Nuclear Slovakia, has been used for the solidification purposes. This technology allows an on-site treatment of various special radioactive waste streams (resins, sludge, sludge/resins and borates) at the room temperature. The SIAL{sup R} matrix and technology were licensed by the Czech State Office for Nuclear Safety in 2007. On-site treatment and solidification of spent ion exchange resins at Dukovany NPP involves process of resin removal from tank using remotely operated manipulator, resin transportation, resin separation from free water, resin filling into 200 dm{sup 3} drums and solidification into SIAL{sup R} matrix in 200 dm{sup 3} drums using the FIZA S 200 facility. The final product is observed for compressive strength, leachability, radionuclide composition, dose rate, solids and total weight. After meeting the requirements for final disposal and consolidation, the drums are being transported for the final disposal to the Repository at Dukovany site. During the 3 month's trial operation in 2010, and the normal operation in 2011 and 2012, 189 tons of dewatered resins have been treated into 1960 drums, with total activity higher than 920 GBq. At the end of trial run (2010), 22 tons of dewatered resins were treated into 235 drums. During standard operation approximately 91 tons in 960 drums (2011) and 76 tons in 765 drums (2012) were treated. The weights of resins in the drum ware in the range from 89 - 106 kg and compressive strength limit (10 MPa) has already been achieved 24 hours after fixation. The final measured strength values ranged from 19.0 - 34.7 MPa and real leachability values ranged from 0.03 - 0.65%, far below the 4% limit value. Collective effective dose of all workers in 2012 was 7.7 mSv (12.6 mSv in 2011, 6.2 mSv in 2010). Average individual effective dose in 2012 was 0.55 mSv (14 workers), and maximal individual effective dose was 2.25 mSv. This approach allows fast, safe and cost effective immobilization and transformation of dangerous radioactive waste such as sludge and resins into the solid form, which is suitable for long term storage or disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Synthesis and Characterization of Tc(I) Carbonyl Nitrosyl Species Relevant to the Hanford Tank Waste: FY 2016 Status Report

Among long-lived radioactive constituents in the Hanford tank waste, Tc presents a unique challenge in that it exists predominantly in the liquid phase, generally in the anionic form of pertechnetate, TcO 4 - , which is highly volatile at low-activity waste (LAW) vitrification melter temperatures and mobile in the Hanford site’s subsurface environment. The complex behavior of Tc under storage, treatment, and immobilization conditions significantly affects its management options, which to-date remain uncertain. In strongly alkaline environments, Tc exists as pertechnetate, TcO 4 - (oxidation state +7), and in the reduced forms (oxidation state < +7) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized, anionic Tc species that can be removed from LAW by anion exchange or solvent extraction methods. There is no definitive information on the origin of the non-pertechnetate Tc species, nor is there a comprehensive description of their composition and behavior. It has been recently proposed that the non-pertechnetate species can comprise Tc(I) metal center and carbonyl or mixed carbonyl nitrosyl ligands stabilizing low-valent Tc. Recent work by our group has significantly expanded this previous work, generating a series of Tc(I) carbonyl compounds and demonstrating that they can be generated from reduction of TcO 4 - in the simulated Hanford tank waste in presence of CO at elevated temperature (Levitskaia et al. 2014). These results are consistent with the previous proposal that [Tc(CO) 3 ] + species can be present in the Hanford tank waste and suggest that the low Tc(I) oxidation state is stabilized by the π-accepting ability of the CO ligands. The continuation work has been initiated to develop model Tc carbonyl nitrosyl compounds and investigate their potential presence in the Hanford tank wastes. This report summarizes our to-date results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Chloride-Based Volatility for Waste Reduction and/or Reuse of Metallic-, Oxide- and Salt-Based Reactor Fuels

The objective of the chloride based volatility project (CBV) was to demonstrate the ability to separate uranium from used fuel to enable process improvements resulting in 10x reduction of waste volume, while maintaining safeguards standards and global backend costs at $\$$1/MW-hr. Current industrial practices perform separations of used fuel using solvent media in the form of aqueous and molten salt processes, resulting in contaminated process waste. The CBV approach utilized solid state chemistry with no solvent media and was successful in chlorinating uranium and fission product oxides that sublimed into the gas phase and were collected in targeted condensation zones based on temperature gradients of chlorinated products. Recovery of better than 95% of initial uranium in the form of UCl4 was demonstrated when simulated used nuclear fuel was used. Laser induced breakdown spectroscopy, LIBS, and ultraviolet-visible spectroscopy, UV-Vis, were combined into a high temperature flow cell design and utilized as process monitoring techniques to observed chlorinated products leave the reaction vessel in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Performance of the Fluidized Bed Steam Reforming product under hydraulically unsaturated conditions

