Darrell S. Dunn Southwest Research Institute
Christophe Poinssot CEA Saclay
Bruce Begg Australian Nuclear Science & Technology Organisation (ANSTO)
NN1: Spent Fuel
Monday AM, November 27, 2006
Constitution A (Sheraton)
9:30 AM - **NN1.1
Corrosion of Nuclear Fuel Inside a Failed Copper Nuclear Waste Container.
David Shoesmith 1 Show Abstract
1 Chemistry, University of Western Ontario, London, Ontario, Canada
10:00 AM - NN1.2
Alteration Behavior of High Burnup Spent Fuel in Salt Brine Under Hydrogen Overpressure and in Presence of Bromide.
Andreas Loida 1 , Bernhard Kienzler 1 , Volker Metz 1 Show Abstract
1 Institut für Nukleare Entsorgung, Forschungszentrum Karlsruhe, Karlsruhe Germany
10:15 AM - NN1.3
Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel.
Jeffrey Fortner 1 , A. Kropf 1 , James Cunnane 1 Show Abstract
1 Chemical Engineering Division, Argonne National Laboratory, Argonne, Illinois, United States
Neptunium is an important radionuclide in dose contribution according to many performance assessment models of the U.S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin (~ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U4+ environment. Available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. This work was funded by the U.S. Department of Energy (USDOE), Office of Civilian Radioactive Waste Management (OCRWM), Office of Repository Development, and the OCRWM Office of Science and Technology and International. Use of the APS was supported by the U. S. Department of Energy, Basic Energy Sciences, Office of Energy Research (DOE-BES-OER), under Contract W-31-109-Eng-38.
10:30 AM - NN1.4
Modeling the Distribution of Acidity within Nuclear Fuel (UO2) Corrosion Product Deposits and Porous Sites.
Woo-Jae Cheong 1 , Zack Qin 1 , J. Wren 1 , David Shoesmith 1 Show Abstract
1 , The University of Western Ontario, London, Ontario, Canada
10:45 AM - NN1.5
Gamma-Radiolysis of NaCl Brine in the Presence of UO2(s): Effects of Hydrogen and Bromide
Volker Metz 1 , Elke Bohnert 1 , Kelm Manfred 1 , Schild Dieter 1 , Juergen Reinhardt 2 , Kienzler Bernhard 1 Show Abstract
1 Institute for Radioactive Waste Disposal, Research Center Karlsruhe, FZK-INE, Karlsruhe Germany, 2 , Leibniz-Institut f. Oberflaechenmodifizierung, Leipzig Germany
In the safety case of a geological disposal for spent nuclear fuel, SNF, radiolysis of groundwater is of high importance at least in the first 10.000s years of the post closure phase. In the case of a repository in rock salt, access of brines to the waste is considered as an accident scenario. Radiolytic decomposition of NaCl brine is accompanied by the formation of H2 and O2. Hydrogen, e.g. formed by corrosion of Fe-bearing waste containers, inhibits significantly the radiolytic decomposition of aqueous solution and consequently the oxidative SNF dissolution. Previous studies on gamma-radiolysis in homogeneous salt brine have shown that radiation chemically active contaminants such as Br- compete with hydrogen for oxidative radiolytic products. As a consequence a reliable protective H2 effect with respect to radiolysis and SNF corrosion is reduced if not annihilated. Still, there is a lack of experimental data to determine the radiation chemical reactions which control the competing effects of H2 and Br-. In the present study, we measure radiolytic decomposition of concentrated NaCl solution under external gamma-irradiation. Most experiments are conducted in autoclaves at elevated pressures, to suppress formation of a gas phase. One series of autoclave experiments is conducted in homogenous solution; in a 2nd series of experiments UO2(s) pellets are present in solution. Initial concentrations of dissolved H2 and Br- vary between 0 < [H2] < 0.01 M and 0 < [Br-] < 0.001 M, respectively. The bromide concentration range is relevant to natural brines as well as to groundwater of argillaceous and crystalline formations, whereas the hydrogen concentration range is representative to a pressure built-up by corrosively formed H2 in a deep geological repository. In addition to these autoclave experiments, radiolysis of NaCl solution in presence of UO2(s) is studied at ambient conditions.In autoclave experiments with Br-, steady state concentrations of radiolytically formed H2 and O2 are significantly higher compared to those in experiments in the absence of bromide. The presence of UO2(s) does not affect the radiolysis yield at the studied conditions. In presence UO2(s), concentration of dissolved U increased to [U] > 10e-7 M, which corresponds to the solubility of schoepite / Na-diuranate transition. Our results indicate that an H2 inhibition effect on radiolysis gas production and on UO2(s) corrosion is significantly reduced by Br-, even at a concentration ratio of [H2] / [Br-] > 100. The present observations are related to the competitive reactions of OH radicals with H2, Cl- and Br-. It is expected that H2 and Br- have similar effects on the yield of gamma-radiolysis products in solutions of low Cl- concentration.
11:30 AM - NN1.6
Reactivity of UO2 in Solutions of Variable Bicarbonate: Results from Vertical Scanning Interferometry (VSI)
Jonathan Icenhower 1 , Michael Vinson 2 , Andreas Luttge 2 , Julia Glovack 1 , Dawn Wellman 1 , Eric Pierce 1 Show Abstract
1 Applied Geology and Geochemistry, Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Department of Earth Science, Rice University, Houston, Texas, United States
11:45 AM - NN1.7
Dissolution Behaviour of UO2 in Anoxic Conditions Comparison of Ca-bentonite and Boom Clay.
Christelle Cachoir 1 , Karel Lemmens 1 , Thierry Mennecart 1 Show Abstract
1 , SCK-CEN, Mol Belgium
A general concept for the disposal of spent fuel in clay formations is based on the multibarrier principle. In such concept, the barriers for radionuclides released into the environment are the clay host rock, the backfill, the canister overpack and the fuel itself. The innermost barrier is the dissolution of UO2 matrix, which is the main component of spent fuel. However the dissolution of UO2 upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. In order to determine in how far the clay properties influence the dissolution of spent fuel, two different kinds of clay were considered: Ca-bentonite which presents an initial oxidizing environment, and Boom Clay which is characterized by its strong reducing capacity. The experiments were carried out with depleted UO2 in presence of either compacted dry Ca-bentonite with Boom Clay groundwater or compacted dry Boom Clay with Boom Clay groundwater. The leach tests were performed at 25°C in anoxic atmosphere (glove box under 0.4%CO2/99.6%Ar) for 2 years. The U concentrations were sampled during these 2 years, once every 6 months and the amount of U was determined in the clay after 2 years in order to determine the dissolution rate. After 2 years, an unexpected uranium concentration was found 50 times higher in the system Boom Clay with Boom Clay water (2.10-7mol/L) than in the system Ca-bentonite with Boom Clay water (4.10-9 mol/L), maybe resulting from a larger colloidal fraction in the system Boom Clay with Boom Clay water. Final results are expected to allow the comparison of the U retention capacity of Ca- bentonite and Boom Clay in anoxic conditions with the U retention of Boom Clay found in reducing conditions .- S.Salah, C.Cachoir, K. Lemmens and N. Maes. Static dissolution tests of alpha-doped UO2 under repository relevant conditions: Influence of Boom Clay and alpha-activity on fuel dissolution rates. MRS 2005, 12-16 September 2005, Ghent-Belgium.
12:00 PM - NN1.8
Coffinite (USiO4) Under Reducing and Oxidizing Conditions: Implications for Spent Nuclear Fuel (SNF) Disposal.
Artur Deditius 1 , Satoshi Utsonomiya 1 , Rodney Ewing 1 Show Abstract
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Coffinite (USiO4 ● nH2O; n~2) is the dominant alteration product of UO2 under reducing conditions, however, the conditions and mechanisms for its formation are still poorly understood. This study presents detailed observations of a micro-geochemical system involving coffinite and organic matter, under both reducing and oxidizing conditions in order to investigate: (i) the stability of coffinite, (ii) the effects of organic matter on formation of coffinite and U6+ alteration phases, (iii) association of fission product elements with the coffinite. Samples of a hydrothermal uranium deposit from Grants Region, New Mexico, USA, were analyzed by means of scanning electron microscopy (SEM) electron microprobe analyses (EMPA) and high-resolution transmission electron microscopy (HRTEM). The coffinite occurs at the grain boundaries of a sandstone, and this nearly closed system provides a unique opportunity to investigate the stability of coffinite under both oxidizing and reducing conditions. The SEM and HRTEM results showed that radial concentrations of well crystalline, coarse-grained coffinite precipitated between the layers of organic matter (10-20 μm thick) that contain randomly oriented 10 nm nanocrystals of coffinite. Na-boltwoodite (Na,K)(UO2)(SiO3OH)(H2O)1.5, soddyite (UO2)2(SiO4)(H2O)2, jáchymovite (UO2)(SO4)(OH)14(H2O)13 and minor uranopilite (UO2)6(SO4)(OH)10(H2O)12 and Na-zippeite (Na)4(UO2)6(SO4)3(OH)10(H2O)4 precipitated subsequently as alteration phases that formed at the expense of coffinite on the surface of a broken “screen” or covering of organic matter. In addition, numerous clausthalite (PbSe) inclusions (~1μm) occurred in coffinite grains, suggesting that coffinite can also retard some fission product elements (e.g., Se) from further migration. The organic matter appears to act as an effective “screen” and to prevent coffinite from altering, even under oxidizing conditions. Under both reducing and oxidizing conditions, the presence of organic matter may accelerate the precipitation of U-minerals, possibly by the complexation of UO22+ by humic acids. Our results suggest that the organic matter with high silica concentrations influences the mobility of U(IV) and U(VI) under the hydrothermal conditions by precipitation of coffinite, uranyl silicates and sulfates. These results support the concept of using organic matter as a part of multi-barrier systems of SNF repository sites.
12:15 PM - NN1.9
Thermodynamics of Formation of Soddyite ((UO2)2(SiO4)(H2O)2).
Lena Mazeina 1 , Drew Gorman-Lewis 2 , Jeremy Fein 2 , Jennifer Szymanowski 2 , Peter Burns 2 , Alexandra Navrotsky 1 Show Abstract
1 Thermochemistry Facility, University of California at Davis, Davis, California, United States, 2 Civil Engineering and Geological Sciences, University of Notre Dame, Notre Dame, Indiana, United States
12:30 PM - NN1.10
Hydrothermal Phase Relations Among Uranyl Minerals at the Nopal I Natural Analog Site.
William Murphy 1 Show Abstract
1 Geological and Environmental Sciences, California State University, Chico, Chico, California, United States
Isolation of nuclear wastes in a geologic repository at Yucca Mountain, Nevada, will be affected by the properties of the relatively stable alteration products of spent nuclear fuel and other engineered barrier components. Prior studies have established similarities in the paragenesis of secondary uranyl minerals at the Nopal I natural analog site, Chihuahua, Mexico, and in experiments representing alteration of spent nuclear fuel in conditions like Yucca Mountain but at elevated temperature. Geochemical conditions varied at Nopal I, and published interpretations of mineralization at Nopal I suggest processes ranging from moderate hydrothermal alteration to low temperature leaching and precipitation. Empirical and estimated thermodynamic data permit evaluations of uranyl mineral stabilities as functions of environmental conditions including water chemistry, partial pressures of reactive gases, and temperature. Calculated thermodynamic phase relations can help interpret the conditions of uranyl mineral formation at Nopal I and to refine the analogy between Nopal I and Yucca Mountain. Negative standard enthalpies of dissolution indicate that uranyl minerals have decreasing solubility constants with increasing temperature, which is countered in environments such as those at Nopal I and Yucca Mountain by the increasing stability of aqueous uranyl carbonate complexes with increasing temperature. The stability of uranophane, the predominant uranium mineral at Nopal I, increases sharply with increasing temperature relative to other common uranyl minerals, such as metaschoepite and soddyite. Soddyite is calculated to be stable relative to metaschoepite at moderately high aqueous silica activities characteristic of water in silicic rocks at Nopal I or Yucca Mountain. The paragenetic sequence of mineral formation from schoepite to soddyite to uranophane, which is observed petrographically in Nopal I samples, may reflect increasing activities of environmental components, silica and calcium. The sequence of schoepite or soddyite to uranophane also may represent prograde hydrothermal alteration. The occurrence and long-term stability of schoepite at Nopal I, as well as the persistence of primary uraninite, are indications that strong geochemical gradients affect mineral stabilities on time scales of millions of years. The latest forming uranium bearing minerals at Nopal I comprise opal and calcite. These phases are likely to be tertiary alteration products of secondary uranyl minerals formed under conditions similar to those that slowly produce opal and calcite vein mineralization at ambient conditions in the unsaturated zone at Yucca Mountain.
12:45 PM - NN1.11
Influence of Temperature on the Corrosion of Uranium Dioxide Nuclear Fuel.
