Raul B. Rebak GE Global Research
Neil C. Hyatt The University of Sheffield
David A. Pickett Southwest Research Institute
Q1: National Programmes and Advanced Fuel Cycles
Monday PM, December 01, 2008
Back Bay D (Sheraton)
10:00 AM - **Q1.1
Long-Term Peformance of the Proposed Yucca Mountain Repository, USA.
Peter Swift 1 Show Abstract
1 , Sandia National Laboratories, Albuquerque , New Mexico, United States
In its role as the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Lead Laboratory for Repository Systems, Sandia National Laboratories (SNL) has completed a quantitative assessment of the long-term performance of the proposed Yucca Mountain repository for spent nuclear fuel and high-level radioactive waste. This performance assessment is based on more than two decades of scientific investigations of the engineered and natural systems that comprise the disposal system, and provides estimates of radiation releases from the disposal system and mean annual radiation doses to a hypothetical “reasonably maximally exposed individual” for one million years following closure of the facility, considering releases from all pathways. This presentation will summarize the technical basis for the performance assessment and review the results presented to the US Nuclear Regulatory Commission (NRC) as part of the demonstration of compliance with applicable long-term radiation standards. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The statements expressed in this article are those of the authors and do not necessarily reflect the views or policies of the United States Department of Energy or Sandia National Laboratories.
10:30 AM - **Q1.2
Integration of Postclosure Safety Analysis with Repository Design for the Yucca Mountain Repository through the Selection of Design Control Parameters.
Gerald Nieder-Westermann 1 , Robert Andrews 1 , Neil Brown 2 , Robert Spencer 1 Show Abstract
1 , Bechtel SAIC Company, LLC, Las Vegas, Nevada, United States, 2 , Los Alamos National Laboratory, Las Vegas, Nevada, United States
11:30 AM - **Q1.3
A Trade Study for Waste Concepts to Minimize HLW Volume.
Dirk Gombert 1 , Tim Trickel 4 1 , Steven Piet 2 , Gretchen Matthern 1 , John Vienna 5 , William Ebert 3 Show Abstract
1 Environmental Engineering & Technology Department, Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 Nuclear Engineering, North Carolina State University, Raleigh, North Carolina, United States, 2 Reactor Physics Analysis & Design, Idaho National Laboratory, Idaho Falls, Idaho, United States, 5 Process Development, Pacific Northwest National Laboratory, Richland, Washington, United States, 3 Chemical Sciences & Engineering, Argonne National Laboratory, Chicago, Illinois, United States
12:00 PM - **Q1.4
The National Nuclear Laboratory and Collaborative University Research in the UK.
Graham Fairhall 1 Show Abstract
1 , Nexia Solutions Ltd, Sellafield United Kingdom
The UK has recently established its first National Nuclear Laboratory (NNL). This has been formed out of the Nexia Solutions organisation, formally the Research and Technology subsidiary of BNFL. Over the next year the NNL will develop so that it can fulfil its mission, which will include the development and maintenance of key skills and undertaking strategic R&D programmes both in the UK and in international collaborations. A key role of the NNL will be to enhance its interactions with Universities to facilitate skills transfer into the nuclear industry as well as support its R&D programmes. Over the past decade the NNL and its predecessor has already established close relationships with leading Universities in the UK including Sheffield, Manchester, Leeds and Imperial College London. This paper will describe the future plans for working with Universities as the NNL develops, building on the success to date.One of the key objectives of the NNL has been to work in an integrated way with University researchers with programmes spanning fundamental research through to applied R&D.The Immobilisation Science Laboratory (ISL) at Sheffield University was established by the NNL predecessors and a range of joint research has been undertaken over the past 8 years. This includes support for the UK Pu disposition programme where R&D has included investigating a range of ceramic and glass wasteforms which has allowed a number of ceramic wasteforms to be down selected for detailed evaluation. Cementation research has included understanding wasteform performance, for example long-term durability. This has involved work on determining free water and the implications for immobilisation of reactive wastes.Other work involving the NNL and Universities, in particular at Leeds and Manchester, has considered the characteristics and behaviour of intermediate and high level waste sludges. This has included determining the chemical speciation of actinides and fission products, and physical properties of active and simulated sludges using experimental and modelling techniques.A significant programme on environmental work has been undertaken by the NNL. In research applicable to low level waste disposal and contaminated land reactive transport modelling is utilised to apply experimental and field based research in environmental geochemistry and radiochemistry undertaken by university collaborators. This includes research into Sr-90 interactions with wastes and soils and synchtrotron studies of green-rust corrosion products and their potential to uptake contaminants. This paper will describe a number of examples of R&D carried out by NNL in association with its University partners.
12:30 PM - Q1.5
DIAMOND: A New Research Programme to Support UK Decomissioning, Immobilisation and Management of Nuclear Wastes for Disposal.
Neil Hyatt 1 , Simon Biggs 2 , Francis Livens 3 , Jim Young 2 Show Abstract
1 Engineering Materials, University of Sheffield, Sheffield United Kingdom, 2 School of Process, Environmental & Materials Engineering, University of Leeds, Leeds United Kingdom, 3 Department of Chemistry, University of Manchester, Manchester United Kingdom
Q2: Spent Nuclear Fuel
Monday PM, December 01, 2008
Back Bay D (Sheraton)
2:30 PM - **Q2.1
Key Scientific Issues Related to the Sustainable Management of the Spent Nuclear Fuel in the Back-end of the Fuel Cycle.
Christophe Poinssot 1 , Jean-Marie Gras 2 Show Abstract
1 Department of RadioChemistry and Processes, CEA, Nuclear Energy Division, Bagnols / Ceze France, 2 R&D division, EDF, Electricité de France, Moret-Sur-Loing France
Direct geological disposal has for a long time being considered in many countries as a reference solution to ultimately manage Spent Nuclear Fuel (SNF). However, the recent concerns about the global climate change and the strong increase of energy demands in the world leads to a remarkable renaissance of nuclear energy for a few years with anticipated new reactor and plant construction. In this new context, resources have to be preserved and recycled as far as possible. Directly disposing SNF in deep underground may not be the most sustainable solution since 96% is still recyclable (U, Pu, minor actinides). Recycling part of the actinides is therefore an option of growing interest for many countries instead of direct disposal. This significant policy evolution has obviously to influence the relative importance of the different scientific research fields, in particular regarding SNF. This papers aims to depict the scientific key issues related to the different options considered for managing spent nuclear fuel, direct disposal and recycling.Studies on SNF long term evolution in direct disposal has significantly developed in the last decade. They allow deriving reliable and scientifically-sounded radionuclides source term to be used in safety analyses, especially for the Instant Release Fraction and the radiolytic dissolution which is more deeply understood. However, key scientific areas are still to be addressed, in particular on the IRF inventory and secondary phases precipitation.Regarding the closed cycle scenario, spent nuclear fuel is supposed to be ultimately digested in highly acidic conditions. SNF dissolution properties have to be significantly improved. However, SNF is also probably to be stored for some time in order to optimise the global U and Pu stockpiles in the whole cycle (to avoid any Pu accumulation). In long term storage, questions regarding long term cladding performance are of importance for the safety demonstrations. In particular, long term creep and rupture criteria in dry storages, as well as cladding embrittlement and potential rupture in wet storage (French choice) have to be accurately defined. Partitioning actinides is also a field of significant research area in order to improve the robustness and proliferation-resistance of the current processes and allow future recovering of minor actinides (MA). After partitioning of U, Pu +/- MA, long-lived radionuclides are confined by a dedicated matrix, the nuclear glass. Its performance has been proven to last long enough (> 300ky.) to allow a safe disposal in reducing deep geological environment. Main improvements are currently anticipated in the field of (i) the confinement capabilities of nuclear glass (higher alpha loading …) and (ii) elucidation of glass dissolution / environment interactions in the geological disposal. In conclusion, this paper will focus the main scientific inputs which are needed to support the global optimisation of the SNF sustainable cycle.
3:00 PM - Q2.2
Influence of the Evolution of the Surface Area Value on the Spent Nuclear Fuel Dissolution Rate for Performance Assessment Studies.
Eduardo Iglesias 1 , Javier Quinones 1 , Juan Manuel Nieto 1 , Nieves Rodriguez 1 Show Abstract
1 , CIEMAT, Madrid Spain
3:15 PM - Q2.3
UO2 Corrosion in an Iron Waste Package.
Elizabeth D. A. Ferriss 1 , Katheryn Helean 2 , Charles Bryan 2 , Patrick Brady 2 , Rodney Ewing 1 Show Abstract
1 geological sciences, University of Michigan, Ann Arbor, MI, Michigan, United States, 2 , Sandia National Laboratories, Albuquerque, New Mexico, United States
Understanding the corrosion of spent nuclear fuel (SNF) and the subsequent mobilization of released radionuclides, particularly under oxidizing conditions, is one of the key issues in evaluating the long-term performance of a nuclear waste repository. However, the large amounts of iron in the metal waste package may create locally reducing conditions that would lower corrosion rates for the SNF, as well as reduce the solubility of some key radionuclides, e.g., Tc, Np, and U. In order to investigate the interactions among SNF-waste package-fluids, six small-scale (~1:40 by length) waste package mockups were constructed using metals similar to those proposed for use in waste packages at the proposed repository at Yucca Mountain (YMR). Each mockup experiment differed with respect to water input, exposure to the atmosphere, and temperature, and two of the mockups contained 0.1 g of UO2. Simulated Yucca Mountain process water (YMPW) was injected into five of the mockups at a rate of 200 μL per day for five days a week using a calibrated needle syringe. The YMPW was prepared by equilibrating 50 mg/L silica as sodium metasilicate with air and adding enough HCl to lower the pH to 7.6 in contact with an excess of powdered calcite. X-ray powder diffraction, scanning electron microscopy, and electron microprobe analysis confirm that, despite interactions with air outside the waste package, the dominant corrosion product in all cases was the Fe(II)-bearing magnetite. In the high temperature (60 degrees celsius) experiment, hematite and a fibrous, Fe-O-Cl phase were also identified. The Fe(II)/Fe(III) ratios measured in the corrosion products using a wet chemistry technique indicate extremely low oxygen fugacities (10-36 to 10-59 bar). The uranium observed associated with the corrosion products is still UO2, suggesting that even after two years conditions were sufficiently reducing to minimize oxidation. The small amounts of U that were dissolved and removed from the original grains can be found in the water (0.5-5 ppb U) and associated with the steel corrosion products surrounding the UO2 grains. Although these experiments were two years in duration, they still do not address the long-term behavior of a breached waste package; however, these results do support the possibility of a transient period of reducing conditions within a breached waste package.
3:30 PM - Q2.4
The Influence of Canister Material on Radionuclide Retention During Spent Fuel Leaching.
Daqing Cui 1 , Jeanett Low 1 , Vincenzo Rondinella 2 , Jinshan Pan 3 , Kastriot Spahiu 4 Show Abstract
1 Spent Fuel Chemistry, Studsvik AB, Nykoping Sweden, 2 Hot Cells , Inst. For Transuranium Elements, Karlsruhe Germany, 3 Div. of Corrosion Science, Inst. For Transuranium Elements, Stockholm Sweden, 4 R & D, SKB, Stockholm Sweden
In this experimental work, the corrosion behaviours of SF and canister materials (cast iron and copper), as well as immobilization reactions of radionuclides on iron canister material were investigated under simulated early canister failure conditions.The leaching solution, water with 10 mM NaCl and 2 mM NaHCO3, was deoxygenated by bubbling with pre-deoxygenated 99.97%Ar + 0.03% CO2 gas mixture before and during the experimental period. The gamma-radiation level experienced by the leaching system in the hot cell during the whole experimental period was 850 mGy / hour. During the initial leaching period, the concentration for all radionuclides dissolved from a SF pin increased with time, but after introducing small foils of Fe-Cu-cast iron in the leaching solution the concentration of U, Tc and Np dropped sharply, suggesting the precipitation of insoluble oxides on the surface of iron foils. pH, Eh and Ecorrosion for Fe, Cu and cast iron were recorded during the whole experiment. The polarization resistance Rp was measured. The corrosion rates of iron, cast iron and copper were calculated based electrochemical measurement and also estimated from the thickness of the corrosion layers after two years of corrosion. Similarities in Ecorr values and corrosion product layers (non-uniform, 20 - 50 micro m thick after two years) for pure Fe and cast-iron, indicate that the two materials exhibited similar corrosion behaviour. The corrosion rate of copper was found to be 200 times slower than that measured under air saturated conditions.An iron-silica rich corrosion layer containing 1-2 micro m sized U-Si rich particles was observed by SEM-WDS analysis on the cross section of reacted iron foils. Trace nuclides, Np, Pu, Tc, Sr and Cs were also detected by sensitive SIMS method.The effects of adding 10% H2 in the purging gas mixture on the SF leaching and radionuclides sorption on glass vessel were also investigated. The ratios of various radionuclides in the leaching solutions at end of the experiments and the corresponding radionuclides sorbed/precipitated in glass vessel wall were interpreted as consequences of redox and sorption/desorption reactions.After the leaching experiments, the SF pin was reacted in a saturated Fe(II) solution (FeCO3) solution for 450 days. The formation of ferric precipitates on SF surface was observed. The leaching rates of all radionuclides were found to be largely decreased as compared to the initial leaching experiment.Information obtained in this work is useful for discussing the behaviour of SF and canister materials under early-canister-failure conditions at deep geological repositories.