Several candidates for supplemental low-activity waste (LAW) immobilization at the Hanford site in Washington State, USA are being considered. One waste sequestering technology considered is Fluidized Bed Steam Reforming (FBSR). The granular product resulting from the FBSR process is composed primarily of an insoluble sodium aluminosilicate matrix with the dominant phases being feldspathoid minerals with a 1:1:1 molar ratio of Na, Al and Si. To demonstrate the durability of the product, which can be disposed of at the unsaturated Integrated Disposal Facility (IDF) at Hanford, a series of tests has been performed using the Pressurized Unsaturated Flow (PUF) system, which allows for the accelerated weathering of the solid materials. The system maintains hydraulically unsaturated conditions, thus mimicking the open-flow and transport properties that will be present at the IDF. Two materials were tested using the system: 1) the FBSR granular product and 2) the FBSR granular product encapsulated in a geopolymer to form a monolith. Results of the experiments show a trend of relatively constant effluent concentration of Na, Si, Al, and Cs as a function of time from both materials. The elements I and Re show a steady release throughout the yearlong test from the granular material but their concentrations seem to be increasing at one year from the monolith material. This result suggests that these two elements may be present in the sodalite cage structure rather than in the predominant nepheline phase because their release occurs at a different rate compared to nepheline phase. Also, these elements to not seem to reprecipitate when released from the starting material. Calculated one-year release rates for Si are on the order of 10 6 g/(m2 d) for the granular material and 10 5 g/(m2 d) for the monolith material while Re release is seen to be two orders of magnitude higher than Si release rates. SEM imaging and XRD analysis show how the alteration of the two materials is dependent on their depth in the column. This phenomenom is a result of depth-dependent solution concentrations giving rise chemical environments that may be supersaturated with respect to a number of mineral phases.

Fluidized Bed Steam Reforming↗

DOE OSTI2018

Graphene oxide supported tin dioxide: synthetic approaches and electrochemical characterization as anodes for lithium- and sodium-ion batteries

The review addresses synthetic approaches to composite materials based on graphene oxide and nano tin dioxide and their electrochemical properties as anodes for lithium- and sodiumion batteries. The introduction of a carbon matrix into the composite material improves the electrochemical characteristics of the anodes. In most methods, the synthesis of graphene oxide–tin dioxide composites is based on the use of tin(II,IV) chlorides as the starting compounds, and the most efficient electrode materials were obtained by the hydrothermal or solvothermal routes. Thermal processing is much more economic than the gas phase deposition protocols but requires heating of a large volume of dilute tin oxide dispersions in an autoclave. Mechanochemistry (ball milling) is also economically unfavorable for the synthesis of composite materials. In addition, large volumes of acidic wastes that should be neutralized and safely discarded are formed when tin chlorides are used. An alternative environmentally friendly technique based on the use of aqueous peroxide solutions can be applied for the production of efficient anode materials based on graphene oxide and tin dioxide. This process does not involve acidic wastes, uses hydrogen peroxide and ethanol as reagents, and accomplishes film deposition (coating) at room temperature. Final thermal treatment is required only for the active material, which minimizes energy expenses and equipment costs.

25 ENERGY STORAGE↗

DOE OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Synthesis of branched, nano channeled, ultrafine and nano carbon tubes from PET wastes using the arc discharge method