Michael Broczkowski 1 , Jamie Noel 1 , David Shoesmith 1 Show Abstract
1 Chemistry, The University of Western Ontario, London, Ontario, Canada
NN2: Performance Assessment and Models
Monday PM, November 27, 2006
Constitution A (Sheraton)
2:30 PM - **NN2.1
Regulatory Perspective on Implementation of a Dose Standard for a One-Million Year Compliance Period.
Timothy McCArtin 1 Show Abstract
1 , U.S. Nuclear Regulatory Commission, Washington, District of Columbia, United States
The disposal of high-level radioactive wastes in a potential geologic repository at Yucca Mountain, Nevada is governed by the U.S. Environmental Protection Agency (EPA) standards and U.S. Nuclear Regulatory (NRC) regulations. The U.S. Department of Energy (DOE) must comply with these regulations for NRC to authorize construction and license operation of a potential repository at Yucca Mountain. As mandated by the Energy Policy Act of 1992, NRC's regulations are to be consistent with the radiation protection standards issued by EPA. In 2005, as a result of a Court decision, EPA proposed changes to its standards for Yucca Mountain for doses that could occur after 10,000 years up to 1 million years. The EPA's proposed standards for this unprecedented regulatory period includes both a limit for the peak dose after 10,000 years, as well as criteria DOE must use in the treatment of specific features, events and processes (FEPs) in performance assessments for estimating dose. Because of the uncertainties associated with estimating performance over very long times, EPA proposed specific constraints on the consideration of FEPs. For example, EPA proposed that the nature and degree of climate change may be represented by constant climate conditions (NRC is to specify the values to represent the climate change). Consistent with EPA’s proposal, NRC proposed the use of a deep percolation rate from 13 to 64 mm/year (0.5 to 2.5 inches/year) to represent the effect of future climate in performance assessments after 10,000 years. This ``average'' deep percolation rate represents the average amount of water flowing to the repository horizon. NRC is currently reviewing its performance assessment models and techniques to assure they are consistent with EPA’s proposed requirements for the period after 10,000 years to assist the review of a potential license application from the DOE. Consideration is being given to what types of performance assessment subsystem results would be most useful for understanding the behavior of the repository over very long times. For example, the failure time of the waste package may be useful for understanding doses over a 10,000 year period, however, over much longer time periods(e.g., one million years) understanding the impact of the waste package as a barrier may require futher information related to the long-term degradation of the waste package (e.g., increased water flux into the waste package as waste package degradation processes continue over long times). Additionally, as the performance assessment tools and techniques are revised, the NRC will be conducting sensitivity analyses to identify additional sensitivities associated with estimating doses over very long time periods that may indicate a benefit of futher subsystem results to improve understanding of performance of a potential repository at Yucca Mountain.
3:00 PM - NN2.2
Analysis of the Maximum Disposal Capacity for Commercial Spent Nuclear Fuel in a Yucca Mountain Repository.
Michael Apted 1 , John Kessler 2 , Frank Schwartz 3 , Wei Zhou 1 , Fraser King 4 , John Kemeny 5 , Ben Ross 6 , Alan Ross 7 Show Abstract
1 , Monitor Scientific LLC, Denver, Colorado, United States, 2 , Electric Power Research Institute, Charlotte, North Carolina, United States, 3 Department of Geology, The Ohio State University, Columbus, Ohio, United States, 4 , Integrity Corrosion Consulting Ltd., Calgary, Alberta, Canada, 5 Department of Mining Engineering, University of Arizona, Tucson, Arizona, United States, 6 , Disposal Safety Inc., Washington, District of Columbia, United States, 7 , Alan Ross and Associates, San Jose, California, United States
3:15 PM - NN2.3
A Probabilistic-Micromechanical Methodology for Assessing Zirconium Alloy Cladding Failure.
Yi-Ming Pan 1 , Kwai Chan 2 , David Riha 2 , Vijay Jain 1 Show Abstract
1 , CNWRA-Southwest Research Institute, San Antonio, Texas, United States, 2 , Southwest Research Institute, San Antonio, Texas, United States
4:30 PM - **NN2.4
Nuclear Waste Package Performance in European Geological Disposal Concepts.
Bernd Grambow 1 Show Abstract
1 SUBATECH, Ecole des Mines de Nantes, Nantes France
5:15 PM - NN2.6
New Trends in the Field of Spent Nuclear Fuel Radionuclides Release in a Deep Geological Repository.
Christophe Poinssot 1 , Cécile Ferry 1 , Arnaud Poulesquen 1 Show Abstract
1 Nuclear Energy Division, Department of Physics and Chemistry, CEA, Gif sur Yvette France
5:30 PM - NN2.7
The Brag and GM2003 Models for Glass Dissolution.
Marc Aertsens 1 , Bernd Grambow 2 Show Abstract
1 Waste&Disposal, SCKCEN, Mol Belgium, 2 , Subatech, Nantes France
5:45 PM - NN2.8
Microstructural Characterization of U Coprecipitated Phases Formed in bentonic-granitic Groundwater and under Anoxic Conditions.
Javier Quinones 1 , Eduardo Iglesias 1 , Aurora Martinez Esparza 2 , Jose Gomez de Salazar 3 Show Abstract
1 Energy, Ciemat, Madrid Spain, 2 , Enresa, Madrid Spain, 3 Matrials Science, Universidad Complutense de Madrid, Madrid Spain
NN3: Poster Session: Spent Fuel, Corrosion and Waste Characterisation
Monday PM, November 27, 2006
Exhibition Hall D (Hynes)
9:00 PM - NN3.1
Flowsheet Development for Dissolution and Disposition of Savannah River Site Legacy Plutonium and Uranium-Containing Materials.
Ann Visser 1 , Glen Kessinger 1 , Robert Pierce 1 Show Abstract
1 Actinide Technology Section, Savannah River National Laboratory, Aiken, South Carolina, United States
The U.S. Department of Energy (DOE) has legacy plutonium and uranium containing materials from processes dating back to the 1950s. Some of these materials have been consolidated at the DOE's Savannah River Site (SRS) and are being dispositioned as the result of SRS facility deactivation, deinventory, and decommissioning. The majority of these materials contain Pu and/or U in alloys, metallic phases, and oxides in the form of test materials, nuclear reactor fuel fabrication residues, and scrap metal materials. These legacy materials were tested for flowsheet development to determine dissolution rates and, for some, hydrogen generation, prior to starting dissolution campaigns in SRS processing facilities. This presentation will highlight the characterization of samples of Pu and/or U metal alloys, oxides, and fluorides as well as SRS glovebox sweepings.
9:00 PM - NN3.10
Synthesis and Performance of Fe-based Amorphous Alloys for Nuclear Waste Applications.
Kjetil Hildal 1 , John Perepezko 1 , Larry Kaufman 2 Show Abstract
1 Materials Science and Engineering, University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 , CALPHAD Inc., Brookline, Massachusetts, United States
Recent developments in multicomponent Fe-based amorphous alloys have shown that these novel materials exhibit outstanding corrosion resistance compared to typical crystalline alloys such as high-performance stainless steels and Ni-based C-22 alloy. During the past decade, amorphous alloy synthesis has advanced to allow for the casting of bulk metallic glasses. In several Fe-based alloy systems it is possible to produce glasses with cooling rates as low as 100 K/s. At such low cooling rates, there is an opportunity to produce amorphous solids through industrial processes such as thermal spray-formed coatings. Moreover, since cooling rates in typical thermal spray processing exceed 1000 K/s, novel alloy compositions can be synthesized to maximize corrosion resistance (i.e. adding Cr and Mo) and to improve radiation compatibility (adding B) and still maintain glass forming ability. The applicability of Fe-based amorphous coatings in typical environments where corrosion resistance and thermal stability are critical issues has been examined in terms of amorphous phase stability and glass-forming ability. For example, a wedge casting technique has been applied to examine bulk glass forming alloys by combining multiple thermal probes with a measurement based kinetics analysis and a computational thermodynamics evaluation to elucidate the phase selection competition and critical cooling rate conditions. Based upon direct measurements and kinetics modeling it is evident that a critical cooling rate range should be considered to account for nucleation behavior and that the relative heat flow characteristics as well as nucleation kinetics are important in judging ease of glass formation. Also, the synthesis and characterization of alloys with increased cross-section for thermal neutron capture will be outlined to demonstrate that through careful design of alloy composition it is possible to tailor the material properties of the thermally spray-formed amorphous coating to accommodate the challenges experienced in typical nuclear waste storage applications. This work was co-sponsored by the Defense Advanced Projects Agency (DARPA) Defense Science Office (DSO) and the United States Department of Energy (DOE) Office of Science and Technology and International (OSTI). The work was done under the auspices of the U.S. DOE by Lawrence Livermore National Laboratory (LLNL) under Contract Number W-7405-Eng-48, subcontract B529197.
9:00 PM - NN3.12
The Effect of Temperature on Microstructure, Mechanical Properties, and Devitrification of Corrosion Resistant Fe-Cr-Mo-B-Containing Alloys.
Nancy Yang 1 , Gene Lucadamo 1 , Tom Headley 1 , Joseph Farmer 2 , Dan Day 2 Show Abstract
1 Analytical Materials Science, Sandia National Laboratories, Livermore, California, United States, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States
Fe-based structural amorphous metals (SAM) have been studied intensively for wear and corrosion resistance. We are actively engaged in developing a corrosion resistant, amorphous Fe-Cr-Mo-B alloy coating for the safe, long-term storage of spent nuclear fuel and naval applications. Preliminary cyclic polarization tests show that corrosion resistance of Fe-Cr-Mo-B-based SAM alloys is sensitive to devitrification. The results of corrosion testing in seawater or 5M CaCl2 consistently show high corrosion resistance of the SAM alloys with an amorphous structure and poor corrosion resistance with a devitrified, crystalline structure. These findings motivate investigations of the relationship between temperature and the onset of devitrification of the corrosion resistant SAM alloys. Understanding such an evolution of the microstructure enables us to design alloys and processes to yield corrosion resistant amorphous SAM coatings for the targeted applications.The microstructure, mechanical properties, and devitrification of the SAM alloys upon thermal annealing through the glassy transition temperature (Tg) are investigated using several complementary techniques including: X-ray diffraction (XRD), scanning electron microscopy (SEM), focused ion beam imaging (FIB), transmission electron microscopy (TEM), and electron probe microanalysis (EPMA). The starting materials for the investigations are fast cooled melt-spun ribbons (MSR) to ensure a 100% amorphous structure. Due to the small dimensions of the MSR, the mechanical strength is measured using a Vickers micro hardness indentation technique. Current results show that the as-received amorphous ribbons are extremely hard, >1000 MPa, and have a uniform chemical composition, as expected. All of the ribbons retain their amorphous structure and high strength upon annealing to 500°C. At 600°C and higher, all of the ribbons begin to devitrify. The devitrified ribbons usually are nanocomposites containing crystalline phases of Cr-Mo-depleted α-ferrite, metal-carbides, and carborides. Examination of corroded ribbons suggests that the presence of the Cr-Mo-depleted α-ferrite is primarily responsible for lowering the corrosion resistance of the devitrified SAM alloys. In this presentation, we will present the detailed experimental results. The correlation between microstructure and corrosion performance also will be discussed.
9:00 PM - NN3.15
Corrosion Behavior of Carbon Steel in Compacted Bentonite under Gamma Ray Irradiation.
Satoshi Iwasa 1 , Masashi Haginuma 1 , Shoichi Ono 1 , Kazunori Suzuki 1 Show Abstract
1 , Institute of Research and Innovation, Kashiwa-shi, Chiba Japan
9:00 PM - NN3.16
Computational Investigation of the Thermodynamics of Mixing in Rare-Earth Pyrochlore Minerals using First-Principles and Monte-Carlo Methods.
Darius Dixon 1 2 , Udo Becker 1 , Rodney Ewing 1 2 3 Show Abstract
1 Geological Sciences, University of Michigan- Ann Arbor, Ann Arbor, Michigan, United States, 2 Materials Science and Engineering, University of Michigan- Ann Arbor, Ann Arbor, Michigan, United States, 3 Nuclear Engineering and Radiological Sciences, University of Michigan- Ann Arbor, Ann Arbor, Michigan, United States
Extensive irradiation damage studies have been performed on a suite of synthetic pyrochlore minerals, A23+B24+O7 (A= REE and B= Ti, Zr, Sn, Hf), for the properties that lend them as potential hosts to radioactive waste material . One of the interesting aspects of these materials is that the thermodynamic properties of pyrochlore minerals vary considerably with changes in composition. For example, Gd2Sn2O7 has a critical amorphization temperature (Tc) of 350K and Gd2Ti2O7 has a Tc value of ~1120K whereas Gd2Zr2O7 does not amorphize at all but instead undergoes a phase change to a defect-fluorite structure [1,2]. The mechanism behind this transformation is not well understood, mainly due to the fact that the pyrochlore structure itself possesses a considerable degree of flexibility and is generally difficult to synthesize. The structural and electronic properties of several pyrochlore compounds are reported here using first-principles calculations using density functional theory (DFT). Throughout the course of this investigation the commercially available programs, CASTEP, VASP and DMOL3, were used to perform the necessary quantum mechanical calculations to evaluate the atomic, electronic and spin structures and the energetics of different cation configurations. Subsequent Monte-Carlo methods are also performed in order to examine thermodynamic properties of solid solutions between a few of the end-member compositions through interaction parameters and configuration analysis. These theoretical results are compared with those of thermochemical measurements compiled for several stannate pyrochlore compositions . Trends between the experimental and theoretical results are discussed.