4:15 PM - **Q2.5
Spent Fuel Stability during Storage and Final Disposal.
Vincenzo Rondinella 1 , Thierry Wiss 1 , Joaquin Cobos 1 Show Abstract
1 Hot Cells, JRC-ITU, Karlsruhe Germany
Spent nuclear fuel and other high level nuclear wasteforms are subjected to radiation damage. Fission damage accumulated during in-pile irradiation amounts to thousand of displacements per atom (dpa). Spent fuel also accumulates alpha-decay damage and He during storage. The dose rates and the temperatures experienced in this case are lower than for in-pile operation, and depend on composition, history and activity of the fuel: however, the duration of the storage is longer (up to a few hundred years if extended interim storage concepts are considered); if final disposal in the repository is considered, the time interval in which radiation damage accumulates is open-ended.This presentation shows highlights from studies on irradiated fuels and analogue materials (both natural and tailor-made). The effects due to accumulating alpha decay damage and He on the corrosion resistance and overall durability of spent fuel will be discussed. The final goal of these studies is to assess the long-term stability of spent fuel.In order to simulate long-term accumulation of alpha-decay damage within timeframes suitable for laboratory experiments, alpha-doped materials can be used, i.e. matrices loaded with short-lived alpha-emitters (like e.g. Pu-238, U-233). The evolution of microstructural defects and corresponding macroscopic properties as a function of accumulated dose for spent fuel and alpha-doped UO2 shows remarkable similarity of behaviour. The activity levels for possible dose rate effects (or artefacts) to occur will be discussed. Experiments combining annealing studies using calorimetric techniques and He-release as a function of temperature with microstructure examination using TEM were performed to investigate the correlation among the annealing of defects in the microstructure, the release behaviour of He, and the heat effects associated with these processes in the material.Alpha-doped UO2 containing different fractions of alpha-emitter simulates the level of activity of spent fuel after different storage times, and can be used to study the effects of water radiolysis on the corrosion behaviour of aged spent fuel exposed to groundwater in a geologic repository. Alpha-doped UO2 with specific activities spanning over five orders of magnitude was used in static leaching experiments at room temperature in deionized water or in groundwater under various nominal redox conditions and using different Surface/Volume (S/V) ratio. Within the context of the experimental conditions considered, these experiments allowed establishing significance and extent of the alpha-radiolysis enhancement of fuel dissolution. The radiolytic enhancement of the corrosion process was measured for the alpha-doped materials compared to undoped UO2. At relatively low S/V conditions, an activity dependence of the concentration levels in the leachate was evident, while saturation and precipitation of oxidized phases characterized the evolution of the leaching tests at high S/V.
4:45 PM - Q2.6
A New Criterion for the Degradation of a Defective Spent Fuel Rod under Dry Storage Conditions Based on Nuclear Ceramic Cracking.
Lionel Desgranges 1 , Cécile Ferry 2 , Jean Radwan 2 Show Abstract
1 DEN/DEC, CEA, Saint-Paul lez Durance France, 2 DEN/DPC, CEA, Saclay France
An accident scenario for nuclear spent fuel dry storage consists in cask and fuel rod simultaneous failures that will put nuclear ceramic in contact with air. The swelling associated to UO2 oxidation in U3O8 might lead to the rod ruin. In literature interpretation, U3O8 is associated to the sigmoid part of the UO2 oxidation weight gain curve, and its 36% crystalline swelling is taken for responsible of the rod degradation. That is why previous criterions for safe behaviour of defected fuel rod in contact with air took into account U3O8 molar fraction. Recently the sigmoid part of the UO2 oxidation weight gain curves was reinterpreted as partially resulting from ceramic cracking and increased reactive surface; ceramic cracking induces sample bulking leading up to 300% swelling, one order of magnitude higher than the crystalline swelling induced by the UO2 to U3O8 transformation. This new interpretation led us to propose a new criterion of safety in which no significant damage is expected in the defective spent fuel rod before cracking apparition. The apparition of significant cracking and consecutive bulking is associated to the formation of a critical depth of U4O9+γ the time, at which the U4O9+gamma critical depth is formed, is calculated thanks to a finite difference model derived from a previously published model dedicated to un-irradiated UO2 oxidation. Although this new criterion significantly reduces the margin before cladding damage because cracking occurs before U3O8 formation, it leads to similar evaluations for the duration of safety than some previous ones.
5:00 PM - Q2.7
Corrosion Studies with High Burn-up LWR Fuel in Simulated Groundwater.
Ella Ekeroth 1 , Jeanett Low 1 , Hans-Urs Zwicky 2 , Kastriot Spahiu 3 Show Abstract
1 Studsvik Nuclear AB, Hot Cell Laboratory, SE-611 82 Nykoping Sweden, 2 Zwicky Consulting GmbH, Mönthalerstr. 44, CH-5236 Remigen Switzerland, 3 SKB, Box 250, SE-101 24 Stockholm Sweden
The release of toxic and radioactive species from spent fuel in contact with water is expected to depend mainly on the rate of dissolution of the UO2 matrix. In Sweden, the spent fuel will be disposed at 500-700 m below ground level, where the conditions are reducing and UO2 has very low solubility. However, ionizing radiation emitted from the spent fuel will produce oxidants (and reductants) and alter the otherwise reducing conditions and thus enhance the rate of spent fuel dissolution. The burn-up of future spent fuel to be disposed will be higher than the burn-up of today’s fuel. Actinides accumulate in the rim zone and the content of lanthanides and other fission products will also increase in spent fuel as a consequence of higher burn-up. Furthermore, the formation of metallic particles will be enhanced. The increased actinide content in spent fuel at higher burn-ups will lead to a higher α-dose rate in the surrounding water and the higher content of fission products will also contribute to a higher β- and γ-dose rate initially. The dissolution rate is expected to increase with higher burn-up due to higher dose rates. On the other hand, the presence of fission products like lanthanides in the UO2 matrix has been shown to have an inhibiting effect on UO2 dissolution. The outcome of the study will be discussed taking the above mentioned as well as other physical properties into account. Previous static corrosion tests on spent fuel with a burn-up range of 27 to 49 MWd/kg U showed that the cumulative release fractions increase with burn-up to reach a maximum at approximately 40-45 MWd/kg U. At higher burn-up (up to 49 MWd/kg U) the release rates decrease. The study has now been extended to comprise spent fuel with even higher burn-up. Static leaching of spent fuel with 60 and 75 MWd/kg U burn-up has been started. From each spent fuel rod, a fuel pin segment, containing one complete and two half pellets, is leached under oxidizing conditions in synthetic groundwater. Results from five contact periods, for a cumulative contact time of one year, will be presented and compared with previous results.
5:15 PM - Q2.8
Catalysis of the Reaction Between Hydrogen Peroxide and Hydrogen on Epsilon Particles in Spent Nuclear Fuel.
M. Broczkowski 1 , P. Keech 1 , J. Noel 1 , David Shoesmith 1 Show Abstract
1 , University of western ontario, London, Ontario, Canada
Raul B. Rebak GE Global Research
Neil C. Hyatt The University of Sheffield
David A. Pickett Southwest Research Institute
Q3: Nuclear Waste Glasses and Vitrification
Tuesday AM, December 02, 2008
Back Bay D (Sheraton)
9:45 AM - **Q3.1
A New Model for Nuclear Waste Borosilicate Glass Alteration.
Thorsten Geisler 1 , Arne Janssen 1 , Jasper Berndt 1 , Andrew Putnis 1 Show Abstract
1 Institut fuer Mineralogie, University of Muenster, Muenster Germany
Understanding the mechanism of aqueous corrosion of nuclear waste borosilicate glasses is essential to reliably predict their long-term aqueous durability in a geologic repository. Here we report the results of corrosion experiments with borosilicate glass cuboids with edge lengths of about 2.5 mm in a HCl solution of pH = 0 at 150°C for 6 to 336 hours, including two experiments with 18O and 26Mg as isotope tracers. Several observations were made in this study such as (i) the occurrence of chemical oscillations in the corrosion rim that is composed of silica, (ii) an enrichment of 18O and 26Mg in the corrosion rim formed in the isotopically enriched solutions without observable diffusion profiles, (iii) a sharp phase boundary of the corrosion rim towards the pristine glass, (iv) a high porosity in the corrosion rim, and (v) the occurrence of silica spherules at the surface while the overall shape of the cuboids was retained. These features are not at all compatible with classical theories about the formation of the corrosion or "gel" layer that are based on diffusion-controlled hydration and ion exchange reactions and subsequent solid-state re-condensation of the hydrolyzed glass network (Grambow 2006, and references therein). We propose a new mechanistic model for glass corrosion that is based on congruent dissolution of the glass that is spatially and temporally coupled to the precipitation of amorphous silica at an inward moving reaction interface. The free energy difference needed to drive a dissolution-reprecipitation reaction is provided by the difference of the solubility of the parent and product phase in solution, which in turn may result from textural and/or chemical energy differences. Because both reactions take place at the borosilicate glass-silica interface, only a small amount of material needs to be in solution in the boundary layer at the interface at any one time. This implies that the relative solubility of the two solids in the solution is more important than their absolute solubility. The model provides, for the first time, a reasonable thermodynamic and mechanistic basis to understand (i) the formation of chemically self-organized corrosion zones in naturally altered glasses, (ii) why glass corrosion proceeds with relatively fast rates even in silica saturated solutions, (iii) why long-term corrosion rates remain constant as observed in a number of experimental studies, and (iv) why the initial glass corrosion process are often found to be congruent in laboratory experiments (see Grambow 2006, and references therein). It thus provides a novel framework to evaluate the long-term performance of nuclear waste glasses.
Grambow B. (2006) Nuclear Waste Glasses - How Durable? Elements 2, 357-364.
10:15 AM - Q3.2
Iron Redox Reactions in Nuclear Waste Glasses and Melts.
Benjamin Cochain 4 1 , Daniel Neuville 1 , Dominique De Ligny 2 , Laurent Cormier 3 , Olivier Pinet 4 , Pascal Richet 1 Show Abstract
4 CEA, CEA Valhro SECM-LDMC, Marcoule France, 1 PMM-IPGP-CNRS, IPGP-CNRS, Paris, 75005, France, 2 LPCML, UCBL, Lyon, 69000, France, 3 IMPMC, CNRS, Paris, 75000, France
Silicate glasses are privileged material for nuclear waste immobilization because they allow incorporation of a large range of elements and high volume reduction. Understanding of their physical properties thus is an important goal. In this respect, multivalent oxides raise special difficulties. Specifically, they can affect melt and final glass properties in a complex way because they occur in glasses in different redox state. One of the major multivalent elements in these glasses is iron. It occurs as Fe2+ and Fe3+ ions whose abundances markedly depend on intensive thermodynamic variables. The objective of the present work is to arrive at a deeper knowledge of iron redox reactions in alumino-boro-silicate glasses and melts whose compositions are relevant to nuclear waste storage glasses.In this work our goal has been to determine the iron redox state and to derive structural information about iron through Raman spectroscopy and XANES experiments at the K edge of iron. Redox equilibria were examined as a function of time and just above Tg and at superliquidus temperatures. We present here some experiments up to 2100K in the liquid and glass states. With the concept of « redox diffusivity » defined from the time required to achieve redox equilibrium at a given temperature, comparisons can be made with the diffusivities of oxygen and various cations. These comparisons allow us to distinguish the temperature and compositions ranges where a given redox mechanism predominates, namely, diffusion of network-modifying cations at low temperatures, and diffusion of oxygen above the liquidus.