Highlights: • Polymer wastes are converted into ultrafine and nano carbon tubes and spheres. • Simple process with a minimal processing time. • It is a catalyst free and solvent free approach. • This process forms branched ultrafine carbon tubules with nano channels. - Abstract: Upcycling polymer wastes into useful, and valuable carbon based materials, is a challenging process. We report a novel catalyst-free and solvent-free technique for the formation of nano channeled ultrafine carbon tubes (NCUFCTs) and multiwalled carbon nanotubes (MWCNTs) from polyethylene terephthalate (PET) wastes, using rotating cathode arc discharge technique. The soot obtain from the anode contains ultrafine and nano-sized solid carbon spheres (SCS) with a mean diameter of 221 nm and 100 nm, respectively, formed at the lower temperature region of the anode where the temperature is approximately 1700 °C. The carbon spheres are converted into long “Y” type branched and non-branched NCUFCTs and MWCNTs at higher temperature regions where the temperature is approximately 2600 °C, with mean diameters of 364 nm and 95 nm, respectively. Soot deposited on the cathode is composed of MWCNTs with a mean diameter of 20 nm and other nanoparticles. The tubular structures present in the anode are longer, bent and often coiled with lesser graphitization compared to the nanotubes in the soot on the cathode.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1993

Reclamation with Recovery of Radionuclides and Toxic Metals from Contaminated Materials, Soils, and Wastes

A process has been developed at Brookhaven National Laboratory (BNL) for the removal of metals and radionuclides from contaminated materials, soils, and waste sites. In this process, citric acid, a naturally occurring organic complexing agent, is used to extract metals such as Ba, Cd, Cr, Ni, Zn, and radionuclides Co, Sr, Th, and U from solid wastes by formation of water soluble, metal-citrate complexes. Citric acid forms different types of complexes with the transition metals and actinides, and may involve formation of a bidentate, tridentate, binuclear, or polynuclear complex species. The extract containing radionuclide/metal complex is then subjected to microbiological degradation followed by photochemical degradation under aerobic conditions. Several metal citrate complexes are biodegraded, and the metals are recovered in a concentrated form with the bacterial biomass. Uranium forms binuclear complex with citric acid and is not biodegraded. The supernatant containing uranium citrate complex is separated and upon exposure to light, undergoes rapid degradation resulting in the formation of an insoluble, stable polymeric form of uranium. Uranium is recovered as a precipitate (polyuranate) in a concentrated form for recycling or for appropriate disposal. This treatment process, unlike others which use caustic reagents, does not create additional hazardous wastes for disposal and causes little damage to soil which can then be returned to normal use.

Francis, A. J.↗

DOE OSTI2015

Hazardous waste status of discarded electronic cigarettes

Highlights: • Electronic cigarettes were tested using TCLP and WET. • Several electronic cigarette products leached lead at hazardous waste levels. • Lead was the only element that exceeded hazardous waste concentration thresholds. • Nicotine solution may cause hazardous waste classification when discarded unused. - Abstract: The potential for disposable electronic cigarettes (e-cigarettes) to be classified as hazardous waste was investigated. The Toxicity Characteristic Leaching Procedure (TCLP) was performed on 23 disposable e-cigarettes in a preliminary survey of metal leaching. Based on these results, four e-cigarette products were selected for replicate analysis by TCLP and the California Waste Extraction Test (WET). Lead was measured in leachate as high as 50 mg/L by WET and 40 mg/L by TCLP. Regulatory thresholds were exceeded by two of 15 products tested in total. Therefore, some e-cigarettes would be toxicity characteristic (TC) hazardous waste but a majority would not. When disposed in the unused form, e-cigarettes containing nicotine juice would be commercial chemical products (CCP) and would, in the United States (US), be considered a listed hazardous waste (P075). While household waste is exempt from hazardous waste regulation, there are many instances in which such waste would be subject to regulation. Manufactures and retailers with unused or expired e-cigarettes or nicotine juice solution would be required to manage these as hazardous waste upon disposal. Current regulations and policies regarding the availability of nicotine-containing e-cigarettes worldwide were reviewed. Despite their small size, disposable e-cigarettes are consumed and discarded much more quickly than typical electronics, which may become a growing concern for waste managers.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Evaluation of GlassNet for physics-informed machine learning of glass stability and glass-forming ability