 R.C. Ewing, Earth and Plan. Sci. Letters (2005) 165.
 J. Lian, J. Chen, L.M. Wang, R.C. Ewing, J.M. Farmer, L.A. Boatner, K.B. Helean, Phys. Rev. B (2003) 134107.
 J. Lian, K.B. Helean, B.J. Kennedy, L.M. Wang, A. Navrotsky, R.C. Ewing, J. Phys. Chem. B (2006) 2343.
9:00 PM - NN3.17
Synthesis and Characterization of Technetium-Zirconium Alloys Waste Forms for the UREX+1 Process
Ken Czerwinski 1 2 , Frederic Poineau 2 , Thomas Hartmann 2 , Gordon Jarvinen 3 Show Abstract
1 Chemistry, University of Nevada, Las Vegas, Las Vegas, Nevada, United States, 2 , UNLV-Harry Reid Center for Environmental Studies , Las Vegas, Nevada, United States, 3 Nuclear Materials Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
In the US advanced fuel cycle program the UREX+1 process is proposed to separate transuranic elements from spent nuclear fuel. The fission product 99Tc will be extracted into an organic phase containing tributylphosphate together with uranium within the first process steps. Treatment of this stream requires the separation of U from Tc and the placement of the Tc into a suitable waste storage form. A potential candidate waste form is to immobilize the Tc as an alloy with excess metallic zirconium. Alloying metallic Tc with zirconium has potential advantages in terms of the future reuse or treatment of 99Tc. Developing a storage form for Tc waste streams which includes a transmutation option promotes the current US program. However, only a few thermodynamic data in the binary metal system technetium–zirconium exist, and little is available on the synthesis of technetium-zirconium alloys and on their potential performance under temporary or geological storage conditions. This work presents activities on the separation of U from Tc. Waste forms based on the separation techniques are given with examples focusing on the fundamental chemistry of Tc. The routes and conditions for the waste form synthesis are discussed. The waste forms are characterized by X-ray diffraction and microscopy techniques. Results include comparisons of Tc solids with rhenium homologs and evaluation against existing thermodynamic data.
9:00 PM - NN3.18
Characterization of Solids in Residual Wastes from Underground Storage Tanks at the Hanford Site, Washington, U.S.A.
Kenneth Krupka 1 , William Deutsch 1 , H. Schaef 1 , Bruce Arey 1 , Steve Heald 1 , Michael Lindberg 1 , Kirk Cantrell 1 Show Abstract
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Performance assessments will be completed to evaluate the long-term health risks associated with closure of underground storage tanks containing residual radioactive and toxic wastes at the U.S. Department of Energy’s (DOE) Hanford Site in Washington state. The primary contaminants of concern are typically 99Tc, 238U, 129I, and Cr because of their mobility in the environment and long half-lives. PNNL is currently developing source term models that describe the release of contaminants as infiltrating water contacts these residual solids. Given the complexity of waste streams and tank transfers from Hanford operations, contaminant source-term release models can not be developed assuming analogies to the limited information available from Hanford tank simulant studies or prior analyses of sludge and supernatant samples taken from other Hanford tanks. Characterization of the residual waste is thus needed to develop the models that simulate the interactions between the leachants and contaminant-containing solids. Because these wastes are highly radioactive, dispersible powders and are chemically-complex assemblages of crystalline and amorphous solids that contain contaminants as discrete phases and/or sequestered within oxide phases, their detailed characterization offers an extraordinary challenge.X-ray diffraction (XRD), scanning electron microscopy/energy dispersive spectrometry (SEM/EDS), and other techniques, such as synchrotron-based X-ray analysis, are being used to identify phases present in the as-received samples of residual wastes and solids remaining after the water leach and selective extraction studies. Characterization results will be presented and compared for solid phases identified in analyzed samples of residual wastes from Hanford tanks 241-C-106, 241-C-203, 241-C-204, 241-AY-102 and 241-BX-101. Depending on the specific tank, numerous solids [such as ćejkaite; Na2U2O7; clarkeite; gibbsite; boehmite; dawsonite; cancrinite; Fe oxides such as hematite, goethite, and maghemite; rhodochrosite; lindbergite; whewellite; nitratine; and several amorphous phases] have been identified in wastes from different tanks studied to date. XRD and SEM/EDS are the two principal methods used to characterize solid phases and their contaminant associations in these wastes. Because many contaminants of concern are heavy elements, SEM analysis using the backscattered electron (bse) signal has proved invaluable in distinguishing phases containing elements, such as U and Hg, within the complex assemblage of particles that make up each waste. XRD and SEM/EDS also provide different, but complimentary characterization data, which help determine the morphologies, particle sizes, surface coatings, and compositions of phases in the wastes. The impact of these techniques is greatly magnified when each is used in an iterative fashion to cross interpret the results from the other analysis method and then identify additional, more refined analyses to be done.
9:00 PM - NN3.19
European Laboratories for Actinide Research: the ACTINET Pooled Facilities
H. Geckeis 1 , D. Guillaneux 2 , T. Wiss 3 , M. Coeck 4 , A. Scheinost 5 , G. Geipel 5 , A. Scheidegger 6 , Volker Metz 1 , P. Chaix 2 Show Abstract
1 Institute for Radioactive Waste Disposal, Research Center Karlsruhe, FZK-INE, Karlsruhe Germany, 2 , Commissariat a l`Energie Atomique , Paris France, 3 , Institute for Transuranium Elements, Joint Research Centre, Karlsruhe Germany, 4 , Centre D`Etude de l`Energie Nucleaire, SCK-CEN, Mol Belgium, 5 Institut fuer Radiochemie, Forschungszentrum Rossendorf, FZR-IRC, Dresden Germany, 6 , Paul-Scherrer-Institut, Villigen Switzerland
The aim of the European Network on Excellence ACTINET is to promote excellence in Actinide research within the European Community. In order to enhance the potentialities for research on radioactive material, European Actinide laboratories are networked and access for scientists is promoted and supported. Laboratories operated by CEA (France), ITU (Joint Research Centre, EU), INE-FZK (Germany), SCK-CEN (Belgium), IRC-FZR (Germany) and PSI (Switzerland) are coordinated within the ACTINET ‘pooled facilities’. A broad variety of standard analytical and radioanalytical instrumentation, hot cell and glove box equipment thus becomes accessible. Numerous state-of-the-art speciation and characterization techniques are offered for actinide research in the fields of (1) Chemistry and Physics of Actinides in solution and solid phases, (2) Chemistry of Actinides in the geological environment and (3) Chemistry and Physics of Actinide materials under/after irradiation. The ACTINET ‘pooled facilities’ are introduced in the present contribution and possible research activities are outlined.
9:00 PM - NN3.2
Chemical Equilibrium of the Dissolved Uranium in Groundwaters From a Spanish Uranium-ore Deposit.
Antonio Garralon 1 , Paloma Gomez 1 , Maria Jesus Turrero 1 , Belen Buil 1 , Lorenzo Sanchez 1 Show Abstract
1 Medio Ambiente, CIEMAT, Madrid, Madrid, Spain
9:00 PM - NN3.20
Radiolysis and Ageing of C2-BTP in Cinnamaldehyde/hexanol Mixtures.
Anna Fermvik 1 , Christian Ekberg 1 , Teodora Retegan 1 , Gunnar Skarnemark 1 Show Abstract
1 Nuclear Chemistry, Chalmers University of Technology, Gothenburg Sweden
The separation of actinides from lanthanides is an important step of the alternative methods for nuclear waste treatment currently under development. Polycyclic molecules containing nitrogen are synthesised and used for solvent extraction. A potential problem in this process can be the degradation of the molecule due to irradiation or ageing. An addition of nitrobenzene has proved to have an inhibiting effect on degradation when added to an irradiated system containing C2-BTP in hexanol.In this study, C2-BTP was dissolved in different mixtures of cinnamaldehyde and hexanol and the effects on extraction after irradiation was investigated. The solutions were placed in a gamma emitting cobalt source and another set of solutions was placed outside the source to act as reference and at the same time allowing a study of the ageing of the systems. At various times samples were taken and the extraction of Am was investigated and compared to the starting values.Both C2-BTP in cinnamaldehyde and C2-BTP in hexanol seem to be degrading with time. The system with C2-BTP in pure hexanol is relatively stable up until after 17 days but then starts to slowly degrade. The cinnamaldehyde had started to degrade already after ~20 hours. For the degradation due to radiolysis it is the opposite; hexanol systems are more sensitive towards radiolysis than cinnamaldehyde systems. Most of the radiolytic degradation took place during the first days of irradiation, up until a received dose of 4 kGy.
9:00 PM - NN3.21
Application of a Comprehensive Sensitivity Analysis Method on the Safety Assessment of TRU Waste Disposal in JAPAN.
Takao Ohi 1 , Hiroyasu Takase 2 , Manabu Inagaki 2 , Kiyoshi Oyamada 3 , Tomoyuki Sone 1 , Morihiro Mihara 1 , Takeshi Ebashi 1 , Kunihiko Nakajima 4 Show Abstract
1 , JAEA, Tokai-mura, Ibaraki-Ken, Japan, 2 , Quintessa K. K, Yokohama, Kanagawa-ken, Japan, 3 , JGC Corporation, Yokohama, Kanagawa-ken, Japan, 4 , NESI Inc., Tokai-mura, Ibaraki-ken, Japan
In order to gain confidence in safety assessments of the geological disposal of radioactive waste, it is necessary to consider the effects of diverse uncertainties. These uncertainties could be evaluated deterministically by increasing parameter ranges and the number of parameter combinations. However, in such evaluation of uncertainty, the logic of parameter selection and adequate combinations of variants would need to be explained. To this end, the probabilistic approach which can give a risk estimate for the whole system has been developed using probability density functions to model possible parameter variations and parameter combinations in order to complement the deterministic evaluation. However in addition to such probabilistic estimates, there is a need in uncertainty analyses to identify key parameters on the safety and to extract quantitatively conditions that would yield a robust system.In this study, “a comprehensive sensitivity analysis method” was developed to give such quantitative information efficiently. This methodology is composed of the following two components: (1) a stochastic approach used to sample independent parameters randomly and to identify parameters that have a large impact on dose and then to extract threshold values of parameters and/or combinations of parameter thresholds that yield a “successful condition” where maximum dose does not exceed a target value, (2) a nuclide migration model that as far as possible comprehensively incorporates the various phenomena that occur within the repository.This approach was applied in a safety assessment of the geological disposal of TRU waste in Japan. It was found that hydraulic parameters (transmissivity, hydraulic gradient, matrix porosity) significantly affect the maximum dose. From the obtained successful condition, it was confirmed that, if a maximum acceptable target dose of 10μSv/y is assumed, safety is not compromised by the effects of engineered barrier degradation, colloids, gases and initial oxidation. It was also confirmed that the system’s safety would not be compromised in the case that the geological environmental condition is not as favorable as the reference condition which is moderately conservative, owing to improvements in the performance of waste packages.By adopting this approach, it is possible to perform a comprehensive evaluation of the effects of uncertainty on system safety. It is concluded that the method can be used to present the characteristics of system performance in a way that is easy to understand through use of key parameters and successful conditions. This approach depends on the conservativeness of the analytical model for evaluating each phenomenon and variation range of parameters. Hence, further study on phenomena and handling of parameter relationships is necessary to enhance this approach. Moreover, adopting this approach to specific geological conditions and design concepts is also future issue.
9:00 PM - NN3.3
Assessment of the Radionuclide Release from the Near-Field Environment of a Spent Nuclear Fuel Geological Repository
Arnaud Poulesquen 1 , Jean Radwan 1 , Christophe Poinssot 1 , Cécile Ferry 1 Show Abstract
1 Department of Physics and Chemistry, CEA Saclay, Gif-sur-Yvette France
9:00 PM - NN3.4
Synthesis and Characterization of Quadruple-Bonded Technetium Dimers.
Frederic Poineau 1 , Alfred Sattelberger 2 , Ken Czerwinski 1 , Steven Conradson 3 Show Abstract
1 Harry Reid Center for Environmental Studies, University of nevada Las Vegas, Las Vegas, Nevada, United States, 2 Physical Sciences Directorate, Argonne National Laboratory, Argonne, Illinois, United States, 3 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
9:00 PM - NN3.5
Uranium Sequestration by Aluminum Phosphate Minerals in Unsaturated Soils.