10:30 AM - Q3.3
Widening the Envelope of UK HLW Vitrification – Experimental Studies with High Waste Loading Formulations Containing Platinoids.
Carl Steele 1 , Charlie Scales 2 Show Abstract
1 High Level Waste Plants, Sellafield Ltd, Seascale, Cumbria, United Kingdom, 2 , Nexia Solutions, Seascale, Cumbria, United Kingdom
Sellafield Ltd operates 3 vitrification lines to convert highly active concentrate liquor arising as a waste product of reprocessing operations into glass for safe interim storage in the Vitrified Product Store (VPS) prior to long term disposal.Highly Active Liquor (HAL) is stored in Highly Active Storage Tanks and transferred to WVP in batches to the liquid stock tank. It is metered in a semi-continuous batch operation to a calciner (rotating tube furnace) where it is converted into an oxide powder (calcine). Glass frit is fed at the lower end of the calciner where it discharges into an Inconel melter vessel controlled at approximately 1100 οC. The glass and calcine are melted together and then poured into a container as a batch operation. After two pours the container is allowed to cool, a lid is then fitted to the container, which after further cooling is welded to the container. This container is then cleaned and transferred to the VPS.Platinoid species containing ruthenium, rhodium and palladium present in the HAL form insoluble oxide phases in the glass product. The platinoid concentration in the glass will increase with increasing waste oxide loading to an extent that settling of the platinoids in the glass may occur, leading to heel enrichment, poor melter performance and difficulty in draining the melter. The viscosity will also increase, which may require higher melter temperatures to mix and pour the molten glass and could result in enhanced corrosion of the melter. Inactive laboratory scale experiments with different glass frit formulations have been performed to determine whether product quality could be maintained with higher platinoid concentrations. Operational envelopes with existing formulations were expanded to observe laboratory trial performance and determine any changes to resulting glass qualities. Also, glasses with high waste incorporations have been produced to test process capability and to ascertain any potential phase separation or devitrification issues that could affect either the process or product performance. Physical properties of the different glass formulations were performed to measure changes in viscosity, density and the rates of settling to examine the amount of phase separation that can occur.The results have shown that ruthenium, palladium or rhodium were insoluble in the melt and were not evenly distributed throughout the glass but clustered together. These results will be used as a basis for further development work. This paper presents some findings of these experiments.
10:45 AM - Q3.4
Radiation Effects in Glass and Ceramic Waste Forms for the Advanced Fuel Cycle.
William Weber 1 Show Abstract
1 Chemical & Materials Science Division, Pacific Northwest National Laboratory, Richland, Washington, United States
Advanced nuclear fuel cycles may include separate waste streams for minor actinides and various fission products. Thus, radiation effects from alpha decay in actinide-bearing waste forms and beta decay in waste forms for fission product immobilization, particularly Cs and Sr, may need to be considered separately. The results of computer simulations, new models, and experimental studies over the past 30 years will be presented to highlight the current state-of-the-art understanding and predictive models for radiation effects in relevant glass and ceramic materials. Alpha decay generally leads to amorphization, macroscopic swelling and increases in dissolution rates for nearly all phases currently under consideration; furthermore, accelerated test methods and multiscale models have been developed to provide reliable data to predict long-term performance behavior of new actinide-bearing waste forms. Ionization from the beta decay of fission products can cause covalent and ionic bond rupture, valence changes, localized electronic excitations, and significant changes in hydrogen exchange and mobility. In the case of Cs and Sr immobilization, extreme ionization rates may occur in the waste forms, which can only be studied using electron or ion beam irradiation techniques. Under these extreme ionization conditions, ionization-induced electronic excitations can have lifetimes on the order of the time between consecutive events, which creates a steady-state concentration of localized electronic defects that affect local bonding and diffusion. Glasses exposed to extreme ionization are observed to readily undergo decomposition, bubble formation, and enhanced diffusion of alkali elements and oxygen. Extreme ionization in ceramics can enhance phase transformations, precipitation, and diffusion processes. For high Cs and Sr loadings, the change in valence state and ionic radii of the transmutation products may further affect overall structural stability.
11:30 AM - Q3.5
Microwave Processing of Glasses for Waste Immobilisation.
Martin Stennett 1 , Neil Hyatt 1 Show Abstract
1 Engineering Materials, The University of Sheffield, Sheffield, South Yorkshire, United Kingdom
Microwave (or dielectric) heating is recognised as a fast, clean and economical preparation route for a wide range of inorganic solids. The microwave spectrum lies in the frequency range 0.3 to 300 GHz which is mainly used for communication purposes, although a narrow frequency window centred at 2.45 GHz is allowed for microwave heating purposes. Materials in general fall into three categories: microwave reflectors, transmitters, and absorbers. A key requirement for microwave heating is that one or more of the major constituents of the charge must be a microwave absorber and couple strongly to the microwave field at room temperature. Microwave heating has a number of key advantages over conventional heating. Microwave heating is a rapid process and unlike conventional heating the charge is heated from the inside out. This minimises undesirable decomposition, oxidation / reduction, loss of volatile materials, and other kinetically slow processes which can occur during conventional melting. In this work, microwave heating has been applied to processing of different iron phosphate glass compositions reported in the literature as being suitable for immobilisation of radioactive wastes. Chemically homogeneous glasses have been successfully prepared in all systems by microwave heating over timescales of the order of several minutes. The resulting glasses have been characterised by a range of techniques including X-ray diffraction (XRD), optical and scanning electron microscopy (SEM), Raman and IR spectroscopy and differential thermal analysis (DTA).
11:45 AM - Q3.6
In Situ NMR Investigation of Phase Separation in Molybdenum-Bearing Nuclear Waste Glasses.
Scott Kroeker 1 2 , Ian Farnan 2 , Sophie Schuller 3 , Thierry Advocat 4 Show Abstract
1 Chemistry, University of Manitoba, Winnipeg, Manitoba, Canada, 2 Earth Sciences, University of Cambridge, Cambridge United Kingdom, 3 DEN/DTCD/SECM/LDMC, CEA Valrhô Marcoule, Bagnols/céze France, 4 DEN/DANS/DPC/SECR, CEA Saclay, Gif sur Yvette France
Effective disposal of Mo-containing radioactive wastes by vitrification is hampered by the tendency of this element to nucleate soluble crystalline phases. Empirical studies have determined key processing parameters related to this challenging system, and current industrial solutions involve limiting the waste-loading capacity. However, less is known about the detailed compositional and structural evolution of molybdate formation. Recent ex situ EXAFS work established that four-coordinate molybdate species are not directly bonded to the glass network, but remain isolated within alkali-rich environments conducive to macroscopic phase separation and subsequent crystallization. The present work builds on this foundation by probing cesium local environments in borosilicate glasses using nuclear magnetic resonance (NMR) spectroscopy from room temperature through Tg, and into the melt. Distinct 133Cs NMR signals are observed over a large temperature range for cations present in the bulk glass and those associated with Mo, implying that liquid-liquid phase separation persists well above Tg. Crystal growth can also be observed by NMR at high temperatures in a glass nucleated by crystalline cesium molybdates. This method is also used to study the phase transformation behaviour of a crystalline mixed-alkali molybdate phase that may be paradigmatic of the more complex molybdates that form during waste vitrification. This work on model waste glasses demonstrates that high-temperature NMR provides a unique approach to monitoring phase separation processes in situ, with the potential to gain valuable fundamental insight that may guide industrial solutions.
12:00 PM - Q3.7
Glass Development for Vitrification of Wet Intermediate Level Waste (WILW) from Decommissionning of the Hinkley Point `A' Site.
Paul Bingham 1 , Neil Hyatt 1 , Russell Hand 1 , Christopher Wilding 2 Show Abstract
1 Immobilisation Science Laboratory, University of Sheffield, Sheffield, South Yorkshire, United Kingdom, 2 , Magnox South Ltd., Nr Bridgwater, Somerset, United Kingdom
12:15 PM - Q3.8
Infrared Spectroscopy Structural Analysis of Corroded Nuclear Waste Glass K-26.
Sergey Stefanovsky 1 , Alexander Barinov 1 , Irene Startseva 1 , Galina Varlakova 1 , Michael Ojovan 2 Show Abstract
1 , SIA Radon, Moscow Russian Federation, 2 , University of Sheffield, Sheffield United Kingdom
Silicate glasses in near-neutral water solutions have minimal corrosion rates thus requiring long term tests to understanding their behaviour. Long-term field test data on environmental behaviour of a high-alkali-borosilicate nuclear waste glass K-26 revealed high retaining properties of glass in an experimental shallow ground repository. Data on infrared (IR)-spectroscopy for glass K-26 samples taken from both the bulk of glass and the near surface altered glass layers show that the structure of glass K-26 is made up of [SiO4] tetrahedra with boron in the network structure in the form of [BO4] tetrahedra. The IR-spectra show that the glass surface has a highly ordered structure compared the bulk of glass K-26. This is assumed to result from glass corrosion in the disposal environment and formation of secondary crystalline phases on the surface.
12:30 PM - Q3.9
Initiation of Damage in Nuclear Waste by a Nanoscale Pump-probe EELS Technique.
Nan Jiang 1 Show Abstract
1 Physics, Arizona State University, Tempe, Arizona, United States
Here we introduce a nanoscale technique to detect the early stages of radiation damage in nuclear waste glasses and ceramics using in situ electron energy loss spectroscopy (EELS) in transmission electron microscopy (TEM). This method is demonstrated in study of Mg migration in spinel MgAl2O4 by electron irradiation. Assuming that the Mg migrate between different interstitial sites in spinel structure under radiation, the in situ EELS of the Mg core edges (e.g. L23-edge) should show time-dependent changes, which can be directly related to the initial radiation process. Both the parallel detection capability of EELS, and the use of a field emission electron source make this experimental arrangement highly efficient, and provide much higher count rates than are possible using x-ray absorption near edge structure on a synchrotron, with much greater spatial resolution. Unlike photons, electrons are not annihilated in an inelastic interaction, and continue to the detector after an energy-loss event. It is therefore possible to operate every energy-loss detection channel in a magnetic quadrant spectrometer simultaneously. This high detection efficiency allows the time-evolution of spectra to be studied in detail, with the very high spatial and temporal resolutions. The sensitivity of this technique is also studied. We have proved in MgAl2O4 that the migration of Mg under radiation can be seen by in situ EELS much more easily than by other methods, such as diffraction.
12:45 PM - Q3.10
Effects of Silica Particle Size and pH on Rheological Properties of Simulated Melter Feed Slurries for Nuclear Waste Vitrification.
Hong Zhao 1 , Ian Pegg 1 Show Abstract
1 Vitreous State Laboratory, The Catholic University of America, Washington DC, District of Columbia, United States
In the process of nuclear waste vitrification, nuclear waste is usually mixed with glass-forming additives to form an intermediate aqueous slurry, followed by vitrification through a melter at high temperature (typically 1100-1200°C) to form a durable glass product with the desired composition and properties. The rheological behavior of such an intermediate melter feed slurry is important in the slurry mixing stage and in the subsequent slurry transport to the melter. A high solids content is desirable in terms of glass production rate but this must be offset against the requirements for favorable apparent viscosity and yield stress for slurry mixing and transport. In the present work, a model slurry system at solid contents above 70%, prepared by adding commonly used glass-formers into a simulated waste, was studied to investigate the effects of pH and the particle size of silica (as the major glass-forming additive) on the rheological properties of simulated slurries. The results show that whereas silica particle size significantly affects both the apparent viscosity and yield stress, pH changes modify the yield stress while only slightly altering the apparent viscosity. Such relationships between rheological properties and controllable variables are useful for engineering melter feed slurries with desired rheological properties and potentially for recovery from upset conditions that lead to unfavorable slurry rheologies.