Glassy materials form the basis of many modern applications, including nuclear waste immobilization, touch-screen displays, and optical fibers, and also hold great potential for future medical and environmental applications. However, their structural complexity and large composition space make design and optimization challenging for certain applications. Of particular importance for glass processing and design is an estimate of a given composition's glass-forming ability (GFA). However, there remain many open questions regarding the underlying physical mechanisms of glass formation, especially in oxide glasses. It is apparent that a proxy for GFA would be highly useful in glass processing and design, but identifying such a surrogate property has proven itself to be difficult. While glass stability (GS) parameters have historically been used as a GFA surrogate, recent research has demonstrated that most of these parameters are not accurate predictors of the GFA of oxide glasses. Here, in this work, we explore the application of an open-source pre-trained neural network model, GlassNet, that can predict the characteristic temperatures necessary to compute GS with reasonable performance and assess the feasibility of using these physics-informed machine learning (PIML)-predicted GS parameters to estimate GFA. In doing so, we track the uncertainties at each step of the computation—from the original ML prediction errors to the compounding of errors during GS estimation, and finally to the final estimation of GFA. While GlassNet exhibits reasonable accuracy on all individual properties, we observe a large compounding of error in the combination of these individual predictions for the PIML prediction of GS, finding that random forest models offer similar accuracy to GlassNet. We also break down the performance of GlassNet on different glass families and find that the error in GS prediction is correlated with the error in crystallization peak temperature prediction. Lastly, we utilize this finding to assess the relationship between top-performing GS parameters and GFA for two ternary glass systems: sodium borosilicate and sodium iron phosphate glasses. We conclude that to obtain true ML predictive capability of GFA, significantly more data needs to be collected.

36 MATERIALS SCIENCE↗

DOE OSTI2017

A conceptual framework for negotiating public involvement in municipal waste management decision-making in the UK

Highlights: • A review of public involvement in waste management decision-making was undertaken. • Public involvement seen as necessary to create stronger foundations for decisions. • The mode and level of public involvement explored in different decision contexts. • Public involvement can be negotiated based on the technological and social context. • Early public involvement favoured where issues are contentious or uncertain. - Abstract: The technical expertise that politicians relied on in the past to produce cost-effective and environmentally sound solutions no longer provides sufficient justification to approve waste facilities. Local authorities need to find more effective ways to involve stakeholders and communities in decision-making since public acceptance of municipal waste facilities is integral to delivering effective waste strategies. This paper presents findings from a research project that explored attitudes towards greater levels of public involvement in UK waste management decision-making. The study addressed questions of perception, interests, the decision context, the means of engagement and the necessary resources and capacity for adopting a participatory decision process. Adopting a mixed methods approach, the research produced an empirical framework for negotiating the mode and level of public involvement in waste management decision-making. The framework captures and builds on theories of public involvement and the experiences of practitioners, and offers guidance for integrating analysis and deliberation with public groups in different waste management decision contexts. Principles in the framework operate on the premise that the decision about ‘more’ and ‘better’ forms of public involvement can be negotiated, based on the nature of the waste problem and wider social context of decision-making. The collection of opinions from the wide range of stakeholders involved in the study has produced new insights for the design of public engagement processes that are context-dependent and ‘fit-for-purpose’; these suggest a need for greater inclusivity in the case of contentious technologies and high levels of uncertainty regarding decision outcomes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2004

Properties of a Novel Ion-Exchange Film

A new ion-exchange material (based on polyacrylic acid) and some of its analytical applications have been reported. This paper contains data on the ion-exchange properties of the film form of the material and its potential application to the decontamination of waste water and drinking water. The film has a high exchange capacity of 5 to 6 meq/g and a pK(sub a) of 5.7. The calcium form is the most effective for removing metal ions from solution, and the optimum pH range is between 5 and 7. The exchange rates are slower for the film than for bead and powder forms of the ion-exchange material; otherwise, the properties are similar. The film is effective when hard water solutions are employed and also when metal ions are in the complex matrix of waste water from electroplating. The film can be used in flow systems having a flow channel large enough to allow passage of turbid solutions.

Street, Kenneth W.↗

NASA NTRS2002

Properties of a Novel Ion-Exchange Film

Street, Kenneth W.↗

DOE OSTI2017

Reconstruction of radionuclide intakes for the residents of East Urals Radioactive Trace (1957–2011)