James Jerden 1 Show Abstract
1 , Argonne National Laboratory, Argonne , Illinois, United States
A mineralogical and geochemical study of soils developed from the unmined Coles Hill uranium deposit (Virginia) was undertaken to determine how phosphorous influences the speciation of uranium in an oxidizing soil/saprolite system typical of the eastern US. This paper presents mineralogical and geochemical results that identify and quantify the processes by which uranium has been sequestered in these soils. It was found that uranium is not leached from the saturated soil zone (saprolites) overlying the deposit due to the formation of a sparingly soluble uranyl phosphate mineral of the meta-autunite group. The concentration of uranium in the saprolites is approximately 1000 mg uranium per kg of saprolite. It was also found that a significant amount of uranium was retained in the unsaturated soil zone overlying uranium-rich saprolites. The uranium concentration in the unsaturated soils is approximately 200 mg uranium per kg of soil (20 times higher than uranium concentrations in similar soils adjacent to the deposit). Mineralogical evidence indicates that uranium in this zone is sequestered by a barium-strontium-calcium aluminum phosphate mineral of the crandallite group (gorceixite). This mineral is intimately inter-grown with iron and manganese oxides that also contain uranium. The amount of uranium associated with both the aluminum phosphates (as much as 1.4 weight percent) and the iron/manganese oxides has been measured by electron microprobe microanalyses and the geochemical conditions under which these minerals formed has been studied using thermodynamic reaction path modeling. The geochemical data and modeling results suggests the meta-autunite group minerals present in the groundwater saturated saprolites are unstable in the overlying unsaturated soils due to a decrease in soil pH (down to a pH of 4.5) at depths less than 3 meters below the surface. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)-bearing aluminum phosphates.
9:00 PM - NN3.6
Monte Carlo Simulations of the Degradation of the Engineered Barriers System in the Yucca Mountain Repository Using the EBSPA Code
Zack Qin 1 , David Shoesmith 1 Show Abstract
1 , The University of Western Ontario, London, Ontario, Canada
Based on the probabilistic model proposed early [1, 2], a Monte Carlo simulation code (EBSPA) has been developed to predict the lifetime of the engineered barriers system within the Yucca Mountain nuclear waste repository. Corrosion commences once aqueous conditions are established on the surfaces of the Ti Grade 7 drip shield and the Alloy 22 waste package. The degradation modes considered in the EBSPA are passive corrosion and hydrogen-induced cracking for the drip shield, and passive and crevice corrosion and stress corrosion cracking for the waste package. The EBSPA can calculate the cumulative probability of failures (CPF) of the drip shield, the waste package, and the overall engineered barriers system, as well as the ratios of failures due to different degradation modes to failure. Based on the anticipated repository environments and available corrosion data on relevant Ti alloys and C-series of Ni-Cr-Mo alloys, two scenarios have been simulated using the EBSPA code: (a) a realistic scenario for the conditions thought most likely to prevail in the repository, and (b) a worst-case scenario in which the impact of the degradation processes is overstated. The results show that the engineered barriers system should not fail until 255,000 years (CPF = 0.0001) under realistic conditions, and that even for the worst-case scenario, the engineered barriers system would likely survive for more than 115,500 years (CPF = 0.0001). The simulations also predict that the drip shield will always fail due to hydrogen-induced cracking, and crevice corrosion of the waste package is unlikely under realistic conditions.REFERENCES 1.Z. Qin and D. W. Shoesmith, Scientific Basis for Nuclear Waste Management XXVIII, San Franciso, CA, April 13-16, 2004, pp 11-18.2.Z. Qin, L. Yan and D. W. Shoesmith, Probabilistic Safety Analysis, San Francisco, CA, Sept. 11-15, 2005, pp 399-406.
9:00 PM - NN3.9
Study of Susceptibility of Copper OFP to Corrosion in Presence of Cchloride Ion and Simulated Underground Water.
Ivan Escobar 1 , Claudia Lamas 1 , Lars Werme 2 , Virginia Oversby 3 Show Abstract
1 Nuclear Materials, Chilean Commission for Nuclear Energy, Santiago Chile, 2 , Svensk Kärnbränslehantering AB (SKB), Stockholm Sweden, 3 , VMO Konsult, Stockholm Sweden
9:00 PM -
NN3.8 Transferred to NN2.3
Darrell S. Dunn Southwest Research Institute
Christophe Poinssot CEA Saclay
Bruce Begg Australian Nuclear Science & Technology Organisation (ANSTO)
NN4: Waste Forms for Plutonium and Ceramics for HLW
Tuesday AM, November 28, 2006
Constitution A (Sheraton)
9:30 AM - **NN4.1
Development of Ceramic Waste Forms for High-Level Nuclear Waste over the Last 30 years.
Eric Vance 1 Show Abstract
1 , ANSTO, Lucas Heights, New South Wales, Australia
10:00 AM - NN4.2
An Evaluation of Single Phase Ceramic Formulations for Plutonium Disposition.
Martin Stennett 1 , Neil Hyatt 1 , Ewan Maddrell 2 , Charlie Scales 2 , Francis Livens 3 , Matthew Gilbert 3 Show Abstract
1 Engineering Materials, The University of Sheffield, Sheffield, South Yorkshire, United Kingdom, 2 B170, Nexia Solutions Ltd., Seascale, Cumbria, United Kingdom, 3 Chemistry, The University of Manchester, Manchester, Lancashire, United Kingdom
10:15 AM - NN4.3
Survey of Potential Glass Compositions for the Immobilisation of the UK’s Separated Plutonium Stocks.
Mike Harrison 1 , Charlie Scales 1 , Paul Bingham 2 , Russell Hand 2 Show Abstract
1 Nexia Solutions Ltd., Sellafield, Seascale, Cumbria, United Kingdom, 2 Immobilisation Science Laboratory, Sheffield University, Sheffield United Kingdom
10:30 AM - NN4.4
Development of a Phosphate Ceramic as a Host for Halide-contaminated Plutonium Pyrochemical Reprocessing Wastes.
Brian Metcalfe 1 , Shirley Fong 1 , Lee Gerrard 1 , Ian Donald 1 , Denis Strachan 2 , Randall Scheele 2 Show Abstract
1 MSRD, AWE, Reading, Berkshire, United Kingdom, 2 , PNNL, Richland, Washington, United States
10:45 AM - NN4.5
The Benefits of Tailored Glass-Ceramic Waste Forms to Lock up Problematic HLW.
Bruce Begg 1 , Arthur Day 1 , Martin Stewart 1 , Sam Moricca 1 Show Abstract
1 , ANSTO, Menai, New South Wales, Australia
To deliver maximum benefits and optimum performance to the HLW cleanup program, the selection of the waste form should be driven by the characteristics of nuclear waste to be immobilized rather than adopting a single baseline approach. The use of tailored, high-performance waste forms that include ceramics and glass-ceramics, coupled with alternative mature process technologies, such as hot-isostatic pressing (HIPing) and sintering, offer significant performance improvements and efficiency savings for the immobilisation of wastes difficult to incorporate in glass. Advantages include:*higher waste loadings (fewer disposal canisters); *enhanced chemical durability (lower environmental risk); *lower off-gas emissions;*greater processing flexibility.The benefits tailored glass-ceramic waste forms bring to the HLW cleanup program will be briefly reviewed with reference to the immobilisation of HLW calcines at Idaho and also impure plutonium residues in the UK. The INL HLW calcines contain a significant quantity of waste components that are problematic to incorporation in glass. Results will be presented describing a tailored glass-ceramic that can achieve waste loadings in excess of 80% for these calcines whilst maintaining chemical durability at least 10 times EA glass. Consolidation via HIPing results in a dense monolithic waste form, with no high-temperature off-gas emissions, and an overall volume reduction of 30% compared to the as-stored calcines. Previous processing of plutonium containing materials on the Sellafield site has resulted in a range of wastes and residues, from which it is uneconomic to recover the plutonium. The identified wastes are chemically diverse, containing significant quantity of process impurities making them extremely problematic to incorporate in either glass or a straight ceramic. Results will be presented for a tailored zirconolite glass-ceramic waste form, prepared by HIPing, in which the plutonium is overwhelmingly preferentially partitioned into the zirconolite over the glass by factors over 100:1, whilst the process impurities are incorporated in a durable vitreous phase. This approach combines the proliferation resistance of ceramics with the flexibility of glass to immobilise this otherwise problematic waste stream.
11:45 AM - NN4.7
Effect of Stainless Steel Can/Glass-ceramic Interaction Layer on Aqueous Durability.
Peter McGlinn 1 , Yingjie Zhang 1 , Huijun Li 1 , Tim Payne 1 Show Abstract
1 Institute of Materials & Engineering Science, Australian Nuclear Science & Technology Organisation, Lucas Heights, New South Wales, Australia
12:00 PM - NN4.8
Hydrothermal Methods as a New Way of Actinide Phosphate Preparation.
Nicolas Dacheux 1 , Nicolas Clavier 1 , Gilles Wallez 2 , Michel Quarton 2 Show Abstract
1 Radiochemistry Group, Nuclear Physics Institute, Orsay France, 2 , University Pierre et Marie Curie - Paris 6, Paris France
12:15 PM - NN4.9
Na Release from Na5Zr(PO4)3 –rich Ceramics.
Melody Carter 1 , E. Vance 1 , John Hanna 1 Show Abstract
1 , ANSTO, Lucas Heights, New South Wales, Australia
While NaZr2(PO4)3 (NZP)- structured waste forms have been put forward for many years as immobilising agents for High-Level Waste (HLW) from nuclear power plant fuel or weapons production, little attention has been paid to Na release from NZP (but see ). For example, highly active Na-rich wastes can arise from liquid sodium used in primary cooling of fast reactors or radioisotope production and Na immobilisation and speciation in HIPed ceramics are the focus of the current work. For Na immobilisation, it would be clearly advantageous from the point of view of waste loading to utilise the most Na-rich version of the NZP structure, namely Na5Zr(PO4)3. HIPing at ~ 900oC is used in the production to eliminate volatility in the hot consolidation step, and studies of possible can/waste form interactions will be presented. Initial X-ray and scanning electron microscope studies have shown that material containing ~ 50 vol % of Na5Zr(PO4)3 has leach rates of <1 g/m2/day in PCT tests, as found earlier for other NZP-rich preparations  and the preparation of materials much richer in Na5Zr(PO4)3 is under way. The characterisation of these materials by XRD, SEM , leaching, and multinuclear solid-state nuclear magnetic resonance will be reported. 1. V. N. Zyryanov and E. R. Vance (1997), “Comparison of Sodium Zirconium Phosphate-Structured HLW forms and Synroc for High-Level Nuclear Waste Immobilization”, in Scientific Basis for Nuclear Waste Management XX, Ed. W. J. Gray and I.R. Triay, Materials Research Society, PA, pp. 409-16.
12:30 PM - NN4.10
Rare Earth-Doped Murataite Ceramics.
Sergey Stefanovsky 1 , Sergey Yudintsev 1 , Boris Nikonov 1 , Olga Stefanovsky 1 , Natalia Mikhailenko 1 Show Abstract
1 , SIA Radon, Moscow Russian Federation
Phase composition of the murataite-based ceramics containing 10 wt.% lanthanum, cerium, neodymium, europium, gadolinium, yttrium, zirconium oxides was studied. The ceramics were prepared by melting of oxide mixtures in Pt ampoules in air at ~1500 0C. They are composed of predominant murataite-type phases and minor extra phases: rutile, crichtonite, perovskite, ilmenite/ pyrophanite, and zirconolite (in the Zr-bearing sample only). Three murataite-related phases with five- (5C), eight- (8C), and three-fold (3C) elementary fluorite unit cell are normally present in all the ceramics. These phases form core, intermediate zone, and rim of the murataite grains, respectively. They are predominant host phases for the rare earth elements whose concentrations are reduced in a row: M-5C>M-8C>M-3C. Appreciate fraction of La and Ce may enter the perovskite phase.
12:45 PM - NN4.11
Synroc-D Type ceramics Produced by Hot Isostatic Pressing and Cold Crucible Melting for Immobilisation of (Al, U) Rich Nuclear Waste.
Eric Vance 1 , M. La Robina 1 , H. Li 1 , J. Davis 1 Show Abstract
1 , ANSTO, Lucas Heights, New South Wales, Australia
NN5: Spent Fuel
Tuesday PM, November 28, 2006
Constitution A (Sheraton)
2:30 PM - NN5.1
Effect of HElium Accumulation on the Spent Fuel Microstructure.
Ferry Cecile 1 , Piron Jean-Paul 2 , Stout Ray 3 Show Abstract
1 Department of physico-chemistry, CEA Saclay, Gif-sur-Yvette France, 2 Department of Fuel Studies, CEA, Cadarache France, 3 , Rho Beta Sigma Affairs, Livermore, California, United States
2:45 PM - NN5.2
Thermal Diffusion of Helium in UO2.