Q4: Ceramic Wasteforms
Tuesday PM, December 02, 2008
Back Bay D (Sheraton)
2:30 PM - **Q4.1
Hiped Tailored Glass-Ceramic Waste Forms for the Immobilization of High-Level Nuclear Waste.
Melody Carter 1 Show Abstract
1 IME, ANSTO, Lucas Heights, New South Wales, Australia
Hot-isostatic pressing (HIPing) offers significant advantages for the immobilization of nuclear waste. Firstly there are zero off-gas emissions during high-temperature consolidation, which mitigate volatility concerns. Secondly the process places minimal constraints on the waste form chemistry which in turn permits significantly higher waste loadings. In addition, the HIP process readily produces a dense monolithic waste form, which both minimizes disposal volume and reduces surface area available for aqueous attack once emplaced in a repository.Coupling the HIP process with glass-ceramic waste forms results in a superior waste immobilization system. Glass-ceramics seek to combine the process and chemical flexibility of glasses with the excellent chemical durability of ceramics. This can be achieved by exploiting the glass forming components present in the waste, along with appropriate additives, to form a durable glass and the desired crystalline phases. Consequently glass-ceramics are well suited to a number of compositionally diverse waste streams. In this paper the design of HIPed glass-ceramics for the immobilization of uranium –rich wastes, K-basin sludge and waste from Mo99 production are discussed. A detailed discussion on the design rationale for HIPed glass-ceramics is presented. Detailed microstructural, diffraction and spectroscopic characterization of selected glass-ceramic samples have been performed, along with chemical durability these results will be presented.
3:00 PM - Q4.2
Calcium Phosphate: A Potential Host for Halide Contaminated Plutonium Wastes.
Brian Metcalfe 1 , Ian Donald 1 , Shirley Fong 1 , Lee Gerrard 1 , Denis Strachan 2 , Randall Scheele 2 Show Abstract
1 MSRD, AWE, Reading United Kingdom, 2 , PNNL, Richland, Washington, United States
Significant quantities of fluoride and chloride present in four types of legacy wastes produced during plutonium pyrochemical reprocessing necessitated the development of a new wasteform which could adequately immobilize the halides in addition to the Pu and Am. Using Sm as a surrogate a process was developed by AWE to immobilize Type I waste, a chloride-based waste, which used Ca3(PO4)2 as the host material into which the Cl ions were incorporated to produce two crystalline phases, chlorapatite, [Ca5(PO4)3Cl], and spodiosite, [Ca2(PO4)Cl]. Powder prepared at PNNL with 239Pu and 241Am confirmed the AWE durability data obtained using the surrogate. Accelerated radiation-induced damage trials have been carried out on the powder with the substitution of 238Pu (half life 87.7 y) for 239Pu (half life 2.4 x 104 y). These trials were conducted over five years during which time the material experienced an α fluence of 4 x 1018αg-1. Throughout the ageing period the powder was monitored with XRD but no change in the crystalline structure was detected. Initial trials on the durability of the aged powder showed that durability had decreased by approximately two orders of magnitude. The process was applied to three other types of waste (Types II-IV). These differed from Type I in that they were essentially oxide-based, although they contained some Cl and F and required the use of Hf as the surrogate for the Pu(IV) present in these wastes. A more complex phase assemblage of cation-doped whitlockite, β-Ca2P2O7 and chlor-fluorapatite resulted. Durability trials at AWE showed these phases to be quite durable with losses below the limit of detection in the leachate for Hf and Sm and normalized mass losses of 2.5 x 10-4, 2.8 x 10-4 and 7.6 x 10-5g.m-2 for chloride for Types II, III and IV wastes respectively. Durability studies of a monolithic wasteform produced by sintering a mixture of powder and sodium aluminophosphate glass showed that whilst the losses of Sm and Hf remained below the limit of detection, the durability of Cl decreased by approximately two orders of magnitude for Type II and III waste and three orders for Type IV. Monolithic Type II specimens containing either 238Pu or 239Pu in addition to Am have been made at PNNL to allow the radiation-induced damage on the durability and structure to be investigated. These specimens are currently undergoing their initial characterization.
3:30 PM - Q4.4
Thermochemistry of New Fission Product Waste Forms.
Tae-Jin Park 1 , Tina Nenoff 2 , Alexandra Navrotsky 1 Show Abstract
1 NEAT ORU, UC Davis, Davis, California, United States, 2 , Sandia National Laboratories, Albuquerque, New Mexico, United States
To reduce the costs and minimize the risk of contamination to the environment during waste processing and storage, disposal of nuclides with relatively short half lives (e.g., 137Cs and 90Sr) needs fundamental understanding and correlation of a variety of issues related to thermally consolidated Cs, Sr-loaded waste forms. In addition, since Cs and Sr form new elements (e.g., Ba, Y, Zr) by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products merits study. We present new information on the durability and thermodynamic stability of Cs- and Ba-loaded ceramic waste forms and those containing their decay products using non-radioactive analogues. Enthalpies of formation of the synthesized phases from the constituent oxides and from the elements have been determined by drop solution calorimetry into molten lead borate solvent at 974 K. The variation of formation enthalpies when Ba2+ substitutes for Cs1+ is related to the concomitant occurrence of O2- vacancies, which compensate the charge imbalance between Ba2+ and Cs1+ in the structure.
3:45 PM - Q4.5
Discriminate Immobilization of Cs and Sr Ions in the Form of Single-crystalline Titanates.
Hideki Abe 1 , Akira Satoh 1 , Kenji Nishida 1 Show Abstract
1 , National Institute for Materials Science, Tsukuba Japan
Global warming and environmental pollution caused by huge consumption of fossil fuels is a growing challenge. Nuclear fuels are the most efficient energy source as an alternative to fossil fuels. Unlike fossil fuels, the consumption of nuclear fuels is accompanied by neither green-house gases nor toxic chemicals. One of the most urgent issues for the nuclear fuel technology is the development of safe, cost-effective disposal methods of highly radioactive wastes (radwastes) that are unavoidably produced by the use of nuclear fuels. We present an electrolytic method to discriminately immobilize the different kinds of radioactive elements in radwastes.[*] Cs and Sr ions contained in simulated wastes, indeed, have been electrochemically separated in the form of single-crystalline titanates. Simulated wastes of 137Cs- and 90Sr-containing radwastes were prepared by mixing the oxides of Cs and Sr in a different mole ratio of Cs:Sr = 1:x. The simulated wastes were dissolved in a mixture of titania (TiO2) and molybdates that was molten at 1050 oC under ambient pressure. The melts were electrolyzed at a constant current of 10 mA over 1 hour. In the case x < 0.01, several ten milligrams of single crystals of Cs1.35Ti8O16 (Hollandite-type structure; I4/m; a =10.3 Å, c = 2.97 Å) were obtained on the cathode. When x > 0.05, instead of Cs1.35Ti8O16, several milligrams of single crystals of SrTiO3-δ (Perovskite-type structure; Pm3m; a = 3.91 Å) were obtained, again on the cathode. Cs and Sr ions contained in the simulated wastes were electrochemically separated in the different forms of single-crystalline Cs1.35Ti8O16 and SrTiO3-δ, respectively, at a separation efficiency of 99 percents. The electrolytic method demonstrated by this work may pave the way for discriminate solidification of radwastes as an alternative to the conventional vitrification.[*]...H.Abe, A. Satoh, K. Nishida, E. Abe, T. Naka, M. Imai and H. Kitazawa, J. Solid State Chem. 179, 2006, 1521-1524.
4:45 PM - Q4.7
Yttrium Substitution in MTiO3 (M=Ca, Sr, Ba) Perovskites and Implication for Incorporation of Fission Products into Ceramic Waste Forms.
Nissim Navi 1 4 2 , Giora Kimmel 3 , Sergey Ushakov 1 , Alexandra Navrotsky 1 Show Abstract
1 Peter A Rock Thermochemistry Laboratory and NEAT ORU, University of California at Davis, Davis, California, United States, 4 Chemistry, Nuclear Research Center–Negev, Beer-Sheva Israel, 2 Nuclear Engineering, Ben-Gurion University of the Negev, Beer-Sheva Israel, 3 Institutes for Applied Research, Ben-Gurion University of the Negev, Beer-Sheva Israel
Titanate perovskites have the potential to host nuclear waste due to their capability of taking a wide range of elements into stable solid solution. Cs and Sr are short-lived radioisotopes which decay to Ba and Y/Zr respectively and will induce composition changes of the initial ceramic waste form accompanied by volume, microstructure and chemical reactivity changes and radiation effects. In order to characterize phase relations of the decay product Y substitution into perovskites MTiO3 (M=Ca, Sr, Ba) non-radioactive precursor mixtures with atomic composition of M:Y:Ti = 0.75:0.25:1 were synthesized in air at 1500°C to undergo solid-state reaction. Results showed that all M-Y-Ti-O systems formed major perovskite (MTiO3) and pyrochlore (Y2Ti2O7) like phases with limited solubility. The Ca-Y-Ti-O system showed slightly greater solubility than the Ba-Y-Ti-O. In the Ba-Y-Ti-O system a small amount of additional Ba-Ti rich phase was observed. This study indicates a potential tendency of the decay product Y to segregate out from the perovskite waste form into a Y rich phase while in an oxidation environment.
5:00 PM - Q4.8
Chemical Evolution and Structural Stability of Crystalline Nuclear Waste Forms during Transmutation.
Chao Jiang 1 , Christopher Stanek 1 , Nigel Marks 2 , Kurt Sickafus 1 , Blas Uberuaga 1 Show Abstract
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , University of Sydney, Sydney, New South Wales, Australia
A critical component of any closed nuclear fuel cycle, e.g. the concept of the Global Nuclear Energy Partnership, is a comprehensive waste strategy. A possible alternative to the current "burn and bury" strategy is to develop specialized waste forms for individual fission products to be isolated and stored. For example, if the short-lived isotopes of Cs and Sr, as well as the radiotoxic actinides are removed, significantly more waste can be included in a geologic repository. However, and especially for the short lived fission products, when considering the durability of customized immobilization hosts, one must consider the chemical evolution effects.Experimental studies of these compounds are difficult due to the long time over which the transmutation occurs. In this work, we thus use density functional theory (DFT) to study a model crystalline waste form for 137-Cs immobilization, CsCl. While CsCl is a simple system and is unlikely to be used in any real nuclear fuel cycle, we expect that the insights gained into the chemical thermodynamics of this system will apply to more complex and realistic waste form compounds. Furthermore, it has been synthesized in the past for Cs immobilization, which may provide useful information for potential comparison with our modeling results. Using DFT, we explore the effects of 137-Cs transmutation to 137-Ba on the stability of the original crystal as the concentration of Ba increases. We examine the thermodynamic driving force for the decomposition of the material into Ba-rich phases. We also determine the mechanical stability of those phases. Using this information, we evaluate the thermodynamic stability of the waste form during its lifetime. The methodology we employ should be applicable to more complex crystalline waste forms and to other waste species, such as Sr and Tc.
5:15 PM - Q4.9
Experimental and Atomistic Modelling Study of Ion Irradiation Damage in Thin Crystals of the TiO2 Polymorphs.
Gregory Lumpkin 1 , Bronwyn Thomas 1 3 , Katherine Smith 1 , Mark Blackford 1 , Nigel Marks 3 , Nestor Zaluzec 2 , Karl Whittle 1 Show Abstract
1 Institute of Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 3 Physics, University of Sydney, Sydney, New South Wales, Australia, 2 Materials Science Division, Argonne National Laboratory, Chicago, Illinois, United States
Thin crystals of rutile, brookite, and anatase were irradiated in-situ with 1MeV Kr at 50K, using the IVEM-TANDEM facility. The critical amorphisation fluence has revealed a large difference in the radiation tolerance. Synthetic rutile remained crystalline up to 5*1015 ion cm-2, and did not show convincing evidence for onset amorphisation at this level. Natural brookite and anatase became amorphous at 8.1±1.8*1014 and 2.3±0.2*1014 ions cm-2 respectively. The results correlate with the number of shared edges, degree of octahedral distortion and volumes.The octahedral distortion and volumes are directly linked through Ti-Ti repulsive forces in the octahedral frameworks. Static defect calculations have indicated the response of rutile is indicative of the higher defect formation energies and the presence of low energy migration pathways for both Ti and O. Amorphisation of anatase is facilitated by low defect energies of Ti and O. In addition to the results outlined above Tc/Fc curves will be presented showing the changes as a function of structure.