The East Urals Radioactive Trace (EURT) was formed after a chemical explosion in the radioactive waste-storage facility of the Mayak Production Association in 1957 (Southern Urals, Russia) and resulted in an activity dispersion of 7.4 × 10 16 Bq into the atmosphere. Internal exposure due to ingestion of radionuclides with local foodstuffs was the main factor of public exposure at the EURT. The EURT cohort, combining residents of most contaminated settlements, was formed for epidemiological study at the Urals Research Center for Radiation Medicine, Russia (URCRM). For the purpose of improvement of radionuclide intake estimates for cohort members, the following data sets collected in URCRM were used: (1) Total β-activity and radiochemical measurements of 90 Sr in local foodstuffs over all of the period of interest (1958–2011; n = 2200), which were used for relative 90 Sr intake estimations. (2) 90 Sr measurements in human bones and whole body ( n = 338); these data were used for average 90 Sr intake derivations using an age- and gender-dependent Sr-biokinetic model. Non-strontium radionuclide intakes were evaluated on the basis of 90 Sr intake data and the radionuclide composition of contaminated foodstuffs. Validation of radionuclide intakes during the first years after the accident was first carried out using measurements of the feces β-activity of EURT residents ( n = 148). The comparison of experimental and reconstructed values of feces β-activity shows good agreement. 90 Sr intakes for residents of settlements evacuated 7–14 days after the accident were also obtained from 90 Sr measurements in human bone and whole body. Furthermore, the results of radionuclide intake reconstruction will be used to estimate the internal doses for the members of the EURT cohort.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2014

Technologies for Production of Heat and Electricity

Biomass is a desirable source of energy because it is renewable, sustainable, widely available throughout the world, and amenable to conversion. Biomass is composed of cellulose, hemicellulose, and lignin components. Cellulose is generally the dominant fraction, representing about 40 to 50% of the material by weight, with hemicellulose representing 20 to 50% of the material, and lignin making up the remaining portion [4,5,6]. Although the outward appearance of the various forms of cellulosic biomass, such as wood, grass, municipal solid waste (MSW), or agricultural residues, is different, all of these materials have a similar cellulosic composition. Elementally, however, biomass varies considerably, thereby presenting technical challenges at virtually every phase of its conversion to useful energy forms and products. Despite the variances among cellulosic sources, there are a variety of technologies for converting biomass into energy. These technologies are generally divided into two groups: biochemical (biological-based) and thermochemical (heat-based) conversion processes. This chapter reviews the specific technologies that can be used to convert biomass to energy. Each technology review includes the description of the process, and the positive and negative aspects.

09 BIOMASS FUELS↗

DOE OSTI2018

Stability, Composition, and Core-Shell Particle Structure of Uranium(IV)-Silicate Colloids

Uranium is typically the most abundant radionuclide by mass in radioactive wastes and is a significant component of effluent streams at nuclear facilities. Actinide(IV) (An(IV)) colloids formed via various pathways, including corrosion of spent nuclear fuel, have the potential to greatly enhance the mobility of poorly soluble An(IV) forms, including uranium. This is particularly important in conditions relevant to decommissioning of nuclear facilities and the geological disposal of radioactive waste. Previous studies have suggested that silicate could stabilize U(IV) colloids. Here the formation, composition, and structure of U(IV)-silicate colloids under the alkaline conditions relevant to spent nuclear fuel storage and disposal were investigated using a range of state of the art techniques. The colloids are formed across a range of pH conditions (9-10.5) and silicate concentrations (2-4 mM) and have a primary particle size 1-10 nm, also forming suspended aggregates <220 nm. X-ray absorption spectroscopy, ultrafiltration, and scanning transmission electron microscopy confirm the particles are U(IV)-silicates. Additional evidence from X-ray diffraction and pair distribution function data suggests the primary particles are composed of a UO2-rich core and a U-silicate shell. U(IV)-silicate colloids formation correlates with the formation of U(OH)3(H3SiO4)3 2- complexes in solution indicating they are likely particle precursors. Finally, these colloids form under a range of conditions relevant to nuclear fuel storage and geological disposal of radioactive waste and represent a potential pathway for U mobility in these systems.