Guillaume Martin 1 , Pierre Desgardin 1 , Thierry Sauvage 1 , Gaelle Carlot 2 , Philippe Garcia 2 , Hicham Khodja 3 , Marie-France Barthe 1 Show Abstract
1 , CNRS / CERI, Orleans France, 2 , CEA Cadarache / LLCC, St Paul Lez Durance France, 3 , CEA Saclay / LPS, Gif sur Yvette France
3:00 PM - NN5.3
Chlorine Diffusion in UO2 : Thermal Effects versus Radiation Enhanced Effects.
Yves Pipon 1 , Nelly Toulhoat 1 , Nathalie Moncoffre 1 , Nicolas Bererd 1 , Henri Jaffrezic 1 , Louis Raimbault 2 , Andre Scheidegger 3 , Farncois Farges 4 Show Abstract
1 CNRS/IN2P3 Universite Claude Bernard Lyon 1, Institut de Physique Nucleaire de Lyon, Villeurbanne France, 2 Centre d'Informatique Geologique, Ecole des Mines de Paris, Fontainebleau France, 3 Nuclear Energy and Safety Department Laboratory for Waste Management, Paul Scherrer Institut, Villigen Switzerland, 4 Laboratoire des Geomateriaux, Université de Marne la Vallée, Champs sur Marne France
3:15 PM - NN5.4
Modelling of Spent Fuel Oxidation at Low Temperature
Arnaud Poulesquen 1 , Lionel Desgranges 2 , Cécile Ferry 1 Show Abstract
1 Department of Physics and Chemistry, CEA Saclay, Gif-sur-Yvette France, 2 Department of Fuel studies, CEA Cadarache, Saint Paul Lez Durance France
During dry storage, the oxidation of the spent fuel in case of cladding and container failure (accidental scenario) could be detrimental for further handling of the spent fuel rod and for the safety of the facilities. In this accidental context, the oxidation kinetics of spent fuel will be approached in this paper. Depending on whether the uranium dioxide is in the form of powder or pellet, irradiated (spent) fuel or unirradiated fuel, the weight gain curves do not present the same shape. For unirradiated powders, a parabolic shape is noticed at the beginning of oxidation whereas a sigmoidal shape is observed for sintered unirradiated pellet with an incubation period depending on the oxidation temperature. Recently, oxidation experiments in air at 200°C have been performed on high burn-up UOX and MOX and shown a sigmoidal shape instead of parabolic one as reported in literature.To account for these different behaviours, two models have been developed. Firstly, the oxidation of unirradiated powders has been modelled based on the coexistence, during the oxidation, of two intermediate products, U4O9 and U3O7. It was shown that the comparison between the calculation and the literature data is good in terms of weight gain curves and chemical diffusion coefficient of oxygen in the two phases. Secondly, the oxidation of spent fuel fragments may be approached by a convolution procedure between a grain oxidation model and an empirical parameter which may represent the linear oxidation speed of grain boundary or an average distance able to cover the entire spent fuel fragment. This procedure of calculation allows in one hand to account for the incubation period noticed on unirradiated pellets or spent fuel and in another hand to link the empirical parameter to physical or operational parameters as porosity, cracks or linear power, gas release fraction respectively. A comparison of this new modelling with experimental data will be proposed.
3:30 PM - NN5.5
Computational Investigation of the Formation of Hyperstoichiometric Uranium Dioxide (UO2+x).
Frances Skomurski 1 , Rodney Ewing 1 2 3 , Udo Becker 1 Show Abstract
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 2 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States, 3 Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States
NN6: Glass Characterization and Leaching
Tuesday PM, November 28, 2006
Constitution A (Sheraton)
4:30 PM - NN6.1
EELS Spectrum Imaging and Tomography Studies of Simulated Nuclear Waste Glasses.
Guang Yang 1 , Günter Möbus 1 , Russell Hand 1 Show Abstract
1 Engineering Materials, University of Sheffield, Sheffield, South Yorkshire, United Kingdom
Electron energy loss spectroscopy (EELS) has been widely used in analysis of ceramic, minerals and semiconductors. It can provide the unique advantage of sensitivity to composition, coordination and valency, while providing nano-scale spatial resolution. Due to the electron irradiation sensitivity, EELS studies on glass are quite rare. However, using special care in glass-composition selection and by recording time series of EELS-spectra to monitor any structural changes, EELS then becomes a very useful technique to study glass-chemistry and glass structure with the highest spatial resolution of all chemically sensitive techniques. Alkali borosilicate glasses (ABS) doped with Cr2O3 (2 mol%), CeO2 (4 mol%) and ZrO4 (4 mol%) were melted, cooled and annealed in the context of simulating radionuclide immobilisation glasses. Precipitates with diameter in the range of ~20 nm to ~200 nm were found homogeneously distributed in the glass. In preliminary studies we measured the oxidation state in the glass and in precipitates via evaluation of the Ce-M-edge double white line ratio, confirming the crystals as Ce(IV) oxide. We also found Boron K-edge ELNES spectra as a sensitive signal to boron coordination (N4 = BO4/(BO3+BO4)), which is found at N4 ratios of around 35-45%.In the present study we have extended this work to the acquisition of dense line scans (spectrum imaging) across an area of interest on our composite glass with emphasis on three questions:(i)Change of cerium valence upon transition from the glass to the precipitates?(ii)Possible oscillation of boron N4-ratio in the glass within a line scan as a consequence of random medium-order related glass fluctations when scanning the <5nm probe across a seemingly homogeneous area?(iii)N4-changes in the immediate glass layer surrounding the precipitates?Our ongoing research has produced results which prove a systematic change of the Ce oxidation state from precipitates (+IV) to glass (mixed +III/+IV). It is found that boron coordination is constant within the sensitivity given by signal-to-noise in the glass matrix. A change of the Boron K-edge fine structure upon crossing a nanoparticle during one particular linescan has been observed, which could hint to a coordination change of the glass layer surrounding the particle. Other linescan measurements have not shown any change and acquisition of an enlarged better statistic is work in progress. The ABS-glass with distributed nanoparticles provides an ideal example of a 3D nanocomposite. The only proven technique for the three-dimensional reconstruction of such materials on the nanoscale is electron tomography. We present the first results of a 3D reconstructed nuclear waste glass by using a tilt series of ADF STEM images covering a glass fragment of 2000nm field of view containing several tens of nanoparticles distributed around its volume.
4:45 PM - NN6.2
Experimental Study and Monte Carlo Modeling of Calcium Borosilicate Glasses Leaching.
Mehdi Arab 1 , Celine Cailleteau 1 , Frederic Angeli 1 , François Devreux 2 Show Abstract
1 CEA/DEN/DTCD/SECM/Laboratoire d'étude du comportement à long terme des matériaux de conditionnement, CEA, Bagnols sur Ceze France, 2 Laboratoire de Physique de la Matière Condensée, CNRS & Ecole Polytechnique , Palaiseau France
5:00 PM - NN6.3
Tc and Re Behavior in Borosilicate Waste Glass Vapor Hydration Tests.
David McKeown 1 , Andrew Buechele 1 , Wayne Lukens 2 , David Shuh 2 , Ian Pegg 1 Show Abstract
1 Vitreous State Laboratory, Catholic University , Washington, District of Columbia, United States, 2 Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California, United States
Technetium (Tc), found in some radioactive wastes, is of particular concern with regard to long-term waste storage because of its long-lived radioactivity and high mobility in the environment. One method of stabilization of such waste is through vitrification to produce a durable borosilicate glass matrix. The fate of Tc under hydrothermal conditions in the Vapor Hydration Test (VHT) was studied to assess and possibly predict the long-term rate of release of Tc from borosilicate waste glass. For comparison, the fate of rhenium (Re), the preferred non-radioactive surrogate for Tc, was similarly studied. X-ray absorption spectroscopy (XAS) and scanning electron microscopy (SEM) measurements were made on each original borosilicate glass and the corresponding sample after exposure to the VHT. Tc K-edge XAS data and analyses indicate that, despite starting with different Tc(IV) and Tc(VII) distributions for each glass, both corresponding VHT samples have 100% Tc(IV). The Tc reduction within the VHT samples may be driven by simultaneous oxygen depletion within the hydrothermal environment from corrosion of the surrounding stainless steel vessel. From SEM analyses, both of the Tc-containing VHT samples show complete alteration of the original glass, significant Tc enrichment near the sample surface, and nearly complete depletion of Tc toward the sample center. XAS indicates Tc(IV)O6 environments, possibly within gel-like amorphous silicate phases in both VHT samples, where Tc-Tc correlations are observed only in the higher Tc-content VHT sample. Re LII-edge XAS and SEM measurements indicate quite different behavior for Re under VHT conditions. Re speciation is relatively insensitive to the VHT treatment, where perrhenate (Re(VII)) species are dominant in all Re-containing samples investigated. For both Re containing VHT samples, Re2O7 concentrations near the sample surface are low and increase to approach the concentration of the original glass toward the sample center and remaining un-reacted glass.
5:15 PM - NN6.4
Dissimilar Behavior of Technetium and its Non-radioactive Surrogate Rhenium in Waste Glasses.
Wayne Lukens 1 , David McKeown 2 , Andrew Buechele 2 , Ian Pegg 2 , David Shuh 1 Show Abstract
1 Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California, United States, 2 Vitreous State Laboratory, The Catholic University of America, Washington, District of Columbia, United States
NN7: Historical Perspectives and Future Trends
Tuesday PM, November 28, 2006
Constitution A (Sheraton)
8:00 PM - **NN7.1
What Has Changed in Thirty Years?
Rodney Ewing 1 Show Abstract
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
On the occasion of the 30th meeting of the MRS symposium, Scientific Basis for Nuclear Waste Management, I review briefly the changes that have occurred since the first meeting in the Fall of 1978. I believe that there are three broad currents of change in the United States that have had a dramatic effect on our scientific efforts in nuclear waste management: i.) The grand covenant of the Nuclear Waste Policy Act of 1982 has now been essentially dismantled. ii.) The original concept of multiple barriers to radionuclide release has given way to an attempted quantification of the release of radionuclides and resulting risk using a total system performance assessment (TSPA) methodology. iii.) There is a renewed, some say a renaissance, of interest in nuclear power, as evidenced by a series of initiatives: Generation IV reactors, an Advanced Fuel Cycle Initiative, and a Global Nuclear Energy Partnership. This renaissance of interest in all things nuclear should create broad opportunities for research in nuclear materials as evidenced by the recent workshop on “Basic Research Needs for Advanced Nuclear Energy Systems” sponsored by the Office of Science of the U.S. Department of Energy. What remains unchanged, however, is that there is still not an operating geologic repository for the disposal of high-level nuclear waste and spent nuclear fuel. In the United States, a combination of technical, regulatory and policy issues has complicated the future prospects for the proposed nuclear waste repository at Yucca Mountain. The slow progress in the United States may provide a cautionary tale for colleagues in other national programs.
8:30 PM - **NN7.2
Nuclear Waste Disposal in Deep Geological Formations: What are the Major Remaining Scientific Issues?
Pierre Toulhoat 1 Show Abstract
1 Institut des Sciences Analytiques, Université Lyon1, Villeurbanne France
9:00 PM - **NN7.3
A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles.
Mark Peters 1 , Rodney Ewing 2 Show Abstract
1 Applied Science and Technology, Argonne National Laboratory, Argonne, Illinois, United States, 2 Department of Geological Sciences, The University of Michigan, Ann Arbor, Michigan, United States
There are two compelling reasons for the importance of understanding the source term and near-field processes in a geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are no longer important, it is the waste form that controls the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years.The United States Department of Energy’s Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U6+- secondary phases; c) waste form–waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of the source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 100,000 years).Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms “tailored” to specific geologic settings. The development of these “tailored” waste forms is also contingent on an integrated analysis of the closed nuclear fuel cycle. Such a systems analysis would allow for optimization of the entire separations/storage/disposal process. The key elements are to develop the optimal separations process to minimize separations process losses, while producing the required product form for advanced nuclear fuel fabrication and storage form for short-lived fission products, and finally, waste forms tailored to specific geologic settings.
9:30 PM - **NN7.4
Expansion of Nuclear Energy and the Impact on Nuclear Waste Management Issues.