5:30 PM - Q4.10
Porous Alumosilicates as a Filter for High-Temperature Trapping of Cesium-137 Vapour.
Albert Aloy 1 , Alexander Strelnikov 1 Show Abstract
1 , Khlopin Radium Institute, Saint Petersburg Russian Federation
Vitrification and ceramization of high-level radioactive waste are accompanied by volatilization of vapour, containing Cs-137. Previously, it was well established that the amorphous alumosilicate phase of porous inorganic materials (PIM) have a high reactivity and can very fast interact with cesium in solutions forming the stable crystalline phases, namely CsAlSiO4 and CsAlSi2O6 (pollucite) at high temperature.It was suggested to use the same PIM with different content of amorphous phase for Cs-137 vapour trapping.Experiments in static and dynamic conditions were performed to study trapping characteristics and determine the phase composition of the PIM filters after reaction with gaseous cesium generated during CsNO3 calcination at different temperatures To reduce the final volume of spent filters before disposal the cold pressing followed by sintering was used.Microstructure, phase and chemical composition of PIM before and after saturation by cesium were analyzed using XRD, SEM, EPMA and optical microscopy.It was shown that cesium trapping quantity and phase compositions were sensitive to the types of PIM which contained the different amount of alumosilicate amorphous phase and porous microstructure. Results of this study will be presented in the paper.
Raul B. Rebak GE Global Research
Neil C. Hyatt The University of Sheffield
David A. Pickett Southwest Research Institute
Q5: Engineered Barrier Systems, the Near Field and Cementitious
Wednesday AM, December 03, 2008
Back Bay D (Sheraton)
9:45 AM - **Q5.1
EPRI Perspective on Spent Nuclear Fuel Storage, Transportation, and Disposal - Present and Future.
John Kessler 1 Show Abstract
1 , EPRI Inc, Charlotte, North Carolina, United States
10:15 AM - Q5.2
Identification of Oxygen Depleting Components in MX-80 Bentonite.
Torbjorn Carlsson 1 , Arto Muurinen 1 Show Abstract
1 , VTT Technical Research Centre of Finland, Espoo Finland
The near-field of a nuclear waste repository contains after closure large amounts of oxygen in tunnels and deposition holes. The bentonite buffer/backfill will contain oxygen as a gas phase in unsaturated pores and also as dissolved gas in porewater. The redox conditions in the bentonite fillings will after post-closure change towards reducing conditions. In the initial stage, the development of the redox-state is mainly governed by the depletion of oxygen. The main mechanisms of oxygen depletion in the bentonite are: (1) diffusion into the surrounding rock, and (2) reactions with accessory minerals and by microbial aerobic consumption of organic matter (CH2O) (SKB R-06-106 and Posiva 2006-05). The reactions leading to oxygen depletion in bentonite are not, however, well understood. There is a lack of knowledge concerning what minerals or compounds are involved in the oxygen depletion. Often pyrite (FeS2) is mentioned as a mineral that might consume oxygen, but pyrite oxidation may not necessarily be the main process for oxygen depletion (Lazo et al. 2003). The same also holds for other components in the bentonite; many components are potential oxygen scavengers, but their exact roles in the oxygen depletion are not clear. The objective of this work was to improve the knowledge concerning what components take part in the depletion of dissolved oxygen (DO) in MX-80 bentonite. This was done experimentally under N2 by using MX-80 sub-samples from which one or several of the original components had been removed. Briefly, the experiments were carried out by adding known amounts of oxygen to water-saturated sub-samples, which were then let to stay some time. The oxygen depletion was studied in two ways; (i) by measuring the redox state using electrodes during the depletion, and (ii) by measuring the DO concentration in the water phase by an oxygen meter at the end of the experiments. The role of bacteria in the oxygen depletion was to some extent investigated by comparing the DO depletion in unheated samples and in samples that had been heated to temperatures were the bacterial activity was thought to be eliminated. The extent of oxygen depletion in each case is presented and discussed in terms of the composition of the sub-sample.
10:30 AM - Q5.3
Soft X-ray Spectroscopic Characterization of Montmorillonite.
Jan-Erik Rubensson 1 , Franz Hennies 2 , Lars Werme 3 1 , Ola Karnland 4 Show Abstract
1 Physics and Materials Science, Uppsala University, Uppsala Sweden, 2 MAX-Lab, Lund University, Lund Sweden, 3 , Swedish Nuclear Fuel and Waste Management Co (SKB), Stockholm Sweden, 4 , Clay Technology AB, Lund Sweden
The local, atomic specific nature of the X-ray process and the relatively large penetration length of soft X-ray photons make soft X-ray spectroscopy the ideal method for in-situ studies of montmorillonite samples. With soft X-ray emission and absorption spectroscopy at the L edges of Ca, Al, Si, the chemical state locally at those atoms can be readily probed. Quantum confinement effects for the oxides will be monitored, and the site specificity will make it feasible to determine the substitution configurations. At our beamline at MAX-lab synchrotron light source in Lund, Sweden, we recorded X-ray absorption and selectively excited X-ray emission spectra of some dry montmorillonite samples; the experimental set-up also allows the study of wet samples and this will be included in the future. In general, the absorption spectra demonstrate that the electronic structure varies considerably throughout the structure. Thus, it is obvious that only a local probe can give useful electronic structure information. More in detail, the Ca L spectrum, associated with the inter-layer counter ions shows a typical quasi-atomic behavior. On the present level of accuracy we do not see any effects of the ligand field, which are prominent in all ionic Ca compounds. Small deviations from the quasi-atomic behavior will be analyzed in terms of coordination and ligand interactions. The O K spectrum shows a pre-peak at around 532 eV, which is absent in the spectra of both the pure, bulk Al and Si oxides; earlier such a pre-peak has been associated with impurities and/or lattice distortions. Here, the reduced dimensionality in the layers suggests that the pre-peak is associated with the ‘dangling oxygen atoms’ in the outermost atomic layer. Intriguingly, the Si L absorption closely simulates the absorption spectrum of a pure silicon oxide, indicating that the local electronic structure at the Si sites is affected very little, and that therefore the SiO4 tetrahedra in the clay must be essentially undisturbed. The Al L spectrum, on the other hand, is complex and indicates that the chemical state of the Al atoms (in octahedral coordination) may be quite different from what is common in the oxides. Although the physical significance of these results must be further analyzed, this qualitative discussion illustrates the large potential of the methods and indicates some specific research directions.
10:45 AM - Q5.4
Diffusion Behavior of Humic Acid in Compacted Bentonite: Effect of Ionic Strength, Dry Density and Molecular Weight of Humic Acid.
Kazuki Iijima 1 , Seiichi Kurosawa 2 , Minoru Tobita 2 , Satoshi Kibe 2 , Yuji Ouchi 2 Show Abstract
1 , Japan Atomic Energy Agency, Ibaraki Japan, 2 , Inspection Development Company. Ltd., Ibaraki Japan
In the geological disposal system of high level radioactive waste in Japan, it is expected that humic acid (HA), one of the natural organic matters in groundwater, cannot diffuse inside compacted bentonite due to its large size and, therefore, has less effect on solubility and diffusion behavior of radionuclides in this region. On the other hand, a recent paper reported that humic colloids can diffuse through compacted bentonite and some nuclides show higher apparent diffusivities in the presence of HA . HA generally has broad molecular weight distribution, which is estimated to affect on the diffusion behavior. However, it has not been investigated so far. In this paper, through-diffusion (T-D) experiments of HA and neodymium (Nd) in the presence of HA were performed and the effect of ionic strength, dry density of bentonite and molecular weight of HA was investigated.Bentonite used in this study was Kunigel V1® (Kunimine Industry Co. Ltd.), which contains around 50wt% of Na-montmorillonite. It was compacted into 20 mm in diameter and 5 mm in thickness with dry density from 1.2 to 1.6 Mg/m3, and saturated with experimental solution, from 0.001 to 1 M NaCl. After saturation, 500 mg/L of purified Aldrich HA was added into the tracer-reservoir of the T-D cell. In some batches, 1x10-6 M of NdCl3 was also added. The pH was adjusted to 8 and kept at 8±0.2 during experiment. Aliquot of the solution in the sample-reservoir of the T-D cell was taken (sample-solution) at time intervals and same volume of fresh experimental solution was added. At the end of experiment, the bentonite was sliced and each slice was dispersed in fresh experimental solution to extract HA and Nd (extracting-solution). Total organic carbon, absorbance spectrum, molecular weight distribution of HA and Nd concentration in the sample- and the extracting-solution were measured by TOC analyzer, UV-VIS spectrophotometer, size exclusion chromatography and ICP-MS, respectively. All experiments were performed under aerobic condition.Breakthrough of HA is observed in case of dry density from 1.2 to 1.6 in 1M NaCl and from 1.2 to 1.4 in 0.1M NaCl, while no breakthrough is observed in other conditions. In case of dry density 1.2 in 1M NaCl, breakthrough of Nd is observed in the presence of HA, while no breakthrough in the absence of HA. Molecular weight of HA diffusing through bentonite is estimated to be lower than 3,000, comparing with those of HA filtered by ultrafilters with MWCO from 3 kD to 100 kD. It can be considered that only HA with molecular weight lower than 3,000 can diffuse through bentonite under this experimental condition and diffusion flux of Nd increases due to high solubility of Nd-HA complex. Diffusivity and rock capacity factor will be evaluated based on both breakthrough curve and distribution in the compacted bentonite.  Wold, S. and Eriksen, T.: “Diffusion of humic colloids in compacted bentonite”, Physics and Chemistry of the Earth, 32, pp.477-484 (2007).
11:30 AM - Q5.5
Experimental Work Conducted on MgO Characterization and Hydration A, B.
Haoran Deng 1 , Yongliang Xiong 1 , Martin Nemer 2 , Shelly Johnsen 1 Show Abstract
1 Repository Performance Dept. 6712, Sandia National Laboratories, Carlsbad, New Mexico, United States, 2 Performance Assessment and Decision Analysis Dept. 6711, Sandia National Laboratories, Carlsbad, New Mexico, United States
Magnesium oxide (MgO) is the only engineered barrier certified by the EPA for emplacement in the Waste Isolation Pilot Plant (WIPP), a U.S. Department of Energy repository for transuranic waste in southeast New Mexico. MgO reduces actinide solubility by sequestering CO2 generated by the biodegradation of cellulose, plastic, and rubber. Demonstration of the effectiveness of MgO is essential for WIPP recertification. In the past, a series of experiments was conducted at Sandia National Laboratories to verify the efficacy of Premier Chemicals LLC (Premier) MgO as a chemical-control agent in the WIPP. Since the end of 2004, Premier MgO has no longer been available for emplacement in the WIPP. Martin Marietta Magnesia Specialties LLC is the new MgO supplier. MgO characterization, including chemical analysis and reactivity analysis, has been performed to quantify the amount of reactive constituents in Martin Marietta MgO. Characterization results of Premier MgO will be reported for comparison.It was previously determined that (a) particle size, (b) solid-to-liquid ratio, and (c) stirring speed, all affect the rate of carbonation of MgO slurries. Thus, it’s reasonable to hypothesize that these factors would also affect the rate of hydration. Accelerated MgO hydration experiments were carried out to test the effect of the above factors in de-ionized water at 70 Celsius. The Minitab statistical software package was used to design a fractional factorial experiment and analyze the test results. We also fit the accelerated-inundated hydration data to four different kinetic models and calculated the hydration rates. As a result of this study we have determined that different mechanisms may be important for different particle sizes, surface-control for larger particles and diffusion for small particles.A.Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE-AC04-94AL85000.B.This research is funded by WIPP programs administered by the Office of Environmental Management of the U.S Department of Energy.
11:45 AM - Q5.6
Modelling the Spatial and Temporal Evolution of pH in the Cementitious Backfill of a Geological Disposal Facility.