Neill, Thomas↗

DOE OSTI2025

Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

DOE OSTI2017

A Study Of Geological Disposal Facility (GDF) Ageing Phenomena In Bentonite - 17605

Since the 1950's, the UK's nuclear activity in energy, defence and medicine has led to the accumulation of a large amount of higher activity radioactive waste (HAW). Most of nuclear waste arises in a wide variety of physical and chemical forms with different radioisotopes and loadings. Due to complexity in HAW, it is imperative that there be a rigorous management system that ensures population and environmental safety over extended periods. The concept of Deep Geological Disposal of waste in a stable geological formation is, nowadays, an internationally accepted solution as the surest and the most viable long term method for management. Any future Geological Disposal Facility (GDF) will comprise, among other artificial and natural engineered barriers, a barrier immediately around the waste containers compounded by a sealing material. The long-term stability of a GDF will be determined by the design and construction of the engineered barriers, and the mechanical, hydraulic and geochemical interactions that may happen between the different engineered components as a result of combined processes such as the heat produced by radioactive decay and the ion water composition of the surrounding host rock. Bentonite clay has been selected as the most promising buffer and backfill material in many disposal programmes because of its swelling capacity, low permeability, plasticity and cation exchange capacity. In the nuclear scope, bentonite testing at different scale-lengths (laboratory and field scale) has been performed and implemented in order to examine the thermo-hydro-mechanical (THM) and thermo-hydro-geochemical (THG) processes that may happen in the engineered barriers and geological settings associated with a GDF. Two major European field scale tests have been performed to better understand bentonite performance and behaviour immediately after HAW emplacements. The Full-Scale Engineered Barriers Experiment (FEBEX) at the Grimsel Test Sit (Switzerland) and the Alternative Buffer Material (ABM) Project at the Aespoe Hard Rock Laboratory (Sweden). The Febex Experiment was a pilot project designed for 'field research' of HAW repository hosted in crystalline rock where surrogate HAW canisters were placed horizontally in drifts, surrounded by compacted blocks of bentonite, sealed by a concrete plug and maintaining a constant canister surface temperature of 100 deg. C. After five years of operation, a partial dismantling was carried out in 2002. Heater processes occurring in the surrounding bentonite. The dismantling of the heater no. 2 (after 18 years of operation) started in late 2014 and is planned to be finished at the end of 2016 (termed FEBEX-DP). The FEBEX Project utilised bentonite from the Cortijo de Archidona deposit (Almeria, Spain). This 'Febex' bentonite is sourced from calcium-magnesium montmorillonite with good properties in relation to swelling capacity, permeability, adsorption capacity, and thermal conductivity. The ABM Project is a large-scale programme that has been running since 2006. The programme is designed to study a suite of possible buffer materials and correlate the physical and chemical properties of eleven different types of bentonites to their fundamental mineralogical properties. The ABM test consists in three 'packages' containing a central heater tube surrounded by different pre-compacted bentonite blocks and instrumentation. Each 'package' was installed in separate boreholes. The target temperature for the surface of the heated tubes in contact with the bentonite was 130 deg. C. Both FEBEX and ABM test the bentonite was saturated with water percolating from surrounding host rock (FEBEX and ABM) and also via artificial supply systems (ABM), being the water itself acting as a long term corrosion agent. For both experiments and emerging interest is in the interaction of bentonite with metallic components, under geological disposal conditions are not considered to be thermodynamically stable and hence arising corrosion products could interact with the buffer. These interactions depend on various factors such as the redox potential, temperature, pH, solute concentration and buffer material characteristics. During the operational phases of the FEBEX and ABM and after its 'closure' a significant volume of air was found to be retained in the bentonite. Under these conditions a period of initial aerobic corrosion by the air in the bentonite barrier would have occurred followed by establishment of anaerobic corrosion would be the dominant process ongoing. The anaerobic corrosion began when the bentonite became water saturated. Under these conditions, the water acted as the major oxidant. The present project focuses on the characterization of samples arising from the FEBEX and ABM Projects as well as performing additional independent experiments designed to complement these prior studies. Characterization will focus mainly on corrosion related alteration combined with thermal effects. Through detailed sample characterization, the intention is to achieve a greater knowledge regarding the changes to the bentonite under the experimental conditions to which it was exposed, and plausible alterations in its THM and THG properties occurring during the experiment due to blended effects of temperature,water content, engineering structural components and solutes are also to be measured. The fulfilment of a large-scale 'in-situ' test is intricate, time-consuming and with an abundance of external variables that are not always well controlled. For this reason, the University of Bristol is performing a series of well constrained time-resolved laboratory tests which bentonite is being subjected to different possible scenarios that may happen whilst in a HAW repository. In virtue of this experiment, the UoB aims to understand the illitisation and cementation process and their effects on bentonite safety properties. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