Allison Macfarlane 1 Show Abstract
1 Science, Technology & Society, MIT, Cambridge, Massachusetts, United States
Over the next century nuclear power may play a bigger role in generating electricity than it does now, due to pressures exerted by global climate change, increasing electricity needs, and unstable fossil fuel markets. By 2030, the Energy Information Administration is expecting a 20% increase in installed capacity. Some climate-based scenarios suggest a four-fold expansion might be necessary. More nuclear power will mean more nuclear waste. Since no country has yet to open a repository for high-level nuclear waste, the prospect of a large increase in the volume of highly radioactive waste is not comforting. Over the next 20 years, nuclear power expansion will likely be based on existing plant designs, resulting in about 2000 metric tons of spent fuel for every 100 GW of electricity produced on an annual basis. Beyond the next 20 years, a number of scenarios are possible that will result in different impacts on nuclear waste management.One of the most recent suggestions for a new nuclear future is that proposed by President Bush in early 2006, the Global Nuclear Energy Partnership (GNEP). This ambitious plan envisions the reprocessing of existing spent fuel stocks using the UREX+ process, the recycling of plutonium and some actinides as mixed oxide fuel in existing light water reactors, the construction of fast-neutron burner reactors to transmute the remaining actinides, and the eventual reprocessing of fast reactor fuel via pyroprocessing. The goal of the GNEP proposal is a substantial reduction in volume of high-level nuclear waste destined for a Yucca Mountain geologic repository. The GNEP proposal will produce a variety of wastes, not all high-level. In fact, the majority of wastes produced will be low and intermediate level waste from the reprocessing process. Moreover, if successful, the GNEP process will produce high-level waste streams that will require new waste forms for geologic disposal at Yucca Mountain, such as technetium-99. If the GNEP program is not successful, for example, in completing the fast reactor/pyroprocessing step, there will be a whole new set of waste streams to deal with, including wastes from the UREX+ process such as fission products, actinides, technetium, as well as MOX spent fuel. All of these waste streams should be considered in careful planning for future nuclear energy and nuclear waste policy.Finally, a careful examination of the capacity of Yucca Mountain is necessary to determine the ability of one repository site to provide needed storage. Most recent studies simply evaluate the capacity on the basis of thermal calculations and Department of Energy design specifications for a repository. A robust analysis of Yucca Mountain capacity will take into account thermal factors as well as limits imposed by geological and hydrological features and processes. Only a complete examination of all these issues will produce the basis for realistic policy.
Darrell S. Dunn Southwest Research Institute
Christophe Poinssot CEA Saclay
Bruce Begg Australian Nuclear Science & Technology Organisation (ANSTO)
NN8: Containers and Engineered Barriers
Wednesday AM, November 29, 2006
Constitution A (Sheraton)
9:30 AM - **NN8.1
Processes Contributing to the Stifling and Arrest of Localized Corrosion.
Joe Payer 1 Show Abstract
1 Materials Science and Engineering, Case Western Reserve University, Cleveland, Ohio, United States
Corrosion is a primary determinant of waste package performance at the proposed Yucca Mountain Repository and will control the delay time for radionuclide transport from the waste packages. A special feature of the proposed Yucca Mountain Repository is the extremely long time frame of interest, i.e. 10,000’s of years and longer. Thus, the time evolution of corrosion damage is of primary interest in the determination of expected performance. A framework for the analysis of localized corrosion at the proposed Yucca Mountain Repository has been presented earlier. In this paper, the processes that can contribute to the stifling and arrest of localized corrosion are described. Stifling refers to phenomena that result in decreasing penetration rates. The waste packages are emplaced on support pallets and held in air. While they will never be fully immersed in waters, moisture could form on the metal surfaces from condensation, deliquescence, drips and seepage on waste packages. An outcome is that the important time periods when localized corrosion could occur on waste packages are restricted to finite time periods. For these time periods where localized corrosion can be supported based upon the temperature and possible water chemistries, the evolution of corrosion damage is controlled by the initiation, propagation, stifling and arrest phenomena. It is useful to group stifling and arrest processes into categories. These include geometric, ohmic, anodic and cathodic processes. In order for localized corrosion to persist, it is necessary to develop and maintain a critical, aggressive environment. Several of the stifling processes involve the elimination of dissipation of this critical environment. The chemical, electrochemical and transport phenomena are analyzed as they pertain in layers of particulate and deposit on metal surfaces. Much of this is based upon work underway within a multi-university DOE Cooperative for Corrosion and Materials Performance. Support by the Office of Science and Technology and International of the U.S. Department of Energy, Office of Civilian Radioactive Waste Management under DOE Cooperative Agreement Number: DE-FC28-04RW12252 is gratefully acknowledged.
10:00 AM - NN8.2
Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls.
Jor-Shan Choi 1 Show Abstract
1 Energy and Environment, Lawrence Livermore National Laboratory, Livermore, California, United States
J-S. Choi1, J. Farmer1, C. Lee1, S. Day1, M. Boussoufi2, H. Egbert2, H. Liu2, R. G. Flocchini21Lawrence Livermore National Laboratory, 2McClellan Nuclear Radiation Center, UC Davis, D. J. Branagan3, A. D’Amato4, 3The NanoSteel Company, 4Plasma Technology, Inc. Abstract Spent nuclear fuel contains fissionable materials (235U, 239Pu, 241Pu, etc.). Neutron multiplication and the potential for criticality are enhanced by the presence of a moderator during cask loading in water, water incursion in accidents conditions during spent fuel storage or transport. To prevent nuclear criticality in spent fuel storage, transportation, and during disposal, neutron-absorbing materials would have to be applied.The success in demonstrating that the High-Performance Corrosion-Resistant material (HPCRM) can be thermally sprayed as coating onto base metal to provide for corrosion resistance raises the interest in applying the HPCRM to USDOE/OCRWM spent fuel management program. The fact that the HPCRM relies on the presence of boron to make the material amorphous – an essential property for corrosion resistance – and that the boron has to be homogenously distributed in the HPCRM make the material a good candidate to be a neutron absorber.This paper describes the effectiveness of the boron-containing, iron-based amorphous metals as neutron poison for support basket material inside the spent fuel disposal containers. This includes the manufacturing of the amorphous-metal powders (at The Nanosteel Company), the thermal-spraying of the powder onto metallic substrates (SS316L) which are made into a support-basket structure (at Plasma Technology, Inc.), and the examination of the coated substrates, including the radiation measurement of its neutron absorbing effectiveness in a research reactor (at McClellan Nuclear Radiation Center). The use of these advanced boron-containing, iron-based, corrosion-resistant materials to prevent nuclear criticality in long-term spent fuel storage and disposal is very beneficial to DOE’s nuclear waste management programs.The work described here was co-sponsored by the Defense Advanced Projects Agency (DARPA) Defense Science Office (DSO) and the United States Department of Energy (DOE) Office of Science and Technology and International (OSTI). This work was done under the auspices of the U.S. DOE by Lawrence Livermore National Laboratory (LLNL) under Contract Number W-7405-Eng-48. Contact InformationJ.-S. Choi, Lawrence Livermore National Laboratory, firstname.lastname@example.orgM. Boussoufi, email@example.comD. J. Branagan, NanoSteel Company, DBranagan@nanosteelco.comA. D’Amato, Plasma Technology, Inc., firstname.lastname@example.orgNoteAuthorship reflects multi-institutional collaboration. Additional authors may be added in future revisions of the abstract and paper.
10:15 AM - NN8.3
High-Performance Corrosion-Resistant Iron-Based Amorphous Metals: The Effects of Composition, Structure and Environment on Corrosion Resistance.
Joseph Farmer 1 2 3 , Jor-Shan Choi 1 2 , Jeffrey Haslam 1 2 4 , Tiangan Lian 1 2 3 , Sumner Day 1 2 3 , Nancy Yang 5 , Craig Blue 6 , William Peter 6 , Robert Bayles 7 , Joe Payer 11 , John Lewandowski 11 , John Perepezko 10 , Kjetil Hildal 10 , Enrique Lavernia 8 , Leonardo Ajdelsztajn 8 , Olivia Graeve 9 , Daniel Branagan 12 , M. Beardsley 13 , Louis Aprigliano 14 , Lawrence Kaufman 15 , Jay Boudreau 16 Show Abstract
1 Nuclear Science & Systems Engineering Program, Lawrence Livermore National Laboratory, Livermore , California, United States, 2 Energy & Environment Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States, 3 Chemistry & Materials Science Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States, 4 Engineereing Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States, 5 Materials Science, Sandia National Laboratory, Livermore, California, United States, 6 Materials Science, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 7 Materials Science, Naval Research Laboratory, Washington, District of Columbia, United States, 11 Materials Science, Case Western Reserve University, Cleveland, Ohio, United States, 10 Materials Science, University of Wisconsin Madison, Madison, Wisconsin, United States, 8 Materials Science, University of California Davis, Davis, California, United States, 9 Materials Science, University of Nevada Reno, Reno, Nevada, United States, 12 Institute of Nanomaterials Research & Development, The NanoSteel Company, Idaho Falls, Idaho, United States, 13 Advanced Materials Technology, Caterpillar Incorporated, Peoria, Illinois, United States, 14 Consulting, SAINC, Arlington, Virginia, United States, 15 Consulting, CALPHAD, Brookline, Massachusetts, United States, 16 Consulting, BLE, Los Alamos, New Mexico, United States
New corrosion-resistant, iron-based amorphous metals have been identified from published data or developed through combinatorial synthesis, and tested to determine their relative thermal phase stability, microstructure, mechanical properties, damage tolerance, and corrosion resistance. Some alloy additions are known to promote glass formation and to lower the critical cooling rate. Other elements are known to enhance the corrosion resistance of conventional stainless steels and nickel-based alloys and have been found to provide similar benefits to iron-based amorphous metals.Many of these materials can be cast as relatively thick ingots, or applied as coatings with advanced thermal spray technology. A wide variety of thermal spray processes have been developed by industry, and can be used to apply these new materials as coatings. Any of these can be used for the deposition of the formulations discussed here, with varying degrees of residual porosity and crystalline structure. Thick protective coatings have now been made that are fully dense and completely amorphous in the as-sprayed condition. An overview of the High-Performance Corrosion Resistant Materials (HPCRM) Project will be given, with particular emphasis on the corrosion resistance of several different types of iron-based amorphous metals in various environments of interest. The salt fog test has been used to compare the performance of various wrought alloys, melt-spun ribbons, arc-melted drop-cast ingots, and thermal-spray coatings for their susceptibility to corrosion in marine environments. Electrochemical tests have also been performed in seawater. Spontaneous breakdown of the passive film and localized corrosion require that the open-circuit corrosion potential exceed the critical potential. The resistance to localized corrosion is seawater has been quantified through measurement of the open-circuit corrosion potential, the breakdown potential and the repassivation potential. The greater the difference between the open-circuit corrosion potential and the repassivation potential, the more resistant a material is to modes of localized corrosion such as pitting and crevice corrosion. Cyclic polarization was used as a means of measuring the critical potential relative to the open-circuit corrosion potential. Linear polarization has been used to determine the corrosion current and the corresponding corrosion rate. Other aspects of the materials will also be discussed, as well as potential applications, which include use as protective coatings for spent fuel containers and criticality control materials.Work was co-sponsored by DARPA DSO and DOE OSTI. This work was done under the auspices of the DOE by LLNL under Contract Number W-7405-Eng-48.
10:45 AM - NN8.5
A Review of 25 years of Corrosion Studies on HLW Container Materials at the CEA.
Max Helie 1 Show Abstract
1 DPC/SCCME, French Atomic Energy Commission, Gif sur Yvette France
11:30 AM - NN8.6
Comparison between welding experiment and Finite Element Modelling in Friction Stir Welding of Copper joints.
Therese Kallgren 1 , Lai-Zhe Jin 1 , Rolf Sandström 1 Show Abstract
1 Department of Materials Science and Engineering, Royal Institute of Technology, Stockholm Sweden
11:45 AM - NN8.7
Corrosion Resistance of Iron-based Structural Amorphous Metals.
Tiangan Lian 1 , Summer Day 1 , Joe Farmer 1 Show Abstract
1 Chemistry and Materials Science, Lawrence Livermore National Laboratory, Livermore, California, United States
The potential advantages of amorphous metals have been recognized for some time [Latanison 1985]. Iron-based corrosion-resistant, amorphous-metal coatings under development may prove important for maritime applications [Farmer et al. 2005]. Such materials could also be used to coat the entire outer surface of containers for the transportation and long-term storage of spent nuclear fuel, or to protect welds and heat affected zones, thereby preventing exposure to environments that might cause stress corrosion cracking [Farmer et al. 1991 & 2000]. In the future, it may be possible to substitute such high-performance iron-based materials for more-expensive nickel-based alloys, thereby enabling cost savings in a wide variety of industrial applications. It should be noted that thermal-spray ceramic coatings have also been investigated for such applications [Haslam et al. 2005].This report focuses on the corrosion resistance of a yttrium-containing amorphous metal, SAM1651. SAM1651 has a glass transition temperature of ~584°C, a recrystallization temperature of ~653°C, and a melting point of ~1121°C. The measured critical cooling rate for SAM1651 is ≤ 80 K per second, respectively. The yttrium addition to SAM1651 enhances glass formation, as reported by Guo and Poon . The corrosion behavior of SAM1651 was compared with nickel-based Alloy 22 in electrochemical polarization measurements performed in several highly concentrated chloride solutions.Corrosion tests were performed on as-received SAM1651 vacuum arc-melted drop cast ingots. The nominal composition of SAM1651 material is listed in Table 1. Each electrochemical test includes a potentiodynamic polarization measurement after a 24-hour immersion in test solutions. The test solutions included 3.5 m NaCl, 6 m NaCl, 5 M CaCl2, and seawater from Half Moon Bay, California. All tests were conducted at 90°C. The Alloy 22 test data was generated on 5/8 inch diameter disc specimens. The composition of Alloy 22 is also listed in Table 1.Based on preliminary test results, SAM1651 demonstrates a promising corrosion resistance that is comparable to that of Alloy 22, the preferred material for the outer barrier of nuclear waste storage containers.