Joe Small 1 , Olivia Thompson 1 Show Abstract
1 Modelling and Environmental Management, Nexia Solutions Ltd, Warrington United Kingdom
The Nuclear Decommissioning Authority (NDA) disposal concept for the UK’s intermediate-level and low-level radioactive waste utilises a cementitious material (Nirex Reference Vault Backfill, NRVB) for backfilling vaults containing waste packages. The NRVB provides a surface for the sorption of radioelements, and is designed to provide long-term conditioning of the near field porewater to a high pH. This provides an environment in which the solubility of many radioelements is low and their sorption is enhanced. Calculations of the volume of NRVB required to maintain sufficiently high pH have considered the reactions between wastes, groundwater and NRVB and have been based on the assumption that the disposal vault can be considered to be spatially homogeneous.This paper presents an initial investigation under the NDA research programme into the use of multidimensional reactive-transport geochemical models to examine the spatial and temporal effects that may be expected as the NRVB interacts with groundwater and waste packages. The interactions between NRVB and waste packages includes consideration of reactions between differing cement compositions as well as reactions including that with carbon dioxide generated from cellulose containing waste packages. Models have been developed using the geochemical modelling code PHREEQC to represent the incongruent dissolution of calcium silicate hydrate (CSH), which along with Ca(OH)2 are the main cement phases that generate alkaline conditions. The precipitation of secondary mineral phases including brucite, calcite and ettringite is also considered in determining the pH of the system. The modelled changes in mineral assemblage are linked to changes in porosity, which has the potential to affect the water flow properties of the disposal facility. The PHREEQC models are included in a PHAST 3-dimensional groundwater flow and reactive chemical transport model to examine the spatial and temporal changes in pH buffering, mineral reaction and the resulting effect and feedback on porosity and solute transport. Results will be presented to illustrate the modelling approach and the results obtained. In the case of groundwater-NRVB interactions the reactions of carbonate, sulphate and magnesium will be discussed with respect to their effect on pH buffering and porosity. In the case of NRVB-waste package interactions dissolution and re-precipitation of ettringite at the margins of the waste package are of relevance to transport and structural properties. Overall, the models indicate an increase in porosity of the NRVB as calcium is leached from the NRVB and waste package grouts. Examination of the effects of flow variation confirm the strong chemical buffering capacity of the NRVB and waste package cement materials, but also highlight the effect that heterogeneity in flow properties may have on the pattern of NRVB leaching.
12:00 PM - Q5.7
Properties, Composition and Structure of Cemented Radioactive Wastes Extracted from the Mound-type Repository.
Galina Varlakova 1 , Zoya Golubeva 1 , Alexander Barinov 1 , Igor Sobolev 1 , Michael Ojovan 2 Show Abstract
1 Applied Research Centre, Moscow SIA "Radon", Moscow Russian Federation, 2 Immobilisation Science Laboratory, University of Sheffield, Sheffield United Kingdom
Cementitious wasteforms are the most frequently used in immobilisation of radioactive wastes. The most reliable assessment of long-term behaviour and cemented radioactive wastes can be done using data from field tests in conditions close to that envisaged in a real disposal environment. The Moscow SIA “Radon” carries out field tests of cementitious wasteforms since 1965 including tests in near-surface and mound-type repositories as well as in open store conditions without additional engineered barriers. This work analyses cementitious wasteforms of specific radioactivity from 3.4E5 to 1,5E6 Bq/kg extracted from a mound-type repository of cemented low-level radioactive wastes which was set in operation since 1965 [I.A. Sobolev et.al. Mater. Res. Soc. Symp. Proc. 932, 721 (2006).]. Cemented radioactive wastes in form of cubic blocks were disposed of in a mound-type repository covered by loam soil of testing site. After 40 years of storage cement blocks retained a good physical condition without visual mechanical damage, wall saltpetre, structural or colour changes caused by freeze-thaw cycles. The mound-type repository cover layer has significantly delayed the damaging action of atmospheric precipitates and freeze-thaw cycles. Properties of 40-years aged cementitious wasteforms were assessed analysing: radionuclide leaching rates, compressive strength, density and porosity. The density of blocks was in the range 1.65 – 2.9 g/cm3 whereas the porosity was from 17 to 33%. Data obtained showed that the compressive strength (5.1 - 20 MPa) is higher compared with 5 MPa required by regulatory documents. The normalised leaching rates were within 1.6E-6 – 9.4E-4 g/cm2day which are below the regulatory limit 1E-3 g/cm2day. The examination of wasteform structure using SEM/EDX and its composition using XRD have shown that the cementitious material after 40 year is a homogeneous stone-like conglomerate of grains with pores of sizes 10 - 200 mcm. As a rule the pores are filled with needle and rhomb shaped crystals. Needle shaped crystals are also present within intergrain spaces. It was noted that such structures result in an enhanced strength of cements. Recent studies have shown that pore filling with secondary minerals i.e. calcite in cementitious materials suppresses radionuclide migration. All wasteform samples evidenced XRD patterns of: portlandite, calcite, dihydrate tri-calcium silicate and alite. Some of samples have shown XRD patterns of ettringite and phases corresponding to K3Al(SO4)4. Microbiological examination of cementitious wasteforms evidenced presence of viable bacteria of different physiological groups related to Bacillus, Pseudomonas, Mycobacterium species and microscopic mushrooms. These microorganisms are strong destructors of silicate minerals including those present in the cement stone. Some of microorganisms maintain viability and are capable to activate their activity in auspicious conditions.
12:15 PM - Q5.8
RBS and micro-PIXE study of I and Cs Heterogeneous Retention on Concrete.
Ursula Alonso 1 , Tiziana Missana 1 , Miguel Garcia-Gutierrez 1 , Alessandro Patelli 2 , Nairoby Albarran 1 , Trinidad Lopez-Torubia 1 , Valentino Rigato 3 Show Abstract
1 Environmental, CIEMAT, Madrid Spain, 2 , CIVEN, Venezia- Marguera Italy, 3 , LNL - INFN, Legnaro, Padova, Italy
Cement-based materials are being used as major component in the barriers of low- level and high-level radioactive waste repositories, because they are well suited to retain radionuclides (RN) both by physical and chemical processes. To assess the long-term safety of cement barriers, the transport or retention of critical RN have to be study. Cement materials are composed by different phases and RN migration would be highly conditioned by the heterogeneous component distribution.This study proposes a methodology to evaluate the RN retention on cement materials, taking into account their heterogeneity. The methodology was tested on the Spanish reference backfill concrete. Two elements (Cs and I) with different sorption behavior onto the cements, which was evaluated by batch sorption experiments, were selected.A combination of two nuclear ion beam techniques, Rutherford Backscattering Spectrometry (RBS) and micro-Particle Induced X-Ray Emission (micro-PIXE) was applied to evaluate both the diffusion profiles and radionuclide spatial distribution onto the cement surface. Diffusion coefficients and surface distribution coefficient can be measured at micro-meter scale, so that the more relevant and reactive phases could be identified for each studied element.The applicability and limitations of the selected methodology to other relevant radionuclides are discussed.This work was partially supported by ENRESA (E) under the FISQUIA II project (05/190/007/800070) and by the EU under the Research Infrastructure Action under FP6 Contract n. 506065 EURONS- EUROpean Nuclear Structure Research.
12:30 PM - Q5.9
The Inhibition of Aluminium Corrosion in OPC Based Composite Cements.
Nicholas Collier 1 , Neil Milestone 1 Show Abstract
1 Department of Engineering Materials, The University of Sheffield, Sheffield United Kingdom
The hydration of ordinary Portland cement (OPC) produces a paste with a highly alkaline pore solution approximating pH 13.5. When used as an immobilising medium, this can be corrosive to aluminium metal based waste streams arising from the re-processing of nuclear fuel. Partial replacement of OPC by secondary cementitious materials such as blast furnace slag (BFS) and pulverised fuel ash (PFA) reduces the pH of the paste but is still high enough to cause metal corrosion. This corrosion causes the generation of hydrogen gas and the formation of expansive corrosion products, both of which may affect the integrity of the hardened wasteform. This paper reports results obtained from encapsulating aluminium metal in a range of composite cements based on the partial replacement of OPC with BFS and PFA and compares these results with those acquired from the same systems activated with near-neutral sulphate based activators.Initial results show that aluminium corrosion occurs in both the un-activated and activated systems yielding hydrogen gas and a corrosion layer around the aluminium metal. The quantity of hydrogen formed was highest for the un-activated systems. The corrosion layer of the un-activated samples consisted of a combination of C-S-H, aluminium hydroxides (gibbsite and bayerite) and strätlingite whilst the corrosion product of the activated samples contained much more C-S-H and ettringite. Visual and SEM examination suggests that the formation of the corrosion product in the activated samples is less than in the un-activated samples and the formation of the corrosion product in the former ceases at a certain age and protects the aluminium from further corrosion. The prevention of aluminium corrosion increases wasteform durability and reduces the likelihood of radionuclide discharge due to hydrogen gas release and loss of wasteform integrity due to expansion cracking.
12:45 PM - Q5.10
Experimental and Modeling Studies of WIPP Near-Field Chemistry by Sandia National Laboratories since the CRA- 2004A, B.
Laurence Brush 1 , Daniel Clayton 2 , Haoran Deng 1 , James Garner 2 , Ahmed Ismail 2 , Shelly Johnsen 3 , Martin Nemer 2 , Edwin Nowak 1 , Gregory Roselle 1 , Yongliang Xiong 1 Show Abstract
1 Repository Performance Dept. 6712, Sandia National Laboratories, Carlsbad, New Mexico, United States, 2 Performance Assessment and Decision Analysis Dept. 6711, Sandia National Laboratories , Carlsbad, New Mexico, United States, 3 Carlsbad Programs Group 6710, Sandia National Laboratories , Carlsbad, New Mexico, United States
The Waste Isolation Pilot Plant (WIPP) is a U.S. Department of Energy (DOE) repository for defense-related transuranic (TRU) waste in southeast New Mexico. The DOE submitted the WIPP Compliance Certification Application to the U.S. Environmental Protection Agency (EPA) in October 1996. The EPA certified that the WIPP complies with its regulations in May 1998 and the repository opened in March 1999. The DOE submitted the first WIPP Compliance Recertification Application (CRA-2004) to the EPA in March 2004. Sandia National Laboratories (SNL) has carried out extensive studies of WIPP near-field chemistry since the CRA-2004.The DOE is emplacing MgO in the WIPP to serve as the engineered barrier by decreasing the solubilities of the actinide (An) elements in brine. MgO will decrease An solubilities by consuming all CO2 that might be produced by microbial activity in the repository, thus preventing acidification and limiting carbonate complexation. SNL completed studies of MgO from the previous vendor, supported the selection of a new vendor, and characterized and conducted a preliminary hydration study of the new MgO. SNL supported the DOE’s April 2006 request to reduce the amount of excess MgO that it must emplace by quantifying the uncertainties in calculating the MgO excess factor; the EPA approved this request in February 2008. SNL also started a laboratory study of the carbonation of steel and Pb in the repository, which will produce H2. This study includes rigorous control of environmental parameters.SNL re-established the uncertainty ranges and probability distributions for thermodynamic predictions of the solubilities of An elements in the +III, +IV, and +V oxidation states (An(III), An(IV), and An(V)) in the WIPP by comparing experimentally measured solubilities with solubilities predicted by thermodynamic models under identical conditions. SNL conducted 243 An(III), 45 An(IV), and 136 An(V) comparisons. SNL updated the database to incorporate the results of these comparisons, made several other minor changes in the database, and studied the sensitivity of An solubilities to the concentrations of organic ligands in TRU waste. Finally, SNL recalculated An(III), An(IV), and An(V) solubilities for the CRA-2004 Performance Assessment Baseline Calculations (PABC), the current PA baseline. The strongly reducing conditions created by the processes that produce gas, and the low fCO2 and mildly basic pH imposed by MgO will result in low An solubilities.A. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE-AC04-94AL85000.B. This research is funded by WIPP programs administered by the Office of Environmental Management of the U.S Department of Energy.
Q6/O7: Joint Session: Fluorites: Actinide Fuel and Waste Forms
Wednesday PM, December 03, 2008
Back Bay D (Sheraton)
3:45 PM - **Q6.1/O7.1
Computational Approaches to UO2 Defects and Radiation Damage.