12:00 PM - NN8.8
Cermet Spent Nuclear Fuel Casks and Waste Packages.
Charles Forsberg 1 , Leslie Dole 1 Show Abstract
1 Nuclear Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Cermet spent nuclear fuel (SNF) casks made of steel, depleted uranium dioxide (DUO2), and other materials are being developed for storage, transport, and disposal of SNF. The results of a series of experimental and theoretical studies to develop an integrated cermet cask system are described. The performance of a cask ultimately depends upon the materials available to construct casks. Cermets, with ceramics embedded in metals, have potentially better materials properties than any other alternative to meet the multiple, conflicting requirements for SNF casks.As a SNF container for storage, transport, and disposal, a cask has multiple requirements: provide a structurally strong container, provide gamma shielding, provide neutron shielding, resist external events including assault, minimize SNF temperature (hence degradation) by removal of SNF decay heat, minimize weight within transport requirements, and minimize size within transport requirements. As a waste package (WP), the cermet cask would be the internal container with an overpack, such as C-22 for the Yucca Mountain repository. The cermet cask can potentially improve the long-term performance of the repository by creating a local geochemical environment to slow radionuclide releases after the corrosion-resistant overpack fails by control of the chemical composition of the cermet. Iron and DUO2 in the cask provide a locally reducing environment that slows SNF degradation and creates conditions that slow groundwater flow though the degraded SNF. The degradation products of DUO2 sorb neptunium while the equivalent iron products slow the release of technetium. The DU reduces the long-term probability of nuclear criticality caused by enriched uranium migration. The recent experimental and analogue data supporting these preliminary conclusions is described.Three methods are being investigated to produce the cermet casks. Two methods are based on powder metallurgy techniques and can produce a variable composition cermet where the properties of the cermet vary from the inside to the outside of the cask. The third method, casting, has the potential of very low costs.
12:15 PM - NN8.9
Oxide Film Aging on Alloy 22 in Halide Containing Solutions.
Ricardo Carranza 1 , Martín Rodríguez 1 , Raúl Rebak 2 Show Abstract
1 Dpt. Materiales, CNEA, San Martín, Buenos Aires, Argentina, 2 Corrosion Group, LLNL, Livermore, California, United States
The aim of the present work is to study the passive behavior of Alloy 22 in chloride and fluoride containing solutions varying the heat treatment of the alloy, the halide concentration and the pH of the solution at 90°C. General corrosion behavior was studied using electrochemical techniques, which included open circuit potential monitoring over time, potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) measurements carried out at open circuit and at passivity potentials. Corrosion rates obtained by EIS measurements after 24 h immersion were below 0.5 um/year. The corrosion rates were practically independent of the solution pH, short term corrosion potential (Ecorr), alloy heat treatment and halide ion nature and concentration. Polarization resistance (RP) values increased with open circuit potential and polarization time at constant potential in 1M NaCl, pH 6, 90°C. This was attributed to an increase in the oxide film thickness and oxide film aging. Capacitance measurements indicated that passive oxide on Alloy 22 presented a double n-type/p-type semiconductor behavior in the passive potential range.
12:30 PM - NN8.10
Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials.
Darrell Dunn 2 , Yi-Ming Pan 1 , Xihua He 1 , Lietai Yang 1 , Roberto Pabalan 1 Show Abstract
2 , Southwest Research Institute, San Antonio, Texas, United States, 1 , Center for Nuclear Waste Regulatory Analyses, San Antonio, Texas, United States
12:45 PM - NN8.11
The Influence of Pre-oxidation on the Corrosion of Copper Nuclear Waste Canisters in Aqueous Anoxic Sulphide Solutions.
Jared Smith 1 , Zack Qin 1 , David Shoesmith 1 Show Abstract
1 Chemistry, University of Western Ontario, London, Ontario, Canada
It has been proposed that natural groundwater sulphide sources may be of sufficient concentration to render Scandinavian/Canadian high-level waste containers thermodynamically unstable in anticipated granitic repository environments. Repository conditions are expected to evolve from initially warm and oxic to eventually cool and anoxic and at the time of water saturation it is anticipated that the canister surface will be covered with various corrosion products (namely copper oxides/hydroxides containing anions such as carbonate and sulphate). It is therefore necessary to investigate the effect of dissolved sulphide on pre-oxidized copper surfaces.Using a variety of electrochemical and surface analytical techniques, the mechanism and corrosion kinetics of sulphide breakthrough on oxide covered Cu surfaces in anoxic, aqueous environments are being investigated.Corrosion potential and cathodic stripping voltammetric measurements are being used to follow the evolution of oxide/hydroxide covered surfaces on exposure to sulphide. Since the evolution may occur at local sites rather than generally across the exposed surface, we are using electrochemical impedance spectroscopy to characterize the nature of film conversion processes for sulphide concentrations in the range 10-5 to 10-3 mol/L. Changes in surface morphology and composition are being followed using scanning electron microscopy and in-situ Raman spectroscopy. Of critical importance is whether or not a localized attack by sulphide can lead to the maintenance of pitting or not.The primary goal of these studies is to determine whether or not a period of pre-oxidation of a Cu container surface can prevent subsequent reaction of the surface with remotely produced sulphide.
NN9: Waste Form Radiation Damage and Dissolution
Wednesday PM, November 29, 2006
Constitution A (Sheraton)
2:30 PM - NN9.1
Simulations of Fluorite Related Ceramics for use as Waste Forms: A2B2O7 Pyrochlore and A4B3O12 δ–phase.
Antony Cleave 1 , Robin Grimes 1 , Kurt Sickafus 2 Show Abstract
1 Materials Department, Imperial College, London United Kingdom, 2 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
2:45 PM - NN9.2
Radiation Tolerance of A2B2O7 Materials - A Question of Bonding.
Karl Whittle 1 , Gregory Lumpkin 2 , Katherine Smith 2 , Mark Blackford 3 , Nestor Zaluzec 2 , Elizabeth Harvey 4 Show Abstract
1 Engineering Materials, University of Sheffield, Sheffield United Kingdom, 2 Institute of Materials and Engineering Science, Australian Nuclear Science and Technology Organisation, Sydney, New South Wales, Australia, 3 Materials Science Division , Argonne National Laboratory, Chicago, Illinois, United States, 4 Department of Earth Sciences, University of Cambridge, Cambridge United Kingdom
Ceramic waste forms provide attractive alternatives to the direct disposal of spent fuel or the immobilisation of high-level radioactive waste in borosilicate glass waste forms, due to their compliance with the principle nuclear safeguards agreements and relatively high aqueous durability, respectively. Over the design lifetime of the waste form, the actinide waste species will undergo alpha decay releasing alpha particles and alpha recoil nuclei. These particles interact with host lattices. In some cases, this will lead to a crystalline-amorphous transformation, volume expansion, cracking and reduced chemical durability due to increased surface area and decreased thermodynamic stability. Consequently there have been many studies of the radiation damage response of various potential waste form phases: including pyrochlores, defect fluorites, perovskites etc.Some aspects of alpha decay damage can be simulated by irradiation with heavy ions. Consequently many in situ ion irradiation experiments have been conducted on various complex oxides. Two important parameters established in such studies are: the critical amorphisation dose, Dc, at a particular temperature and the critical temperature above which a material remains crystalline, Tc. Recently we have determined a mechanism for predicting Tc, using previously published results for pyrochlore/fluorites based on structural, electronegativity, and energetic parameters. This model has been used to predict the radiation response for III-IV pyrochlore and defect fluorite compounds.A2Ti2O7 compounds, where A is a lanthanide element form two different structure types at room temperature, La-Nd form a perovskite related structure, while Sm-Lu form as pyrochlores. The results show that Tc increases with decreasing rA/rB, with a maximum at Gd, then drops until Lu.Since the nature of the materials change as you proceed across the period, it is unlikely that a simple ionic radius approach is valid. Therefore in order to fully understand the stability of these materials to irradiation an examination of the structure, bonding, and crystal stability has been undertaken. Using the results from such analysis an 'holistic' approach to the stabilty of these materials to radiation damage is discussed.
3:00 PM - NN9.3
Damage Accumulation in Au-irradiated Sr2Nd8(SiO4)6O2
Yanwen Zhang 1 , In-Tae Bae 1 , William Weber 1 , Mikio Higuchi 2 Show Abstract
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 Japan
Single crystal rare-earth silicates with the apatite structure are potentially valuable inertial matrices to actinide and some long-lived fusion product confinement. Accumulation of radiation damage in the host phases may ultimately compromise the physical and chemical durability. Thus, it is important to understand and predict the behavior of the apatite materials in a radiation environment.Alpha decay of actinide elements produces 4.5 to 5.8 MeV alpha particles and 70 to 100 keV recoil nuclei (alpha recoils). The more massive and lower energy alpha recoils that are produced account for most of the crystal damage through elastic scattering collisions. Since the peak in nuclear stopping of heavy ions, such as Au, is similar to the nuclear stopping of alpha recoils, it is expected that damage accumulation and amorphization processes induced by heavy-ion irradiation in the damage peak region do not differ significantly from those produced through alpha decay in the waste materials. In current studies, ion-induced damage has been investigated in single crystalline Sr2Nd8(SiO4)6O2 using 1.0 MeV Au ions at 160 K and room temperature. Rutherford backscattering spectrometry using 2.0 MeV He+ beam was carried out along the channeling direction to study damage accumulation behavior. Damage accumulation at both irradiation temperatures has been determined as the relative disorder on the Sr sublattice at the damage peak as a function of local dose. A disorder accumulation model, with contributions from the amorphous fraction and the crystalline disorder, has been fit to the Sr damage accumulation data. The results indicate that defect-stimulated amorphization is the primary amorphization mechanism in Sr2Nd8(SiO4)6O2.
3:15 PM - NN9.4
Enhancement of Zirconolite Dissolution Due to Water Radiolysis.
Magaly Tribet 1 , Nelly Toulhoat 1 , Nathalie Moncoffre 1 , Pierre Toulhoat 2 , Christophe Jegou 3 , Catherine Corbel 4 , Florence Bart 3 , Isabelle Bardez 3 Show Abstract
1 CNRS/IN2P3 Universite Claude Bernard Lyon 1, Institut de Physique Nucleaire de Lyon, 69622 Villeurbanne cedex France, 2 Universite Claude Bernard Lyon 1, CNRS/ISA, 69622 Villeurbanne cedex France, 3 , CEA/DEN/DTCD/SECM, BP 17171 30207 Bagnols sur Ceze cedex France, 4 CEA/DSM/DRECAM/LSI , Ecole Polytechnique, 91128 Palaiseau France
Zirconolite is one of the matrices foreseen for the confinement of minor actinides in case of deep geological disposal. Zirconolite (general formula : CaZrxTi(3-x)O7 (0.8 < x <1.37)) is able to incorporate rare earth elements (Nd, Ce, La, Hf, Gd) and actinides (U, Np, Am, Cm, Pu) by substitution in calcium and zirconium sites. Moreover, its chemical durability, even under alpha self-irradiation, makes this ceramics a potential candidate for the containment of minor actinides from reprocessing of nuclear spent fuel. In this work we present the effects of water radiolysis induced by charged particles (alphas or protons) on the dissolution of a synthetic sintered zirconolite. The formula of this zirconolite is Ca0.8Nd0.2ZrTi.8Al0.2O7 where Nd simulates the presence of trivalent and tetravalent actinides.We performed the irradiations with external ion beams in two distinct geometries where the fluences ranged from 1E15 to 1E16 ions/cm2. In the first geometry the beam gets through the sample before stopping at the surface/water interface. In the second one the beam stops before the surface/water interface. The use of these different configurations allows to study the respective influence of parameters such as sample irradiation, Linear Energy Transfer at the surface/water interface or total deposited energy.The irradiations were performed both on crystalline and amorphous zirconolites in pure water or with complexing ions such as F-. The sample dissolution has been monitored through the release of cations measured by ICP-MS. The radiolytic production of H2O2 has been measured by UV-VIS spectrophotometry. Our results show that both the irradiation and/or the water radiolysis have an effect on the preferential release of Zr, Ti and Nd. This behavior is enhanced (about 100 times) in presence of complexing species such as fluoride ions. These experiments are discussed with respect to experiments performed out of irradiation in pure water or with H2O2 in order to decouple the effects due to H2O2 from the radical precursors produced during water radiolysis.
3:45 PM - NN9.6
Kinetic and Thermodynamic Study of the Chemistry of Neoformed Phases during the Dissolution of Phosphate Based Ceramics.
Erwan du Fou de Kerdaniel 1 , Nicolas Dacheux 1 , Nicolas Clavier 1 , Renaud Podor 2 Show Abstract
1 , Insitut de Physique Nucléaire d'Orsay, Orsay France, 2 , LCSM, Vandoeuvre lès Nancy France
In the field of the immobilization of tri- and tetravalent actinides coming from the spent nuclear fuels, several phosphate-based ceramics (β-TPD, β-TPD / monazite composites monazite / brabantite solid solutions, or britholites) can be considered as potential candidates. Their chemical durability during aqueous alteration was then studied using under- and over-saturation processes. The retention phases precipitated in the back-end of the dissolution were then extensively characterized (grazing XRD, EPMA, SEM including X-EDS, µ-Raman) and then the associated solubility constant were evaluated. In order to study the thermodynamic stability of such phases, over-saturation experiments first using lanthanides as surrogates for actinides were carried out. Three phosphate – based phases (monazite, rhabdophane and xenotime) were precipitated depending on the ionic radius of the element, the temperature and the heating time. The associated kinetic of precipitation was examined. A similar study involving actinides (Th, U) was undertaken. Unknown (Ln, Ca, AnIV) rhabdophane-type phases were prepared with various compositions. The evolution of this (Ln, Ca, AnIV) rhabdophane phase is observed after few months with the precipitation of lanthanides in monazite, rhabdophane or xenotime, on the one hand and the preferential precipitation of actinides in the already described Thorium Phosphate Hydrogen Phosphate Hydrate (TPHPH), on the other hand. The formation of such phases was also checked at the surface of leached phosphate or phospho-silicate based ceramics using under-saturation experiments. Moreover, other phases such as hydrated silica (SiO2 . nH2O) and Nd1-2xCaxThx-yUyPO4 . 0.5H2O which also crystallized in the rhabdophane structure were also observed at the surface of leached britholite while it was not completely characterized at the surface of leached monazite / brabantite solid solutions consequently to its low quantity even after 1 year of leaching in aggressive conditions. Due to their rapid precipitation and their very low solubility constants, such phases could act as protective layers and are then of great interest for the evaluation of the long term behaviour of phosphate ceramics loaded with actinides.
NN10: Waste Forms for LLW and ILW
Wednesday PM, November 29, 2006
Constitution A (Sheraton)
4:30 PM - NN10.1
Geopolymers as Candidates for Low/Intermediate Level Highly Alkaline Waste.
D. Perera 1 , Eric Vance 1 , S. Kiyama 1 , Z. Aly 1 , P. Yee 1 Show Abstract
1 , ANSTO, Lucas Heights, New South Wales, Australia
4:45 PM - NN10.2
Pretreatment of Tc-Containing Waste and Its Effect on Tc-99 Leaching from Grouts
Albert Aloy 1 , Elena Kovarskaya 1 , John Harbour 2 , Christine Langton 2 , Bill Holtzscheiter 2 Show Abstract
1 , Khlopin Radium Institute, Saint Petersburg Russian Federation, 2 , Savannah River National Laboratory, Aiken, South Carolina, United States
This paper presents results of studies on how pretreatment of alkaline, Tc-99-containing waste affects leach rates from grout compositions based on Portland cement (PC) only, or mixes also containing granulated blast furnace slag (BFS).Pretreatment of the waste simulant was associated with adding a certain quantity of Fe(+2) and Fe(+3) compounds, finely dispersed magnetite powder, as well as slag. We found that the maximum Tc-99 co-precipitation with Fe(OH)3 (Kd=2700-3100 cm3/g) can be achieved with adding 1-2 mg Fe2+/ml (Fe(+2) sulfate ) of radioactive waste . For the waste grouting, we used pure PC (PC-500), or a mixture of PC-500 (25%) and BFS (75%) with a range of water-to-cement ratio of 0.4 to 0.5. The BFS characteristics will be provided in the paper. After the grout mixes cured, we used the modified leaching protocol ASTM D-5233 to measure Tc-99 leaching. The comparison of Tc-99 leaching from various samples shows that preliminary Tc-99 co-precipitation with Fe2+ reduces leaching by approximately a factor of two if pure PC-500 is used. However, if the PC:BFS combination is used, this effect is less noticeable, and on day 27 the difference was only a small fraction of a percent, thereby demonstrating that, in the long term, it is the reducing capability of the BFS that will play the key role in Tc-99 stabilization.Using the ANSI/ANS-16.1 protocol, we studied the kinetics of Tc-99 leaching into distilled water on monolith samples of the slag containing grout compounds. From these results the effective diffusion coefficient (DTc=4.75 x 10-12 cm2/sec) and leach index ( LTc=11.2) were calculated for the cured grout monolith.We used these values, as well as data on porosity and density of the monolith samples to model Tc-99 behavior (oxidation and transport) in large cement subsurface disposal configurations using the commercially available PORFLOW code.
5:00 PM - NN10.3
Frequency Characteristics of Acoustic Emission in Cementitious Wasteforms with Encapsulated Al.
Lyubka Spasova 1 , Michael Ojovan 1 Show Abstract
1 Immobilisation Science Laboratory, Engineering Materials, University of Sheffield, Sheffield United Kingdom
Acoustic emission(AE) frequency characterisation was used to evaluate the mechanical performance of cementitious wasteform with encapsulated Al waste. Fast Fourier Transformation(FFT) analysis has been applied to differentiate the waveforms detected at different stages of AE monitoring on ordinary Portland cement(OPC) structure. Correlations were identified between the frequency spectrum of the AE signals, the time of their appearance in highest population and the initiation and development of cement matrix fracture process. In the UK cementation of low and intermediate level waste (LILW) has been established as a successful immobilisation formula and developed as a commercially available technology [1,2]. The UK Radioactive Wastes Inventory  reports for a large amount of metallic wastes among which 980 tonnes of Al considered to be encapsulated in cement matrix. Durability of the cementitious wasteforms is a substantial factor for the safety of immobilised wastes. Al due to the enhanced corrosion rate in the high pH cement environment has been proved as a process causing wasteform cracking . Degradation of wasteforms is associated with AE events and these can be used to assess the level of damage induced. Indeed AE technique has been proved to be sensitive to formation of large cracks within the cement-based structures as well as to microcrack nucleation and development . Recently we demonstrated that the AE monitoring of cement matrix encapsulating Al bars gives information on different stages of Al corrosion and matrix degradation . This work aims to examine the frequency spectrum of the AE waveforms. A wideband(WD) piezoelectric transducer from Physical Acoustic Corp. was used for detection of AE signals from the OPC sample. The WD transducer has operational range 100-1000 kHz with a resonance at 125 kHz. The results revealed that the frequency band of detected AE signals was spread between 20 and 550 kHz with Fourier peaks at frequencies different from the resonance of the transducer. The FFT power spectrum of the waveforms emitted during different stages of AE monitoring showed that the events in cement matrix are generated by several individual sources. Our results also revealed that frequency range and Fourier peaks distribution can be potentially used to distinguish detected AE activity. Moreover AE waves frequency analysis can be applied as an additional tool for assessment of the mechanical stability of cementitious wasteforms with encapsulated Al. This is of significant importance for developing non-destructive techniques for long-term monitoring of wasteforms. 1. R. Steatfield. Proc. WM’01, Tucson, AZ (2001). 2. 2004 UK Radioactive Waste Inventory, Main Nirex Report N/090, October 2005. 3. A. Setiadi, N.B. Milestone, M. Hayes. Ext. Abstract Cem. Concr. Sci. Conf. 2004, Warwick. 4. K. Wu, B. Chen, W. Yao. J. Cem. Concr. Res., 30, 1495 (2000). 5. L.M. Spasova, M.I. Ojovan. J. Hazardous Materials, 136, No.2 (2006).
5:15 PM - NN10.4
Fluidized Bed Steam Reformed (FBSR) Mineral Waste Forms: Characterization and Durability Testing.
Carol Jantzen 1 , James Marra 1 , John Pareizs 1 , Troy Lorier 2 Show Abstract
1 , Savannah River National Laboratory, Aiken, South Carolina, United States, 2 , Savannah River Site, Aiken, South Carolina, United States
Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates aqueous low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS) to a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The FBSR technology converts organic compounds to CO2 and H2O by pyrolysis, converts nitrate/nitrite species to N2, and produces a solid residue through reactions with superheated steam, the fluidizing media. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750°C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals appear to stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Samples of FBSR mineral waste forms were made in a pilot-scale FBSR at INL with (1) a Hanford Envelope A low-activity waste (LAW) simulant and (2) an INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The objectives were to obtain the overall durability of the FBSR waste form products compared to previous FBSR waste form testing and compared to vitreous LAW waste forms. The SPFT testing was performed at 5 different pH values and 4 different temperatures for 3-14 day test durations. Steady state conditions were achieved between 8-14 day test durations at the high flow rate (288 mL/day) tested. The forward dissolution rate and the activation energies for dissolution for Hanford’s and INL’s FBSR waste forms was determined and compared to the release behavior of FBSR pilot scale products previously tested by PNL for Hanford Envelope C FBSR waste forms made by THOR Treatment Technologies (TTT). A qualitative comparison of the FBSR bed products was assessed against the low-activity reference material (LRM) glass developed for Hanford low activity waste (LAW) qualification. The durability of the FBSR waste form is comparable to that of LAW glass for all of the test responses studied.
5:30 PM - NN10.5
Dissolution Testing of a Metallic Waste Form in Chloride Brine.
Dawn Janney 1 Show Abstract
1 , Argonne National Laboratory--West (current affiliation: Idaho National Laboratory), Idaho Falls, Idaho, United States
5:45 PM - NN10.6
Characterisation and Durability of Plasma Vitrified Simulant Plutonium Contaminated Waste Material.
Neil Hyatt 1 , Suzy Morgan 1 , Charlie Scales 2 Show Abstract
1 Department of Engineering Materials, The University of Sheffield, Sheffield United Kingdom, 2 , Nexia Solutions Ltd., Sellafield United Kingdom
Production of plutonium by reprocessing of nuclear fuel gives rise to a large volume of waste plutonium contaminated material (PCM). This waste is heterogeneous in nature and includes plutonium contaminated metal, polymer, organic and inorganic materials. Plasma vitrification offers considerable potential in the treatment of such compositionally heterogeneous waste, simultaneously vitrifying the non-combustible waste fraction in a glassy-slag and incinerating the combustible waste fraction resulting in a substantial volume reduction. We have examined the phase assemblage, microstructure and durability of the product arising from plasma vitrification of a simulant PCM waste, spiked with 1.23 wt% CeO2 to simulate the presence of PuO2. The simulant PCM waste contained aluminium, steel, glass, masonry, soil, polymers and organic material in appropriate ratios. The simulant PCM waste was vitrified in a calcium aluminosilicate glass (composition: CaO, 29.2 wt%; Al2O3 27.7 wt%; SiO2 43.2 wt%), using a twin torch cold skull plasma reactor. Plasma vitrified compositions containing 30.6 wt%, 40.0 wt% and 50.6 wt% of the simulant PCM waste were found to form a glassy slag-like material. Characterisation of this slag-like material, by X-ray powder diffraction and Scanning Electron Microscopy, confirmed the formation of a calcium aluminosilicate glass matrix with minor crystalline inclusions of alumina, spinel and mullite. The plutonium surrogate, cerium, was found to partition into the calcium aluminosilicate matrix.The durability of the plasma vitrified products was examined using the ASTM Product Consistency Test at 90oC, with a test period between 3 and 28 days. In de-ionised water, the normalised Ca and Si release rates of the plasma vitrified products containing 30.6 wt%, 40.0 wt% and 50.6 wt% of the simulant PCM waste were found to be ~1.0 gm-2d-1, independent of composition. The Ce release rates from all glass compositions were found to be below the limit of detection. Analysis of the recovered glass powder by Scanning Electron Microscopy revealed the formation of calcium and iron aluminosilcate alteration products. The durability of the plasma vitrified product containing 50.4 wt% of the simulant PCM waste was also examined in a pH 11.0 buffer solution of 0.025 M KHCO3 and 0.015 M KOH to simulate the effect of the alkaline plume resulting from a cementitious repository environment. The Ca release rate in this alkaline buffer solution was found to be ~3.0 gm-2d-1; in contrast, the Ce release rate was found to be below the limit of detection.These results demonstrate that plasma vitrification may be utilised to immobilise PCM wastes in a durable calcium aluminosilicate glass, with a wide tolerance for waste loading and waste stream composition.
NN11: Poster Session: Waste Forms, Radionuclides and Disposal
Wednesday PM, November 29, 2006
Exhibition Hall D (Hynes)
9:00 PM - NN11.1
Simulation of Self-irradiation of High-sodium Content Nuclear Waste Glasses.
Michael Ojovan 1 , Olga Batyukhnova 2 , William (Bill) Lee 3 , Alexey Pankov 1
1 Immobilisation Scie