Mark Asta 1 , Sergey Barabash 2 , Anurag Chaudhry 3 , Niels Gronbech-Jensen 3 , Benjamin Hanken 1 , Byoungseon Jeon 3 , Yongduo Liu 2 , Vidvuds Ozolins 2 , Alex Thompson 4 , Pratyush Tiwary 5 , Axel van de Walle 5 , Chris Wolverton 4 Show Abstract
1 Chemical Engineering and Materials Science, University of California at Davis, Davis, California, United States, 2 Materials Science and Engineering, University of California, Los Angeles, Los Angeles, California, United States, 3 Applied Science, University of California at Davis, Davis, California, United States, 4 Materials Science and Engineering, Northwestern University, Evanston, Illinois, United States, 5 Materials Science, California Institute of Technology, Pasadena, California, United States
4:15 PM - Q6.2/O7.2
Combined First Principles and Thermodynamic Calculation of Defect Formation Energies in UO2.
Pankaj Nerikar 1 , Simon Phillpot 1 , Susan Sinnott 1 Show Abstract
1 Materials Science & Engineering, University of Florida, Gainesville, Florida, United States
Uranium oxide is used as the standard nuclear fuel in pressurized water reactors, and defects are created over time by several mechanisms, including self-irradiation. These defects have an important effect on the physical properties of the fuel as they can cause swelling of the material and change the crystal structure thereby reducing the efficiency of the process. Despite their importance, the effect of temperature and oxygen partial pressure on the formation of these defects is not well understood. Here, density functional theory calculations using the Hubbard U correction term is used in combination with thermodynamic approaches to calculate the formation energies associated with intrinsic point defects and fission products. The predicted equilibrium properties and the defect formation energies match trends in the experimental literature quite well. The formation of oxygen interstitials is predicted to become increasingly difficult as higher temperatures and reducing conditions are approached. In addition, the stability of charged defects is predicted to depend to a substantial degree on the position of the Fermi level in the system. This work is supported by a DOE-NERI (DE-FC07-05ID14649).
4:30 PM - Q6.3/O7.3
Oxygen Lattice Distortions and U Oxidation State in UO2+x Fluorite Structures.
Lionel Desgranges 1 , Gianguido Baldinozzi 2 3 Show Abstract
1 DEN/CAD/DEC, CEA, St. Paul-lez-Durance France, 2 SPMS, CNRS, Chatenay-Malabry France, 3 DEN/DANS/DMN/SRMA/LA2M, CEA, Gif-sur-Yvette France
Whether UO2 has a simple fluorite structure, the specific chemical characteristic of oxygen and uranium are largely responsible for the complex features observed in the UO2+x system. In spite of the wide technological interest of uranium oxides, most of the structural features of this system are still unsettled. This is maybe the reason why a lot of speculations about the effective charge of U are still discussed overlooking some the fine structural modifications of the oxygen sub-lattice. Often, the U oxidation state changes are interpreted taking into account only a direct relationship between the lattice parameters and the U ionic radius. Obviously, a better description of the relation between lattice, structure and electronic properties is desirable. In this context, we would like to address structural features induced by different U oxidation states thanks to two examples.The first example deals with oxygen insertion in the fluorite structure during UO2 oxidation at low or intermediate temperature (less than 600K). Following Bevan and Willis pioneering works, the O increase leads only to slight modifications of the X-Ray diffraction pattern, although significant local distortions of the oxygen sub-lattice exist, as it is witnessed by neutron diffraction. Our analysis of U4O9 and U3O7 crystalline structures evidences the deformation of the oxygen coordination polyhedron linked to U5+ and U6+ oxidation state. The second example deals with electron-hole pair formation in UO2 at higher temperature (above 1200 K). In this temperature range, electrical conductivity measurements evidenced the formation of intrinsic carriers, resulting from the formal reaction 2U4+↔U3+ + U5+. Thanks to a coupled analysis of heat capacity experiments, diffraction and in agreement with other results in the literature, evidence is obtained that the U3+-U5+ formation is associated with a significant increase of the fluorite unit cell parameter (about 10%). This unit cell local deformation is in agreement with the structural features expected from the existence U3+ and U5+ coordination polyhedra, replacing the usual U4+ ones.
4:45 PM - Q6.4/O7.4
Theory of Defect Clustering in AnO2+x (An=U, Np or Pu).
David Andersson 1 , Juan Lezama 1 , Steven Conradson 1 , Blas Uberuaga 1 Show Abstract
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
5:00 PM - Q6.5/O7.5
Stress-induced Phase Transformation in Nanocrystalline UO2.
Tapan Desai 1 , Blas Uberuaga 2 , Paul Millett 1 , Dieter Wolf 1 Show Abstract
1 Material Sciences, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
We have performed Molecular Dynamics (MD) simulations using an empirical potential to study stress-induced phase transformation in nanocrystalline UO2 at T = 1000K. The columnar UO2 microstructure consists of 6 grains of identical hexagonal shape and diameter (d = 20 nm) in a three-dimensional periodic simulation cell. Under constant-stress tensile loading conditions, we found a phase transformation from the fluorite to a-PbO2 structure. The heterogeneous nucleation process of this new phase (a-PbO2) occurs at the grain boundaries and the new phase then grows toward the interior of the grain. To verify that this phase transformation seen in MD simulations is physically reasonable, density functional theory (DFT) calculations were performed. The DFT calculations agree that the a-PbO2 structure is energetically favored over the fluorite structure under certain tensile conditions. According to our knowledge, experimental validation of this phase transformation is not yet available. This work was supported by the DOE-BES Computational Materials Science Network.
5:15 PM - Q6.6/O7.6
Helium Behaviour in UO2 doped with 10 wt% of 238PuO2.
Emilio Maugeri 1 , Thierry Wiss 1 , Jean-paul Hiernaut 1 , Jean-Yves Colle 1 , Hartmut Thiele 1 , C. Sabathier 2 , Vincenzo Rondinella 1 , Rudy Konings 1 Show Abstract
1 , Institute for Transuranium Elements, Karlsruhe Germany, 2 , Commissariat à l'Energie Atomique, Centre de Cadarache, St-Paul-lez-Durance France
5:30 PM - Q6.7/O7.7
Cerium Dioxide Surface Characterization.
Nieves Rodriguez 1 , Juan Carlos Marugan 1 , Eduardo Iglesias 1 , Juan Manuel Nieto 1 , Tiziana Missana 1 , Nairoby Albarran 1 , Joaquin Cobos 2 , Javier Quinones 1 Show Abstract
1 , CIEMAT, Madrid Spain, 2 , ITU-JRC. European Commission, Karlsruhe Germany
5:45 PM - Q6.8/O7.8
Order-disorder Phase Transformation and Amorphization of Nd2Zr2O7 in Gd2Zr2O7 in the Electronic Stopping Regime.
Maulik Patel 1 , V. Vijayakumar 1 , Swaminathan Kailas 2 , Devesh Avasthi 3 , Jean-Claud Pivin 4 , Avesh Tyagi 5 Show Abstract
1 High Pressure Physics Division, Bhabha Atomic Research Center, Mumbai, Maharashtra, India, 2 Physics Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India, 3 Materials Science and Radiation Biology, Inter-University Accelerator Center , New Delhi India, 4 , Centre De Spectrometrie Nucleairie Et De Spectrometrie De Masse-In2P3, Paris, Orsay Campus France, 5 Chemistry Division, Bhabha Atomic Research Centre, Mumbai India
Q7: Poster Session: Geological Disposal
Thursday AM, December 04, 2008
Exhibition Hall D (Hynes)
9:00 PM - Q7.1
3D Random Walk Simulation of Migration Behavior of Radionuclide in Granite Core.
Yoshimi Seida 1 , Hiroyasu Takase 2 , Yuichi Niibori 3 , Hiroaki Takahashi 4 Show Abstract
1 Environmental Chemistry and Engineering, Tokyo Institute of Technology, Yokohama, Kanagawa, Japan, 2 , Quintessa Japan, Yokohama, Kanagawa, Japan, 3 Dept. Quantum Science & Energy Engineering, Tohoku University, Sendai, Miyagi, Japan, 4 Nuclear Chemical Engineering, Institute of Reserach and Innovation, Kashiwa, Chiba, Japan
3D random walk simulation was performed to examine the influence of physicochemical heterogeneity of intact rock on breakthrough behavior of diffusate/radionuclide. The internal mineral distribution measured by X-ray CT for Granite core sample (sampled from Inada of Japan, 32mm in diameter x 25mm length) was used to produce simulated cubic model with heterogeneous mineral distribution that is similar to the real granite sample. The volume rendered 3D distribution image produced from the 653 sheets of slice images of CT data indicated the random dispersed distribution of boitite and distribution of alkaline feldspar domain with continuous phase. Refer to the mineral composition and the geometric distribution of mineral in the granite, the 3D rock model was constructed based on 3D cubic cells. Random walk migration analysis in the 3D cell model was performed to investigate the relationship between sorption property of each mineral and macroscopic sorption/retardation property of the rock, and their dependence on scale of rock. The pulse-response breakthrough behavior was simulated and the breakthrough pattern was compared with the result obtained by the calculation with conventional homogeneous transport (diffusion and sorption) model. Importance of the representative elemental volume of sample in the experimental measurement of retardation parameters of rock with physicochemical heterogeneity was shown.
9:00 PM - Q7.10
Evolution of the Geochemical Conditions in the Bentonite Barrier and its Influence on the Corrosion of the Carbon Steel Canister.
Elena Torres Alvarez 1 , Alicia Escribano 1 , Maria Jesus Turrero 1 , Pedro Martín 1 , Javier Pena 1 Show Abstract
1 Engineered and Geological Barriers, CIEMAT, Madrid, Madrid, Spain
9:00 PM - Q7.11
A Thermodynamic Approach on the Effect of Salt Concentration on Swelling Pressure of Water-saturated Bentonite.
Haruo Sato 1 Show Abstract
1 Geological Isolation Research & Development Directorate, Japan Atomic Energy Agency, Horonobe-cho, Teshio-gun Japan
In the safety assessment of geological disposal for high-level radioactive waste in Japan, bentonite is used as the buffer material which is one of the engineered barriers composing the multi-barrier system and as one of the backfill materials. Since the major clay mineral constituent of bentonite is smectite (montmorillonite) which has the nature of swelling, bentonite swells by contacting with groundwater and restricts the groundwater flow. In addition, swelling pressure develops by hydration of the interlayer of smectite. It is known from past studies that swelling pressure of bentonite is affected by salinity, smectite content, interlayer cation, dry density of the bentonite, silica sand content, saturation degree and temperature. However, no general model on swelling pressure has been reported. In this study, the effect of salt concentration on the swelling pressure of compacted bentonite was modelled based on the thermodynamic data of water at smectite surface and of water in solutions of various salt concentrations coming in contact with the smectite. This paper presents a general thermodynamic model to calculate swelling pressure for various kinds of bentonites saturated with solutions of various salinities in addition to measurements of the thermodynamic data of the water at the smectite surface. At first, activities (aw) of the water at the smectite surface were obtained as a function of water content (0-83%) and temperature (288-313K) in a range of smectite density 0.6-0.9Mg/m3 by a vapor pressure method, and the relative partial molar Gibbs free energies (dGw) were determined based on the aw values. Both aw and dGw values decreased with a lowering of water content in the region where water content is about 40%. Since water affected from smectite surface was deduced to be up to a distance of approximately 2 water layers from the correlation between the aw versus water content and the specific surface area (800m2/g) of the smectite, the affected water is regarded as almost all interlayer water. Swelling pressure versus smectite partial density which is the density focused on only part of the smectite in bentonite was estimated for solutions of various salinities ([NaCl]=0.2-3.4M) and compared to data measured under the same salinities for various kinds and different silica sand contents of bentonites. The calculated swelling pressure decreased with increasing salt concentration, while little difference was found in a range of high smectite partial density. The effect of salinity on the swelling pressure was not clear in the measured data and within the range of the scattering in the measured data. This cause is presumed to be due to the increase of the ionic strength of porewater by dissolution of soluble minerals contained in the bentonites, and the effect of salinity on the swelling pressure can be regarded to be at most within the range of the scattering in the measured data.
9:00 PM - Q7.12
Influence of Humic Acid on Sorption of Sedimentary Rock for Se and Th.
Yoshimi Seida 1 , Yukio Tachi 1 , Akira Kitamura 1 , Toshiyuki Nakazawa 2 , Norikazu Yamada 2 Show Abstract
1 Geological Isolation Research Unit, Japan Atomic Energy Agency, Tokai, Ibaraki, Japan, 2 Naka Energy Research Laboratory, Mitsubishi Materials Corporation, Naka, Ibaraki, Japan
Organic substances dissolved in ground water and/or sorbed onto natural rocks is one of uncertainties that may influence the migration of radionuclides in the rocks and the retardation capacity of rocks in deep geological disposal of high level radioactive waste. The influence on the retardation capacity will depend on characteristic of the organic substances, partition in the radionuclide-organic substance-rock system and pore structure of the rock although the details of them have not been well understood, especially for Se and Th. In the present study, influence of humic acid on both solubility of Se(IV) and Th(IV) and their retardation parameters in sedimentary rock was studied using the rock sample obtained at Horonobe generic URL site in simulated ground water under oxygen free condition. The ionic composition of simulated ground water was determined based on the analysis of ground water at the URL site (NaCl and bicarbonate ions dominant). Humic acid supplied from Aldrich was used as a model organic substance after purification by a conventional method. The stability of humic acid dissolved in the simulated ground water was confirmed not to coagulate and precipitate over one month by monitoring total organic carbon (TOC) and pH of the solution that was filtered by 0.45 µm and ultrafiltration (size exclusion: 2 nm, molecular weight 10000) membranes at each time. Then, the influence of humic acid on the solubility of Se(IV) and Th(IV) was examined at the range of 0.1~100 mg dm-3 humic acid and 6~9 of solution pH. The filtrates of 0.45 µm and ultrafiltraion membranes were analyzed for each mixture. The 75 % of humic acid was excluded by both membranes, indicating the existence of large molecules with more than 2 nm in size. The exclusion of humic acid was almost the same in the cases with Se(IV) and Th(IV) at the examined range of pH and the concentration of radionuclides (Se: 1.0x10-4 M, Th: 1.1x10-6 M). This means that the radionuclides did not produce larger molecules through their complexation with the excluded humic acid and/or any aggregation, and the solubility of the elements kept high enough in the experimental range. Sorption behavior of the crushed sedimentary rock (~ 200 µm in diameter) was evaluated by batch sorption method. The partition coefficient, Kd, of the rock for Se(IV) with humic acid was around 4 ml g-1 and little influence of the humic acid on the Kd was observed. The Kd for Th(IV) was estimated to be more than 1000 ml g-1 below 100 mg dm-3 humic acid but the Kd was largely decreased when the concentration of humic acid was 100 mg dm-3. The through diffusion experiment showed that the coexistence of humic acid did not affect the diffusion of Se(IV). In the case of Th(IV), breakthrough of Th(IV) has not been observed for 6 months due to its large Kd.*This study was partly financed by the METI of Japan.
9:00 PM - Q7.13
Sorption and Diffusion of Cs in Horonobe-URL's Sedimentary Rock : Comparison and Model Prediction of Retardation Parameters from Batch Sorption and Diffusion Experiments.
Yukio Tachi 1 , Yoshimi Seida 1 , Reisuke Doi 1 , Xiaobin Xia 1 , Mikazu Yui 1 Show Abstract
1 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency, Tokai, Ibaraki Japan
Sorption and diffusion of radionuclides in deep geological environment are the key processes in the safe geological disposal of high-level radioactive waste. Sorption and diffusion of Cs for the sedimentary rock in the Horonobe generic URL were examined from the viewpoints of reliability of evaluation method and consistency of transport model in the present study. The retardation parameters of the rock, Kd and De, for Cs were measured by both batch sorption and intact diffusion experiments. The obtained sorption parameters were compared between both methods, and were also simulated based on geochemical model calculation to interpret the sorption behavior theoretically under the assumption of dominant sorption minerals and their sorption mechanism. The rock samples were obtained from the Wakkanai formation at depth of 500 and 600 m in borehole HDB-6 of Horonobe URL site. Synthetic ground water used in each experiment was prepared based on the dominant composition of ground water under atmospheric condition (Na, NH4, K are dominant cations). Batch sorption experiments with the crushed rocks were carried out as a function of reaction time and concentration of Cs in the liquid phase. Sorption of Cs reached equilibrium within a short time and sorption isotherms showed Henry type at Cs concentration below 1x10-5M. The partition coefficients, Kd, obtained from the isotherms were around 40 ml/g. Diffusion and sorption behaviors of the intact rock samples were observed based on through-diffusion method. Both tracer depletion and breakthrough curves were monitored until the system attained quasi-steady-state followed by analysis of concentration profile inside the rock sample. The concentration profiles in the rocks were obtained by means of sectioning, crushing and acid-extraction of Cs after the diffusion experiment. The tracer depletion, breakthrough and inner concentration curves were all simulated simultaneously by conventional transport model with diffusion and equilibrium sorption using one set of the fitting retardation parameters. The Kd values obtained from the fitting were consistent with those obtained by the batch experiment with crushed rock. Based on mineralogy of the sedimentary rock, Cs retention is supposed to be dominated by sorption onto clay minerals such as illite and smectite. Selective sorption of Cs onto illite was confirmed by EPMA and SEM observations. The sorption behavior was simulated using the ion-exchange model for illite and smectite using equilibrium adsorption constants published elsewhere. The model predicted the Kds obtained by the series of experiments fairy well. *This study was partly financed by the Ministry of Economy, Trade and Industry of Japan.
9:00 PM - Q7.14
Migration Behavior of Bentonite Colloids through a Fractured Rock.
Yoshio Kuno 1 , Hiroshi Sasamoto 1 Show Abstract
1 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency, Tokai-Mura, Ibaraki, Japan
Bentonite colloids generated from the buffer material by groundwater flow might influence the radionuclide migration from an underground repository for high level radioactive waste because of their high sorption capability for most radionuclides. These radionuclides sorbing onto the colloids may not be retarded due to the exclusion from matrix pores in rock, therefore it is an important issue to clarify the migration behavior of colloids in geological medium to estimate their effects on radionuclide migration. In this study, column experiments with bentonite colloids were carried out using an artificial fracture within granite rock to investigate the migration behavior of bentonite colloids, especially the possibility of colloid filtration through the fractured rock. Bentonite colloids formed from an industrially refined Na-type bentonite 'Kunipia F®' (montmorillonite content >99%) were prepared at the concentration of 70 mg/L. Different solutions, such as distilled water, 10-4 M and 10-3 M NaCl solutions, were used to confirm the effect of the ionic concentrations in the colloidal solutions. These solutions were constantly injected into the granite column with a single artificial fracture (lengths: 15 cm and 30 cm, aperture: 1 mm), using low flow rate (2 cm/hr). Fractions of the effluent were collected by a sampler and the concentrations of Si were measured by ICP-AES analyzer. A portion of each fraction was filtered by ultrafilters (MWCO: 10,000) and the concentration of Si in the filtrate was also measured to exclude the amount of dissolved ions. The colloidal concentrations were determined by the content of Si, which was the major element in the bentonite. As the results, migrating bentonite colloids was not filtered in distilled water and 10-4 M NaCl solution because the colloidal particles passed through the granite column with bulk fluid and normalized effluent concentration (C/C0) readily increased to 1. In the case of 10-3 M NaCl solution, rapid increase of the breakthrough curves was similarly observed in an early stage of the migration experiments. After that, however, the breakthrough curves transited to an almost steady stage where the C/C0 value was less than 1. The same tendency was observed in both cases of column size. It seems that bentonite colloids are filtered through a fractured rock due to the interaction between colloids and rock surfaces that is affected by the ionic concentration. These results indicate that a sufficient fraction of bentonite colloids is expected to be filtered within the fractured granite under comparatively high ionic concentration (10-3 M NaCl). The results also suggest that bentonite colloids may possibly not only enhance the radionuclide migration but also retard it by the sorption onto the filtered colloids. The basic information on filtration of bentonite colloids in the geological medium would contribute to more realistic estimation of colloidal influence on the radionuclide migration.
9:00 PM - Q7.15
Transport of Uranium in a Granite Fracture: Effect of the Presence of Bentonite Colloids.
Nairoby Albarran 1 , Tiziana Missana 1 , Ursula Alonso 1 , Miguel Garcia-Gutierrez 1 , Manuel Mingarro 1 , Trinidad Lopez 1 Show Abstract
1 Departamento de Medioambiente, CIEMAT, Madrid, Madrid, Spain
9:00 PM - Q7.16
Filtration of Gold Colloids in Compacted Bentonite.
Michael Holmboe 1 , Susanna Wold 1 Show Abstract
1 School of Chemical Science and Engineering, Royal Institute of Technology, Stockholm Sweden
In a final repository for nuclear waste, water saturated and compacted bentonite presents both low hydraulic permeability and ability to prevent or at least slow down transport of especially positively charged contaminants from a faulted canister. It is generally believed that the colloid transport in the actual bentonite can be disregarded, even though studies concerning colloid transport in water saturated bentonite are merely a handful. The high swelling pressure of saturated compacted bentonite determines the microstructure by constricting available pores and decreasing the interlamellar distance and thus is seen as an effective filter for colloids. At present there are actually discrepancies in the results between colloid transport studies in compacted bentonite using organic colloids and studies using inorganic colloids. In case of canister failure, radiocolloids in the close proximity of the fuel can be formed. The bentonite also contains a small amount of organic material which if mobilized can enhance transport of radionuclides . Hence, experimental observations from column experiments or in situ experiments are vital for the understanding of colloid diffusion in compacted bentonite, and to quantify the filtration capacity of the compacted bentonite. In this study transport of 2, 5 and 15 nm gold colloids was studied in column experiments during 4 months using MX-80 bentonite with varying dry density, 0.5-2.0 g/cm3
. The concentration of gold colloids in the outlet solutions were analyzed with ICP-OES and the stability of 5 and 15 nm gold colloids in the inlet solutions were monitored by UV/VIS spectroscopy. From the diffusion experiment, breakthrough of gold colloids was not found in any of the three diffusion experiments. However, from concentration profiles in the bentonite, small amounts of gold were found and seen to decrease along the direction of transport. Using a finite difference based model, ANADIFF, the largest gold colloid diffusivities possible according to the actual experimental setup and results was found to be on the order of 10-15
 S. Wold, T. Eriksen. (2007) Diffusion of humic colloids in compacted bentonite. Phys. Chem. Earth, 32, 477-484.
9:00 PM - Q7.17
A Systematic Approach to Evaluate the Importance of Key Uncertainties Affecting the Geological Disposal of Radioactive Wastes.
Takao Ohi 1 , Manabu Inagaki 1 , Makoto Kawamura 1 , Takeshi Ebashi 1 Show Abstract
1 Geological Isolation R&D Directorate, Japan Atomic Energy Agency, Naka-gun Ibaraki-ken Japan
In the safety assessment of a geological disposal system for radioactive wastes, key uncertainties including potentially detrimental factors have been identified and researched extensively. Being able to demonstrate the relative importance of these key uncertainties systematically is a critical issue to conduct the relevant research effectively and to improve the reliability of safety assessment.It is considered that the relative importance of these key uncertainties in the geological disposal research should be judged from the viewpoint of the impact on the safety of the disposal system and that the impact of the key uncertainties on the safety should be evaluated quantitatively. This is based on the consideration of the change of the conditions of the interconnected field on the propagation of the impact. In the current study, a systematic approach was developed to represent the relative importance of the key uncertainties based on the following manner: (1) Establishing a Total Assessment Work Frame (TAWF) for representing the change of the conditions of the interconnected field with unified structure for all scenarios.(2) Establishing an organization procedure to evaluate the relative importance based on the TAWF.In the establishment of TAWF, several “field conditions” affected by key uncertainties, such as “natural phenomena”, “surface condition”, “geological condition”, “design specification”, “disposal condition”, “safety assessment” are set as factors of TAWF. In the establishment of the organaization procedure, the changes in each “field conditions” are classified in terms of T [Thermal], H [Hydrological], M [Mechanical], C [Chemical], and G [Geometrical] behavior. Also, the information with respect to the changes, the treatment of the changes in the assessment and the relationship between the changes and the key uncertainties is organized as quantitative as possible by using typical assessment parameters linked to THMCG behavior. Subsequently, the impacts of the key uncertainties are typified by using the treatment of the changes in the assessment and the value of typical assessment parameters. Furthermore, importance evaluation of the typified parameters is conducted by the feedback of the information obtained from exi